WorldWideScience

Sample records for channels reactor

  1. Inspection of Candu Nuclear Reactor Fuel Channels

    International Nuclear Information System (INIS)

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

  2. Channel-type nuclear reactor with a boiling coolant

    International Nuclear Information System (INIS)

    The invention is aimed at increasing the channel-type reactor safety, in particular, RBMK-type reactors, during accidents resulting in the coolant circulation discontinuation. The reactor core is assembled of vertial technological channels connected in parallel between distributing group collectors and drum-separator. Each technological channel contains a high pressure tube, a fuel assembly with fuel elements and a storage vessel located above the fuel assembly which is filled with water at saturation temperature in the normal operation regime. After dehydration of channels in the course of accident the boiling water from storage vessel is ejected into them. So the device described allows one to reduce the fuel element can temperature in the course of accidents connected with the coolant circulation discontinuation and so to increase the plant safety level

  3. Decommissioning of nuclear reactor fuel channels using laser technology

    Science.gov (United States)

    Panchenko, Vladislav Y.; Zabelin, Alexandre M.; Slepokon, Yu. I.; Ryahin, V. M.; Kuznetsov, P. P.; Panasyuk, V. F.; Korotchenko, A. V.; Kislov, V. S.; Loktev, S. V.

    2000-07-01

    Decommissioning of nuclear reactors using laser remote dismounting and welding was experimentally proved at a nuclear reactor of Kursk Nuclear Power Plant. The main reason of laser beam application in this case is the marked decrease of radioactive exposure of the service personnel. The use of a high-power laser beam provided for laser cutting and welding processes realization at a distance up to 35 m between the laser and the workstation placed behind a radiation shield. By application of laser cutting gas and dust contamination is ten-fold decreased. Some results of decommissioning application of a stationary laser workstation based upon a 5 kW fast-transverse-flow discharge CW CO2 laser TL-5M installed at a nuclear reactor site are presented. A special high-beam- quality model of the laser was developed to satisfy the needs of decommissioning. Laser cutting process was applied to decommissioning of fuel channels (FC) of RBMK-1000 reactor, after their extractor from the reactor active zone during the procedure of channels replacement.

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

    Energy Technology Data Exchange (ETDEWEB)

    Na, Jonggeol; Jung, Ikhwan; Kshetrimayum, Krishnadash S.; Park, Seongho; Park, Chansaem; Han, Chonghun [Seoul National University, Seoul (Korea, Republic of)

    2014-12-15

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-04-15

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

  6. Numerical study of the effects of lamp configuration and reactor wall roughness in an open channel water disinfection UV reactor.

    Science.gov (United States)

    Sultan, Tipu

    2016-07-01

    This article describes the assessment of a numerical procedure used to determine the UV lamp configuration and surface roughness effects on an open channel water disinfection UV reactor. The performance of the open channel water disinfection UV reactor was numerically analyzed on the basis of the performance indictor reduction equivalent dose (RED). The RED values were calculated as a function of the Reynolds number to monitor the performance. The flow through the open channel UV reactor was modelled using a k-ε model with scalable wall function, a discrete ordinate (DO) model for fluence rate calculation, a volume of fluid (VOF) model to locate the unknown free surface, a discrete phase model (DPM) to track the pathogen transport, and a modified law of the wall to incorporate the reactor wall roughness effects. The performance analysis was carried out using commercial CFD software (ANSYS Fluent 15.0). Four case studies were analyzed based on open channel UV reactor type (horizontal and vertical) and lamp configuration (parallel and staggered). The results show that lamp configuration can play an important role in the performance of an open channel water disinfection UV reactor. The effects of the reactor wall roughness were Reynolds number dependent. The proposed methodology is useful for performance optimization of an open channel water disinfection UV reactor.

  7. Numerical study of the effects of lamp configuration and reactor wall roughness in an open channel water disinfection UV reactor.

    Science.gov (United States)

    Sultan, Tipu

    2016-07-01

    This article describes the assessment of a numerical procedure used to determine the UV lamp configuration and surface roughness effects on an open channel water disinfection UV reactor. The performance of the open channel water disinfection UV reactor was numerically analyzed on the basis of the performance indictor reduction equivalent dose (RED). The RED values were calculated as a function of the Reynolds number to monitor the performance. The flow through the open channel UV reactor was modelled using a k-ε model with scalable wall function, a discrete ordinate (DO) model for fluence rate calculation, a volume of fluid (VOF) model to locate the unknown free surface, a discrete phase model (DPM) to track the pathogen transport, and a modified law of the wall to incorporate the reactor wall roughness effects. The performance analysis was carried out using commercial CFD software (ANSYS Fluent 15.0). Four case studies were analyzed based on open channel UV reactor type (horizontal and vertical) and lamp configuration (parallel and staggered). The results show that lamp configuration can play an important role in the performance of an open channel water disinfection UV reactor. The effects of the reactor wall roughness were Reynolds number dependent. The proposed methodology is useful for performance optimization of an open channel water disinfection UV reactor. PMID:27108375

  8. A new safety channel based on ¹⁷N detection in research reactors.

    Science.gov (United States)

    Seyfi, Somayye; Gharib, Morteza

    2015-10-01

    Tehran research reactor (TRR) is a representative of pool type research reactors using light water, as coolant and moderator. This reactor is chosen as a prototype to demonstrate and prove the feasibility of (17)N detection as a new redundant channel for reactor power measurement. In TRR, similar to other pool type reactors, neutron detectors are immersed in the pool around the core as the main power measuring devices. In the present article, a different approach, using out of water neutron detector, is employed to measure reactor power. This new method is based on (17)O (n,p) (17)N reaction taking place inside the core and subsequent measurement of delayed neutrons emitted due to (17)N disintegration. Count and measurement of neutrons around outlet water pipe provides a reliable redundant safety channel to measure reactor power. Results compared with other established channels indicate a good agreement and shows a linear interdependency with true thermal power. Safety of reactor operation is improved with installation & use of this new power measuring channel. The new approach may equally serve well as a redundant channel in all other types of reactors having coolant comprised of oxygen in its molecular constituents. Contrary to existing channels, this one is totally out of water and thus is an advantage over current instrumentations. It is proposed to employ the same idea on other reactors (nuclear power plants too) to improve safety criteria. PMID:26123105

  9. Thermalhydraulic characteristics for fuel channels using burnable poison in the CANDU reactor

    International Nuclear Information System (INIS)

    The power coefficient is one of the most important physics parameters governing nuclear reactor safety and operational stability, and its sign and magnitude have a significant effect on the safety and control characteristics of the power reactor. Recently, for an equilibrium CANDU core, the power coefficient was reported to be slightly positive when newly developed Industry Standard Tool set reactor physics codes were used. Therefore, it is required to find a new way to effectively decrease the positive power coefficient of CANDU reactor without seriously compromising the economy. In order to make the power coefficient of the CANDU reactor negative at the operating power, Roh et al. have evaluated the various burnable poison (BP) materials and its loading scheme in terms of the fuel performance and reactor safety characteristics. It was shown that reactor safety characteristics can be greatly improved by the use of the BP in the CANDU reactor. However, the previous study has mainly focused on the safety characteristics by evaluating the power coefficient for the fuel channel using BP in the CANDU reactor. Together with the safety characteristics, the economic performance is also important in order to apply the newly designed fuel channel to the power plant. In this study, the economic performance has been evaluated by analyzing the thermal hydraulic characteristics for the fuel channel using BP in the CANDU reactor

  10. Neutron field for activation experiments in horizontal channel of training reactor VR-1

    International Nuclear Information System (INIS)

    The experimental channels of nuclear reactors often serve for nuclear data measurement and validation. The dosimetry-foils activation technique was employed to measure neutron field parameters in the horizontal radial channel of the training reactor VR-1, and to test the possibility of using the reactor for scientific purposes. The reaction rates, energy spectral indexes, and neutron spectrum at several irradiation positions of the experimental channel were determined. The experimental results show the feasibility of the radial channel for irradiating experiments and open new possibilities for data validation by using this nuclear facility. - Highlights: • Neutron activation analysis of various samples. • Neutron spectrometry and gamma-spectrometry. • Study of keff for various types of reactor core

  11. High conversion pressurized water reactor with boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Margulis, M., E-mail: maratm@post.bgu.ac.il [The Unit of Nuclear Engineering, Ben Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Shwageraus, E., E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, CB2 1PZ Cambridge (United Kingdom)

    2015-10-15

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–{sup 233}U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–{sup 233}U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm{sup 3}, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore

  12. High conversion pressurized water reactor with boiling channels

    International Nuclear Information System (INIS)

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–233U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–233U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm3, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore, some means of

  13. On the neutron spatial distribution in ionization chamber channels of the WWER type reactors

    International Nuclear Information System (INIS)

    Results of experimental and calculational studies permitting to estimate the neutron flux spatial distribution in ionization chamber channels of the commercial WWER-1000 and WWER-440 reactors and also of the WWER-440 reactor with water biological shield are presented. The integral neutron flux density distribution along the channel cross section approximately at height of the core middle and the corresponding thermal and fast neutron flux density distributions are measured by the activation detectors. It is shown that the difference in fast neutron flux density exceeds that of thermal neutrons. The commercial WWER-1000 type reactor the fast neutron flux density is decreased by the factor of 1.7, and thermal neutron flux density - by the factor of 1.2, for the commercial WWER-440 reactor these values are 1.37 and 1.18, and for the WWER-440 one with water shield - 1.5 and 1.18

  14. Measurement of neutron spectrum in vertical channel of IR-8 nuclear reactor

    CERN Document Server

    Kashchuk, Y A; Sevastyanov, V D; Egorov, E V

    2002-01-01

    Paper presents the experimental results of measurement of neutron flux spectrum and density in a dry vertical channel if the Kurchatov Institute RRC IR-8 research reactor. One obtained data due to application of 5 sets of various activation detectors. The spectrum was reconstructed on the basis of the KASKAD combined computer program used for the method of integral neutron detectors. The derived results enable to design most efficiently experiments to irradiate and to investigate into radiation resistance of the ITER diagnostic system various components in a new irradiation channel of IR-8 research reactor

  15. Experimental study of the IPR-R1 TRIGA reactor power channels responses

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Henrique F.A.; Ferreira, Andrea V., E-mail: hfam@cdtn.br, E-mail: avf@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The IPR-R1 nuclear reactor installed at Centro de Desenvolvimento da Tecnologia Nuclear CDTN/CNEN, Belo Horizonte, Brazil, is a Mark I TRIGA reactor (Training, Research, Isotopes, General Atomics) and became operational on November of 1960. The reactor has four irradiation devices: a rotary specimen rack with 40 irradiation channels, the central tube, and two pneumatic transfer tubes. The nuclear reactor is operated in a power range between zero and 100 kW. The instrumentation for IPR-R1 operation is mainly composed of four neutronic channels for power measurements. The aim of this work is to investigate the responses of neutronic channels of IPR-R1, Linear, Log N and Percent Power channels, and to check their linearity. Gold foils were activated at low powers (0.125-1.000 kW), and cobalt foils were activated at high powers (10-100kW). For each sample irradiated at rotary specimen rack, another one was irradiated at the same time at the pneumatic transfer tube-2. The obtained results allowed evaluating the linearity of the neutronic channels responses. (author)

  16. Reactor core protection system using a 4-channel microcomputer

    International Nuclear Information System (INIS)

    A four channel microcomputer system was fitted in Grafenrheinfeld NPP for local core protection. This system performs continuous on-line monitoring of peak power density, departure from nucleate boiling ratio and fuel duty. The system implements limitation functions with more sophisticated criteria and improved accuracy. The Grafenrheinfeld system points the way to the employment of computer based limitation system, particularly in the field of programming language, demarkation of tasks, commissioning and documentation aids, streamlining of qualification and structuring of the system. (orig.)

  17. Evaluation of surface deposits on the channel wall of trepanned reactor core graphite samples

    Science.gov (United States)

    Heard, P. J.; Payne, L.; Wootton, M. R.; Flewitt, P. E. J.

    2014-02-01

    Samples have been trepanned from the fuel and interstitial channel walls of PGA graphite reactor cores of two Magnox gas cooled power stations after a period of service. These samples have been considered explicitly for the presence of deposits on the channel facing surfaces. A combination of focused ion beam milling and imaging has been used to determine the presence of such deposits and where present to make measurements of the thickness. These thicknesses vary from a few nanometres to tens of micrometres. In addition, both the chemical composition and chemical state have been investigated using energy dispersive X-ray microanalysis in a scanning electron microscope and Raman spectroscopy respectively. EDX measurements showed that surface deposits found on the channel walls of one of the reactors contained increased concentrations of oxygen, iron, chromium and sulphur compared with the underlying material. Raman spectroscopy also suggested that the deposit had a smaller crystallite size than PGA graphite.

  18. Fast Traveling-Wave Reactor of the Channel Type

    CERN Document Server

    Rusov, Vitaliy D; Vashchenko, Volodymyr N; Chernezhenko, Sergei A; Kakaev, Andrei A; Pantak, Oksana I

    2015-01-01

    The main aim of this paper is to solve the technological problems of the TWR based on the technical concept described in our priority of invention reference, which makes it impossible, in particular, for the fuel claddings damaging doses of fast neutrons to excess the ~200 dpa limit. Thus the essence of the technical concept is to provide a given neutron flux at the fuel claddings by setting the appropriate speed of the fuel motion relative to the nuclear burning wave. The basic design of the fast uranium-plutonium nuclear traveling-wave reactor with a softened neutron spectrum is developed, which solves the problem of the radiation resistance of the fuel claddings material.

  19. Effect of Saliva on Measurement of Chemiluminescence by a Micro-Reactor Incorporating a Micro-Channel

    OpenAIRE

    Tsukagoshi, Kazuhiko; Fukumoto, Kazuaki; Nakajima, Riichiro; Yamashita, Kenichi; Maeda, Hideaki

    2007-01-01

    Effect of saliva on measurement of chemiluminescence was examined by a micro-reactor incorporating a micro-channel. Sodium hypochlorite and hydrogen peroxide solutions were delivered into a micro-channel developed in a micro-reactor by a syringe pump, providing a laminar flow liquid-liquid interface in the channel and leading to chemiluminescence from singlet oxygen. It was found under certain conditions including saliva that ca. 5% chemiluminescence of the total chemiluminescence was lost in...

  20. An Experimental Study of Natural Convection in The Hottest Channel of TRIGA 2000 k W Reactor

    International Nuclear Information System (INIS)

    With the increase of radioisotope demand, in 1995, National Nuclear Energy Agency of Indonesia made a decision to upgrade the power of the TRIGA Mark II reactor from 1 MW to 2 MW maximum power. The reactor reached its first criticality on May 13, 2000. To accomplish the safety evaluation of the reactor, a thermal hydraulic analysis was carried out by using thermal hydraulic computer code. This code calculates the natural convection flow through water coolant bounded by vertical cylindrical heat sources. In this paper, it will be reported the experimental study of natural convection in the hottest channel of TRIGA 2000 k W reactor. The purpose of the experimental study is to verify the theoretical analysis, especially the temperature distribution in the hottest coolant channel. In this experiment, a special probe for temperature detection has been designed and inserted to central thimble (CT). In the experiment, eight thermocouples were used to measure the bulk temperature of the water at different position in the cooling channel and simultaneous quantitative measurement of the temperature distribution were done by using a data acquisition cards system. The result obtained theoretically using the STAT code has been verified by this experimental study. (author)

  1. Interaction effects between laminar natural convection and surface thermal radiation in PHWR reactor channels

    Energy Technology Data Exchange (ETDEWEB)

    Chandramouli, D., E-mail: deepthi1223@gmail.com [National Inst. of Tech., Trichy (India); Rajendrakumar, M.; Velusamy, K.; Chellapandi, P., E-mail: mrk@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2011-07-01

    Pressurized Heavy Water Reactor (PHWR) reactor channel consists of a pressure tube (PT), which is concentrically placed inside the calendria tube (CT). After a hypothetical loss of coolant accident, sagging or ballooning of the PT may occur and physical contact between PT and CT takes place. Under this condition, knowledge of temperature distributions in PT and CT are essential to assess their structural integrity. Towards this, a 2-D CFD study has been carried out to understand the natural convection of CO{sub 2} and surface thermal radiation. Detailed parametric study has been carried for various values of temperature difference, emissivity and eccentricity. (author)

  2. Characterization of the neutron flux gradients in typical irradiation channels of a TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    The neutron distribution in a defined volume (gradient) for different matrices (air, water, cellulose, biological material and silicon dioxide) in two typical irradiation channels (pneumatic tube (PT) and IC40-channel in the carousel facility) in the TRIGA Mark II reactor at the Jozef Stefan Institute (IJS) was studied. Experiment was based on inserting Fe wires (flux monitors) into the chosen matrices. The wires were cut into small pieces after irradiation and the induced activities of 59Fe measured. The results showed that for the studied geometry the average spatial thermal neutron flux inhomogeneities (for five studied matrices) are about 2.3% in the PT-channel and about 2.9% in the IC40-channel. (author)

  3. Expected reactivity effect of fuel channel coolant boiling in the Darlington NGS A reactor core

    International Nuclear Information System (INIS)

    We have developed a formalism for estimating the expected reactivity due to channel boiling in any reactor designed to have some quality in the channel. In applying this formalism to the Darlington NGS A equilibrium core, we calculate a value of 0.024 ± 0.003 mk at 100% power operation. In Darlington, the channel feeders are individually sized so that the coolant in each channel has some boiling on reaching the entrance to the reactor outlet header. (Hereafter called the 'ROH quality'). The design is such that when each channel is at its nominal time-averaged 100 percent power, the quality at the ROH should be just under 2%. The day-to-day variation of each channel's power around its time-averaged value (i.e., 'ripple') results in a corresponding variation in the quality and consequently in the reactivity due to boiling. Traditionally, fuel management codes such as SORO, FMDP, RFSP and OHRFSP use fuel properties generated by a lattice code such as POWDERPUFS or LATREP. These fuel properties are functions of fuel irradiation only, with all other core-varying input parameters to the lattice code held constant at core-averaged values. Recently, some work has gone into developing a Pt. Lepreau version of RFSP in which the fuel properties are functions of fuel temperature and coolant density as well as of fuel irradiation. This paper reports the results of a study which was undertaken to quantify the expected variation in core reactivity due to this day-to-day variation in channel power and channel boiling. It could then be determined whether the reactivity effect of this boiling is sufficient to justify the explicit representation of the fuel properties as a function of coolant density

  4. Channel structures in aerobic biofilms of fixed-film reactors treating contaminated groundwater.

    Science.gov (United States)

    Massol-Deyá, A A; Whallon, J; Hickey, R F; Tiedje, J M

    1995-02-01

    Scanning electron microscopy, confocal scanning laser microscopy, and fatty acid methyl ester profiles were used to study the development, organization, and structure of aerobic multispecies biofilm communities in granular activated-carbon (GAC) fluidized-bed reactors treating petroleum-contaminated groundwaters. The sequential development of biofilm structure was studied in a laboratory reactor fed toluene-amended groundwater and colonized by the indigenous aquifer populations. During the early stages of colonization, microcolonies were observed primarily in crevices and other regions sheltered from hydraulic shear forces. Eventually, these microcolonies grew over the entire surface of the GAC. This growth led to the development of discrete discontinuous multilayer biofilm structures. Cell-free channel-like structures of variable sizes were observed to interconnect the surface film with the deep inner layers. These interconnections appeared to increase the biological surface area per unit volume ratio, which may facilitate transport of substrates into and waste products out of deep regions of the biofilm at rates greater than possible by diffusion alone. These architectural features were also observed in biofilms from four field-scale GAC reactors that were in commercial operation treating petroleum-contaminated groundwaters. These shared features suggest that formation of cell-free channel structures and their maintenance may be a general microbial strategy to deal with the problem of limiting diffusive transport in thick biofilms typical of fluidized-bed reactors. PMID:7574613

  5. Fluid modeling and design of gas channels of solar non-stoichiometric redox reactor

    Science.gov (United States)

    Kedlaya, Aditya

    The present numerical study in FLUENT analyzes the fluid flow field within a solar powered reactor designed for syngas production. The thermochemical reactor is based on continuous cycling of cerium oxide (ceria) in a non-stoichiometric reduction/oxidation cycle. The reactor uses a hollow cylinder of porous ceria which rotates through a high-temperature zone, by exposure to concentrated sunlight and partially reduced in an inert atmosphere due to flow of the sweep gas (N2), and then through a lower temperature zone where the reduced ceria is re-oxidized with a flow of CO2 and/or H2O, to produce CO and/or H2. In terms of fluid flow modeling, the issue of crossover of species (leakage) within the reactor is critical for proper functioning of the reactor. The first part of the work relates to the geometry and placement of the inlet/outlet gas channels for the reactor optimized to minimize crossover of the species. This is done by conducting a parametric study of geometric variables associated with the inlet/outlet geometry. A simplified 2D fluid flow reactor model which incorporates multi-species flow is used for this study. Further, 2D and 3D reactor models which capture the internal structure more accurately are used to refine the inlet/outlet design. The optimized reactor model is found to have an O2 crossover of 2%-6% and oxidizer crossover of 8%-21% at different flow rates of the sweep gas and the oxidizer studied. In the second part of the work, the reactor model is simulated under varying test conditions. Different working conditions include morphologies of the reactive material, rotational speed of the ceria ring and the recuperator, flow rates of sweep gas and the oxidizer, types of oxidizer (CO2, H2O). The 3D reactor model is also tested using one, two and three discrete inlet/outlet ports and compared with slot configuration.

  6. Hydrogen production in a zigzag and straight catalytic wall coated micro channel reactor by CFD modeling

    Energy Technology Data Exchange (ETDEWEB)

    Fazeli, Ali; Behnam, Mohsen [Gas Research Division, Research Institute of Petroleum Industry (RIPI), P.O. Box 14665-137, Tehran (Iran)

    2010-09-15

    Hydrogen production from steam reforming of methanol for fuel cell application was modeled in a wall coated micro channel reactor by CFD approach. Heat of steam reforming (SR) was supplied from catalytic total oxidation (TOX) of methanol on Cu/ZnO/Al{sub 2}O{sub 3} catalyst and Heat conducts from TOX to SR zone through Steel divider wall between two channels. Heat integration was compared in zigzag and straight geometry of microreactor by CFD modeling. The model is two dimensional, steady state and containing five zones: TOX fluid, TOX catalyst layer, steel wall of the channel, SR catalyst layer and SR fluid. Set of partial differential equations (PDEs) including x and y momentum balance, continuity, partial mass balances and energy balance was solved by finite volume method. Stiff reaction rates were considered for methanol total oxidation (TOX), methanol steam reforming (SR), water gas shift (WGS) and methanol decomposition (MD) reactions. The results show that zigzag geometry is better than straight one because heat and mass transfer in zigzag reactor are more than straight. Conversion of methanol in zigzag geometry is greater than straight one. In the outlet of zigzag micro channels, carbon monoxide selectivity is less and hydrogen mole fraction is more than straight one. (author)

  7. Simplified model for the thermo-hydraulic simulation of the hot channel of a PWR type nuclear reactor

    International Nuclear Information System (INIS)

    The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs

  8. Effects of cooling channel blockage on fuel plate temperature in Tehran Research Reactor

    Institute of Scientific and Technical Information of China (English)

    TABBAKH Farshid

    2009-01-01

    In this study, the variation of the temperature distribution of the fuel plate in Tehran Research Reactor core was studied in case of coolant channels blockage. While the experimental method is not possible, both the analytical and simulation methods were used to obtain the more reliable data. The results show that one channel blockage will increase the fuel temperature to about 100%, but it does not lead to clad melt down still. With further calculation and simulation it is understood that if the coolant velocity drops to 90% of its nominal value, it may causes the clad melt-ing down. At least two channels with complete blockage even at the positions far from the core center can also melt down the clad.

  9. Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN

    Energy Technology Data Exchange (ETDEWEB)

    Ammirabile, Luca, E-mail: luca.ammirabile@ec.europa.e [European Commission, JRC, Institute for Energy, Westerduinweg 3, 1755 LE Petten (Netherlands)

    2010-10-15

    In the Generation IV International Forum (GIF) program, the supercritical water reactor (SCWR) concept is among the six innovative reactor types selected for development in the near future. In principle the higher efficiency and better economics make the SCWR concept competitive with the current reactor design. Due to different technical challenges that, however exist, fuel assembly design represents a crucial aspect for the success of this concept. In particular large density variations, low moderation, heat transfer enhancement and deterioration have a strong effect on the core design parameters. Only a few computational tools are currently able to perform sub-channel thermal-hydraulic analysis under supercritical water conditions. At JRC-IE the existing sub-channel code COBRA-EN has been improved to work above the critical pressure of water. The water properties package of the IAPWS Industrial Formulation 1997 was integrated in COBRA-EN to compute the Thermodynamic Properties of Water and Steam. New heat transfer and pressure drop correlations more indicated for the supercritical region of water have also been incorporated in the code. As part of the efforts to appraise the new code capabilities, a code assessment was carried out on the hexagonal fuel assembly of a fast supercritical water reactor. COBRA-EN was also applied in combination with the neutronic code MCNP to investigate on the use of hydride fuel in the HPLWR supercritical water fuel assembly. The results showed that COBRA-EN was able to reproduce the results of similar studies with acceptable accuracy. Future activities will focus on the validation of the code against experimental data and the implementation of new features (counter-current moderator channel, wall, and wire-wrap models).

  10. Method Experimental Definition of Efficiency Fuelburn in a Gas Reactors with Mini Channels

    Directory of Open Access Journals (Sweden)

    Slesareva Ekaterina

    2016-01-01

    Full Text Available Approbation of a method of experimental definition of efficiency of a fuelburn in minichanneling gas reactors has been carried out. The method of visualisation of a field of temperatures by the temperature transducer made of fine-meshed with low-inertia use for definition of thermal structure of gas streams. A field of temperatures on an outlet the simulator of a gas reactor with minichannels registered on thermogram by infrared imager. Experiments have shown that the thermal imaging method provides the good resolutions on spatial and on time for a field of temperatures both in stationary and in non-stationary regime of gas flow and of heat release. Qualitative and quantitative character a change a temperatures on thermograms correspond to classical physical representations about nature of а gas flow in lengthy channels. The study showed that the method of thermal imaging allows in a regime real time defined a parameters of gas flow on outlet a channels of reactor which are necessary for definition of fuelburn efficiency.

  11. Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

    2015-07-01

    Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

  12. Development of the installation for zirconium alloy corrosion testing in superheated steam flow in a reactor channel

    International Nuclear Information System (INIS)

    The technological flowsheet of an installation for corrosion testing of structural materials in a reactor channel under conditions typical for the RBMK type reactors is described. The structural specific features of a capsule for zirconium alloy in-pile testing at superheated steam pressure of 2MPa and temperature upto 550 deg C are considered. The peculiarities of such capsule parameter calculation are discussed. On the base of the results of the installation performance testing in a reactor channel at power density of 1-1.5 W/g during 3500 h the conclusion on its high operational reliability is made

  13. Improved catalytic performance of Ni catalysts for steam methane reforming in a micro-channel reactor

    Institute of Scientific and Technical Information of China (English)

    Bozhao Chu; Nian Zhang; Xuli Zhai; Xin Chen; Yi Cheng

    2014-01-01

    Milliseconds process to produce hydrogen by steam methane reforming (SMR) reaction, based on Ni catalyst rather than noble catalyst such as Pd, Rh or Ru, in micro-channel reactors has been paid more and more attentions in recent years. This work aimed to further improve the catalytic performance of nickel-based catalyst by the introduction of additives, i.e., MgO and FeO, prepared by impregnation method on the micro-channels made of metal-ceramic complex substrate. The prepared catalysts were tested in the same micro-channel reactor by switching the catalyst plates. The results showed that among the tested catalysts Ni-Mg catalyst had the highest activity, especially under harsh conditions, i.e., at high space velocity and/or low reaction temperature. Moreover, the catalyst activity and selectivity were stable during the 12 h on stream test even when the ratio of steam to carbon (S/C) was as low as 1.0. The addition of MgO promoted the active Ni species to have a good dispersion on the substrate, leading to a better catalytic performance for SMR reaction.

  14. Installation of permanent cadmium-lined channel as a means for increasing epithermal NAA capabilities of miniature neutron source reactors

    International Nuclear Information System (INIS)

    Highlights: • High demand for epithermal neutrons necessitated the need of a permanent cadmium-line. • We reported the design specifications, preliminary studies done and steps followed. • Reactivity worth of the old channel = 0.12 mk and the new channel = 0.336 mk. • Temperature coefficient = −0.1 mk/°C and control rod worth coefficient = 0.023 mk/mm. • The work is a useful tool to the MNSR community for upgrading their reactors. -- Abstract: High demand for epithermal neutrons by the clients of the Nigerian Research Reactor-1 (NIRR-1), a Miniature Neutron Source Reactor (MNSR) has necessitated the need to explore avenues for increasing epithermal Neutron Activation Analysis (NAA) capabilities of the reactor. Safety and flux stability simulations were done by our group using Monte Carlo Transport Code MCNP5 for permanent cadmium line inside the irradiation channel of NIRR-1 and compared with the ones reported by other MNSR groups. The results of all these simulations revealed that the effect of cadmium-line on safety and flux stability is very minimal in the outer channel than in the inner channel. We have reported here the design specifications, preliminary studies done, steps followed in installation and measurements done in the pre and post installation of the permanent cadmium-line in outer channel of the reactor. We measured the reactivity worth of the old and new channel and readjusted the reactor's core excess reactivity after the installation. Results obtained are: reactivity worth of the old channel (0.12 mk), reactivity worth of the new channel = 0.336 mk, temperature coefficient = −0.1 mk/°C, control rod worth coefficient = 0.023 mk/mm and the core excess reactivity = 3.85 mk. We have also measured the radial and axial flux distribution in the channels of the reactor after the installation. The installation of the permanent cadmium-lined channel reported here will not only boost the sample handling capabilities of NIRR-1 but will also

  15. Combination of preferential CO oxidation and methanation in hybrid MCR (micro-channel reactor) for CO clean-up

    International Nuclear Information System (INIS)

    CO in the hydrogen stream must be reduced to extremely low levels, under 10 ppm, because the Pt electrode is detrimentally affected by residual CO in the H2 stream. Therefore removal of carbon monoxide from the H2-rich stream during fuel generation from hydrocarbons is a critical challenge, especially for PEMFC (proton exchange membrane fuel cell) applications. Herein, CO an initial concentration of 1.0 vol.% was successfully removed from a H2-rich stream to a residual level below 10 ppm, within the wide operating temperature range from 92 to 235 °C by utilizing a hybrid channel reactor comprising a micro-channel heat exchanger and mini-packed bed reactor. The mini-packed bed reactor contained two kinds of catalysts that promote preferential oxidation and methanation of CO in series. The HMCR (hybrid micro- and mini- channel reactor) offers not only ultimately safe operation but also easy scale-up and is adaptable to mass production of CO clean-up units. - Highlights: • Hybrid reactor (micro-channel heat exchanger/mini-packed bed reactor) developed. • Serial bi-catalysis allowed preferential oxidation and methanation of CO. • PrOx-methanation catalysts had higher activity in HMCR, with 143 °C HMCR window. • CO below 10 ppm achieved by combined PrOx and methanation in HMCR system. • Lower operation temperature and wider temperature range for HMCR achieved

  16. Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

    Science.gov (United States)

    Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

    2011-01-01

    The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1.

  17. CFD analysis of the 37-element fuel channel for CANDU6 reactor

    International Nuclear Information System (INIS)

    We analyzed the thermal-hydraulic behavior of coolant flow along fuel bundles with appendages of end support plate, spacer pad, and bearing pad, which are the CANDU6 characteristic design. The computer code used is a commercial CFD code, CFX-12. The present CFD analysis model calculates the conjugate heat transfer between the fuel and coolant. Using the same volumetric heat source as the O6 channel, the CFD predictions of the axial temperature distributions of the fuel element are compared with those by the CATHENA (one-dimensional safety analysis code for CANDU6 reactor). It is shown that CFX-12 predictions are in good agreement with those by the CATHENA code for the single liquid convection region (especially before the axial position of the first half of the channel length). However, the CFD analysis at the second half of the fuel channel, where the two-phase flow is expected to occur, over-predicts the fuel temperature, since the wall boiling model is not considered in the present CFD model. (author)

  18. Design and production process of bushing-type fuel elements for channel research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Afanasiev, V.L.; Aleksandrov, A.B.; Enin, A.A. [NZHK, Novosibirsk (Russian Federation)

    1998-07-01

    The design of bushing-type fuel elements (FEs) based on the dioxide fuel composition UO{sub 2}+Al for channel research reactors is described. Commercial technological process for bushing-type FEs with up to 0.8 g/cm{sup 3} uranium concentration in the fuel core is presented. This technology is based on fuel core production using powder metallurgy with subsequent chemical treatment of its surface and enclosing into the finished cladding. Commercial technological process for bushing-type FEs with 0.8-3.8 g/cm{sup 3} uranium concentration in the fuel composition is considered. This process is based on fuel core production by means of extrusion technology followed by fuel core enclosing into the cladding. (author)

  19. Thermal-hydraulics numerical analyses of Pebble Bed Advanced High Temperature Reactor hot channel

    International Nuclear Information System (INIS)

    Background: The thermal hydraulics behavior of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) hot channel was studied. Purpose: We aim to analyze the thermal-hydraulics behavior of the PB-AHTR, such as pressure drop, temperature distribution of coolant and pebble bed as well as thermal removal capacity in the condition of loss of partial coolant. Methods: We used a modified FLUENT code which was coupled with a local non-equilibrium porous media model by introducing a User Defined Scalar (UDS) in the calculation domain of the reactor core and subjoining different resistance terms (Ergun and KTA) to calculate the temperature of coolant, solid phase of pebble bed and pebble center in the core. Results: Computational results showed that the resistance factor has great influence on pressure drop and velocity distribution, but less impact on the temperature of coolant, solid phase of pebble bed and pebble center. We also confirmed the heat removal capacity of the PB-AHTR in the condition of nominal and loss of partial coolant conditions. Conclusion: The numerical analyses results can provide a useful proposal to optimize the design of PB-AHTR. (authors)

  20. Verification of Monte Carlo calculations of the neutron flux in typical irradiation channels of the TRIGA reactor, Ljubljana

    NARCIS (Netherlands)

    Jacimovic, R; Maucec, M; Trkov, A

    2003-01-01

    An experimental verification of Monte Carlo neutron flux calculations in typical irradiation channels in the TRIGA Mark II reactor at the Jozef Stefan Institute is presented. It was found that the flux, as well as its spectral characteristics, depends rather strongly on the position of the irradiati

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    N. Dharmaraju; Rama Rao, A.

    2008-01-01

    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 ...

  3. The ability to create NTD silicon technology in the IRT-T reactor in a horizontal experimental channel with one-side access

    Science.gov (United States)

    Varlachev, V. A.; Golovatsky, A. V.; Emets, E. G.; Butko, Ya A.

    2016-06-01

    The article shows the ability of creation of neutron transmutation doping (NTD) of monocrystalline silicon technology in the reactor's channel, which has a one-side access. In the article a distribution of thermal neutron flux through the length of channel and it's radius, neutron spectrum were obtained which confirmed that horizontal experimental channel HEC-1 is suitable for NTD.

  4. Conceptual design of a self-sustainable pressurized water reactor with boiling channels

    International Nuclear Information System (INIS)

    Parametric studies have been performed on a seed-blanket Th-U233 fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts required substantial reduction of the core power density in order to operate under nominal PWR system conditions. Boiling flow regime in the seed area allows better heat removal, which in turn, may potentially allow increasing the power density of the core. In addition, the reduced moderation improves the breeding performance. A 2-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to 104 W/cc, created a map of designs with their corresponding fissile inventory ratio (FIR) values. It was found that several options have the potential to achieve the main objective - a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. (author)

  5. Characterization of typical irradiation channels of CNESTEN's TRIGA MARK II reactor (Rabat, Morocco) using NAA k0-Method

    International Nuclear Information System (INIS)

    The aim of this work is the use of neutron activation analysis using k0-standardization method to characterize some typical irradiation channels of the Moroccan TRIGA Mark II research reactor. The two parameters of neutron flux in the selected irradiation channels used for elemental concentration calculation, f (thermal-to-epithermal ratio) and α (deviation from the 1/E distribution), have been determined as well in the pneumatic tube (PT) as in the carousel facility (CR1) using a set of Al (99.9%), Au (0.l %), Zn (99.99%) and Zr (99.8%) monitors. Results obtained for f and α in two irradiation channels show that f parameter determined in this way is different in the carousel facility (CRl) and the PT channel. This can be explained by the fact that the CR1 channel is situated in a graphite reflector and is relatively far from the reactor core, while the PT is in the core. Parameter α in the CR1 has a positive value, as expected, indicating that the neut ron spectrum is relatively well thermalized. Parameter α in the PT has a negative value, which is very small and can not significantly influence the final result obtained by k0-method. The method in our laboratory is validated by analyzing the elemental concentrations of the IAEA Standard Reference Material (Soil-7). All calculations were performed using Kay Win Software.

  6. Simulation of a channel blockage transient in the Angra 2 Nuclear Reactor using a RELAP5-3D model

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez-Mantecon, Javier; Costa, Antonella L.; Veloso, Maria Auxiliadora F.; Pereira, Claubia; Reis, Patricia A.L.; Scari, Maria E., E-mail: mantecon1987@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: patricialire@yahoo.com.br, E-mail: melizabethscari@yahoo.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear

    2015-07-01

    The Angra 2 Nuclear Power Plant (NPP) is a Pressurized Water Reactor (PWR) type with electrical output of about 1350 MW. The RELAP5-3D code was used to develop a detailed thermal hydraulic model of such reactor using reference data from the Angra 2 Final Safety Analysis Report (FSAR). In this work, a blockage transient has been investigated at full power operation. The transient herein considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

  7. Effect of channel wall conductance on the performance characteristics of self-cooled liquid metal fusion reactor blankets

    International Nuclear Information System (INIS)

    One of the critical issues in self-cooled liquid metal tritium breeding blankets in magnetically confined fusion reactors is strong MHD effects particularly when the channel walls are not electrically insulated from the flowing liquid metals. Another critical issue is the cooling of the first wall which is subjected to intense heat load from the fusion plasma. In this work we investigate the effect of channel wall conductance on the friction factor and Nusselt number. It is shown by solving the indication and linear momentum equations that even for relatively small channel wall conductance ratios, the friction factor increases by an order of magnitude for the typical Hartmann numbers encountered in fusion reactor blankets. Furthermore, by solving the temperature equation, it is shown that channel wall conductance has negligible effect on Nusselt number in spite of high velocity jets developing near the side walls. Taking into account these limitations, it is shown however, that the self-cooled liquid metal blankets remain a feasible proposition for both first wall heat extraction and bulk heat removal from the blanket. The most important thermal-hydraulic performance parameter -the heat removal rate to pumping power ratio- can still be kept quite high by suitably choosing the design variables of the liquid metal cooling system. The results are presented and compared for the three prime candidates for self-cooled liquid metal breeding blankets, i.e., lithium, lead-lithium, and tin-lithium alloys. (author)

  8. 3D thermal hydraulic simulation of the hot channel of a typical material testing reactor under normal operation conditions

    Energy Technology Data Exchange (ETDEWEB)

    El-Morshedy, Salah El-Din; Salama, Amgad [Atomic Energy Authority, Cairo (Egypt). Reactors Dept.

    2010-09-15

    The hot channel in a typical Material Testing Reactor (MTR) is subjected to 3D simulation. Because of the existence of similarity planes, only a quarter of the hot channel including meat thickness, clad, and coolant channel is considered for CFD analysis using the FLUENT code. For the simulation, steady state normal operation regime at the reactor nominal power is assumed. In order to build confidence in our modeling approach, the results obtained in this work are compared with those obtained from the one-dimensional simulation code, MTRTHA. That is, modified variables were generated in order to match those obtained by MTRTHA and to allow comparisons. Quite good agreement is generally observed, however, the maximum clad surface temperature predicted by the 3D calculations, located at the clad mid-width, is higher than the 1D prediction by about 8 C but still below the onset of subcooled boiling by adequate safety margin. The results show quite interesting 3D patterns in both the flow field and the heat transfer. Temperature profiles, velocity profiles and contours are all presented to highlight the essential 3D features of this system. (orig.)

  9. Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

    Science.gov (United States)

    Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor.

  10. Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

    Science.gov (United States)

    Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor. PMID:27552124

  11. Erythorbic acid promoted formation of CdS QDs in a tube-in-tube micro-channel reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liang, Yan; Tan, Jiawei; Wang, Jiexin; Chen, Jianfeng [State Key Laboratory of Organic–Inorganic Composites, Beijing University of Chemical Technology, Beijing 100029 (China); Research Center of the Ministry of Education for High Gravity Engineering and Technology, Beijing University of Chemical Technology, Beijing 100029 (China); Sun, Baochang, E-mail: sunbc@mail.buct.edu.cn [State Key Laboratory of Organic–Inorganic Composites, Beijing University of Chemical Technology, Beijing 100029 (China); Research Center of the Ministry of Education for High Gravity Engineering and Technology, Beijing University of Chemical Technology, Beijing 100029 (China); Shao, Lei, E-mail: shaol@mail.buct.edu.cn [State Key Laboratory of Organic–Inorganic Composites, Beijing University of Chemical Technology, Beijing 100029 (China); Research Center of the Ministry of Education for High Gravity Engineering and Technology, Beijing University of Chemical Technology, Beijing 100029 (China)

    2014-12-15

    Erythorbic acid assistant synthesis of CdS quantum dots (QDs) was conducted by homogeneous mixing of two continuous liquids in a high-throughput microporous tube-in-tube micro-channel reactor (MTMCR) at room temperature. The effects of the micropore size of the MTMCR, liquid flow rate, mixing time and reactant concentration on the size and size distribution of CdS QDs were investigated. It was found that the size and size distribution of CdS QDs could be tuned in the MTMCR. A combination of erythorbic acid promoted formation technique with the MTMCR may be a promising pathway for controllable mass production of QDs.

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

    Science.gov (United States)

    Alman, David E.; Wilson, Rick D.; Davis, Daniel L.

    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.

  13. Fuel Temperature Characteristics for Fuel Channels using Burnable Poison in the CANDU reactor

    International Nuclear Information System (INIS)

    Although the CANFLEX RU fuel bundle loaded 11.0 wt% Er2O3 are originally designed focused on the safety characteristics, the fuel temperature characteristics is revealed to be not deteriorated but rather is slightly enhanced by the decreased fuel temperature in the outer ring compared with that of standard 37 fuel bundle. Recently, for an equilibrium CANDU core, the power coefficient was reported to be slightly positive when newly developed Industry Standard Tool set reactor physics codes were used. Therefore, it is required to find a new way to effectively decrease the positive power coefficient of CANDU reactor without seriously compromising the economy. In order to make the power coefficient of the CANDU reactor negative at the operating power, Roh et al. have evaluated the various burnable poison (BP) materials and its loading scheme in terms of the fuel performance and reactor safety characteristics. It was shown that reactor safety characteristics can be greatly improved by the use of the BP in the CANDU reactor. In a view of safety, the fuel temperature coefficient (FTC) is an important safety parameter and it is dependent on the fuel temperature. For an accurate evaluation of the safety-related physics parameters including FTC, the fuel temperature distribution and its correlation with the coolant temperature should be accurately identified. Therefore, we have evaluated the fuel temperature distribution of a CANFLEX fuel bundle loaded with a burnable poison and compared the standard 37 element fuel bundle and CANFELX-NU fuel bundle

  14. Reactor building indoor wireless network channel quality estimation using RSSI measurement of wireless sensor network

    International Nuclear Information System (INIS)

    Expanding wireless communication network reception inside reactor buildings (RB) and service wings (SW) has always been a technical challenge for operations service team. This is driven by the volume of metal equipment inside the Reactor Buildings (RB) that blocks and somehow shields the signal throughout the link. In this study, to improve wireless reception inside the Reactor Building (RB), an experimental model using indoor localization mesh based on IEEE 802.15 is developed to implement a wireless sensor network. This experimental model estimates the distance between different nodes by measuring the RSSI (Received Signal Strength Indicator). Then by using triangulation and RSSI measurement, the validity of the estimation techniques is verified to simulate the physical environmental obstacles, which block the signal transmission. (author)

  15. Header feedwater supply and power distribution stability in channel boiling water cooled reactors

    International Nuclear Information System (INIS)

    Boundaries of radial-azimuthal instability of the reactor neutron field during the supply of all feedwater and a part of it (25%) to downtake pipes of the separating drum (75% of feedwater come to distributive group headers) are found out for NPP with a RBMK type reactor. Results of computer calculation of the transient process at NPP caused by 2% step increase of nominal pressure in a head collector of a feedwater electric pump are also presented for comparison of the above methods of feed-water supply. Calculation is carried out according to the OKA program with provision for the control system of the reactor total power. It is shown that the boundary of ''mean period'' instability does not change but the reserve in respect to the ''fast'' space instability slightly increases when header feedwater supply at NPP from RBMK is used. It is noted that requirements to the pressure regulator system quick action in a separating drum are increased when the header feedwater supply is used. This fact is explained by the fact that considerable pressure drop in a separating drum occurs during some accidents (for example, at false operation of the emergensy protective system)

  16. Computer aided design (CAD) for electronics improvement of the nuclear channels of TRIGA Mark III reactor of the ININ

    International Nuclear Information System (INIS)

    The 4 neutron measurement channels of the digital control console (CCD) of the TRIGA Mark III reactor (RTMIII) of the ININ, its were designed and built with the corresponding Quality Guarantee program, being achieved the one licensing to replace the old console. With the time they were carried out some changes to improve and to not solve some problems detected in the tests, verification and validation, requiring to modify the circuits originally designed. In this work the corrective actions carried out to eliminate the Non Conformity generated by these problems, being mentioned the advantages of using modern tools, as the software applied to the Attended Engineering by Computer, and those obtained results are presented. (Author)

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

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2015-05-01

    Full Text Available The objective of this paper is to present a possible solution for the designing of a device for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The decommissioning activities are dismantling, demolition, controlled removal of equipment, components, conventional or hazardous waste (radioactive, toxic in compliance with the international basic safety standards on radiation protection. One as the most important operation in the final phase of the nuclear reactor dismantling is the decommissioning of fuel channels. For the fuel channels decommissioning should be taken into account the detailed description of the fuel channel and its components, the installation documents history, adequate radiological criteria for decommissioning guidance, safety and environmental impact assessment, including radiological and non-radiological analysis of the risks that can occur for workers, public and environment, the description of the proposed program for decommissioning the fuel channel and its components, the description of the quality assurance program and of the monitoring program, the equipments and methods used to verify the compliance with the decommissioning criteria, the planning of performing the final radiological assessment at the end of the fuel channel decommissioning. These will include also, a description of the proposed radiation protection procedures to be used during decommissioning. The dismantling of the fuel channel is performed by one device which shall provide radiation protection during the stages of decommissioning, ensuring radiation protection of the workers. The device shall be designed according to the radiation protection procedures. The decommissioning device assembly of the fuel channel components is composed of the device itself and moving platform support for coupling of the selected channel to be dismantled. The fuel channel decommissioning device is an autonomous device designed for

  18. 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.

  19. Application of perturbation theory to sensitivity calculations of PWR type reactor cores using the two-channel model

    International Nuclear Information System (INIS)

    Sensitivity calculations are very important in design and safety of nuclear reactor cores. Large codes with a great number of physical considerations have been used to perform sensitivity studies. However, these codes need long computation time involving high costs. The perturbation theory has constituted an efficient and economical method to perform sensitivity analysis. The present work is an application of the perturbation theory (matricial formalism) to a simplified model of DNB (Departure from Nucleate Boiling) analysis to perform sensitivity calculations in PWR cores. Expressions to calculate the sensitivity coefficients of enthalpy and coolant velocity with respect to coolant density and hot channel area were developed from the proposed model. The CASNUR.FOR code to evaluate these sensitivity coefficients was written in Fortran. The comparison between results obtained from the matricial formalism of perturbation theory with those obtained directly from the proposed model makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations (author). 23 refs, 4 figs, 7 tabs

  20. Micro-channel catalytic reactor integration in CAPER and research/development on highly tritiated water handling and processing

    International Nuclear Information System (INIS)

    The CAPER facility of the Tritium Laboratory Karlsruhe has demonstrated the technology for the tokamak exhaust processing. CAPER has been significantly upgraded to pursue research/development programs towards highly tritiated water (HTW) handling and processing. The preliminary tests using a metal oxide reactor producing HTW afterward de-tritiated with PERMCAT were successful. In a later stage, a micro-channel catalytic reactor was installed in view of long term research program on HTW. The integration of this new system in CAPER was carried out along with a careful safety analysis due to high risk associated with such experiments. First experiments using the μ-CCR were performed trouble free, and HTW up to 360 kCi/kg was produced at a rate of 0.5 g/h. Such HTW was collected into a platinum zeolite bed (2 g of HTW for 20 g of Pt-zeolite), and in-situ detritiation was performed via isotopic exchange with deuterium. These first experimental results with tritium confirmed the potential for the capture and exchange method to be used for HTW in ITER. (authors)

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

    International Nuclear Information System (INIS)

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

  2. On The Deign And Construction Of A Radiation Shielding System For Development Of Neutron Beams Based On The Horizontal Channel No.2 Of Dalat Reactor

    International Nuclear Information System (INIS)

    An optimal structural system of filtered neutron beam and radiation shielding has been designed and calculated using the Monte-Carlo code MCNP5. The system was constructed and installed into the horizontal channel No. 2 of the Dalat reactor. The neutron beam is applied for experimental studies on nuclear physics, nuclear data measurements, and personal training. (author)

  3. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 7 - FUNCTIONING OF THE DECOMMISSIONING DEVICE

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2015-05-01

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

  4. Comparison of the PLTEMP code flow instability predictions with measurements made with electrically heated channels for the advanced test reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Feldman, E. (Nuclear Engineering Division)

    2011-06-09

    When the University of Missouri Research Reactor (MURR) was designed in the 1960s the potential for fuel element burnout by a phenomenon referred to at that time as 'autocatalytic vapor binding' was of serious concern. This type of burnout was observed to occur at power levels considerably lower than those that were known to cause critical heat flux. The conversion of the MURR from HEU fuel to LEU fuel will probably require significant design changes, such as changes in coolant channel thicknesses, that could affect the thermal-hydraulic behavior of the reactor core. Therefore, the redesign of the MURR to accommodate an LEU core must address the same issues of fuel element burnout that were of concern in the 1960s. The Advanced Test Reactor (ATR) was designed at about the same time as the MURR and had similar concerns with regard to fuel element burnout. These concerns were addressed in the ATR by two groups of thermal-hydraulic tests that employed electrically heated simulated fuel channels. The Croft (1964), Reference 1, tests were performed at ANL. The Waters (1966), Reference 2, tests were performed at Hanford Laboratories in Richland Washington. Since fuel element surface temperatures rise rapidly as burnout conditions are approached, channel surface temperatures were carefully monitored in these experiments. For self-protection, the experimental facilities were designed to cut off the electric power when rapidly increasing surface temperatures were detected. In both the ATR reactor and in the tests with electrically heated channels, the heated length of the fuel plate was 48 inches, which is about twice that of the MURR. Whittle and Forgan (1967) independently conducted tests with electrically heated rectangular channels that were similar to the tests by Croft and by Walters. In the Whittle and Forgan tests the heated length of the channel varied among the tests and was between 16 and 24 inches. Both Waters and Whittle and Forgan show that the cause

  5. Calibration of the nuclear power channels of the IPEN/MB-01 reactor obtained from the measurements of the spatial thermal neutron flux distribution in the reactor core through the irradiation of infinitely diluted gold foils

    International Nuclear Information System (INIS)

    Several nuclear parameters are obtained through the gamma spectrometry of targets irradiated in a research reactor core and this is the case of the activation foils which make possible, through the measurements of the activity induced, to determine the neutron flux in the place where they had been irradiated. The power level operation of the reactor is a parameter directly proportional to the average neutron flux in the core. This work aims to get the power operation of the reactor through of spatial neutron flux distribution in the core of IPEN/MB-01 reactor by the irradiation of infinitely diluted gold foils and prudently located in its interior. These foils were made in the form of metallic alloy in concentration levels such that the phenomena of flux disturbance, as the self-shielding factors to neutrons become worthless. These activation foils has only 1% of dispersed gold atoms in an aluminium matrix content of 99% of this element. The irradiations of foils have been carried through with and without cadmium plate. The total correlation between the average thermal neutron flux obtained by irradiation of infinitely diluted activation foils and the average digital value of current of the nuclear power channels 5 and 6 (non-compensated ionization chambers - CINC), allow the calibration of the nuclear channels of the IPEN/MB-01 reactor. (author)

  6. Application of gamma scanning and neutron radiography methods to control fuel element state as tested in MR reactor loop channels

    International Nuclear Information System (INIS)

    The gamma-scanning and neutron radiography methods are described used for non-destructive control of fuel elements after their test in the loop channels of MR reactor, and also the equipment utilized. The techniques and the results obtained while studying fuel elements using the above mentioned methods are provided. It is established that gamma-scanning method can only indicate the presence of defect in the continuity of the fuel element core without identifying its type whereas the advantage of neutron radiography method is in obtaining visual results. At the same time gamma-scanning method makes it possible to determine energy release on the length of fuel elements, to find the burn up fraction, to study the phenomenon of fusion products migration which is difficult or impossible with neutron radiography method. A conclusion is drawn that gamma-scanning and neutron radiography methods successfully supplement each other and make it possible to obtain important information on fuel state in the irradiated fuel elements

  7. Development of a steady-state sub-channel code for small reactor on the basis of combined cross momentum and non-linear conduction

    Energy Technology Data Exchange (ETDEWEB)

    Huandong, Chen; Xiaoying, Zhang, E-mail: zxy1119@scut.edu.cn

    2015-08-15

    Highlights: • Combining equations to have a more stable and faster convergence solution. • Taking account of non-linear conduction of fuel rods. • Validating code with COBRA code. • Applying code to the small reactor “MUTSU” and comparing result with its design conditions. - Abstract: For purpose of thermal hydraulic analysis in small nuclear reactors, a sub-channel code with an improved convergence has been developed based on the homogenous flowing model. A combined lateral momentum equation coupling with continuity and axial momentum equation has been used to substitute the original lateral momentum equation. The Gauss iteration method has been adopted to solve the Kirchhoff's transformation equation of nonlinear heat conduction of fuel rod, a temperature dependent conducting has been considered. The code has been validated by using experimental data from the NUPEC PWR Sub-channel and Bundle Tests (PSBT) and then applied to the “MUTSU” reactor. Results show that the code can predict the experimental data with acceptable accuracy and has ability to analyze the small PWR reactor.

  8. Development of a steady-state sub-channel code for small reactor on the basis of combined cross momentum and non-linear conduction

    International Nuclear Information System (INIS)

    Highlights: • Combining equations to have a more stable and faster convergence solution. • Taking account of non-linear conduction of fuel rods. • Validating code with COBRA code. • Applying code to the small reactor “MUTSU” and comparing result with its design conditions. - Abstract: For purpose of thermal hydraulic analysis in small nuclear reactors, a sub-channel code with an improved convergence has been developed based on the homogenous flowing model. A combined lateral momentum equation coupling with continuity and axial momentum equation has been used to substitute the original lateral momentum equation. The Gauss iteration method has been adopted to solve the Kirchhoff's transformation equation of nonlinear heat conduction of fuel rod, a temperature dependent conducting has been considered. The code has been validated by using experimental data from the NUPEC PWR Sub-channel and Bundle Tests (PSBT) and then applied to the “MUTSU” reactor. Results show that the code can predict the experimental data with acceptable accuracy and has ability to analyze the small PWR reactor

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

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2015-05-01

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

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

    International Nuclear Information System (INIS)

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

  11. Dispositivo de posicionamiento de muestras biológicas para su irradiación en un canal radial de un reactor nuclear // Biological samples positioning device for irradiations on a radial channel at the nuclear research reactor

    Directory of Open Access Journals (Sweden)

    Maritza Rodríguez - Gual

    2010-05-01

    Full Text Available ResumenPor la demanda de un dispositivo experimental para el posicionamiento de las muestras biológicaspara su irradiación en un canal radial de un reactor nuclear de investigaciones en funcionamiento, seconstruyó y se puso en marcha un dispositivo para la colocación y retirada de las muestras en laposición de irradiación de dicho canal. Se efectuaron las valoraciones económicas comparando conotro tipo de dispositivo con las mismas funciones. Este trabajo formó parte de un proyectointernacional entre Cuba y Brasil que abarcó el estudio de los daños inducidos por diferentes tipos deradiación ionizante en moléculas de ADN. La solución propuesta es comprobada experimentalmente,lo que demuestra la validez práctica del dispositivo. Como resultado del trabajo, el dispositivoexperimental para la irradiación de las muestras biológicas se encuentra instalado y funcionando yapor 5 años en el canal radial # 3(BH#3 Palabras claves: reactor nuclear de investigaciones, dispositivo para posicionamiento de muestras,___________________________________________________________________________AbstractFor the demand of an experimental device for biological samples positioning system for irradiationson a radial channel at the nuclear research reactor in operation was constructed and started up adevice for the place and remove of the biological samples from the irradiation channels withoutinterrupting the operation of the reactor. The economical valuations are effected comparing withanother type of device with the same functions. This work formed part of an international projectbetween Cuba and Brazil that undertook the study of the induced damages by various types ofionizing radiation in DNA molecules. Was experimentally tested the proposed solution, whichdemonstrates the practical validity of the device. As a result of the work, the experimental device forbiological samples irradiations are installed and operating in the radial beam hole #3(BH#3

  12. Thermal-hydraulic behavior of physical quantities at critical velocities in a nuclear research reactor core channel using plate type fuel

    Directory of Open Access Journals (Sweden)

    Sidi Ali Kamel

    2012-01-01

    Full Text Available The thermal-hydraulic study presented here relates to a channel of a nuclear reactor core. This channel is defined as being the space between two fuel plates where a coolant fluid flows. The flow velocity of this coolant should not generate vibrations in fuel plates. The aim of this study is to know the distribution of the temperature in the fuel plates, in the cladding and in the coolant fluid at the critical velocities of Miller, of Wambsganss, and of Cekirge and Ural. The velocity expressions given by these authors are function of the geometry of the fuel plate, the mechanical characteristics of the fuel plate’s material and the thermal characteristics of the coolant fluid. The thermal-hydraulic study is made under steady-state; the equation set-up of the thermal problem is made according to El Wakil and to Delhaye. Once the equation set-up is validated, the three critical velocities are calculated and then used in the calculations of the different temperature profiles. The average heat flux and the critical heat flux are evaluated for each critical velocity and their ratio reported. The recommended critical velocity to be used in nuclear channel calculations is that of Wambsganss. The mathematical model used is more precise and all the physical quantities, when using this critical velocity, stay in safe margins.

  13. Effects of transient and non-uniform distribution of heat flux on intensity of heat transfer and burnout conditions in the channels of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vitaly Osmachkin [Russian Research Center ' Kurchatov Institute' 1, Kurchatov sq, Moscow 123182 (Russian Federation)

    2005-07-01

    Full text of publication follows: The influence of power transient, changes of flow rate, inlet temperatures or pressure in cores of nuclear reactors on heat transfer and burnout conditions in channels depend on rate of such violations. Non-uniform distribution of the heat flux is also important factor for heat transfer and development of crisis phenomenon. Such effects may be significant for NPPs safety. But they have not yet generally accepted interpretation. Steady state approach is often recommended for use in calculations. In the paper a review of experimental observed so-called non-equilibrium effects is presented. The effects of space and time factors are displaying due delay in reformation turbulence intensity, velocity, temperatures or void fraction profiles, water film flow on the surface of heated channels. For estimation of such effect different methods are used. Modern computer codes based on two or three fluids approaches are considered as most effective. But simple and clear correlations may light up the mechanics of effects on heat transfer and improve general understanding of scale and significance of the transient events. In the paper the simplified methods for assessment the influence of lags in the development of distributions of parameters of flow, the relaxation of temporal or space violations are considered. They are compared with more sophisticated approaches. Velocities of disturbance fronts moving along the channels are discussed also. (author)

  14. High Temperature Stress Analysis on 61-pin Test Assembly for Reactor Core Sub-channel Flow Test

    International Nuclear Information System (INIS)

    In this study, a high temperature heat transfer and stress analysis of a 61-pin test fuel assembly scaled down from the full scale 217-pin sub-assembly was conducted. The reactor core subchannel flow characteristic test will be conducted to evaluate uncertainties in computer codes used for reactor core thermal hydraulic design. Stress analysis for a 61-pin fuel assembly scaled down from Prototype Generation IV Sodium-cooled Fast Reactor was conducted and structural integrity in terms of load controlled stress limits was conducted. In this study, The evaluations on load-controlled stress limits for a 61-pin test fuel assembly to be used for reactor core subchannel flow distribution tests were conducted assuming that the test assembly is installed in a Prototype Generation IV Sodium-cooled fast reactor core. The 61-pin test assembly has the geometric similarity on P/D and H/D with PGSFR and material of fuel assembly is austenitic stainless steel 316L. The stress analysis results showed that 4.05MPa under primary load occurred at mid part of the test assembly and it was shown that the value of 4.05Mpa was far smaller than the code allowable of 127MPa. , it was shown that the stress intensity due to due to primary load is very small. The stress analysis results under primary and secondary loads showed that maximum stress intensity of 84.08MPa occurred at upper flange tangent to outer casing and the value was well within the code allowable of 268.8MPa. Integrity evaluations based on strain limits and creep-fatigue damage are underway according to the elevated design codes

  15. High Temperature Stress Analysis on 61-pin Test Assembly for Reactor Core Sub-channel Flow Test

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dongwon; Kim, Hyungmo; Lee, Hyeongyeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this study, a high temperature heat transfer and stress analysis of a 61-pin test fuel assembly scaled down from the full scale 217-pin sub-assembly was conducted. The reactor core subchannel flow characteristic test will be conducted to evaluate uncertainties in computer codes used for reactor core thermal hydraulic design. Stress analysis for a 61-pin fuel assembly scaled down from Prototype Generation IV Sodium-cooled Fast Reactor was conducted and structural integrity in terms of load controlled stress limits was conducted. In this study, The evaluations on load-controlled stress limits for a 61-pin test fuel assembly to be used for reactor core subchannel flow distribution tests were conducted assuming that the test assembly is installed in a Prototype Generation IV Sodium-cooled fast reactor core. The 61-pin test assembly has the geometric similarity on P/D and H/D with PGSFR and material of fuel assembly is austenitic stainless steel 316L. The stress analysis results showed that 4.05MPa under primary load occurred at mid part of the test assembly and it was shown that the value of 4.05Mpa was far smaller than the code allowable of 127MPa. , it was shown that the stress intensity due to due to primary load is very small. The stress analysis results under primary and secondary loads showed that maximum stress intensity of 84.08MPa occurred at upper flange tangent to outer casing and the value was well within the code allowable of 268.8MPa. Integrity evaluations based on strain limits and creep-fatigue damage are underway according to the elevated design codes.

  16. Computer aided design (CAD) for electronics improvement of the nuclear channels of TRIGA Mark III reactor of the ININ; Diseno asistido por computadora (DAC) para mejorar la electronica de los canales nucleares del reactor TRIGA Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J.L.; Rivero G, T.; Aguilar H, F. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: jlgm@nuclear.inin.mx

    2007-07-01

    The 4 neutron measurement channels of the digital control console (CCD) of the TRIGA Mark III reactor (RTMIII) of the ININ, its were designed and built with the corresponding Quality Guarantee program, being achieved the one licensing to replace the old console. With the time they were carried out some changes to improve and to not solve some problems detected in the tests, verification and validation, requiring to modify the circuits originally designed. In this work the corrective actions carried out to eliminate the Non Conformity generated by these problems, being mentioned the advantages of using modern tools, as the software applied to the Attended Engineering by Computer, and those obtained results are presented. (Author)

  17. Thermal study of a non adiabatic differential calorimeter used for nuclear heating measurements inside an experimental channel of the Jules Horowitz Reactor

    Science.gov (United States)

    Reynard-Carette, C.; Lyoussi, A.; Brun, J.; Muraglia, M.; Carette, M.; Janulyte, A.; Zerega, Y.; André, J.; Bignan, G.; Chauvin, J.-P.; Fourmentel, D.; Gonnier, C.; Guimbal, P.; Malo, J.-Y.; Villard, J.-F.

    2012-11-01

    New online in-pile measurement methods are crucial during irradiations in Material Testing Reactors (MTR) for a better understanding of accelerated material ageing and nuclear fuel behaviour. In particular, instrumentation for measurements of one relevant parameter: nuclear heat deposition rate, called nuclear heating, has to be improved. The knowledge of this quantity is a great interest for various safety, scientist and end-user requirements (design of specific irradiation devices and associated cooling systems with imposed conditions). This paper focuses on thermal experimental and numerical studies carried out under non irradiation conditions on an in-pile calorimeter dedicated to nuclear heating quantification inside a new experimental device which will be dedicated to the experimental condition mapping (neutron and photon fluxes and nuclear heating) inside the JHR experimental channels. Experimental results concerning the calorimeter response during its electrical calibration (calorimeter design is compared to a single calorimeter which gives promising results (miniaturization, higher sensitivity).

  18. Testing measurements at horizontal channels of the MARIA reactor performed using neutron spectrometers; Pomiary testowe przy kanalach poziomych reaktora MARIA wykonane przy uzyciu spektrometrow neutronow

    Energy Technology Data Exchange (ETDEWEB)

    Murasik, A. [Institute of Atomic Energy, Otwock-Swierk (Poland)

    1997-12-31

    By means of neutron diffraction, using the standard polycrystalline sample of Al{sub 2}O{sub 3}, measurements on three (of four spectrometers) already installed in the front of horizontal channels of MARIA reactor have been performed. Basing on these experiments as well as on activation measurements carried out earlier, the fluxes of monoenergetic neutrons have been estimated. These experiments allowed to determine (for a given geometry and kind of monochromators chosen) the resolution efficiency of instruments and high order contamination in the reflected beam. With the help of polycrystalline vanadium and TbBr{sub 3} sample, the possibility of studies using the inelastic scattering process have been tested. (author) 7 refs, 15 figs, 7 tabs

  19. Connected analysis nuclear-thermo-hydraulic of parallel channels of a BWR reactor using distributed computation; Analisis acoplado nuclear-termohidraulico de canales paralelos de un reactor BWR empleando computacion distribuida

    Energy Technology Data Exchange (ETDEWEB)

    Campos Gonzalez, Rina Margarita

    2007-07-15

    This work consists of the integration of three models previously developed which are described widely in Literature: model of the thermo-hydraulic channel, model of the modal neutronic and the model of the recirculation bows. The tool used for this connection of models is the PVM system, Parallel Virtual Machine that allowed paralleling the model by means of the concept of distributed computation. The purpose of making this connection of models is the one of obtaining a more complete tool than better represents the real configuration and the phenomenology of the nucleus of a BWR reactor, thus obtaining better results. In addition to maintaining the flexibility to improve the resulting model at any time, since the very complex or sophisticated models are difficult to improve being impossible to modify the equations they use and can include variables that are not of primary importance in the tackled problem or that mask relations among variables due to the excess of results. Also maintaining the flexibility for adding component of models or systems of the BWR reactor, all of this following the modeling needs. The Swedish Ringhals power plant was chosen to characterize the resulting connected model for counting on a Stability Benchmark that offers the opportunity to count on real plant data. Besides that in case 9 of cycle 14 of this Benchamark oscillations outside phase appeared, which are from great interest because the detection systems that register the average of the power of the nucleus do not detect them. Additionally in this work the model of the recirculation bows as an independent module is obtained in an individual way, since this model belongs to another work and works connected to the reactor vessel. The model of the recirculation bows is able to model several transients of interest, as it is shown in the Appendix A of this work, among which are found the tripping of recirculation pumps or the transference at low or high velocity of them. The scope of the

  20. The design and construction of the input and output isolation system for power channel and control rod position in the research reactor SR4 with analog device AD210

    International Nuclear Information System (INIS)

    The design and construction of analog input-output system with device AD210, SR4 (ICS Kartini Reactor) is used as interface and output signal isolation system for power channel meter NLW-2, NP1000 and control measurement with the slave computer system so that the measuring tools and computer operate independently. Device AD210 provides a very compact insulators and economically with highly accurate system performance. It is a DIP chip with 3-port model, input, output and integrated power supply. It can be applied as a multichannel data acquisition, or a single channel and others. Analog Device AD210 is placed between the output channel power measurement and control rod position measurement system / other support tools such as Computer Slave, ADC, multiplexer, and others. The design and construction of Analog Devices AD210 insulator use the input-output mode with the gain of 1, so that pulses or output voltage are equal to the input voltage pulses. The test results in graphs of output versus input are excellent with the linearity close to 100%. The use of this Device AD210 for NLW2 and NP1000 which are a vital instrument for the continuity of the operation of reactor, is expected to increase the Main Time Between Failure (MTBF) of the reactor as a whole. (author)

  1. NEUTRONIC REACTORS

    Science.gov (United States)

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  2. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THECANDU 6 NUCLEAR REACTOR. PART 8 - PRESENTATION OF THE CUTTING AND EXTRACTING DEVICE

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2015-05-01

    Full Text Available This paper present a constructive solution proposed by the authors in order to achieve of a cutting and extracting device for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. One of the most important part of the decommissioning device is the Cutting and Extraction Device (CED which perform the dismantling, cutting and extraction of the fuel channel components. This flexible and modular device is designed to work inside the fuel channel. The main operations performed by the Cutting and Extraction Device (CED are dismantling and extraction of the channel closure plug and shield plug, cutting and extraction of the pressure tube. The Cutting and Extraction Device (CED consists of following modules: guiding-fixing module, traction modules, cutting module, guiding-extracting module and articulated elements for modules connecting. The guiding-fixing module is equipped with elastic guiding rollers and fixing claws in working position, the traction modules are provided with variable pitch rollers for allowing travel speed change through the fuel channel. The cutting module is positioned in the middle of the device and it is equipped with three roll knives for pressure tube cutting, having a system for cutting place video surveillance and pyrometers for cutting place monitoring temperature. The operations performed by the Cutting and Extraction Device (CED of fuel channel are as follows: unblock and extract the channel closure plug, unblock and extract the channel shield plug, block and cut the middle of the pressure tube, block and cut the end of the pressure tube, block and extract the half of pressure tube. The Cutting and Extraction Device (CED is fully automated, connected by wires to a Programmable Logic Controller (PLC and controlled from a Human Machine Interface (HMI. The design of the Cutting and Extraction Device (CED shall be achieved according to the particular features of the fuel channel components to be

  3. 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

  4. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

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

    Directory of Open Access Journals (Sweden)

    Constantin D. STANESCU,

    2015-05-01

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

  6. Reactor vessel support system. [LMFBR

    Science.gov (United States)

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  7. Microfluidic electrochemical reactors

    Science.gov (United States)

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  8. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THECANDU 6 NUCLEAR REACTOR. PART 11 - PRESENTATION OF THE CUTTING AND EXTRACTING DEVICE OPERATING

    Directory of Open Access Journals (Sweden)

    Constantin D. STANESCU

    2015-05-01

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

  9. Estimative of core damage frequency in IPEN IEA-R1 research reactor due to the initiating events of loss of flow caused by channel blockage and loss of coolant caused by a large rupture in the pipe of the primary circuit - PSA level 1

    Energy Technology Data Exchange (ETDEWEB)

    Hirata, Daniel Massami [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Sabundjian, Gaiane, E-mail: gdjian@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) Sao Paulo, SP (Brazil)

    2011-07-01

    This work applies the methodology of Probabilistic Safety Assessment Level 1 to the research reactor IEA-R1 IPEN-CNEN/SP. Two categories of identified initiating events of accidents in the reactor are studied: loss of flow and loss of primary coolant. Among the initiating events, blockage of flow channel and loss of cooling fluid caused by large pipe rupture in the primary circuit are chosen for a detailed analysis. The event tree technique is used to analyze the evolution of the accident, including the actuation or the fail of actuation of the safety systems and the reactor damages. Using the fault tree the reliability of the following reactor safety systems is evaluated: reactor shutdown system, isolation of the reactor pool, Emergency Core Cooling System (ECCS) and the electric system. Estimative for the frequency of damage to the reactor core and the probability of failure of the analyzed systems are calculated. The estimated values for the frequencies of core damage are within the expected margins and are of the same order of magnitude as those found for similar reactors. The reliability of the reactor shutdown system, isolation of the reactor pool and ECCS are satisfactory for the conditions in which these systems are required. However, for the electric system it is suggested an upgrade to increase its reliability. (author)

  10. Optimization of the core of a 600 MV HTGR reactor

    International Nuclear Information System (INIS)

    Through a thermal analysis, several reactor core parameters are considered, viz.: cooling channel diameter, juel channel diameter, distance between two channels power generated for lenght unit, etc. Using several criteria, the best solution or solutions are chosen

  11. Tasks related to increase of RA reactor exploitation and experimental potential, Independent CO2 loop for cooling the samples irradiated in RA reactor vertical experimental channels, (I-IV), part I

    International Nuclear Information System (INIS)

    This volume contains the description of the design project of the head of the low-temperature coolant loops needed for cooling the samples to be irradiated in the RA vertical experimental channels. The thermal and mechanical calculations are included as well as calculation of antireactivity and activation of the construction materials. Cost estimation data are included as well. The drawings included are: head of the coolant loop; diagram of CO2 coolant temperature dependence; diagrams of weight of the loop tubes in the channels; axial distribution of the thermal neutron flux. Engineering drawings of two design solutions of the low-temperature loops with details are part of this volume

  12. Chemical-vapor-deposition reactor

    Science.gov (United States)

    Chern, S.

    1979-01-01

    Reactor utilizes multiple stacked trays compactly arranged in paths of horizontally channeled reactant gas streams. Design allows faster and more efficient deposits of film on substrates, and reduces gas and energy consumption. Lack of dead spots that trap reactive gases reduces reactor purge time.

  13. Nuclear Reactor RA Safety Report, Vol. 4, Reactor

    International Nuclear Information System (INIS)

    RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices

  14. Catalytic reaction in confined flow channel

    Energy Technology Data Exchange (ETDEWEB)

    Van Hassel, Bart A.

    2016-03-29

    A chemical reactor comprises a flow channel, a source, and a destination. The flow channel is configured to house at least one catalytic reaction converting at least a portion of a first nanofluid entering the channel into a second nanofluid exiting the channel. The flow channel includes at least one turbulating flow channel element disposed axially along at least a portion of the flow channel. A plurality of catalytic nanoparticles is dispersed in the first nanofluid and configured to catalytically react the at least one first chemical reactant into the at least one second chemical reaction product in the flow channel.

  15. 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

    International Nuclear Information System (INIS)

    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

  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. N Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The last of Hanfordqaodmasdkwaspemas7ajkqlsmdqpakldnzsdflss nine plutonium production reactors to be built was the N Reactor.This reactor was called a dual purpose...

  18. Turbulent heat transfer of liquid metal inside the sub-channels of reactor core%液态金属在堆芯子通道内的湍流换热

    Institute of Scientific and Technical Information of China (English)

    葛志浩; 彭勇升; 吕逸君; 邓维平; 赵平辉

    2015-01-01

    采用 Speziale-Sarkar-Gatski (SSG)雷诺应力模型对液态金属在堆芯子通道内的流动、传热过程进行计算流体动力学(Computational Fluid Dynamics, CFD)模拟,研究雷诺数(Re)、分子普朗特数(Pr)、格拉晓夫数(Gr)、节径比(P/D)等无量纲参数对湍流换热的影响。比较无量纲对流换热系数(Nu)可以看出,CFD 预测值与实验值及经验关系式符合得较好。对各种不同无量纲参数下的计算结果进行分析发现:在 P/D 和 Re 数相同条件下,三角形子通道的壁面温度分布比方形更均匀,换热情况更好;提高 Re 数,增大 P/D,选用 Pr 数大的冷却剂,可有效改善温度和换热的周向分布不均情况;在 Re 数大于10000的条件下,浮力对液态金属换热的影响可忽略不计。%Background: Liquid metal has been proposed as the coolant of the fourth generation nuclear reactor and the accelerator driven sub-critical system. Due to its low molecular Prandtl number (Pr), liquid metal differs from other coolants like water or gas in heat transfer. Purpose: This study aims to investigate the character of heat transfer of liquid metal inside the reactor core. Methods: Speziale-Sarkar-Gatski (SSG) Reynolds stress model was applied to the Computational Fluid Dynamics (CFD) prediction of liquid metal flow and heat transfer inside the sub-channels of the reactor core. Effect of different dimensionless parameters, e.g. Reynolds number (Re), Pr, Grashof number (Gr) and pitch-to-diameter ratio (P/D) on the turbulent heat transfer calculated results was investigated. Results: The dimensionless convective heat transfer coefficient (Nu), predicted by the CFD method, agrees well with the experimental data and the empirical relations. Conclusion: Based on the analysis of various dimensionless parameters, it is found that the heat exchange performs better in triangular fuel assembly sub-channels than that in square sub-channels, under the same

  19. Current development in data acquision and processing system for reactor noise analysis in PUSPATI

    International Nuclear Information System (INIS)

    A data acquisition and processing system for reactor noise analysis is described. It consists of four-channel isolation amplifier, a seven-channel DC amplifier, a four-channel analog to digital converter, analog filters, a microcomputer system and a plotter. This system is being applied to investigate the reactor dynamics of the PUSPATI TRIGA MK II reactor. (author)

  20. Pressure-dependent calibration of the OH and HO2 channels of a FAGE HOx instrument using the Highly Instrumented Reactor for Atmospheric Chemistry (HIRAC)

    Science.gov (United States)

    Winiberg, F. A. F.; Smith, S. C.; Bejan, I.; Brumby, C. A.; Ingham, T.; Malkin, T. L.; Orr, S. C.; Heard, D. E.; Seakins, P. W.

    2015-02-01

    The calibration of field instruments used to measure concentrations of OH and HO2 worldwide has traditionally relied on a single method utilising the photolysis of water vapour in air in a flow tube at atmospheric pressure. Here the calibration of two FAGE (fluorescence assay by gaseous expansion) apparatuses designed for HOx (OH and HO2) measurements have been investigated as a function of external pressure using two different laser systems. The conventional method of generating known concentrations of HOx from H2O vapour photolysis in a turbulent flow tube impinging just outside the FAGE sample inlet has been used to study instrument sensitivity as a function of internal fluorescence cell pressure (1.8-3.8 mbar). An increase in the calibration constants CHO and CHO2 with pressure was observed, and an empirical linear regression of the data was used to describe the trends, with ΔCHO = (17 ± 11) % and ΔCHO2 = (31.6 ± 4.4)% increase per millibar air (uncertainties quoted to 2σ). Presented here are the first direct measurements of the FAGE calibration constants as a function of external pressure (440-1000 mbar) in a controlled environment using the University of Leeds HIRAC chamber (Highly Instrumented Reactor for Atmospheric Chemistry). Two methods were used: the temporal decay of hydrocarbons for calibration of OH, and the kinetics of the second-order recombination of HO2 for HO2 calibrations. Over comparable conditions for the FAGE cell, the two alternative methods are in good agreement with the conventional method, with the average ratio of calibration factors (conventional : alternative) across the entire pressure range, COH(conv)/COH(alt) = 1.19 ± 0.26 and CHO2(conv)/CHO2(alt) = 0.96 ± 0.18 (2σ). These alternative calibration methods currently have comparable systematic uncertainties to the conventional method: ~ 28% and ~ 41% for the alternative OH and HO2 calibration methods respectively compared to 35% for the H2O vapour photolysis method; ways in

  1. 蒸汽发生器下封头对核主泵入口流场影响%Influence of steam generator channel head on reactor coolant pump inflow field

    Institute of Scientific and Technical Information of China (English)

    侯向陶; 王鹏飞; 许忠斌; 阮晓东

    2016-01-01

    为了研究 AP1000蒸汽发生器(SG)下封头对反应堆冷却剂泵(RCP)入口流场的影响,将SG 下封头与 RCP 统一建模,采用 CFD 方法对其耦合模型进行全三维流场计算,分别研究了在稳态和瞬态情况下 SG 下封头对 RCP 入口流场的影响.稳态计算时,将均匀入流下缩尺泵的数值计算结果与试验结果进行对比,以验证数值计算方法的正确性;瞬态计算时,进行了时间步长无关性验证,以准确分析压力脉动特性.结果表明:在 SG 下封头的影响下,RCP 入口处产生了周向不均匀的轴向速度并且形成了2个回转方向相反的旋涡;瞬态下,核主泵入口处形成了2个低压区,与均匀入流情况相比,监测点的压力系数标准差增幅达53%~90%;SG 下封头使 RCP 入流产生预旋,且与叶轮形成较大的冲角,使得 RCP 扬程、效率分别下降了1.5%~7.7%,2.6%~4.1%.%To study the effect of AP1000 steam generator (SG)channel head on reactor coolant pump (RCP)inflow field,through modeling integrally the channel head of SG and RCP,the three-dimen-sional flow field of the above coupling models were simulated by the CFD method and the influence of SG channel head on RCP inflow field was studied on steady and transient conditions.On the steady condition,the accuracy of the numerical model was verified through comparing numerical simulation results with experimental results of the scaled RCP under uniform inflow.On the transient condition, time step independence was verified to study accurately the pressure fluctuation characteristics.Results show that at RCP inlet,SG channel head causes non-uniform circumferential axial velocity and two vor-tices which rotate in the opposite direction.On the transient condition,there are two-low pressure re-gions at RCP inlet and standard deviation of monitor points′pressure coefficient is increased by 53% -90%,compared with the condition of

  2. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  3. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    2013-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  4. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  5. RFI channels

    Science.gov (United States)

    Mceliece, R. J.

    1980-01-01

    A class of channel models is presented which exhibit varying burst error severity much like channels encountered in practice. An information-theoretic analysis of these channel models is made, and conclusions are drawn that may aid in the design of coded communication systems for realistic noisy channels.

  6. Thermal-hydraulic interfacing code modules for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  7. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  8. Research reactors

    International Nuclear Information System (INIS)

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  9. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  10. Reactor building

    International Nuclear Information System (INIS)

    The whole reactor building is accommodated in a shaft and is sealed level with the earth's surface by a building ceiling, which provides protection against penetration due to external effects. The building ceiling is supported on walls of the reactor building, which line the shaft and transfer the vertical components of forces to the foundations. The thickness of the walls is designed to withstand horizontal pressure waves in the floor. The building ceiling has an opening above the reactor, which must be closed by cover plates. Operating equipment for the reactor can be situated above the building ceiling. (orig./HP)

  11. Nuclear reactor building

    Science.gov (United States)

    Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  12. TREAT Reactor Control and Protection System

    International Nuclear Information System (INIS)

    The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab

  13. Compact Reactor

    International Nuclear Information System (INIS)

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date

  14. NEUTRONIC REACTOR

    Science.gov (United States)

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  15. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  16. Chemical Reactors.

    Science.gov (United States)

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  17. NUCLEAR REACTOR

    Science.gov (United States)

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  18. Nuclear reactor

    International Nuclear Information System (INIS)

    In order to reduce neutron embrittlement of the pressue vessel of an LWR, blanked off elements are fitted at the edge of the reactor core, with the same dimensions as the fuel elements. They are parallel to each other, and to the edge of the reactor taking the place of fuel rods, and are plates of neutron-absorbing material (stainless steel, boron steel, borated Al). (HP)

  19. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Iwashige, Kengo

    1996-06-21

    In an LMFBR type reactor, partitions are disposed to a coolant channel at positions lower than the free liquid level, and the width of the partitions is adapted to have a predetermined condition. Namely, when low temperature fluid overflowing the wall of the coolant channel, flows down and collided against the free liquid surface in the coolant channel, since the dropping speed thereof is reduced abruptly, large pressure waves are caused by kinetic force of the low temperature fluid. However, if appropriate numbers of partitions having an appropriate shape are formed, the dropping speed of the low temperature fluid is moderated to reduce the pressure waves. In addition, since the pressure waves are dispersed to the circumferential and lateral directions of the coolant flow channel respectively, the propagation of the pressure waves can be prevented effectively. Further, when the flow of the low temperature fluid is changed to the circumferential direction, for example, by earthquakes, since the partitions act as members resisting against the circumferential change of the low temperature fluid, the change of the direction can be suppressed. (N.H.)

  20. Catalytic reactor

    NARCIS (Netherlands)

    Sie, S.T.; Cybulski, A.; Moulijn, J.A.

    2000-01-01

    PCT No. PCT/NL93/00231 Sec. 371 Date Jul. 21, 1995 Sec. 102(e) Date Jul. 21, 1995 PCT Filed Nov. 4, 1993 PCT Pub. No. WO94/09901 PCT Pub. Date May 11, 1994There is described a catalyst element (1) consisting of an integral whole having channels (2) extending therethrough. These channels (2) have, in

  1. New technology for reactor protection system of CAREM reactor

    International Nuclear Information System (INIS)

    The use of FPGA in safety functions in a nuclear power plant, increase the reliability of software based systems, without loose any of the function required by the supervision and control systems. In this work the architecture of a Reactor Protection System is described, it use four independent measurement channels in 2 oo 4 configuration, each channel is based on diverse approach in 1 oo 2 configuration, the reliability of this system is near the same than the hardwired logic, with full performance like software based system. (author)

  2. Thermal-hydraulic modeling of reactivity accidents in MTR reactors

    Directory of Open Access Journals (Sweden)

    Khater Hany

    2006-01-01

    Full Text Available This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactors during a reactivity insertion accident. The model is formulated for coupling reactor kinetics with feedback reactivity and reactor core thermal-hydraulics. To represent the reactor core, two types of channels are considered, average and hot channels. The developed computer program is compiled and executed on a personal computer, using the FORTRAN language. The model is validated by safety-related benchmark calculations for MTR-TYPE reactors of IAEA 10 MW generic reactor for both slow and fast reactivity insertion transients. A good agreement is shown between the present model and the benchmark calculations. Then, the model is used for simulating the uncontrolled withdrawal of a control rod of an ETRR-2 reactor in transient with over power scram trip. The model results for ETRR-2 are analyzed and discussed.

  3. Material test reactor fuel research at the BR2 reactor

    International Nuclear Information System (INIS)

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  4. Evaluation of the thermal neutron flux in samples of Al–Au alloy irradiated in the carrousel channels of the TRIGA MARK I IPR-R1 reactor using MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Salomé, J.A.D.; Guerra, B.T. [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Av. Antônio Carlos, 6627 – PCA1 – Anexo Engenharia – Pampulha, CEP 31270-901, Belo Horizonte, MG (Brazil); Pereira, C., E-mail: claubia@nuclear.ufmg.br [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Av. Antônio Carlos, 6627 – PCA1 – Anexo Engenharia – Pampulha, CEP 31270-901, Belo Horizonte, MG (Brazil); Menezes, M.Â.B.C. de [Centro de Desenvolvimento da Tecnologia Nuclear, Comissão Nacional de Energia Nuclear, Campus da UFMG, Av. Antônio Carlos, 6627 31270-901, P.O. Box 941, Belo Horizonte, MG (Brazil); Silva, C.A.M. da [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Av. Antônio Carlos, 6627 – PCA1 – Anexo Engenharia – Pampulha, CEP 31270-901, Belo Horizonte, MG (Brazil); Dalle, H.M. [Centro de Desenvolvimento da Tecnologia Nuclear, Comissão Nacional de Energia Nuclear, Campus da UFMG, Av. Antônio Carlos, 6627 31270-901, P.O. Box 941, Belo Horizonte, MG (Brazil)

    2014-07-01

    Highlights: • The TRIGA IPR-R1 was modelled using MCNP. • The thermal neutron flux through the samples in eleven irradiation channels was obtained. • The simulated results were compared to experimental values. • The relative error, the relative trend, the z-score test and uncertainty were analysed. - Abstract: The TRIGA IPR-R1 was modelled using MCNP. The model consists of a cylinder filled with water, fuel elements, radial reflectors, central tube, control rods and neutron source. Around the core is placed the Rotary Specimen Rack (RSR) with adequate groove to insert the samples to irradiation. The values of the thermal neutron flux through the samples in eleven irradiation channels were simulated and compared to the experimental results to validate the model. After that, the values of the thermal neutron flux, in the same channels, were simulated on two horizontal planes at different heights and compared to validate the model. These channels were characterized as representative channels of the neutron flux distribution in the RSR. To evaluate the results, the relative errors, the relative trend, the z-score test and the relevance to a confidence interval of 95% were analysed. Good agreement has been obtained for the most channels when compared with the experimental results.

  5. The alpha channeling effect

    Energy Technology Data Exchange (ETDEWEB)

    Fisch, N. J.

    2015-12-10

    Alpha particles born through fusion reactions in a tokamak reactor tend to slow down on electrons, but that could take up to hundreds of milliseconds. Before that happens, the energy in these alpha particles can destabilize on collisionless timescales toroidal Alfven modes and other waves, in a way deleterious to energy confinement. However, it has been speculated that this energy might be instead be channeled into useful energy, so as to heat fuel ions or to drive current. Such a channeling needs to be catalyzed by waves Waves can produce diffusion in energy of the alpha particles in a way that is strictly coupled to diffusion in space. If these diffusion paths in energy-position space point from high energy in the center to low energy on the periphery, then alpha particles will be cooled while forced to the periphery. The energy from the alpha particles is absorbed by the wave. The amplified wave can then heat ions or drive current. This process or paradigm for extracting alpha particle energy collisionlessly has been called alpha channeling. While the effect is speculative, the upside potential for economical fusion is immense. The paradigm also operates more generally in other contexts of magnetically confined plasma.

  6. Method and apparatus for a catalytic firebox reactor

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Lance L. (North Haven, CT); Etemad, Shahrokh (Trumbull, CT); Ulkarim, Hasan (Hamden, CT); Castaldi, Marco J. (Bridgeport, CT); Pfefferle, William C. (Madison, CT)

    2001-01-01

    A catalytic firebox reactor employing an exothermic catalytic reaction channel and multiple cooling conduits for creating a partially reacted fuel/oxidant mixture. An oxidation catalyst is deposited on the walls forming the boundary between the multiple cooling conduits and the exothermic catalytic reaction channel, on the side of the walls facing the exothermic catalytic reaction channel. This configuration allows the oxidation catalyst to be backside cooled by any fluid passing through the cooling conduits. The heat of reaction is added to both the fluid in the exothermic catalytic reaction channel and the fluid passing through the cooling conduits. After discharge of the fluids from the exothermic catalytic reaction channel, the fluids mix to create a single combined flow. A further innovation in the reactor incorporates geometric changes in the exothermic catalytic reaction channel to provide streamwise variation of the velocity of the fluids in the reactor.

  7. Using thermal balance model to determine optimal reactor volume and insulation material needed in a laboratory-scale composting reactor.

    Science.gov (United States)

    Wang, Yongjiang; Pang, Li; Liu, Xinyu; Wang, Yuansheng; Zhou, Kexun; Luo, Fei

    2016-04-01

    A comprehensive model of thermal balance and degradation kinetics was developed to determine the optimal reactor volume and insulation material. Biological heat production and five channels of heat loss were considered in the thermal balance model for a representative reactor. Degradation kinetics was developed to make the model applicable to different types of substrates. Simulation of the model showed that the internal energy accumulation of compost was the significant heat loss channel, following by heat loss through reactor wall, and latent heat of water evaporation. Lower proportion of heat loss occurred through the reactor wall when the reactor volume was larger. Insulating materials with low densities and low conductive coefficients were more desirable for building small reactor systems. Model developed could be used to determine the optimal reactor volume and insulation material needed before the fabrication of a lab-scale composting system.

  8. Gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    The gas temperature of a hot gas loop in gas-cooled nuclear reactor plants shall be able to be modified without influencing the gas temperature of the other loops. If necessary, it should be possible to stop the loop. This is possible by means of a mixer which is places below the heat absorbing component in the hot channel and which is connected to a cold gas line. (orig.)

  9. Multi-channel gas-delivery system

    Energy Technology Data Exchange (ETDEWEB)

    Rozenzon, Yan; Trujillo, Robert T.; Beese, Steven C.

    2016-09-13

    One embodiment of the present invention provides a gas-delivery system for delivering reaction gas to a reactor chamber. The gas-delivery system includes a main gas-inlet port for receiving reaction gases and a gas-delivery plate that includes a plurality of gas channels. A gas channel includes a plurality of gas holes for allowing the reaction gases to enter the reactor chamber from the gas channel. The gas-delivery system further includes a plurality of sub-gas lines coupling together the main gas-inlet port and the gas-delivery plate, and a respective sub-gas line is configured to deliver a portion of the received reaction gases to a corresponding gas channel.

  10. Sonochemical Reactors.

    Science.gov (United States)

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  11. Reactor water level control system

    International Nuclear Information System (INIS)

    A BWR type reactor comprises a control valve disposed in a reactor water draining pipelines and undergoing an instruction to control the opening degree, an operation board having a setting device for generating the instruction and a control board for giving the instruction generated by the setting device to the control valve. The instruction is supplied from the setting device to the control valve by way of a control circuit to adjust the opening degree of the control valve thereby controlling the water level in the reactor. In addition, a controller generating an instruction independent of the setting device and a signal transmission channel for signal-transmitting the instruction independent of the control circuit are disposed, to connect the controller electrically to the signal transmission. The signal transmission channel and the control circuit are electrically connected to the control valve switchably with each other. Since instruction can be given to the control valve even at a periodical inspection or modification when the setting device and the control circuit can not be used, the reactor water level can be controlled automatically. Then, operator's working efficiency upon inspection can be improved remarkably. (N.H.)

  12. Brands & Channels

    Institute of Scientific and Technical Information of China (English)

    Alice Yang

    2009-01-01

    @@ "Brands" and "Channels" are the two most important things in Ku-Hai Chen's eyes when doing business with Main-land China. Ku-Hai Chen, Executive Director of the International Trade Institute of Taiwan External Trade Development Council (TAITRA), flies frequently between Chinese Taipei and Mainland China, and was in Beijing earlier this month for his seminar.

  13. The WWR-SM-20 research reactor

    International Nuclear Information System (INIS)

    In this paper the design features and experimental capabilities of the WWR-SM-20 research reactor are described. The reactor uses fuel assemblies consisting of six coaxial fuel tubes with a square cross-section. IRT-3M fuel assemblies can be used with both 90% enriched and 36% enriched uranium. The main characteristics of the IRT-3M fuel assemblies are given, as are the technical and physical parameters of the WWR-SM-20 reactor. The core can hold up to ten ampoule-type channels with a diameter of up to 68 mm. For irradiation purposes, up to 22 26-mm-diameter channels in the fuel assemblies, and up to 48 42-mm-diameter channels in the beryllium blocks of the reflector can be used. In the graphite blanket between the horizontal channels, channels with a diameter of up to 130 mm can be used. The thermal neutron flux density has a maximum value of 1.5 X 1018 m-2 · s-1 in the core and 2.3 X 1018 m-2 · s-1 in the reflector, and the fast neutron flux density (cE > 0.821 MeV) a maximum of 1.9 X 1018 m-2 · s-1. A number of design features have been incorporated in the WWR-SM-20 reactor to make it effectively safe

  14. RA reactor exploitation, task 3.08/01

    International Nuclear Information System (INIS)

    During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report

  15. Pressurized water reactor flow skirt apparatus

    Science.gov (United States)

    Kielb, John F.; Schwirian, Richard E.; Lee, Naugab E.; Forsyth, David R.

    2016-04-05

    A pressurized water reactor vessel having a flow skirt formed from a perforated cylinder structure supported in the lower reactor vessel head at the outlet of the downcomer annulus, that channels the coolant flow through flow holes in the wall of the cylinder structure. The flow skirt is supported at a plurality of circumferentially spaced locations on the lower reactor vessel head that are not equally spaced or vertically aligned with the core barrel attachment points, and the flow skirt employs a unique arrangement of hole patterns that assure a substantially balanced pressure and flow of the coolant over the entire underside of the lower core support plate.

  16. Channel Power in Multi-Channel Environments

    NARCIS (Netherlands)

    M.G. Dekimpe (Marnik); B. Skiera (Bernd)

    2004-01-01

    textabstractIn the literature, little attention has been paid to instances where companies add an Internet channel to their direct channel portfolio. However, actively managing multiple sales channels requires knowing the customers’ channel preferences and the resulting channel power. Two key compon

  17. Reactor container

    International Nuclear Information System (INIS)

    A reactor container has a suppression chamber partitioned by concrete side walls, a reactor pedestal and a diaphragm floor. A plurality of partitioning walls are disposed in circumferential direction each at an interval inside the suppression chamber, so that independent chambers in a state being divided into plurality are formed inside the suppression chamber. The partition walls are formed from the bottom portion of the suppression chamber up to the diaphragm floor to isolate pool water in a divided state. Operation platforms are formed above the suppression chamber and connected to an access port. Upon conducting maintenance, inspection or repairing, a pump is disposed in the independent chamber to transfer pool water therein to one or a plurality of other independent chambers to make it vacant. (I.N.)

  18. Specifications for reactor physics experiments on CANFLEX-RU fuel

    International Nuclear Information System (INIS)

    This is to describe reactor physics experiments to be performed in the ZED-2 reactor to study CANFLEX-RU fuel bundles in CANDU-type fuel channels. The experiments are to provide benchmark quality validation data for the computer codes and associated nuclear databases used for physics calculations, in particular WIMS-AECL. Such validation data is likely to be a requirement by the regulator as condition for licensing a CANDU reactor based on an enriched fuel cycle

  19. Development of Multi-channel Amplifier-discriminator Based on Measurement System of Neutron Fluence Rate Relative Distribution in Reactor%反应堆中子注量率相对分布测量装置中多通道放大甄别器研制

    Institute of Scientific and Technical Information of China (English)

    赵艳辉; 刘丽艳; 黄顺; 刘才学; 踪训成; 赵修良

    2014-01-01

    为了测量反应堆内中子注量率分布,保证反应堆内活化55 M n-58 Ni合金探测片γ计数测量的可靠性,本文研制了中子注量率分布测量装置中9通道放大甄别器。多通道放大甄别器性能指标测试与应用测试结果表明:每个通道放大器增益1~21连续可调、甄别器阈值独立连续可调,具有最大计数率高、灵敏度高、稳定性好、系统抗串扰能力强等优点;放大器增益长期稳定性≤1%,甄别器最小输入脉冲宽度≥0.1μs ,甄别器最大计数率≤4×106 s-1,能用于实时长期稳定测量反应堆内中子注量率分布。%For the measurement of neutron fluence rate in reactor and reliable assurance of γcount measurement of activated 55 M n-58 Ni alloy irradiated foils in reactor ,9-channel amplifier-discriminator was developed . The main technical parameter test and application test show that the gain of each channel amplifier-discriminator is continuously adjustable from 1 to 21 , the threshold of each discriminator circuit is continuously adjustable ,the maximum count rate and sensitivity of discriminator circuit are high , and the system has stable property and excellent anti-interference . In conclusion ,relevant technical parameters can guarantee the real-time and long-term stable measurement of neutron fluence rate relative distribution in reactor , with the technical parameters that gain stability of amplifier is less than 1% ,the minimum input pulse width of discriminator circuit is greater than 0.1 μs ,and the maximum count rate of discriminator is less than 4 × 106 s-1 .

  20. Channel Networks

    Science.gov (United States)

    Rinaldo, Andrea; Rodriguez-Iturbe, Ignacio; Rigon, Riccardo

    This review proceeds from Luna Leopold's and Ronald Shreve's lasting accomplishments dealing with the study of random-walk and topologically random channel networks. According to the random perspective, which has had a profound influence on the interpretation of natural landforms, nature's resiliency in producing recurrent networks and landforms was interpreted to be the consequence of chance. In fact, central to models of topologically random networks is the assumption of equal likelihood of any tree-like configuration. However, a general framework of analysis exists that argues that all possible network configurations draining a fixed area are not necessarily equally likely. Rather, a probability P(s) is assigned to a particular spanning tree configuration, say s, which can be generally assumed to obey a Boltzmann distribution: P(s) % e^-H(s)/T, where T is a parameter and H(s) is a global property of the network configuration s related to energetic characters, i.e. its Hamiltonian. One extreme case is the random topology model where all trees are equally likely, i.e. the limit case for T6 4 . The other extreme case is T 6 0, and this corresponds to network configurations that tend to minimize their total energy dissipation to improve their likelihood. Networks obtained in this manner are termed optimal channel networks (OCNs). Observational evidence suggests that the characters of real river networks are reproduced extremely well by OCNs. Scaling properties of energy and entropy of OCNs suggest that large network development is likely to effectively occur at zero temperature (i.e. minimizing its Hamiltonian). We suggest a corollary of dynamic accessibility of a network configuration and speculate towards a thermodynamics of critical self-organization. We thus conclude that both chance and necessity are equally important ingredients for the dynamic origin of channel networks---and perhaps of the geometry of nature.

  1. Nuclear research reactors

    International Nuclear Information System (INIS)

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.)

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    A nuclear reactor is described in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assemblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters in the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters in the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance

  3. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  4. Candu reactors with thorium fuel cycles

    International Nuclear Information System (INIS)

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

  5. Experimental study on turbulent mixing in two parallel channels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C.M. [Graduate School, Yonsei University, Seoul (Korea); Yang, S.K.; Chun, S.Y.; Chung, M.K. [Korea Atomic Energy Research Institute, Taejon (Korea); Rhim, Y.C. [Yonsei University, Seoul (Korea)

    1998-11-01

    The hydraulic characteristics in two parallel channels were measured using two-component laser-doppler velocimeter(LDV). Parallel rectangular channels in different sizes were simply simulated from subchannels in rod bundles of a nuclear reactor. Time mean velocity, turbulence intensity, skewness factor, and flatness factor in axial and transverse components were measured. From the measured results, turbulent mixing phenomena between two parallel channels were investigated quantitatively. (author). 7 refs., 12 figs.

  6. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  7. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  8. Fast leak of a channel filled with helium at a pressure of 2 bars (channel H5)

    International Nuclear Information System (INIS)

    The loss of seal of a helium-filled channel opening the entire cross section of the front part leads to a fast leak. The channel fills to the upper generatrix of the leak orifice and part of the helium contained in the channel escapes into the circuit. The pressure drop in the reflector can lead to reactor and main pump shutdown. On the other hand, the Cooling Circuit Shutdown Bar circuit pumps remain in operation. This paper evaluates the consequences of an incident of this nature for the reactor and the surrounding experimental zones

  9. Channeling experiment

    International Nuclear Information System (INIS)

    Channeling of water flow and tracer transport in real fractures in a granite body at Stripa have been investigated experimentally. The experimental site was located 360 m below the ground level. Two kinds of experiments were performed. In the single hole experiments, 20 cm diameter holes were drilled about 2.5 m into the rock in the plane of the fracture. Specially designed packers were used to inject water into the fracture in 5 cm intervals all along the fracture trace in the hole. The variation of the injection flowrates along the fracture were used to determine the transmissivity variations in the fracture plane. Detailed photographs were taken from inside the hole and the visual fracture aperture was compared with the injection flowrates in the same locations. Geostatistical methods were used to evaluate the results. Five holes were measured in great detail. In addition 7 holes were drilled and scanned by simpler packer systems. A double hole experiment was performed where two parallel holes were drilled in the same fracture plane at nearly 2 m distance. Pressure pulse tests were made between the holes in both directions. Tracers were injected in 5 locations in one hole and monitored for in many locations in the other hole. The single hole experiment and the double hole experiment show that most of the fracture planes are tight but that there are open sections which form connected channels over distances of at least 2 meters. It was also found in the double hole experiment that the investigated fracture was intersected by at least one fracture between the two holes which diverted a large amount of the injected tracers to several distant locations at the tunnel wall. (authours)

  10. Reactor container

    International Nuclear Information System (INIS)

    Purpose: To prevent shocks exerted on a vent head due to pool-swell caused within a pressure suppression chamber (disposed in a torus configuration around the dry well) upon loss of coolant accident in BWR type reactors. Constitution: The following relationship is established between the volume V (m3) of a dry well and the ruptured opening area A (m2) at the boundary expected upon loss of coolant accident: V >= 30340 (m) x A Then, the volume of the dry well is made larger than the ruptured open area, that is, the steam flow rate of leaking coolants upon loss of coolant accident to decrease the pressure rise in the dry well at the initial state where loss of coolant accident is resulted. Accordingly, the pressure of non-compressive gases jetted out from the lower end of the downcomer to the pool water is decreased to suppress the pool-swell. (Ikeda, J.)

  11. Modernization of control instrumentation and security of reactor IAN - R1

    International Nuclear Information System (INIS)

    The program to modernize IAN-R1 research reactor control and safety instrumentation has been carried out considering two main aspects: updating safety philosophy requirements and acquiring the newest reactor control instrumentation controlled by computer, following the present criteria internationally recognized, for safety and reliable reactor operations and the latest developments of nuclear electronic technology. The new IAN-R1 reactor instrumentation consist of two wide range neutron monitoring channels, commanded by microprocessor a data acquisition system and reactor control, (controlled by computers). The reactor control desk is providing through two displays; all safety and control signals to the reactor operators; furthermore some signals like reactor power, safety and period signals are also showed on digital bar graphics, which are hard wired directly from the neutron monitoring channels

  12. Study of instabilities in phase by using the tool {sup D}ynamics{sup :} analysis of the evolution space temporary of the waves of density in channels of reactors BWR; Estudio de las Inestabilidades en Fase Mediante la Herramienta Dinamics: analisis de la Evolucion Espacio Temporal de las Ondas de Densidad en Canales de Reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J. L.; Escriva, R.; Merino, R.; Melara, J.

    2013-07-01

    This paper presents the basics of Dynamics V2 to code It allows calculations of stability for oscillations in phase in BWR reactors in the time domain. The equations of the model are exposed and is the integration of the equations. The model can be used in a large number of nodes thrust for the calculations to an acceptable computational cost, it has simplified dynamics of recirculation loop and the code has been incorporated the Oscillation in phase boundary conditions. The code incorporates the equations of boiling sub-cooled which allows to make more realistic calculations as well as subroutines to calculate the subroutines-based properties of the MATPRO and ASME.

  13. Researches on oxidation of ethyl benzene in continuous micro-channel reactor%微通道反应器内乙苯连续氧化反应工艺研究

    Institute of Scientific and Technical Information of China (English)

    严生虎; 沈卫; 张跃; 刘建武; 沈介发

    2012-01-01

    The preparation of acetophenone by oxidation of ethyl benzene is investigated in "heart cell" microchannel reactor. The effects of mole ratio of catalyst to ethyl benzene, resident time, amount of solvent and reaction temperature on single conversion and product selectivity are performed. Under the typical conditions: 0. 13 of n(acetate cobalt)/n(ethyl-benzene) ,1. 75 of promoting agents Br/Co,10 of solvent/ethylbenzene,70 s of resident time and 100℃ of reaction temperature, the conversion of ethyl benzene reaches 30.7% and the acetophenone selectivity is 100%.%研究了在“心型”结构微通道反应器内乙苯与过氧化氢反应连续合成苯乙酮的氧化工艺过程,考察了乙苯与催化剂醋酸钴的摩尔比、停留时间、溶剂量、反应温度等对单程转化率及选择性的影响.在典型的工艺条件下,n(醋酸钴)∶n(乙苯)=13%.停留时间70 s,促进剂Br/Co=1.75,V(冰乙酸)/V(乙苯)=10∶1,温度为100℃,乙苯的转化率达到30.7%,苯乙酮的选择性达到100%.

  14. Research Nuclear Reactors

    International Nuclear Information System (INIS)

    Published in English and in French, this large report first proposes an overview of the use and history of research nuclear reactors. It discusses their definition, and presents the various types of research reactors which can be either related to nuclear power (critical mock-ups, material test reactors, safety test reactors, training reactors, prototypes), or to research (basic research, industry, health), or to specific particle physics phenomena (neutron diffraction, isotope production, neutron activation, neutron radiography, semiconductor doping). It reports the history of the French research reactors by distinguishing the first atomic pile (ZOE), and the activities and achievements during the fifties, the sixties and the seventies. It also addresses the development of instrumentation for research reactors (neutron, thermal, mechanical and fission gas release measurements). The other parts of the report concern the validation of neutronics calculations for different reactors (the EOLE water critical mock-up, the MASURCA air critical mock-up dedicated to fast neutron reactor study, the MINERVE water critical mock-up, the CALIBAN pulsed research reactor), the testing of materials under irradiation (OSIRIS reactor, laboratories associated with research reactors, the Jules Horowitz reactor and its experimental programs and related devices, irradiation of materials with ion beams), the investigation of accident situations (on the CABRI, Phebus, Silene and Jules Horowitz reactors). The last part proposes a worldwide overview of research reactors

  15. Channel strategy adaptation

    OpenAIRE

    Rangan, V. Kasturi; Nueno, Jose L

    1999-01-01

    Using transaction cost theory, considerable research in marketing has focused on the conditions under which firms would use direct or vertically integrated versus indirect or arms length channels of distribution. Data from the field, however, indicate that channel configurations are more varied and complex, with multiple channels and composite channels being just as common as direct and indirect channels. In an attempt to explain this variety, this paper revisits the influence on channel stru...

  16. Conceptual Study on Dismantling of CANDU Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  17. Reactor Physics Training

    International Nuclear Information System (INIS)

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  18. A digital data acquisition and display system for ITU TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Full text: In this study, a digital data acquisition and display system realized for ITU TRIGA Mark-II Reactor is described. This system is realized in order to help the reactor operator and to increase reactor console capacity. The system consists of two main units, which are host computers and RTI-815F, analog devices, data acquisition card. RTI-815F is multi-function analog/digital input/output board that plugs into one of the available long expansion slots in the IBM-PC, PC/XT, PC/AT, or equivalent personal computers. It has 16 analog input channels for single-ended input signals or 8 analog input channels for differential input signals. But its channel capacity can be increased to 32 input channels for single-ended input signals or 16 input channels for differential input signals. RTI-815F board contains 2 analog output channels, 8 digital input channels and 8 digital output channels. In the ITD TRIGA Mark-II Reactor, 6 fuel temperature channels, 3 water temperature channels, 3 control rod position channels and 4 power channels are chosen as analog input signals for RTI-815F. Its digital outputs are assigned to cooling tower fan, primary and secondary pump reactor scram, control rod rundown. During operation, data are automatically archived to disk and displayed on screen. The channel selection time and sampling time can be adjusted. The simulated movement and position of control rods in the reactor core can be noted and displayed. The changes of power, fuel temperature and water temperature can be displayed on the screen as a graphic. In this system both period and reactivity are calculated and displayed on the screen. (authors)

  19. The AECL reactor development programme

    International Nuclear Information System (INIS)

    The modem CANDU-PHWR power reactor is the result of more than 50 years of evolutionary design development in Canada. It is one of only three commercially successful designs in the world to this date. The basis for future development is the CANDU 6 and CANDU 9 models. Four of the first type are operating and four more will go an line before the end of this decade. The CANDU 9 is a modernized single-unit version of the twelve large multi-unit plants operated by Ontario Hydro. All of these plants use proven technology which resulted from research, development, design construction, and operating experience over the past 25 years. Looking forward another 25 years, AECL plans to retain all of the essential features that distinguish today's CANDU reactors (heavy water moderation, on-power fuelling simple bundle design, horizontal fuel channels, etc.). The end product of the planned 25-year development program is more than a specific design - it is a concept which embodies advanced features expected from ongoing R and D programs. To carry out the evolutionary work we have selected seven main areas for development: Safety Technology, Fuel and Fuel Cycles, Fuel Channels, Systems and Components, Heavy Water and Tritium Information Technology, and Construction. There are three strategic measures of success for each of these work areas: improved economics, advanced fuel cycle utilization, and enhanced safety/plant robustness. The paper describes these work programs and the overall goals of each of them. (author)

  20. Controlled beta-quenching of fuel channels using inert gas

    International Nuclear Information System (INIS)

    The trend towards higher fuel assembly discharge burnups poses new challenges for fuel channels in terms of their dimensional behavior and corrosion resistance. This led AREVA NP to develop a new technique for beta quenching of fuel channels that combines the effect of beta-quenching with the optimization of the microstructure. The first set of fuel channels with these optimized material properties have been placed in the core of a German boiling water reactor (BWR) nuclear power plant in spring of 2004. Some more channels have been sited in the core of a Scandinavian BWR in fall of 2007 to broaden the in-pile experience with these channels. Dimensional stability is the major requirement that is applied to fuel channels. High corrosion resistance and low hydrogen pickup are certainly required as well. However, corrosion and hydrogen pickup are usually not life limiting factors due to the large wall thickness of the material. Since thick layers of oxide may spall off extensively at high burnup and cause increase of the dose rate for the personnel, high corrosion resistance of fuel channels is mandatory. The fuel channels which surround BWR fuel assemblies are exposed to neutron irradiation as well as to loads induced by the reactor coolant flowing through them. These service conditions induce material growth and creep which cause permanent changes in the dimensions of the channels. Especially, fuel channel bow is of certain interest as increased channel bow may lead to some friction with control blades. Fuel channel bow is mainly induced by fluence gradients. However, there may be additional influences such as oxidation and hydrogen uptake to cause increased channel bow. The effect of hydrogen is currently discussed in the nuclear community to explain the unexpected high fuel channel bow that has been observed in some nuclear power plants. (orig.)

  1. Safeguarding research reactors

    International Nuclear Information System (INIS)

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  2. Dimensional Behavior of Fuel Channels - Recent Experience and Consequences

    International Nuclear Information System (INIS)

    Fuel channels in boiling-water reactors (BWR) undergo distortions like bow, bulge, and twist due to their operating conditions. These distortions may adversely impact planned operating strategy, and therefore need to be adequately addressed during various stages of fuel channel design and manufacturing, core design and operation monitoring. Fuel channel distortion may lead to interference between the fuel channel and adjacent control blade. If severe, such interference can impair both positioning of control blades during normal operations and rapid control blade insertion during a reactor scram. During the last five years, unexpectedly high fuel channel distortions leading to problems in control blade operations have been observed in some C- and S-lattice BWR plants in the U.S. operating on 18 - 24 month cycles. As a result, U.S. operators have implemented costly surveillance programs to detect the onset of high distortions and have declared control blades inoperable when unacceptable control blade operation occurs. This unusual fuel channel distortion has not been observed with AREVA NP fuel supplied in Europe in this scale. Nevertheless fuel channel distortion-related problems were recently observed in reactors outside the U.S. with early control blade operation. The mechanisms causing this unexpected fuel-channel distortion and the influencing factors are not completely understood worldwide for the time being. Therefore, a prediction of channels which will exhibit high bow is very challenging. A status is given from the AREVA NP perspective on: - The existing fuel channel distortion database, - The understanding of the phenomenon, - Measures to gather further information and improve existing tools, materials, and designs, and - Customer actions to reduce potential high channel bow and associated control blade issues. (authors)

  3. Development of demonstration advanced thermal reactor

    International Nuclear Information System (INIS)

    The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported. (Kako, I.)

  4. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  5. Mobile radio channels

    CERN Document Server

    Pätzold, Matthias

    2011-01-01

    Providing a comprehensive overview of the modelling, analysis and simulation of mobile radio channels, this book gives a detailed understanding of fundamental issues and examines state-of-the-art techniques in mobile radio channel modelling. It analyses several mobile fading channels, including terrestrial and satellite flat-fading channels, various types of wideband channels and advanced MIMO channels, providing a fundamental understanding of the issues currently being investigated in the field. Important classes of narrowband, wideband, and space-time wireless channels are explored in deta

  6. Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

    KAUST Repository

    Zhao, Zhenlong

    2013-01-17

    Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

  7. Eddy current and ultrasonic fuel channel inspection at Karachi Nuclear Power Plant

    International Nuclear Information System (INIS)

    In November of 1993 and in-service inspection was performed on eight fuel channels in the Karachi Nuclear Power Plant (KANUPP) reactor. The workscope included ultrasonic and eddy current volumetric examinations, and eddy current measurement of pressure-to calandria tube gap. This paper briefly discusses the planning strategy of the ultrasonic and eddy current examinations, and describes the equipment developed to meet the requirements, followed by details of the actual channel inspection campaign. The presented nondestructive examinations assisted in determining fitness for service of KANUPP reactor channels in general, and confirmed that the problems associated with channel G12 were not generic in nature. (author)

  8. Channel nut tool

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Marvin

    2016-01-12

    A method, system, and apparatus for installing channel nuts includes a shank, a handle formed on a first end of a shank, and an end piece with a threaded shaft configured to receive a channel nut formed on the second end of the shaft. The tool can be used to insert or remove a channel nut in a channel framing system and then removed from the channel nut.

  9. Ship propulsion reactors technology

    International Nuclear Information System (INIS)

    This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

  10. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  11. Reactor System Design

    International Nuclear Information System (INIS)

    SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

  12. Oregon State TRIGA Reactor (OSTR) console upgrading

    International Nuclear Information System (INIS)

    It was decided in the summer of 1977 to replace and upgrade part of the electronics of the OSTR console. The console was the original system installed in 1967 when the reactor first went critical. Although it was generally quite reliable, maintenance was becoming more frequent, and locating spare and replacement parts was getting very difficult. The upgrading would replace the majority of the system with new, state- of-the-art electronics. The new, upgrading package consisted of replacing the left-hand console electronics drawer; specifically: 1. The present multirange linear channel using an ion chamber was replaced by a new 9.5-decade linear channel driven by a fission chamber. No scram features are on the new linear channel. 2. The present multirange log channel with a period circuit using an ion chamber was replaced by a new 10-decade wide- range log channel, also with a period circuit, driven by a fission chamber. The same fission chamber drives the new linear and wide-range log channels. 3. The present count-rate (startup) channel using a fission chamber was removed. Its function was taken over by the new wide-range log and linear channels. 4. A new safety channel with scram capability, driven by an ion chamber, was added. 5. A new fuel element temperature circuit with scram capability was added. The upgrading package was ordered in February 1978 and installation was completed in January 1980. One of the biggest time delays in the process was the NRC review time of the Technical Specifications amendment that was requested for this change. The actual installation of the new package required five weeks, including functional testing. The linearity of the new instrument systems is excellent, and the wide-range capability of the new log and linear channels provides increased operational flexibility and accuracy, especially when a low power run immediately follows a high power run. (author)

  13. Electromechanical drive for a reactor control system

    International Nuclear Information System (INIS)

    The invention is related to control systems of nuclear researche swimming pool-type reactors. The presented electromechanical drive for a nuclear reactor control system consists of an electromagnetic grip of control element with magnet power supply cable, drum and flexible element, e.g., wire rope. Two branches of the rope which are separated from the electromagnet to the core and the drum form the closed loop. To decrease the dimensions of the drive, the magnet power supply cable serves as a loop flexible element which goes from the electromagnet to the core. For a particular reactor the drive, made according to the invention is 100 mm shorter and 20 mm narrower as compared with the known one, and that is rather significant in cases when a drive is to be installed directly on a control system channel

  14. Plant life management strategies for pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

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

  15. An advanced method of heterogeneous reactor theory

    International Nuclear Information System (INIS)

    Recent approaches to heterogeneous reactor theory for numerical applications were presented in the course of 8 lectures given in JAERI. The limitations of initial theory known after the First Conference on Peacefull Uses of Atomic Energy held in Geneva in 1955 as Galanine-Feinberg heterogeneous theory:-matrix from of equations, -lack of consistent theory for heterogeneous parameters for reactor cell, -were overcome by a transformation of heterogeneous reactor equations to a difference form and by a development of a consistent theory for the characteristics of a reactor cell based on detailed space-energy calculations. General few group (G-number of groups) heterogeneous reactor equations in dipole approximation are formulated with the extension of two-dimensional problem to three-dimensions by finite Furie expansion of axial dependence of neutron fluxes. A transformation of initial matrix reactor equations to a difference form is presented. The methods for calculation of heterogeneous reactor cell characteristics giving the relation between vector-flux and vector-current on a cell boundary are based on a set of detailed space-energy neutron flux distribution calculations with zero current across cell boundary and G calculations with linearly independent currents across the cell boundary. The equations for reaction rate matrices are formulated. Specific methods were developed for description of neutron migration in axial and radial directions. The methods for resonance level's approach for numerous high-energy resonances. On the basis of these approaches the theory, methods and computer codes were developed for 3D space-time react or problems including simulation of slow processes with fuel burn-up, control rod movements, Xe poisoning and fast transients depending on prompt and delayed neutrons. As a result reactors with several thousands of channels having non-uniform axial structure can be feasibly treated. (author)

  16. Validation of reactor core protection system

    International Nuclear Information System (INIS)

    Reactor COre Protection System (RCOPS), an advanced core protection calculator system, is a digitized one which provides core protection function based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels adapted a two-out-of-four trip logic. System configuration, hardware platform and an improved algorithm of the newly designed core protection calculator system are described in this paper. One channel of RCOPS was implemented as a single channel facility for this R and D project where we performed final integration software testing. To implement custom function blocks, pSET is used. Software test is performed by two methods. The first method is a 'Software Module Test' and the second method is a 'Software Unit Test'. New features include improvement of core thermal margin through a revised on-line DNBR algorithm, resolution of the latching problem of control element assembly signal and addition of the pre-trip alarm generation. The change of the on-line DNBR calculation algorithm is considered to improve the DNBR net margin by 2.5%-3.3%. (author)

  17. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kawakami, Hiroto

    1995-02-07

    A reactor container of the present invention has a structure that the reactor container is entirely at the same temperature as that at the inlet of the reactor and, a hot pool is incorporated therein, and the reactor container has is entirely at the same temperature and has substantially uniform temperature follow-up property transiently. Namely, if the temperature at the inlet of the reactor core changes, the temperature of the entire reactor container changes following this change, but no great temperature gradient is caused in the axial direction and no great heat stresses due to axial temperature distribution is caused. Occurrence of thermal stresses caused by the axial temperature distribution can be suppressed to improve the reliability of the reactor container. In addition, since the laying of the reactor inlet pipelines over the inside of the reactor is eliminated, the reactor container is made compact and the heat shielding structures above the reactor and a protection structure of container walls are simplified. Further, secondary coolants are filled to the outside of the reactor container to simplify the shieldings. The combined effects described above can improve economical property and reliability. (N.H.).

  18. Fossil nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Maurette, M.

    1976-01-01

    The discussion of fossil nuclear reactors (the Oklo phenomenon) covers the earth science background, neutron-induced isotopes and reactor operating conditions, radiation-damage studies, and reactor modeling. In conclusion possible future studies are suggested and the significance of the data obtained in past studies is summarized. (JSR)

  19. Computational Fluid Dynamics Analysis of Canadian Supercritical Water Reactor (SCWR)

    Science.gov (United States)

    Movassat, Mohammad; Bailey, Joanne; Yetisir, Metin

    2015-11-01

    A Computational Fluid Dynamics (CFD) simulation was performed on the proposed design for the Canadian SuperCritical Water Reactor (SCWR). The proposed Canadian SCWR is a 1200 MW(e) supercritical light-water cooled nuclear reactor with pressurized fuel channels. The reactor concept uses an inlet plenum that all fuel channels are attached to and an outlet header nested inside the inlet plenum. The coolant enters the inlet plenum at 350 C and exits the outlet header at 625 C. The operating pressure is approximately 26 MPa. The high pressure and high temperature outlet conditions result in a higher electric conversion efficiency as compared to existing light water reactors. In this work, CFD simulations were performed to model fluid flow and heat transfer in the inlet plenum, outlet header, and various parts of the fuel assembly. The ANSYS Fluent solver was used for simulations. Results showed that mass flow rate distribution in fuel channels varies radially and the inner channels achieve higher outlet temperatures. At the outlet header, zones with rotational flow were formed as the fluid from 336 fuel channels merged. Results also suggested that insulation of the outlet header should be considered to reduce the thermal stresses caused by the large temperature gradients.

  20. The Earliest Ion Channels

    Science.gov (United States)

    Pohorille, A.; Wilson, M. A.; Wei, C.

    2009-12-01

    Supplying protocells with ions required assistance from channels spanning their membrane walls. The earliest channels were most likely short proteins that formed transmembrane helical bundles surrounding a water-filled pore. These simple aggregates were capable of transporting ions with efficiencies comparable to those of complex, contemporary ion channels. Channels with wide pores exhibited little ion selectivity but also imposed only modest constraints on amino acid sequences of channel-forming proteins. Channels with small pores could have been selective but also might have required a more precisely defined sequence of amino acids. In contrast to modern channels, their protocellular ancestors had only limited capabilities to regulate ion flux. It is postulated that subsequent evolution of ion channels progressed primarily to acquire precise regulation, and not high efficiency or selectivity. It is further proposed that channels and the surrounding membranes co-evolved.

  1. Gramicidin Channels: Versatile Tools

    Science.gov (United States)

    Andersen, Olaf S.; Koeppe, Roger E., II; Roux, Benoît

    Gramicidin channels are miniproteins in which two tryptophan-rich subunits associate by means of transbilayer dimerization to form the conducting channels. That is, in contrast to other ion channels, gramicidin channels do not open and close; they appear and disappear. Each subunit in the bilayer-spanning channel is tied to the bilayer/solution interface through hydrogen bonds that involve the indole NH groups as donors andwater or the phospholipid backbone as acceptors. The channel's permeability characteristics are well-defined: gramicidin channels are selective for monovalent cations, with no measurable permeability to anions or polyvalent cations; ions and water move through a pore whose wall is formed by the peptide backbone; and the single-channel conductance and cation selectivity vary when the amino acid sequence is varied, even though the permeating ions make no contact with the amino acid side chains. Given the plethora of available experimental information—for not only the wild-type channels but also for channels formed by amino acid-substituted gramicidin analogues—gramicidin channels continue to provide important insights into the microphysics of ion permeation through bilayer-spanning channels. For similar reasons, gramicidin channels constitute a system of choice for evaluating computational strategies for obtaining mechanistic insights into ion permeation through the more complex channels formed by integral membrane proteins.

  2. Thermohydraulic assessment of the RP-10 reactor core to determine the maximum power

    International Nuclear Information System (INIS)

    Thermohydraulic parameters assessment of the RP-10 reactor core from the most thermally demanded (hot channel). Determination of the operation thermal maximum power considering security margins and statistical treatment of uncertainty factors

  3. Multi-Channel Retailing

    Directory of Open Access Journals (Sweden)

    Dirk Morschett, Dr.,

    2005-01-01

    Full Text Available Multi-channel retailing entails the parallel use by retailing enterprises of several sales channels. The results of an online buyer survey which has been conducted to investigate the impact of multi-channel retailing (i.e. the use of several retail channels by one retail company on consumer behaviour show that the frequently expressed concern that the application of multi-channel systems in retailing would be associated with cannibalization effects, has proven unfounded. Indeed, the appropriate degree of similarity, consistency, integration and agreement achieves the exact opposite. Different channels create different advantages for consumers. Therefore the total benefit an enterprise which has a multi-channel system can offer to its consumers is larger, the greater the number of available channels. The use of multi-channel systems is associated with additional purchases in the different channels. Such systems are thus superior to those offering only one sales channel to their customers. Furthermore, multi-channel systems with integrated channels are superior to those in which the channels are essentially autonomous and independent of one another. In integrated systems, consumers can achieve synergy effects in the use of sales-channel systems. Accordingly, when appropriately formulated, multi-channel systems in retailing impact positively on consumers. They use the channels more frequently, buy more from them and there is a positive customer-loyalty impact. Multi-channel systems are strategic options for achieving customer loyalty, exploiting customer potential and for winning new customers. They are thus well suited for approaching differing and varied target groups.

  4. Nuclear reactor repairing device

    International Nuclear Information System (INIS)

    Purpose: To enable free repairing of an arbitrary position in an LMFBR reactor. Constitution: A laser light emitted from a laser oscillator installed out of a nuclear reactor is guided into a portion to be repaired in the reactor by using a reflecting mirror, thereby welding or cutting it. The guidance of the laser out of the reactor into the reactor is performed by an extension tube depending into a through hole of a rotary plug, and the guidance of the laser light into a portion to be repaired is performed by the transmitting and condensing action of the reflecting mirror. (Kamimura, M.)

  5. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  6. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  7. Spinning fluids reactor

    Science.gov (United States)

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

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

    International Nuclear Information System (INIS)

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

  9. Present status and future prospects of research reactors in the Soviet Union

    International Nuclear Information System (INIS)

    The research reactors which are currently in use in the USSR are employed in a wide range of research in various scientific fields, as well as for certain applied tasks. Most of these reactors are pool-type reactors. Since it is significantly cheaper to upgrade research reactors rather than to build new ones, the vast majority of them have been upgraded and their experimental capabilities significantly expanded. In the USSR the future of research reactors lies in the continued modernization of currently operating research reactors and the building of new powerful research reactors for which designs are being developed. Some are already under construction (for example, the PIK reactor). These designs are developing Soviet research reactor concepts which centre around pressure-vessel-type reactors and channel-type reactors in tanks. Other technical ideas are also being used. Research reactor safety meets current requirements on the whole; however, their long operating life, their proximity to heavily populated areas, and several other features of research reactors make safety a higher priority. A series of organizational and technical measures are being undertaken to improve research reactor safety

  10. Research nuclear reactor start-up simulator

    International Nuclear Information System (INIS)

    This work presents the design and FPGA implementation of a research nuclear reactor start-up simulator. Its aim is to generate a set of signals that allow replacing the neutron detector for stimulated signals, to feed the measurement electronic of the start-up channels, to check its operation, together with the start-up security logic. The simulator presented can be configured on three independent channels and adjust the shape of the output pulses. Furthermore, each channel can be configured in 'rate' mode, where you can specify the growth rate of the pulse frequency in %/s. Result and details of the implementation on FPGA of the different functional blocks are given. (author)

  11. Gas/liquid separator for BWR type reactor

    International Nuclear Information System (INIS)

    A two phase gas/liquid flow generated at a heating portion of a nuclear reactor is swirled by inlet vanes. The phase gas/liquid flow uprises as a vortex flow in a vortex cylinder, and a liquid phase of a high density gathers at the outer circumference of the vortex cylinder. The liquid phase gathered at the outer circumference is collected at the inlet of a discharge flow channel which protrude into the vortex cylinder and in a three-step structure, and introduced into a recycling liquid phase passing through the discharge flow channel for liquid phase. There is provided a structure that separated liquid collected at the lowermost state in the inlet of the three-step discharge flow channel inlet descends in the discharge flow channel, then uprises in an uprising flow channel and is introduced into the recycling liquid phase by way of a discharge flow channel exit. The height of the discharge flow channel exit is determined equal to that of a liquid level of the recycling liquid phase during rated operation of the reactor. Accordingly, even in a case where the liquid level in the recycling liquid phase is lowered, the liquid level of the uprising flow channel is kept equal to that during rated operation. (I.N.)

  12. NEUTRONIC REACTOR HAVING LOCALIZED AREAS OF HIGH THERMAL NEUTRON DENSITIES

    Science.gov (United States)

    Newson, H.W.

    1958-06-01

    A nuclear reactor for the irradiation of materials designed to provide a localized area of high thermal neutron flux density in which the materials to be irradiated are inserted is described. The active portion of the reactor is comprised of a cubicle graphite moderator of about 25 feet in length along each axis which has a plurality of cylindrical channels for accommodatirg elongated tubular-shaped fuel elements. The fuel elements have radial fins for spacing the fuel elements from the channel walls, thereby providing spaces through which a coolant may be passed, and also to serve as a heatconductirg means. Ducts for accommnodating the sample material to be irradiated extend through the moderator material perpendicular to and between parallel rows of fuel channels. The improvement is in the provision of additional fuel element channels spaced midway between 2 rows of the regular fuel channels in the localized area surrounding the duct where the high thermal neutron flux density is desired. The fuel elements normally disposed in the channels directly adjacent the duct are placed in the additional channels, and the channels directly adjacent the duct are plugged with moderator material. This design provides localized areas of high thermal neutron flux density without the necessity of providing additional fuel material.

  13. Preliminary hazards review overboring Hanford reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nilson, R.; Carlson, P.A.

    1962-07-25

    The General Electric Company, as prime contractor to the AEC at Hanford, is proposing to modify the lattice characteristics of the 8 3/8-inch lattice reactors for the purposes of improving the conversion ratio of these reactors. The proposed overbore modification of the reactors would remove the existing aluminum process tubes, enlarge the diameters of the graphite channels by about one-half inch, insert smooth-bore Zircaloy-2 process tubes and refuel the reactor with larger size, self-supported fuel elements. The overbore fuel will remain the internally-and-externally-cooled cylindrical type, but the weight per foot will be about twice that of the present fuel element. The removal of the inlet and outlet piping connections which would be required in the overboring process will permit the replacement of the existing fittings with ones of improved design. Furthermore, new orifices and venturis which are compatible with the hydraulic characteristics of the overbore tube and fuel geometry and the pumping system will be installed. No basic changes are proposed in the pumping system though the reactor flaw rate may be increased 5--10 percent by changes in hydraulic characteristics depending on the water plant flow capacity.

  14. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  15. Calcium channel blocker overdose

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/002580.htm Calcium channel blocker overdose To use the sharing features on this page, please enable JavaScript. Calcium channel blockers are a type of medicine used ...

  16. The influence of fuel assembly characteristics on reactor safety

    International Nuclear Information System (INIS)

    To improve fuel utilization and nuclear plant economy, most nuclear plants of China adopt increased fuel enrichment and long cycle analysis. Core power distribution will be worse with these advanced items. Radial and axial peak increase too. This is a challenge to reactor safety. Since the fuel assembly is the most important part of a reactor core, fuel assembly characteristics affect reactor safety a lot. A few aspects of influence on reactor safety are discussed in this paper as a reference for fuel assembly design. A better fuel assembly design can increase heat exchange ability, especially in cold wall cells. The grids nearby core outlet can efficiently mix the flow of hot channel and average channel to decrease DNBR. In safety analysis, we always suppose the center of center assembly is the hot channel, but sometimes based on actual power distribution the hot channel occurs at side cell or corner cell. So the distribution of grids pressure drop coefficients can affect the minimum DNBR. A better fuel assembly design can help to spread core power distribution, decrease radial and axial peak efficiently. To spread core power distribution, different neutronic poisons are added into fuel pellet by different ways, and then the relative effects on reactor safety are different. At the same time, better fuel assembly design should leave enough margins for reactor safety to handle high burnup condition and so on. Fuel pellet and clad capabilities are getting worse versus increasing fuel burnup. This is a challenge to reactor safety, so more attentions should be paid to fuel burnup characteristics. (author)

  17. Dynamic channel allocation

    OpenAIRE

    Kaminsky, Andrew D.

    2003-01-01

    Approved for public release; distribution in unlimited. Dynamic Channel Allocation (DCA) offers the possibility of capturing unused channel capacity by allocating unused resources between competing network nodes. This can reduce or possibly eliminate channels sitting idle while information awaits transmission. This holds potential for increasing throughput on bandwidth constrained networks. The purpose of this thesis is to examine the techniques used to allocate channels on demand and acc...

  18. Investigation of techniques for the application of safeguards to a continuously fuelled reactor

    International Nuclear Information System (INIS)

    The following three areas of prime safeguards concern are analyzed: 1) at the reactor face, where spent fuel is first removed from the channel, 2) along the system for transporting spent fuel from the reactor to the storage bay, and 3) in the spent fuel storage bay. The program is designed to develop techniques and instruments to safeguard these areas

  19. Neutronic of heterogenous gas cooled reactors

    International Nuclear Information System (INIS)

    At present, one of the main technical features of the advanced gas cooled reactor under development is its fuel element concept, which implies a neutronic homogeneous design, thus requiring higher enrichment compared with present commercial nuclear power plants.In this work a neutronic heterogeneous gas cooled reactor design is analyzed by studying the neutronic design of the Advanced Gas cooled Reactor (AGR), a low enrichment, gas cooled and graphite moderated nuclear power plant.A search of merit figures (some neutronic parameter, characteristic dimension, or a mixture of both) which are important and have been optimized during the reactor design stage is been done, to aim to comprise how a gas heterogeneous reactor is been design, given that semi-infinity arrangement criteria of rods in LWRs and clusters in HWRs can t be applied for a solid moderator and a gas refrigerator.The WIMS code for neutronic cell calculations is been utilized to model the AGR fuel cell and to calculate neutronic parameters such as the multiplication factor and the pick factor, as function of the fuel burnup.Also calculation is been done for various nucleus characteristic dimensions values (fuel pin radius, fuel channel pitch) and neutronic parameters (such as fuel enrichment), around the design established parameters values.A fuel cycle cost analysis is carried out according to the reactor in study, and the enrichment effect over it is been studied.Finally, a thermal stability analysis is been done, in subcritical condition and at power level, to study this reactor characteristic reactivity coefficients.Present results shows (considering the approximation used) a first set of neutronic design figures of merit consistent with the AGR design.

  20. Liquid-cooled nuclear reactor, especially a boiling water reactor

    International Nuclear Information System (INIS)

    A nuclear reactor with a special arrangement of fuel rods in the core is designed. Each fuel element has its shaft which is made of sheets, has the same cross section as the fuel element and protrudes at least the length of the control rod above the reactor core. Made of a zirconium alloy in the core area and of stainless steel above it, the shaft is equipped with channels for sliding the rods in and out and serves to spatially secure the position of the rods. Coolant flow is provided by the chimney effect. The shaft can conveniently enclose the control rod drive. It can also serve to bear the water separator. Moreover, it can constitute a part of the casing which surrounds the fuel rods and keeps the fuel in an intimate contact with the coolant; the other part of this casing is constituted by inserted sheets which can conveniently have the shape of angles. The walls of neighboring shafts form a compartment accommodating a neutron absorber plate. (M.D.). 11 figs

  1. Microstructured reactors for hydrogen production

    Energy Technology Data Exchange (ETDEWEB)

    Aartun, Ingrid

    2005-07-01

    Small scale hydrogen production by partial oxidation (POX) and oxidative steam reforming (OSR) have been studied over Rh-impregnated microchannel Fecralloy reactors and alumina foams. Trying to establish whether metallic microchannel reactors have special advantages for hydrogen production via catalytic POX or OSR with respect to activity, selectivity and stability was of special interest. The microchannel Fecralloy reactors were oxidised at 1000 deg C to form a {alpha}-Al2O3 layer in the channels in order to enhance the surface area prior to impregnation. Kr-BET measurements showed that the specific surface area after oxidation was approximately 10 times higher than the calculated geometric surface area. Approximately 1 mg Rh was deposited in the channels by impregnation with an aqueous solution of RhCl3. Annular pieces (15 mm o.d.,4 mm i.d., 14 mm length) of extruded {alpha}-Al2O3 foams were impregnated with aqueous solutions of Rh(NO3)3 to obtain 0.01, 0.05 and 0.1 wt.% loadings, as predicted by solution uptake. ICP-AES analyses showed that the actual Rh loadings probably were higher, 0.025, 0.077 and 0.169 wt.% respectively. One of the microchannel Fecralloy reactors and all Al2O3 foams were equipped with a channel to allow for temperature measurement inside the catalytic system. Temperature profiles obtained along the reactor axes show that the metallic microchannel reactor is able to minimize temperature gradients as compared to the alumina foams. At sufficiently high furnace temperature, the gas phase in front of the Rh/Al2O3/Frecralloy microchannel reactor and the 0.025 wt.% Rh/Al2O3 foams ignites. Gas phase ignition leads to lower syngas selectivity and higher selectivity to total oxidation products and hydrocarbon by-products. Before ignition of the gas phase the hydrogen selectivity is increased in OSR as compared to POX, the main contribution being the water-gas shift reaction. After gas phase ignition, increased formation of hydrocarbon by

  2. SIMULATE-3K: Enhancements and Application to Boiling Water Reactor Transients

    International Nuclear Information System (INIS)

    The SIMULATE-3K (S-3K) reactor analysis code has been applied to a variety of pressurized water reactor (PWR) and boiling water reactor (BWR) transients since 1993. Over the years, many changes have occurred in the S-3K channel hydraulics and ex-core component modeling. This paper summarizes those changes and outlines the status of existing vessel and steam line models. Examples are given for BWR transients that can be analyzed with S-3K

  3. Desynched channels on IRCnet

    CERN Document Server

    Hansen, Michael

    2008-01-01

    In this paper we describe what a desynchronised channel on IRC is. We give procedures on how to create such a channel and how to remove desynchronisation. We explain which types of desynchronisation there are, what properties desynchronised channels have, and which properties can be exploited.

  4. Quantum Binary Symmetric Channels

    Institute of Scientific and Technical Information of China (English)

    陈小余; 仇佩亮

    2001-01-01

    Quantum binary symmetric channels are defined via the invariance of fidelity under unitary transformations ofthe input density operators. In this definition, they not only include the most studied case of the depolarizingchannel but also other channels. We investigate the character of the latter and find the maximum of the coherentinformation to estimate the capacities of the channels.

  5. Quantum Multiple Access Channel

    Institute of Scientific and Technical Information of China (English)

    侯广; 黄民信; 张永德

    2002-01-01

    We consider the transmission of classical information over a quantum channel by many senders, which is a generalization of the two-sender case. The channel capacity region is shown to be a convex hull bound by the yon Neumann entropy and the conditional yon Neumann entropies. The result allows a reasonable distribution of channel capacity over the senders.

  6. Cardiac potassium channel subtypes

    DEFF Research Database (Denmark)

    Schmitt, Nicole; Grunnet, Morten; Olesen, Søren-Peter

    2014-01-01

    About 10 distinct potassium channels in the heart are involved in shaping the action potential. Some of the K(+) channels are primarily responsible for early repolarization, whereas others drive late repolarization and still others are open throughout the cardiac cycle. Three main K(+) channels...

  7. KV7 potassium channels

    DEFF Research Database (Denmark)

    Stott, Jennifer B; Jepps, Thomas Andrew; Greenwood, Iain A

    2014-01-01

    Potassium channels are key regulators of smooth muscle tone, with increases in activity resulting in hyperpolarisation of the cell membrane, which acts to oppose vasoconstriction. Several potassium channels exist within smooth muscle, but the KV7 family of voltage-gated potassium channels have been...

  8. In-service inspection and testing of TAPS square fuel channels

    International Nuclear Information System (INIS)

    Tarapur Atomic Power Station is a twin unit Boiling Water Reactor. The initial each unit design was for 210 MWe. Subsequently due to Secondary Steam Generator tube leak problem, the units were de-rated to 160 MWe in the year 1985-86. Since then each unit is operating at 160 MWe. The station has completed 32 years of successful commercial operation. Presently each reactor is re-rated to 530 MWth. There are 284 fuel assemblies in each reactor. Each fuel assembly utilizes fuel channel made of Zircaloy-4 material. The fuel channel is a square tube and it surrounds the fuel bundle. The channel is secured to the fuel bundle by means of channel fastener assembly. The fuel channel length is 158.625 and wall thickness is 0.060. The fuel channel directs the coolant flow to fuel rods. It is also used as a guide for control blade movement inside the core. It provides for the structural stability of the fuel assembly. Initially during fabrication of the fuel channel, care is taken to control its dimensions very stringently by following quality assurance plan. However, once the channels are loaded into core along with fuel bundle, it undergoes dimensional changes due to neutron exposure. The fuel channels are monitored for its dimensions due to the neutron exposure and taken out of core at appropriate time. The paper prescribes the methodology adopted for inspecting the channels and the findings of the inspection. (author)

  9. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  10. Multipurpose research reactors

    International Nuclear Information System (INIS)

    The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  11. On 1-qubit channels

    OpenAIRE

    Uhlmann, Armin

    2000-01-01

    The entropy H_T(rho) of a state rho with respect to a channel T and the Holevo capacity of the channel require the solution of difficult variational problems. For a class of 1-qubit channels, which contains all the extremal ones, the problem can be significantly simplified by associating an Hermitian antilinear operator theta to every channel of the considered class. The concurrence of the channel can be expressed by theta and turns out to be a flat roof. This allows to write down an explicit...

  12. On 1-qubit channels

    CERN Document Server

    Uhlmann, A

    2001-01-01

    The entropy H_T(rho) of a state rho with respect to a channel T and the Holevo capacity of the channel require the solution of difficult variational problems. For a class of 1-qubit channels, which contains all the extremal ones, the problem can be significantly simplified by associating an Hermitian antilinear operator theta to every channel of the considered class. The concurrence of the channel can be expressed by theta and turns out to be a flat roof. This allows to write down an explicit expression for H_T. Its maximum would give the Holevo (1-shot) capacity.

  13. Reactor BR2

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2000-07-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

  14. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

  15. Source and Channel Coding for Correlated Sources Over Multiuser Channels

    OpenAIRE

    Gunduz, Deniz; Erkip, Elza; Goldsmith, Andrea; Poor, H. Vincent

    2008-01-01

    Source and channel coding over multiuser channels in which receivers have access to correlated source side information is considered. For several multiuser channel models necessary and sufficient conditions for optimal separation of the source and channel codes are obtained. In particular, the multiple access channel, the compound multiple access channel, the interference channel and the two-way channel with correlated sources and correlated receiver side information are considered, and the o...

  16. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

  17. VIPRE modeling of VVER-1000 reactor core for DNB analyses

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Y.; Nguyen, Q. [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Cizek, J. [Nuclear Research Institute, Prague, (Czech Republic)

    1995-09-01

    Based on the one-pass modeling approach, the hot channels and the VVER-1000 reactor core can be modeled in 30 channels for DNB analyses using the VIPRE-01/MOD02 (VIPRE) code (VIPRE is owned by Electric Power Research Institute, Palo Alto, California). The VIPRE one-pass model does not compromise any accuracy in the hot channel local fluid conditions. Extensive qualifications include sensitivity studies of radial noding and crossflow parameters and comparisons with the results from THINC and CALOPEA subchannel codes. The qualifications confirm that the VIPRE code with the Westinghouse modeling method provides good computational performance and accuracy for VVER-1000 DNB analyses.

  18. C. elegans TRP channels.

    Science.gov (United States)

    Xiao, Rui; Xu, X Z Shawn

    2011-01-01

    Transient receptor potential (TRP) channels represent a superfamily of cation channels found in all eukaryotes. The C. elegans genome encodes seventeen TRP channels covering all of the seven TRP subfamilies. Genetic analyses in C. elegans have implicated TRP channels in a wide spectrum of behavioral and physiological processes, ranging from sensory transduction (e.g. chemosensation, touch sensation, proprioception and osmosensation) to fertilization, drug dependence, organelle biogenesis, apoptosis, gene expression, and neurotransmitter/hormone release. Many C. elegans TRP channels share similar activation and regulatory mechanisms with their vertebrate counterparts. Studies in C. elegans have also revealed some previously unrecognized functions and regulatory mechanisms of TRP channels. C. elegans represents an excellent genetic model organism for the study of function and regulation of TRP channels in vivo. PMID:21290304

  19. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru

    1997-04-04

    An LMFBR type reactor comprises a plurality of reactor cores in a reactor container. Namely, a plurality of pot containing vessels are disposed in the reactor vessel and a plurality of reactor cores are formed in a state where an integrated-type fuel assembly is each inserted to a pot, and a coolant pipeline is connected to each of the pot containing-vessel to cool the reactor core respectively. When fuels are exchanged, the integrated-type fuel assembly is taken out together with the pot from the reactor vessel in a state where the integrated-type fuel assembly is immersed in the coolants in the pot as it is. Accordingly, coolants are supplied to each of the pot containing-vessel connected with the coolant pipeline and circulate while cooling the integrated-type fuel assembly for every pot. Then, when the fuels are exchanged, the integrated type fuel assembly is taken out to the outside of the reactor together with the pot by taking up the pot from the pot-containing vessel. Then, neutron economy is improved to thereby improve reactor power and the breeding ratio. (N.H.)

  20. PROCESS INTENSIFICATION: MICROWAVE INITIATED REACTIONS USING A CONTINUOUS FLOW REACTOR

    Science.gov (United States)

    The concept of process intensification has been used to develop a continuous narrow channel reactor at Clarkson capable of carrying out reactions under isothermal conditions whilst being exposed to microwave (MW) irradiation thereby providing information on the true effect of mi...

  1. Reliability Analysis of I and C Architecture of Research Reactors Using Bayesian Networks

    International Nuclear Information System (INIS)

    The objective of this research project is to identify a configuration of architecture which gives highest availability with maintaining low cost of manufacturing. In this regard, two configurations of a single channel of RPS are formulated in the current article and BN models were constructed. Bayesian network analysis was performed to find the reliability features. This is a continuation of study towards the standardization of I and C architecture for low and medium power research reactors. This research is the continuation of study to analyze the reliability of single channel of Reactor Protection System (RPS) using Bayesian networks. The focus of research was on the development of architecture for low power research reactors. What level of reliability is sufficient for protection, safety and control systems in case of low power research reactors? There should be a level which should satisfy all the regulatory requirements as well as operational demands with optimized cost of construction. Scholars, researchers and material investigators from educational and research institutes are demanding for construction of more research reactors. In order to meet this demand and construct more units, it is necessary to do more research in various areas. The research is also needed to make a standardization of research reactor I and C architectures on the same lines of commercial power plants. The research reactors are categorized into two broad categories, Low power research reactors and medium to high power research reactors. According to IAEA TECDOC-1234, Research reactors with 0.250-2.0 MW power rating or 2.5-10 Χ 1011 n/cm2.s. flux are termed low power reactor whereas research reactors ranging from 2-10 MW power rating or 0.1-10 Χ 1013 n/cm2.s. are considered as Medium to High power research reactors. Some other standards (IAEA NP-T-5.1) define multipurpose research reactor ranging from power few hundred KW to 10 MW as low power research reactor

  2. AECL experience in fuel channel inspection

    International Nuclear Information System (INIS)

    Inspection of CANDU fuel channels (FC) is performed to ensure safe and economic reactor operation. CANDU reactor FCs have features that make them a unique non-destructive testing (NDT) challenge. The thin, 4 mm pressure-tube wall means flaws down to about 0.1 mm deep must be reliably detected and characterized. This is one to two orders of magnitude smaller than is usually considered of significant concern for steel piping and pressure vessels. A second unique feature is that inspection sensors must operate in the reactor core--often within 20 cm of highly radioactive fuel. Work on inspection of CANDU reactor FCs at AECL dates back over three decades. In that time, AECL staff have provided equipment and conducted or supervised in-service inspections in about 250 FCs, in addition to over 8000 pre-service FCs. These inspections took place at every existing CANDU reactor except those in India and Romania. Early FC inspections focussed on measurement of changes in dimensions (gauging) resulting from exposure to a combination of neutrons, stress and elevated temperature. Expansion of inspection activities to include volumetric inspection (for flaws) started in the mid-1970s with the discovery of delayed hydride cracking in Pickering 3 and 4 rolled joints. Recognition of other types of flaw mechanisms in the 1980s led to further expansion in both pre-service and in-service inspections. These growing requirements, to meet regulatory as well as economic needs, led to the development of a wide spectrum of inspection technology that now includes tests for hydrogen concentration, structural integrity of core components, flaws, and dimensional change. This paper reviews current CANDU reactor FC inspection requirements. The equipment and techniques developed to satisfy these requirements are also described. The paper concludes with a discussion of work in progress in AECL aimed at providing state-of-the-art FC inspection services. (author)

  3. Investigation of neutron distribution in training reactor VR-1

    International Nuclear Information System (INIS)

    The VR-1 training reactor is a pool-type light-water reactor with the low-enriched uranium and maximum thermal power of 1 kW. The reactor is mainly used for students' training. The training is aimed to areas such as the reactor physics, neutronics, dosimetry, nuclear safety and I and C systems. Since neutron flux in the VR-1 core is well measured, this work focuses on one part of the reactor - its Radial experimental Channel (RC). This paper deals with the measurement of the neutron distribution by means of gold-foil neutron-activation technique and continual measurement with 3He-filled detector. Obtained experimental results were verified with the simulation in the Monte-Carlo N-Particle Transport Code. Results and conclusions from this paper will be used for further investigation of neutrons and their spatial distribution inside the low-power training reactor. Also, the data obtained in this paper can be used as a basis for future detailed measurements of neutron flux and its distribution in other hard accessible areas inside the reactor. The paper gives a simple theoretical introduction concerning neutron measurement procedures and available techniques in this field, which is particularly important for improving training courses and a content of offered experiments in the VR-1 reactor. (author)

  4. Compound Wiretap Channels

    Directory of Open Access Journals (Sweden)

    Kramer Gerhard

    2009-01-01

    Full Text Available Abstract This paper considers the compound wiretap channel, which generalizes Wyner's wiretap model to allow the channels to the (legitimate receiver and to the eavesdropper to take a number of possible states. No matter which states occur, the transmitter guarantees that the receiver decodes its message and that the eavesdropper is kept in full ignorance about the message. The compound wiretap channel can also be viewed as a multicast channel with multiple eavesdroppers, in which the transmitter sends information to all receivers and keeps the information secret from all eavesdroppers. For the discrete memoryless channel, lower and upper bounds on the secrecy capacity are derived. The secrecy capacity is established for the degraded channel and the semideterministic channel with one receiver. The parallel Gaussian channel is further studied. The secrecy capacity and the secrecy degree of freedom ( are derived for the degraded case with one receiver. Schemes to achieve the for the case with two receivers and two eavesdroppers are constructed to demonstrate the necessity of a prefix channel in encoder design. Finally, the multi-antenna (i.e., MIMO compound wiretap channel is studied. The secrecy capacity is established for the degraded case and an achievable is given for the general case.

  5. Compound Wiretap Channels

    Directory of Open Access Journals (Sweden)

    Shlomo Shamai (Shitz

    2009-01-01

    Full Text Available This paper considers the compound wiretap channel, which generalizes Wyner's wiretap model to allow the channels to the (legitimate receiver and to the eavesdropper to take a number of possible states. No matter which states occur, the transmitter guarantees that the receiver decodes its message and that the eavesdropper is kept in full ignorance about the message. The compound wiretap channel can also be viewed as a multicast channel with multiple eavesdroppers, in which the transmitter sends information to all receivers and keeps the information secret from all eavesdroppers. For the discrete memoryless channel, lower and upper bounds on the secrecy capacity are derived. The secrecy capacity is established for the degraded channel and the semideterministic channel with one receiver. The parallel Gaussian channel is further studied. The secrecy capacity and the secrecy degree of freedom (s.d.o.f. are derived for the degraded case with one receiver. Schemes to achieve the s.d.o.f. for the case with two receivers and two eavesdroppers are constructed to demonstrate the necessity of a prefix channel in encoder design. Finally, the multi-antenna (i.e., MIMO compound wiretap channel is studied. The secrecy capacity is established for the degraded case and an achievable s.d.o.f. is given for the general case.

  6. Refurbishment of the safety system at the CROCUS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Girardin, Gaetan; Frajtag, Pavel; Braun, Laurent; Pautz, Andreas [Ecole Polytechnique Federale de Lausanne, Lausanne (Switzerland)

    2013-07-01

    This report discusses the partial refurbishment of the first channel (VS-I) of the Reactor Protection System (RPS) at the teaching reactor CROCUS operated at the Swiss Federal Institute of Technology (EPFL) in Lausanne. The CROCUS facility is a zero-power reactor and it is mainly used for educational purposes for undergraduate and master students. The RPS uses two fully redundant and independent channels: VS-I and VS-II. These contain both the nuclear instrumentation and control units that were developed in-house during the reactor commissioning in the 80's. The nuclear instrumentation and control used was provided by Merlin-Gerin for flux measurements and the reactor SCRAM function. The neutron flux is measured by means of fission chambers connected to IS-I and IS-II. The reactor can be in different states, in particular the startup phases, for example the progressive auxiliary and reactor tanks water filling phase, the safety rods pull-up phase, etc. The logic functions corresponding to these states are designed and implemented in SS-I and SS-II. The refurbishment of the reactor SS-I and SS-II was necessary due to the lack of spare parts for some circuits and the difficulty of finding simple logic circuits in the market. The replacement of both safety channels SS-I and SS-II was performed with the resources available in-house at the reactor service laboratory at EPFL. The nuclear instrumentation is not directly impacted by the reported refurbishment activity. The first phase of the refurbishment project consists of the replacement of the first channel (VS-I) keeping the reactor available for operation services at EPFL. The paper focusses on the description of this technical project and the review and approval process conducted by the Swiss Federal Nuclear Inspectorate (ENSI). Details are provided concerning each regulatory phase of the project and also the technological choices (CPLD over TTL) for the newly developed system. The latter were specifically made

  7. Analysis of tritium production in TRIGA Mark II reactor at JSI for the needs of fusion research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jazbec, Anze; Zerovnik, Gasper; Snoj, Luka; Trkov, Andrej [Jozef Stefan Institute, Ljubljana (Slovenia)

    2013-12-15

    In future, electricity could be produced in fusion power plants. One of the steps towards development of fusion power plants is the construction of an experimental fusion reactor ITER where deuterium (D) and tritium (T) will be fused and energy will be released. As natural concentrations of T are extremely low, the T as fusion fuel will have to be produced artificially. A series of calculations were made to investigate the possibility of producing small quantities of T for experimental fusion reactors such as JET and ITER in a small research reactor like the TRIGA Mark II reactor at the Jozef Stefan Institute (JSI). The T production is the largest if all irradiation channels in reactor's reflector are filled with LiAlO{sub 2} samples. When samples are inserted, the excess reactivity decreases by around 200 pcm. In the second part of the work an estimate was made of how long the reactor can operate with current fuel supplies. Calculations were made with the TRIGLAV computer code. TRIGA can operate at full power for at least 2,860 days, during which 152 mg of T could be produced. We conclude that small TRIGA reactors can not produce any significant quantities of T for the needs of the future experimental fusion reactors. (orig.)

  8. Analysis of tritium production in TRIGA Mark II reactor at JSI for the needs of fusion research reactors

    International Nuclear Information System (INIS)

    In future, electricity could be produced in fusion power plants. One of the steps towards development of fusion power plants is the construction of an experimental fusion reactor ITER where deuterium (D) and tritium (T) will be fused and energy will be released. As natural concentrations of T are extremely low, the T as fusion fuel will have to be produced artificially. A series of calculations were made to investigate the possibility of producing small quantities of T for experimental fusion reactors such as JET and ITER in a small research reactor like the TRIGA Mark II reactor at the Jozef Stefan Institute (JSI). The T production is the largest if all irradiation channels in reactor's reflector are filled with LiAlO2 samples. When samples are inserted, the excess reactivity decreases by around 200 pcm. In the second part of the work an estimate was made of how long the reactor can operate with current fuel supplies. Calculations were made with the TRIGLAV computer code. TRIGA can operate at full power for at least 2,860 days, during which 152 mg of T could be produced. We conclude that small TRIGA reactors can not produce any significant quantities of T for the needs of the future experimental fusion reactors. (orig.)

  9. Operation and maintenance of the RA reactor in 1965

    International Nuclear Information System (INIS)

    It has been planned for 1965 that the RA reactor would be operated each month for 20 days at nominal power of 6.5 MW, at lower power for 5 days, meaning production of 27 400 MWh. The plan was fulfilled since reactor produced 28809 MWh, i.e. 5% more than planned. Reactor was used for irradiation in the vertical experimental channels according to the demand of 1264 users from the Institute and 191 external users. Two groups of experiments done: at nominal power simultaneously with isotope production and experiments which demanded particular power levels and temperatures. Three fuel exchanges were done during this year, meaning that 40 fuel channels were changed in total. Vertical experimental channels VEK-1 and VEK-9 having diameter 100 mm were changed by channels having diameter 50 mm and shortened by 435 mm. Channel VEK-5 with diameter 110 mm was changed shortened by 430 mm. This enabled better fuel economy, the burnup was increased from 4500 MWd/t to 5000 MWd/t. This report contains the action plan for 1966

  10. One piece reactor removal

    International Nuclear Information System (INIS)

    Japan Research Reactor No.3 (JRR-3) was the first reactor consisting of 'Japanese-made' components alone except for fuel and heavy water. After reaching its initial critical state in September 1962, JRR-3 had been in operation for 21 years until March 1983. It was decided that the reactor be removed en-bloc in view of the work schedule, cost and management of the reactor following the removal. In the special method developed jointly by the Japanese Atomic Energy Research Institute and Shimizu Construction Co., Ltd., the reactor main unit was cut off from the building by continuous core boring, with its major components bound in the block with biological shield material (heavy concrete), and then conveyed and stored in a large waste store building constructed near the reactor building. Major work processes described in this report include the cutting off, lifting, horizontal conveyance and lowering of the reactor main unit. The removal of the JRR-3 reactor main unit was successfully carried out safely and quickly by the en-block removal method with radiation exposure dose of the workers being kept at a minimum. Thus the high performance of the en-bloc removal method was demonstrated and, in addition, valuable knowhow and other data were obtained from the work. (Nogami, K.)

  11. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  12. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  13. Channel capacity and error exponents of variable rate adaptive channel coding for Rayleigh fading channels

    OpenAIRE

    Lau, KN

    1999-01-01

    We have evaluated the information theoretical performance of variable rate adaptive channel coding for Rayleigh fading channels. The channel states are detected at the receiver and fed back to the transmitter by means of a noiseless feedback link. Based on the channel state informations, the transmitter can adjust the channel coding scheme accordingly. Coherent channel and arbitrary channel symbols with a fixed average transmitted power constraint are assumed. The channel capacity and the err...

  14. Quantum broadcast channels

    CERN Document Server

    Yard, J; Devetak, I; Yard, Jon; Hayden, Patrick; Devetak, Igor

    2006-01-01

    We analyze quantum broadcast channels, which are quantum channels with a single sender and many receivers. Focusing on channels with two receivers for simplicity, we generalize a number of results from the network Shannon theory literature which give the rates at which two senders can receive a common message, while a personalized one is sent to one of them. Our first collection of results applies to channels with a classical input and quantum outputs. The second class of theorems we prove concern sending a common classical message over a quantum broadcast channel, while sending quantum information to one of the receivers. The third group of results we obtain concern communication over an isometry, giving the rates at quantum information can be sent to one receiver, while common quantum information is sent to both, in the sense that tripartite GHZ entanglement is established. For each scenario, we provide an additivity proof for an appropriate class of channels, yielding single-letter characterizations of the...

  15. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 1014 n/cm2/sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  16. TRIGA research reactors

    International Nuclear Information System (INIS)

    TRIGA (Training, Research, Isotope production, General-Atomic) has become the most used research reactor in the world with 65 units operating in 24 countries. The original patent for TRIGA reactors was registered in 1958. The success of this reactor is due to its inherent level of safety that results from a prompt negative temperature coefficient. Most of the neutron moderation occurs in the nuclear fuel (UZrH) because of the presence of hydrogen atoms, so in case of an increase of fuel temperature, the neutron spectrum becomes harder and neutrons are less likely to fission uranium nuclei and as a consequence the power released decreases. This inherent level of safety has made this reactor fit for training tool in university laboratories. Some recent versions of TRIGA reactors have been designed for medicine and industrial isotope production, for neutron therapy of cancers and for providing a neutron source. (A.C.)

  17. Mirror reactor surface study

    International Nuclear Information System (INIS)

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included

  18. Iris reactor conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Carelli, M.D.; Conway, L.E.; Petrovic, B.; Paramonov, D.V. [Westinghouse Electric Comp., Pittsburgh, PA (United States); Galvin, M.; Todreas, N.E. [Massachusetts Inst. of Tech., Cambridge, MA (United States); Lombardi, C.V.; Maldari, F.; Ricotti, M.E. [Politecnico di Milano, Milan (Italy); Cinotti, L. [Ansaldo SpA, Genoa (Italy)

    2001-07-01

    IRIS (International Reactor Innovative and Secure) is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., proliferation resistance, enhanced safety, improved economics and fuel cycle sustainability. It relies on the proven technology of light water reactors and features innovative engineering, but it does not require new technology development. This paper discusses the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and five-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. (author)

  19. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  20. Compact torsatron reactors

    International Nuclear Information System (INIS)

    Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

  1. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  2. HIPPI and Fibre Channel

    International Nuclear Information System (INIS)

    The High-Performance Parallel Interface (HIPPI) and Fibre Channel are near-gigabit per second data communications interfaces being developed in ANSI standards Task Group X3T9.3. HIPPI is the current interface of choice in the high-end and supercomputer arena, and Fibre Channel is a follow-on effort. HIPPI came from a local area network background, and Fibre Channel came from a mainframe to peripheral interface background

  3. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

  4. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1983 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  5. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

  6. Coupled Transport Phenomena in Corrugated Photocatalytic Reactors

    Institute of Scientific and Technical Information of China (English)

    Adam A. Donaldson; ZHANG Zisheng

    2011-01-01

    Corrugated reactors are known for their use in applications requiring UV-exposure, whereby media flowing within the corrugated channel react with a photo-active catalyst impregnated on the surface (i.e. TiO2). The performance in these systems is dependent on catalyst properties and reactivity for a given light source, in conjunc-tion with the coupled transport of reactants within the media and photons falling incident to the catalyst surface. Experimental and computational analyses of local mass transfer and radiation pattems for a broad range of corrugation angles, depths, and non-idealities introduced during manufacture (i.e. fold curvature) are thus integrated to the design and optimization of these systems. This work explores techniques for determining incident energy distribu-tions on the surface of corrugated reactor geometries with non-ideal cross-sectional profiles, and the local and overall mass transfer rates obtained using computational fluid dynamics and experimental analysis. By examining the reaction kinetics for the photo-degradation of 4-chlorophenol over a TiO2 catalyst, the effects of surface area, energy incidence with photon recapture, and local mass transfer on overall reactor performance are presented to highlight ootimization concerns for these tvoes of reactors.

  7. Alpha Channeling in a Rotating Plasma

    Energy Technology Data Exchange (ETDEWEB)

    Abraham J. Fetterman and Nathaniel J. Fisch

    2008-09-23

    The wave-particle α-channeling effect is generalized to include rotating plasma. Specifically, radio frequency waves can resonate with α particles in a mirror machine with E × B rotation to diffuse the α particles along constrained paths in phase space. Of major interest is that the α-particle energy, in addition to amplifying the RF waves, can directly enhance the rotation energy which in turn provides additional plasma confinement in centrifugal fusion reactors. An ancillary benefit is the rapid removal of alpha particles, which increases the fusion reactivity.

  8. Steam-Reheat Option for Supercritical-Water-Cooled Reactors

    Science.gov (United States)

    Saltanov, Eugene

    SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. Main objectives of the development are to increase thermal efficiency of a Nuclear Power Plant (NPP) and to decrease capital and operational costs. The first objective can be achieved by introducing nuclear steam reheat inside a reactor and utilizing regenerative feedwater heaters. The second objective can be achieved by designing a steam cycle that closely matches that of the mature supercritical fossil-fuelled power plants. The feasibility of these objectives is discussed. As a part of this discussion, heat-transfer calculations have been performed and analyzed for SuperCritical-Water (SCW) and SuperHeated-Steam (SHS) channels of the proposed reactor concept. In the calculations a uniform and three non-uniform Axial Heat Flux Profiles (AHFPs) were considered for six different fuels (UO2, ThO 2, MOX, UC2, UC, and UN) and at average and maximum channel power. Bulk-fluid, sheath, and fuel centerline temperatures as well as the Heat Transfer Coefficient (HTC) profiles were obtained along the fuel-channel length. The HTC values are within a range of 4.7--20 kW/m2·K and 9.7--10 kW/m2·K for the SCW and SHS channels respectively. The main conclusion is that while all the mentioned fuels may be used for the SHS channel, only UC2, UC, or UN are suitable for a SCW channel, because their fuel centerline temperatures are at least 1000°C below melting point, while that of UO2, ThO2 , and MOX may reach melting point.

  9. C Reactor overbore test facility review

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, P.A.; Nilson, R.

    1964-04-24

    In 1961, large-size, smooth-bore, Zircaloy process tubes were installed in C-Reactor graphite channels that had been enlarged to 2.275 inches. These tubes were installed to provide a test and demonstration facility for the concept of overboring as a means of securing significant improvement in the production capability of the reactors, After two years of facility operation, it is now appropriate to consider the extent to which original objectives have been achieved, to re-examine the original objectives, and to consider the best future use of this unique facility. This report presents the general results of such a review and re-examination in more detail.

  10. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

    The principles of nuclear, thermal and hydraulic performance calculations for water cooled reactors are discussed. The principles are illustrated by describing their implementation in the UKAEA PATRIARCH scheme of computer codes. This material was originally delivered as a course of lectures at the Technical University of Helsinki in Summer of 1969.

  11. High power density reactors based on direct cooled particle beds

    Science.gov (United States)

    Powell, J. R.; Horn, F. L.

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

  12. Safety requirements applied to research reactors in France

    International Nuclear Information System (INIS)

    Full text: In France, there are currently some twenty research reactors in operation with a thermal powers up to a hundred megawatts. General safety requirements such as the redundancy and separation of protection system channels, continuous monitoring of confinement barriers and containment building leak tightness with respect to underlying soils and the underground water have been gradually established and applied. Regarding the seismic risk and those risks relating to the industrial environment and transportation of hazardous materials, the rules applying to research reactor design are the same as those applying to power reactors, albeit with some adaptations due to the specific features of certain reactors (short operating time or low radioactive product inventory). The following safety requirements applying specifically to the confinement barriers of pool-type research reactors should be noted: there must be no fuel cladding dryout under the various operating conditions; in the case of plate type fuels, this requirement implies checking the absence of flow redistribution in the hottest cooling channel; reactors must not be operated with a fuel element affected by clad failure; in such situations, the reactor must be automatically shut down and the fuel element in question removed and stored in a leaktight container; the core must not be uncovered in the event of a pipe break in the reactor coolant system or a window failure in neutron beam channels; this requirement is met through the integrated design of the reactor primary coolant system, which is installed in a 'water block', and through the implementation of automatic isolation valves on the neutron beam channels. The most significant specific approach adopted in France for the design of pool-type reactors using uranium and aluminum metal fuels is to take into account a BORAX-type explosive reactivity accident. For this type of accident, which is supposed to lead to total meltdown of the core under water

  13. Transients analysis by reactivity insertion in research reactors

    International Nuclear Information System (INIS)

    PARET code was used to simulate accidental situations arising from positive reactivity insertions, in order to analyze the behavior of RP-10 reactor. The simulations considered three different cases: First is for the reactor operating at 10 Mw nominal power with 3 pumps in use, the second, at 6.6 Mw with only one pump working. In all cases the reactor trip was assumed when a 12 Mw power level is reached. An additional simulation for the reactor operating at 50w before the reactivity insertion, showed to be the worst accidental situation of all cases because of the higher temperature and power rise. Hot channel thermohydraulic and kinetic parameters have been evaluated at each axial mesh point and transient time step. None of the cases showed melting of fuel plates

  14. A study on TRIGA core reconfiguration with new irradiation channels

    International Nuclear Information System (INIS)

    Highlights: ► TRIGA reactor core has been studied to achieve enhanced irradiation facilities. ► Two neutronic performance parameters and three safety parameters are calculated. ► Results are compared with reference configuration with single irradiation channel. ► Simultaneous introduction of three irradiation channels inside the core is positive. - Abstract: This study is being carried on as a provision of achieving enhanced irradiation facilities in the TRIGA MARK II research reactor core. Two reconfigured cores, each one having three irradiation channels are being analyzed with the Monte Carlo code MVP and the nuclear data library JENDL-3.3. The existing core has only one irradiation channel at the center. The graphite dummies are rearranged to surround the new irradiation channels in the new configurations. The results for each reconfigured core are compared with the existing core as reference. From this study important information are obtained for enhanced as well as economic utilization of TRIGA core. The results show positively the possibility to introduce new irradiation channels without disturbing the core excess reactivity and neutron flux significantly as well as keeping the shutdown margin and peaking factor close to the reference.

  15. Safety of research reactors

    International Nuclear Information System (INIS)

    The number of research reactors that have been constructed worldwide for civilian applications is about 651. Of the reactors constructed, 284 are currently in operation, 258 are shut down and 109 have been decommissioned. More than half of all operating research reactors worldwide are over thirty years old. During this long period of time national priorities have changed. Facility ageing, if not properly managed, has a natural degrading effect. Many research reactors face concerns with the obsolescence of equipment, lack of experimental programmes, lack of funding for operation and maintenance and loss of expertise through ageing and retirement of the staff. Other reactors of the same vintage maintain effective ageing management programmes, conduct active research programmes, develop and retain high calibre personnel and make important contributions to society. Many countries that operate research reactors neither operate nor plan to operate power reactors. In most of these countries there is a tendency not to create a formal regulatory body. A safety committee, not always independent of the operating organization, may be responsible for regulatory oversight. Even in countries with nuclear power plants, a regulatory regime differing from the one used for the power plants may exist. Concern is therefore focused on one tail of a continuous spectrum of operational performance. The IAEA has been sending missions to review the safety of research reactors in Member States since 1972. Some of the reviews have been conducted pursuant to the IAEA' functions and responsibilities regarding research reactors that are operated within the framework of Project and Supply Agreements between Member States and the IAEA. Other reviews have been conducted upon request. All these reviews are conducted following procedures for Integrated Safety Assessment of Research Reactors (INSARR) missions. The prime objective of these missions has been to conduct a comprehensive operational safety

  16. CFD simulation with detailed chemistry of steam reforming of methane for hydrogen production in an integrated micro-reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Xuli; Cheng, Yinhong; Jin, Yong; Cheng, Yi [Department of Chemical Engineering, Beijing Key Laboratory of Green Chemical Reaction Engineering and Technology, Tsinghua University, Beijing 100084 (China); Ding, Shi [Department of Chemical Engineering, Beijing Key Laboratory of Green Chemical Reaction Engineering and Technology, Tsinghua University, Beijing 100084 (China); Research Institute of Petroleum Processing, SINOPEC, Beijing 100083 (China)

    2010-06-15

    micro-reactor has drawn more and more attention in recent years due to the process intensification on basic transport phenomena in micro-channels, which would often lead to the improved reactor performance. Steam reforming of methane (SRM) in micro-reactor has great potential to realize a low-cost, compact process for hydrogen production via an evident shortening of reaction time from seconds to milliseconds. This work focuses on the detailed modeling and simulation of a micro-reactor design for SRM reaction with the integration of a micro-channel for Rh-catalyzed endothermic reaction, a micro-channel for Pt-catalyzed exothermic reaction and a wall in between with Rh or Pt-catalyst coated layer. The elementary reaction kinetics for SRM process is adopted in the CFD model, while the combustion channel is described by global reaction kinetics. The model predictions were quantitatively validated by the experimental data in the literature. For the extremely fast reactions in both channels, the simulations indicated the significance of the heat conduction ability of the reactor wall as well as the interplay between the exothermic and endothermic reactions (e.g., the flow rate ratio of fuel gas to reforming gas). The characteristic width of 0.5 mm is considered to be a suitable channel size to balance the trade-off between the heat transfer behavior in micro-channels and the easy fabrication of micro-channels. (author)

  17. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  18. Reactor engineering department annual report

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  19. Analysis of a homogenous and heterogeneous stylized half core of a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    EL-Khawlani, Afrah [Physics Department, Sana' a (Yemen); Aziz, Moustafa [Nuclear and radiological regulatory authority, Cairo (Egypt); Ismail, Mahmud Yehia; Ellithi, Ali Yehia [Cairo Univ. (Egypt). Faculty of Science

    2015-03-15

    The MCNPX (Monte Carlo N-Particle Transport Code System) code has been used for modeling and simulation of a half core of CANDU (CANada Deuterium-Uranium) reactor, both homogenous and heterogeneous model for the reactor core are designed. The fuel is burnt in normal operation conditions of CANDU reactors. Natural uranium fuel is used in the model. The multiplication factor for homogeneous and heterogeneous reactor core is calculated and compared during fuel burnup. The concentration of both uranium and plutonium isotopes are analysed in the model. The flux and power distributions through channels are calculated.

  20. A linearization of quantum channels

    Science.gov (United States)

    Crowder, Tanner

    2015-06-01

    Because the quantum channels form a compact, convex set, we can express any quantum channel as a convex combination of extremal channels. We give a Euclidean representation for the channels whose inverses are also valid channels; these are a subset of the extreme points. They form a compact, connected Lie group, and we calculate its Lie algebra. Lastly, we calculate a maximal torus for the group and provide a constructive approach to decomposing any invertible channel into a product of elementary channels.

  1. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  2. Test reactor technology

    International Nuclear Information System (INIS)

    The Reactor Development Program created a need for engineering testing of fuels and materials. The Engineering Test Reactors were developed around the world in response to this demand. The design of the test reactors proved to be different from that of power reactors, carrying the fuel elements closer to the threshold of failure, requiring more responsive instrumentation, more rapid control element action, and inherent self-limiting behavior under accident conditions. The design of the experimental facilities to exploit these reactors evolved a new, specialized, branch of engineering, requiring a very high-lvel scientific and engineering team, established a meticulous concern with reliability, the provision for recovery from their own failures, and detailed attention to possible interactions with the test reactors. This paper presents this technology commencing with the Materials Testing Reactor (MTR) through the Fast Flux Test Facility, some of the unique experimental facilities developed to exploit them, but discusses only cursorily the experiments performed, since sample preparation and sample analyses were, and to some extent still are, either classified or proprietary. The Nuclear Engineering literature is filled with this information

  3. RFI channels, 2

    Science.gov (United States)

    Mceliece, R. J.

    1981-01-01

    The cutoff parameters for a class of channel models exhibiting burst noise behavior were calculated and the performance of interleaved coding strategies was evaluated. It is concluded that, provided the channel memory is large enough and is properly exploited, interleaved coding is nearly optimal.

  4. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts

  5. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  6. Licensed operating reactors

    International Nuclear Information System (INIS)

    The Operating Units Status Report --- Licensed Operating Reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff on NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non- power reactors in the US

  7. nuclear reactor design calculations

    International Nuclear Information System (INIS)

    In this work , the sensitivity of different reactor calculation methods, and the effect of different assumptions and/or approximation are evaluated . A new concept named error map is developed to determine the relative importance of different factors affecting the accuracy of calculations. To achieve this goal a generalized, multigroup, multi dimension code UAR-DEPLETION is developed to calculate the spatial distribution of neutron flux, effective multiplication factor and the spatial composition of a reactor core for a period of time and for specified reactor operating conditions. The code also investigates the fuel management strategies and policies for the entire fuel cycle to meet the constraints of material and operating limitations

  8. Course on reactor physics

    International Nuclear Information System (INIS)

    In Germany only few students graduate in nuclear technology, therefore the NPP operating companies are forced to develop their own education and training concepts. AREVA NP has started together with the Technical University of Dresden a one-week course ''reactor physics'' that includes the know-how of the nuclear power plant construction company. The Technical University of Dresden has the training reactor AKR-2 that is retrofitted by modern digital instrumentation and control technology that allows the practical training of reactor control.

  9. PWR type reactor

    International Nuclear Information System (INIS)

    From a PWR with a primary circuit, consisting of a reactor pressure vessel, a steam generator and a reactor coolant pump, hot coolant is removed by means of an auxiliary system containing h.p. pumps for feeding water into the primary circuit and being connected with a pipe, originating at the upper part, which has got at least one isolating value. This is done by opening an outlet in a part of the auxiliary system that has got a lower pressure than the reactor vessel. Preferably a water jet pump is used for mixing with the water of the auxiliary system. (orig.)

  10. Athermalized channeled spectropolarimeter enhancement.

    Energy Technology Data Exchange (ETDEWEB)

    Jones, Julia Craven; Way, Brandyn Michael; Mercier, Jeffrey Alan; Hunt, Jeffery P.

    2013-09-01

    Channeled spectropolarimetry can measure the complete polarization state of light as a function of wavelength. Typically, a channeled spectropolarimeter uses high order retarders made of uniaxial crystal to amplitude modulate the measured spectrum with the spectrally-dependent Stokes polarization information. A primary limitation of conventional channeled spectropolarimeters is related to the thermal variability of the retarders. Thermal variation often forces frequent system recalibration, particularly for field deployed systems. However, implementing thermally stable retarders, made of biaxial crystal, results in an athermal channeled spectropolarimeter that relieves the need for frequent recalibration. This report presents experimental results for an anthermalized channeled spectropolarimeter prototype produced using potassium titanyl phosphate. The results of this prototype are compared to the current thermal stabilization state of the art. Finally, the application of the technique to the thermal infrared is studied, and the athermalization concept is applied to an infrared imaging spectropolarimeter design.

  11. Channelling versus inversion

    DEFF Research Database (Denmark)

    Gale, A.S.; Surlyk, Finn; Anderskouv, Kresten

    2013-01-01

    . Within this channel were smaller erosional structures (<10 m deep) that truncate originally horizontal bedding, are floored by hardgrounds, and locally have a basal fill of granular phosphorite. The entire channel system was progressively infilled by chalk, as demonstrated by the expanded succession of...... the lower Campanian Culver Chalk Formation. The beds of the channel fill are cut by small step faults, resulting from gravitational collapse. Complete burial had taken place by the base of the upper Campanian Portsdown Chalk Formation, which is of even thickness across the region. The structures are......Evidence from regional stratigraphical patterns in Santonian−Campanian chalk is used to infer the presence of a very broad channel system (5 km across) with a depth of at least 50 m, running NNW−SSE across the eastern Isle of Wight; only the western part of the channel wall and fill is exposed...

  12. Modification of the Core Cooling System of TRIGA 2000 Reactor

    Science.gov (United States)

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-01

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  13. Management of distortion channels in the Cofrentes NPP; Gestion de la deformacion de canales en la central nuclear de Cofrentes

    Energy Technology Data Exchange (ETDEWEB)

    Albendea, J. C.; Garcia, P. J.; Iglesias, J.; Mascarell, R.

    2015-07-01

    Fuel channels distortion in BWR (Boiling Water Reactor) reactors may have implication for safety. This phenomenon is complex and, at the present time it is not known in detail. This article provides the Iberdrola Generacion Nuclear SAU ongoing activities to know, predict and mitigate the consequences that this phenomenon may cause in Cofrentes Nuclear Power Plant. (Author)

  14. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

  15. NEUTRONIC REACTOR FUEL COMPOSITION

    Science.gov (United States)

    Thurber, W.C.

    1961-01-10

    Uranium-aluminum alloys in which boron is homogeneously dispersed by adding it as a nickel boride are described. These compositions have particular utility as fuels for neutronic reactors, boron being present as a burnable poison.

  16. Pulsed fusion reactors

    International Nuclear Information System (INIS)

    This summer school specialized in examining specific fusion center systems. Papers on scientific feasibility are first presented: confinement of high-beta plasma, liners, plasma focus, compression and heating and the use of high power electron beams for thermonuclear reactors. As for technological feasibility, lectures were on the theta-pinch toroidal reactors, toroidal diffuse pinch, electrical engineering problems in pulsed magnetically confined reactors, neutral gas layer for heat removal, the conceptual design of a series of laser fusion power plants with ''Saturn'', implosion experiments and the problem of the targets, the high brightness lasers for plasma generation, and topping and bottoming cycles. Some problems common to pulsed reactors were examined: energy storage and transfer, thermomechanical and erosion effects in the first wall and blanket, the problems of tritium production, radiation damage and neutron activation in blankets, and the magnetic and inertial confinement

  17. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2002-04-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

  18. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2001-04-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  19. Reactor Neutrino Spectra

    CERN Document Server

    Hayes, A C

    2016-01-01

    We present a review of the antineutrino spectra emitted from reactors. Knowledge of these and their associated uncertainties are crucial for neutrino oscillation studies. The spectra used to-date have been determined by either conversion of measured electron spectra to antineutrino spectra or by summing over all of the thousands of transitions that makeup the spectra using modern databases as input. The uncertainties in the subdominant corrections to beta-decay plague both methods, and we provide estimates of these uncertainties. Improving on current knowledge of the antineutrino spectra from reactors will require new experiments. Such experiments would also address the so-called reactor neutrino anomaly and the possible origin of the shoulder observed in the antineutrino spectra measured in recent high-statistics reactor neutrino experiments.

  20. Experience with Kamini reactor

    International Nuclear Information System (INIS)

    Kamini is a 233U fuelled, 30 kW(th) research reactor. It is one of the best neutron source facility with a core average flux of 1012 n/cm2/s in IGCAR used for neutron radiography of active and nonradioactive objects, activation analysis and radiation physics research. The core consists of nine plate type fuel elements with a total fuel inventory of 590 g of 233U. Two safety control plates made of cadmium are used for start up and shutdown of the reactor. Three beam tubes, two-thimble irradiation site outside reflector and one irradiation site nearer to the core constitute the testing facilities of Kamini. Kamini attained first criticality on 29th October 96 and nominal power of 30 kW in September 1997. This paper covers the design features of the reactor, irradiation facilities and their utilities and operating experience of the reactor. (author)

  1. Reactor pressure boundary materials

    International Nuclear Information System (INIS)

    With a long-term operation of nuclear power plants, the component materials are degraded under severe reactor conditions such as neutron irradiation, high temperature, high pressure and corrosive environment. It is necessary to establish the reliable and practical technologies for improving and developing the component materials and for evaluating the mechanical properties. Especially, it is very important to investigate the technologies for reactor pressure boundary materials such as reactor vessel and pipings in accordance with their critical roles. Therefore, this study was focused on developing and advancing the microstructural/micro-mechanical evaluation technologies, and on evaluating the neutron irradiation characteristics and radiation effects analysis technology of the reactor pressure boundary materials, and also on establishing a basis of nuclear material property database

  2. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2002-04-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

  3. New reactor type proposed

    CERN Multimedia

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  4. Bruce 'A' Unit 4 fuel channel feeder coupling leakage

    International Nuclear Information System (INIS)

    CANDU reactor fuel channels are connected to the primary heat transport system by mechanical joints known as feeder couplings. In 1991, three feeder coupling leakages were discovered in Bruce 4. These leakages required the unit to be shut down for repair. An investigation showed that the leakage was caused by a small 'separation' at the couplings. The assessment was that this was unlikely to be a generic issue, because the later reactors have stronger capscrews. At the time of the conference, further improvements were being considered for the seal ring and capscrew materials, and more accurate capscrew preload measurements. 3 tabs., 10 figs

  5. Experimental studies in a single-phase parallel channel natural circulation system. Preliminary results

    Energy Technology Data Exchange (ETDEWEB)

    Bodkha, Kapil; Pilkhwal, D.S.; Jana, S.S.; Vijayan, P.K. [Bhabha Atomic Research Centre, Mumbai (India). Reactor Engineering Div.

    2016-03-15

    Natural circulation systems find extensive applications in industrial engineering systems. One of the applications is in nuclear reactor where the decay heat is removed by natural circulation of the fluid under off-normal conditions. The upcoming reactor designs make use of natural circulation in order to remove the heat from core under normal operating conditions also. These reactors employ multiple vertical fuel channels with provision of on-power refueling/defueling. Natural circulation systems are relatively simple, safe and reliable when compared to forced circulation systems. However, natural circulation systems are prone to encounter flow instabilities which are highly undesirable for various reasons. Presence of parallel channels under natural circulation makes the system more complicated. To examine the behavior of parallel channel system, studies were carried out for single-phase natural circulation flow in a multiple vertical channel system. The objective of the present work is to study the flow behavior of the parallel heated channel system under natural circulation for different operating conditions. Steady state and transient studies have been carried out in a parallel channel natural circulation system with three heated channels. The paper brings out the details of the system considered, different cases analyzed and preliminary results of studies carried out on a single-phase parallel channel system.

  6. Special lecture on nuclear reactor

    International Nuclear Information System (INIS)

    This book gives a special lecture on nuclear reactor, which is divided into two parts. The first part has explanation on nuclear design of nuclear reactor and analysis of core with theories of integral transports, diffusion Nodal, transports Nodal and Monte Carlo skill parallel computer and nuclear calculation and speciality of transmutation reactor. The second part deals with speciality of nuclear reactor and control with nonlinear stabilization of nuclear reactor, nonlinear control of nuclear reactor, neural network and control of nuclear reactor, control theory of observer and analysis method of Adomian.

  7. Jet-Stirred Reactors

    OpenAIRE

    Herbinet, Olivier; Guillaume, Dayma

    2013-01-01

    The jet-stirred reactor is a type of ideal continuously stirred-tank reactor which is well suited for gas phase kinetic studies. It is mainly used to study the oxidation and the pyrolysis of hydrocarbon and oxygenated fuels. These studies consist in recording the evolution of the conversion of the reactants and of the mole fractions of reaction products as a function of different parameters such as reaction temperature, residence time, pressure and composition of the inlet gas. Gas chromatogr...

  8. Generation IV reactors: economics

    International Nuclear Information System (INIS)

    The operating nuclear reactors were built over a short period: no more than 10 years and today their average age rounds 18 years. EDF (French electricity company) plans to renew its reactor park over a far longer period : 30 years from 2020 to 2050. According to EDF this objective implies 3 constraints: 1) a service life of 50 to 60 years for a significant part of the present operating reactors, 2) to be ready to built a generation 3+ unit in 2020 which infers the third constraint: 3) to launch the construction of an EPR (European pressurized reactor) prototype as soon as possible in order to have it operating in 2010. In this scheme, generation 4 reactor will benefit the feedback experience of generation 3 and will take over in 2030. Economic analysis is an important tool that has been used by the generation 4 international forum to select the likely future reactor systems. This analysis is based on 4 independent criteria: the basic construction cost, the construction time, the operation and maintenance costs and the fuel cycle cost. This analysis leads to the evaluation of the global cost of electricity generation and of the total investment required for each of the reactor system. The former defines the economic competitiveness in a de-regulated energy market while the latter is linked to the financial risk taken by the investor. It appears, within the limits of the assumptions and models used, that generation 4 reactors will be characterized by a better competitiveness and an equivalent financial risk when compared with the previous generation. (A.C.)

  9. OECD Halden reactor project

    International Nuclear Information System (INIS)

    This report summarizes the activities of the OECD Halden Reactor Project for the year 1976. The main items reported on are: a) the process supervision and control which have focused on core monitoring and control, and operator-process communication; b) the fuel performance and safety behavior which have provided data and analytical descriptions of the thermal, mechanical and chemical behavior of fuel under various operating conditions; c) the reactor operations and d) the administration and finance

  10. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  11. Department of Reactor Technology

    DEFF Research Database (Denmark)

    Risø National Laboratory, Roskilde

    The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included.......The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included....

  12. Steam Generator Group Project. Task 6. Channel head decontamination

    International Nuclear Information System (INIS)

    The Steam Generator Group Project utilizes a retired-from-service pressurized-water-reactor steam generator as a test bed and source of specimens for research. An important preparatory step to primary side research activities was reduction of the radiation field in the steam generator channel head. This task report describes the channel head decontamination activities. Though not a programmatic research objective it was judged beneficial to explore the use of dilute reagent chemical decontamination techniques. These techniques presented potential for reduced personnel exposure and reduced secondary radwaste generation, over currently used abrasive blasting techniques. Two techniques with extensive laboratory research and vendors prepared to offer commercial application were tested, one on either side of the channel head. As indicated in the report, both techniques accomplished similar decontamination objectives. Neither technique damaged the generator channel head or tubing materials, as applied. This report provides details of the decontamination operations. Application system and operating conditions are described

  13. Reactor TRIGA at the J.Stefan institute in Ljubljana

    International Nuclear Information System (INIS)

    The TRIGA Mark II Reactor began its operation on May 1966. The power of the reactor is 250 kW. TRIGA utilizes solid fuel elements in which the zirconium hydride moderator is homogeneously mixed 20% or 70% enriched uranium. The inique featUre of these fuel - moderator elements is the prompt negative temperature coefficient of reactivity, which gives TRIGA its built-in safety. The reactor core consist of a lattice of cylindrical fuel-moderator elements and graphite (dummy) elements at the bottom of the 6 m high tank full of light water which is used for cooling and radiation protection. The reactor has the following experimental and irradiation facilities: 2 radial beam channels, 2 tangential beam channels, 2 thermal colomns, 40 position rotary specimen rack, pneumatic transfer tube and central thimble. The reactor operates about 2.500 hours per year and it is utilized for the production of isotopes, as a source of neutrons for various experiments and for the training of personnel for the nuclear power station in Krsko

  14. Moon base reactor system

    Science.gov (United States)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  15. BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Purpose: To simplify the structure of an emergency core cooling system while suppressing the flow out of coolants upon rapture accidents in a coolant recycling device of BWR type reactors. Constitution: Recirculation pumps are located at a position higher than the reactor core in a pressure vessel, and the lower plenum is bisected vertically by a partition plate. Further, a gas-liquid separator is surrounded with a wall and the water level at the outer side of the wall is made higher than the water level in the inside of the wall. In this structure, coolants are introduced from the upper chamber in the lower plenum into the reactor core, and the steams generated in the reactor core are separated in the gas-liquid separator, whereby the separated liquid is introduced as coolants by way of the inner chamber into the lower chamber of the lower plenum and further sent by way of the outer chamber into the reactor core. Consequently, idle rotation of the recycling pumps due to the flow-in of saturated water is prevented and loss of coolants in the reactor core can also be prevented upon raptures in the pipeway and the driving section of the pump connected to the pressure vessel and in the bottom of the pressure vessel. (Horiuchi, T.)

  16. OECD Halden reactor project

    International Nuclear Information System (INIS)

    This is the nineteenth annual Report on the OECD Halden Reactor Project, describing activities at the Project during 1978, the last year of the 1976-1978 Halden Agreement. Work continued in two main fields: test fuel irradiation and fuel research, and computer-based process supervision and control. Project research on water reactor fuel focusses on various aspects of fuel behavior under normal, and off-normal transient conditions. In 1978, participating organisations continued to submit test fuel for irradiation in the Halden boiling heavy-water reactor, in instrumented test assemblies designed and manufactured by the Project. Work included analysis of the impact of fuel design and reactor operating conditions on fuel cladding behavior. Fuel performance modelling included characterization of thermal and mechanical behavior at high burn-up, of fuel failure modes, and improvement of data qualification procedures to reduce and quantify error bands on in-reactor measurements. Instrument development yielded new or improved designs for measuring rod temperature, internal pressure, axial neutron flux shape determination, and for detecting cladding defects. Work on computer-based methods of reactor supervision and control included continued development of a system for predictive core surveillance, and of special mathematical methods for core power distribution control

  17. Home-made refurbishment of the instrumentation and control system of the TRIGA reactor of the University of Pavia

    International Nuclear Information System (INIS)

    The Instrumentation and Control (I and C) System of the TRIGA reactor of the University of Pavia was dated and, in order to grant a safe and continuous reactor operation for the future, it became necessary to substitute or to upgrade the system. Since the substitution of the I and C system with a new-made one was very difficult to be performed due to long authorization procedures, an home-made refurbishment was planned. Using commercial components of high quality, almost a complete substitution, channel-by-channel, of the I and C system was realized without changing the operating and safety logics. The system includes: - the Reactor Linear Power Channel and Chart Recorder; - the Reactor Percent Power Safety Channel; - the High Voltage and Low Voltage Power Supply; - the Automatic Reactor Power Control; - the Fuel Elements and Cooling-Water Temperatures Measuring Channels; - the Water Conductivity Measuring Channel. The refurbished I and C system shows a very good operational behavior and reliability and will assure a continuous operation of the reactor for the future

  18. Nondestructive testing of PWR type fuel rods by eddy currents and metrology in the OSIRIS reactor pool

    International Nuclear Information System (INIS)

    The Saclay Reactor Department has developed a nondestructive test bench, now installed above channel 1 of the OSIRIS reactor. As part of investigations into the dynamics of PWR fuel degradation, a number of fuel rods underwent metrological and eddy current inspection, after irradiation

  19. Safety analysis of 5 MW IEAR-1 reactor; Analise de seguranca do reator IEA-R1 a 5 MW

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Antonio T. e; Maprelian, Eduardo; Rodrigues, Antonio C.I.; Cabral, Eduardo L.L.; Molnary, Leslie de; Mesquita, Ricardo N.; Mendonca, Arlindo G. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Dept. de Reatores. E-mail: teixeira@net.ipen.br

    2000-07-01

    This paper presents the methods and procedures utilized in the safety analysis of IEA-R1 research reactor. Four postulated accidents are quantitatively analyzed, being the fuel channel blockage accident considered as the Maximum Credible Accident for the reactor. The potential accident consequences and the criteria for radiological doses acceptance are evaluated and discussed. (author)

  20. Thorium utilization in Candu reactors

    International Nuclear Information System (INIS)

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

  1. Current tendencies and perspectives of development research reactors of Russia

    International Nuclear Information System (INIS)

    ; - reactor PIK, satisfying Russian and foreign requirements for beam experiments in the range of nuclear physics, condensed matter physics and other fundamental investigations; - pulsed reactors; - eventually, ADS draw growing attention lately; these plants can compete with RR as neutron sources for some task solutions. In the report the results of new engineering developments by the designs of reactors under reconstruction and modernization are presented, as well as the results for experimental arrangements of different purposes for these reactors: - IREN; - Steam-water loop PVP-3 (reactor MIR.M1); - Medical channel for BNCT (reactor IRT, MEPhI). Problem of selection of high-powered universal Research Reactor type and design of next generation also exists in the face of Russian experts. They have 5-10 years to solve this problem. Russia is ready for cooperation in the field of development, construction and utilization in future of new research reactor, which can be built in Europe instead of decommissioned ones

  2. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  3. Application at Marcoule of Punched-Card Techniques to Problems of Programming Reactor Unloading During Operation

    International Nuclear Information System (INIS)

    Continuous unloading during operation presents special problems for the unloading programme - problems far more complex than when the reactor is unloaded during shutdown. The yield of the fuel is linked with the homogenization of its irradiation and this in turn is a function of the radial and the axial flux. Axial homogenization of the irradiation depends on the method of unloading. In reactors that can be unloaded during operation it can be achieved under more favourable conditions than in reactors that must be unloaded during shut-down. In reactors unloaded during operation, radial homogenization of the irradiation results from the execution of an unloading programme established per channel - not per zone, as in reactors unloaded during shut-down. The unloading programmes are prepared at the Marcoule Centre by a punched-card machine - electronic computer system, which also carries out various accessory operations connected with the functioning of the reactors and control of the fuel. (author)

  4. Channel Choice: A Literature Review

    DEFF Research Database (Denmark)

    Østergaard Madsen, Christian; Kræmmergaard, Pernille

    2015-01-01

    The channel choice branch of e-government studies citizens’ and businesses’ choice of channels for interacting with government, and how government organizations can integrate channels and migrate users towards the most cost-efficient channels. In spite of the valuable contributions offered no sys...... no systematic overview exist of channel choice. We present a literature review of channel choice studies in government to citizen context identifying authors, countries, methods, concepts, units of analysis, and theories, and offer suggestionsfor future studies....

  5. Deleterious Thermal Effects Due To Randomized Flow Paths in Pebble Bed, and Particle Bed Style Reactors

    Science.gov (United States)

    Moran, Robert P.

    2013-01-01

    A review of literature associated with Pebble Bed and Particle Bed reactor core research has revealed a systemic problem inherent to reactor core concepts which utilize randomized rather than structured coolant channel flow paths. For both the Pebble Bed and Particle Bed Reactor designs; case studies reveal that for indeterminate reasons, regions within the core would suffer from excessive heating leading to thermal runaway and localized fuel melting. A thermal Computational Fluid Dynamics model was utilized to verify that In both the Pebble Bed and Particle Bed Reactor concepts randomized coolant channel pathways combined with localized high temperature regions would work together to resist the flow of coolant diverting it away from where it is needed the most to cooler less resistive pathways where it is needed the least. In other words given the choice via randomized coolant pathways the reactor coolant will take the path of least resistance, and hot zones offer the highest resistance. Having identified the relationship between randomized coolant channel pathways and localized fuel melting it is now safe to assume that other reactor concepts that utilize randomized coolant pathways such as the foam core reactor are also susceptible to this phenomenon.

  6. A Simple Water Channel

    Science.gov (United States)

    White, A. S.

    1976-01-01

    Describes a simple water channel, for use with an overhead projector. It is run from a water tap and may be used for flow visualization experiments, including the effect of streamlining and elementary building aerodynamics. (MLH)

  7. 28-Channel rotary transformer

    Science.gov (United States)

    Mclyman, W. T.

    1981-01-01

    Transformer transmits power and digital data across rotating interface. Array has many parallel data channels, each with potential l megabaud data rate. Ferrite-cored transformers are spaced along rotor; airgap between them reduces crosstalk.

  8. Channelized Streams in Iowa

    Data.gov (United States)

    Iowa State University GIS Support and Research Facility — This draft dataset consists of all ditches or channelized pieces of stream that could be identified using three input datasets; namely the1:24,000 National...

  9. CHANNEL ESTIMATION TECHNIQUE

    DEFF Research Database (Denmark)

    2015-01-01

    A method includes determining a sequence of first coefficient estimates of a communication channel based on a sequence of pilots arranged according to a known pilot pattern and based on a receive signal, wherein the receive signal is based on the sequence of pilots transmitted over the communicat......A method includes determining a sequence of first coefficient estimates of a communication channel based on a sequence of pilots arranged according to a known pilot pattern and based on a receive signal, wherein the receive signal is based on the sequence of pilots transmitted over the...... communication channel. The method further includes determining a sequence of second coefficient estimates of the communication channel based on a decomposition of the first coefficient estimates in a dictionary matrix and a sparse vector of the second coefficient estimates, the dictionary matrix including...

  10. Sensing with Ion Channels

    CERN Document Server

    Martinac, Boris

    2008-01-01

    All living cells are able to detect and translate environmental stimuli into biologically meaningful signals. Sensations of touch, hearing, sight, taste, smell or pain are essential to the survival of all living organisms. The importance of sensory input for the existence of life thus justifies the effort made to understand its molecular origins. Sensing with Ion Channels focuses on ion channels as key molecules enabling biological systems to sense and process the physical and chemical stimuli that act upon cells in their living environment. Its aim is to serve as a reference to ion channel specialists and as a source of new information to non specialists who want to learn about the structural and functional diversity of ion channels and their role in sensory physiology.

  11. TRP channels in disease.

    Science.gov (United States)

    Jordt, S E; Ehrlich, B E

    2007-01-01

    The transient receptor potential (TRP) channels are a large family of proteins with six main subfamilies termed the TRPC (canonical), TRPV (vanilloid), TRPM (melastatin), TRPP (polycystin), TRPML (mucolipin), and TRPA (ankyrin) groups. The sheer number of different TRPs with distinct functions supports the statement that these channels are involved in a wide range of processes ranging from sensing of thermal and chemical signals to reloading intracellular stores after responding to an extracellular stimulus. Mutations in TRPs are linked to pathophysiology and specific diseases. An understanding of the role of TRPs in normal physiology is just beginning; the progression from mutations in TRPs to pathophysiology and disease will follow. In this review, we focus on two distinct aspects of TRP channel physiology, the role of TRP channels in intracellular Ca2+ homeostasis, and their role in the transduction of painful stimuli in sensory neurons. PMID:18193640

  12. Development of defueling device for CANDU fuel channel (modeling)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. Y.; Yu, K. H.; Yang, J. S.; Lee, H. S.; Chang, K. J.; Kim, Y. J. [CNEC Technical Office, Taejon (Korea, Republic of); Lee, S. K. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    Commercial CANDU reactors use D{sub 2}O for moderator and heat transfer material and also have Fueling Machines(F/M) and related system equipment in order to assist on-power refueling operation. A Defuelling Device(DFD) is developed for the proper defuelling of all fuels in all fuel channels during shutdown condition of plant. This device is considered more efficient in defuelling compared to the existing Fuel Grapple System for its use of existing D{sub 2}O flow in the fuel channel. In this study, computational fluid dynamic software is used for optimize and evaluation of the design for its applicability.

  13. Chloride channels in stroke

    Institute of Scientific and Technical Information of China (English)

    Ya-ping ZHANG; Hao ZHANG; Dayue Darrel DUAN

    2013-01-01

    Vascular remodeling of cerebral arterioles,including proliferation,migration,and apoptosis of vascular smooth muscle cells (VSMCs),is the major cause of changes in the cross-sectional area and diameter of the arteries and sudden interruption of blood flow or hemorrhage in the brain,ie,stroke.Accumulating evidence strongly supports an important role for chloride (Clˉ) channels in vascular remodeling and stroke.At least three Clˉ channel genes are expressed in VSMCs:1) the TMEM16A (or Ano1),which may encode the calcium-activated Clˉ channels (CACCs); 2) the CLC-3 Clˉ channel and Clˉ/H+ antiporter,which is closely related to the volume-regulated Clˉ channels (VRCCs); and 3) the cystic fibrosis transmembrane conductance regulator (CFTR),which encodes the PKA-and PKC-activated Clˉ channels.Activation of the CACCs by agonist-induced increase in intracellular Ca2+ causes membrane depolarization,vasoconstriction,and inhibition of VSMC proliferation.Activation of VRCCs by cell volume increase or membrane stretch promotes the production of reactive oxygen species,induces proliferation and inhibits apoptosis of VSMCs.Activation of CFTR inhibits oxidative stress and may prevent the development of hypertension.In addition,Clˉ current mediated by gammaaminobutyric acid (GABA) receptor has also been implicated a role in ischemic neuron death.This review focuses on the functional roles of Clˉ channels in the development of stroke and provides a perspective on the future directions for research and the potential to develop Clˉ channels as new targets for the prevention and treatment of stroke.

  14. Physics of Ion Channels

    OpenAIRE

    Kuyucak, Serdar; Bastug, Turgut

    2003-01-01

    We review the basic physics involved in transport of ions across membrane channels in cells. Electrochemical forces that control the diffusion of ions are discussed both from microscopic and macroscopic perspectives. A case is made for use of Brownian dynamics as the minimal phenomenological model that provides a bridge between experiments and more fundamental theoretical approaches. Application of Brownian and molecular dynamics methods to channels with known molecular structures is discussed.

  15. Identification Via Quantum Channels

    OpenAIRE

    Winter, Andreas

    2012-01-01

    We review the development of the quantum version of Ahlswede and Dueck's theory of identification via channels. As is often the case in quantum probability, there is not just one but several quantizations: we know at least two different concepts of identification of classical information via quantum channels, and three different identification capacities for quantum information. In the present summary overview we concentrate on conceptual points and open problems, referring the reader to the ...

  16. Quantum Feedback Channels

    OpenAIRE

    Bowen, Garry

    2002-01-01

    In Shannon information theory the capacity of a memoryless communication channel cannot be increased by the use of feedback. In quantum information theory the no-cloning theorem means that noiseless copying and feedback of quantum information cannot be achieved. In this paper, quantum feedback is defined as the unlimited use of a noiseless quantum channel from receiver to sender. Given such quantum feedback, it is shown to provide no increase in the entanglement--assisted capacities of a memo...

  17. Chaos in quantum channels

    OpenAIRE

    Hosur, Pavan; Qi, Xiao-Liang; Roberts, Daniel; Yoshida, Beni(Institute for Quantum Information & Matter and Walter Burke Institute for Theoretical Physics, California Institute of Technology, 1200 E. California Blvd., Pasadena, CA, 91125, U.S.A.)

    2016-01-01

    We study chaos and scrambling in unitary channels by considering their entanglement properties as states. Using out-of-time-order correlation functions to diagnose chaos, we characterize the ability of a channel to process quantum information. We show that the generic decay of such correlators implies that any input subsystem must have near vanishing mutual information with almost all partitions of the output. Additionally, we propose the negativity of the tripartite information of the channe...

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

    International Nuclear Information System (INIS)

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

  19. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    Energy Technology Data Exchange (ETDEWEB)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  20. Data acquisition system for nuclear reactor environment

    International Nuclear Information System (INIS)

    We have designed an online real time data acquisition system for nuclear reactor environment monitoring. Data acquisition system has eight channels of analog signals and one channel of pulsed input signal from detectors like GM Tube, or any other similar input. Connectivity between the data acquisition system and environmental parameters monitoring computer is made through a wireless data communication link of 151 MHz/100 mW RF power and 10 km maximum communication range for remote data telemetry. Sensors used are gamma ionizing radiation sensor made from CsI:Tl scintillator, atmospheric pressure sensor with +/-0.1 mbar precision, temperature sensor with +/-l milli degree Celsius precision, relative humidity with +/-0.1RH precision, pulse counts with +/-1 count in 0-10000 Hz count rate measurement precision and +/-1 count is accumulated count measurement precision. The entire data acquisition system and wireless telemetry system is 9 V battery powered and the device is to be fitted on a wireless controlled mobile robot for scanning the nuclear reactor zone from remote. Wireless video camera has been planned for integration into the existing system on a later date for moving the robotics environmental data acquisition system beyond human vision reach. System development cost is Rs.25 Lacs and has been developed for Department of Atomic Energy, Government of India and Indian Defense use. (author)

  1. Fast breeder reactor research

    International Nuclear Information System (INIS)

    Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power

  2. Course on Ionic Channels

    CERN Document Server

    1986-01-01

    This book is based on a series of lectures for a course on ionic channels held in Santiago, Chile, on November 17-20, 1984. It is intended as a tutorial guide on the properties, function, modulation, and reconstitution of ionic channels, and it should be accessible to graduate students taking their first steps in this field. In the presentation there has been a deliberate emphasis on the spe­ cific methodologies used toward the understanding of the workings and function of channels. Thus, in the first section, we learn to "read" single­ channel records: how to interpret them in the theoretical frame of kinetic models, which information can be extracted from gating currents in re­ lation to the closing and opening processes, and how ion transport through an open channel can be explained in terms of fluctuating energy barriers. The importance of assessing unequivocally the origin and purity of mem­ brane preparations and the use of membrane vesicles and optical tech­ niques in the stUGY of ionic channels a...

  3. BR2 Reactor: Introduction

    International Nuclear Information System (INIS)

    The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3He, 6Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

  4. Scaleable, High Efficiency Microchannel Sabatier Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — A Microchannel Sabatier Reactor System (MSRS) consisting of cross connected arrays of isothermal or graded temperature reactors is proposed. The reactor array...

  5. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Takeshi; Iida, Masaaki; Moriki, Yasuyuki

    1994-10-18

    A reactor core is divided into a plurality of coolants flowrate regions, and electromagnetic pumps exclusively used for each of the flowrate regions are disposed to distribute coolants flowrates in the reactor core. Further, the flowrate of each of the electromagnetic pumps is automatically controlled depending on signals from a temperature detector disposed at the exit of the reactor core, so that the flowrate of the region can be controlled optimally depending on the burning of reactor core fuels. Then, the electromagnetic pumps disposed for every divided region are controlled respectively, so that the coolants flowrate distribution suitable to each of the regions can be attained. Margin for fuel design is decreased, fuels are used effectively, as well as an operation efficiency can be improved. Moreover, since the electromagnetic pump has less flow resistance compared with a mechanical type pump, and flow resistance of the reactor core flowrate control mechanism is eliminated, greater circulating flowrate can be ensured after occurrence of accident in a natural convection using a buoyancy of coolants utilizable for after-heat removal as a driving force. (N.H.).

  6. Reactor coolant cleanup facility

    International Nuclear Information System (INIS)

    A depressurization device is disposed in pipelines upstream of recycling pumps of a reactor coolant cleanup facility to reduce a pressure between the pressurization device and the recycling pump at the downstream, thereby enabling high pressure coolant injection from other systems by way of the recycling pumps. Upon emergency, the recycling pumps of the coolant cleanup facility can be used in common to an emergency reactor core cooling facility and a reactor shutdown facility. Since existent pumps of the emergency reactor core cooling facility and the reactor shutdown facility which are usually in a stand-by state can be removed, operation confirmation test and maintenance for equipments in both of facilities can be saved, so that maintenance and reliability of the plant are improved and burdens on operators can also be mitigated. Moreover, low pressure design can be adopted for a non-regenerative heat exchanger and recycling coolant pumps, which enables to improve the reliability and economical property due to reduction of possibility of leakage. (N.H.)

  7. EBT reactor analysis

    International Nuclear Information System (INIS)

    This report summarizes the results of a recent ELMO Bumpy Torus (EBT) reactor study that includes ring and core plasma properties with consistent treatment of coupled ring-core stability criteria and power balance requirements. The principal finding is that constraints imposed by these coupling and other physics and technology considerations permit a broad operating window for reactor design optimization. Within this operating window, physics and engineering systems analysis and cost sensitivity studies indicate that reactors with approx. 6 to 10%, P approx. 1200 to 1700 MW(e), wall loading approx. 1.0 to 2.5 MW/m2, and recirculating power fraction (including ring-sustaining power and all other reactors auxiliaries) approx. 10 to 15% are possible. A number of concept improvements are also proposed that are found to offer the potential for further improvement of the reactor size and parameters. These include, but are not limited to, the use of: (1) supplementary coils or noncircular mirror coils to improve magnetic geometry and reduce size, (2) energetic ion rings to improve ring power requirements, (3) positive potential to enhance confinement and reduce size, and (4) profile control to improve stability and overall fusion power density

  8. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

  9. Application of assemblies of in-core instruments of the emergency process instrumentation system. The reactor pressure vessel coolant level sensor in Pressurized-Water Reactors

    International Nuclear Information System (INIS)

    Using as an example the coolant level sensor within SVRD.KNITU assembly, the report deals with a possibility of using assemblies of in-core instruments (SVRD), which ensure the application of in-core monitoring system in normal operation conditions, in the emergency process instrumentation system. Assemblies portrayed in the present report are designed and operated in the Russian-built pressurized water reactors and uranium-graphite channel reactors. However, the philosophy of their design is such that an assembly can be easily adapted to reactors of other types

  10. Reactor Structural Materials: Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R

    2000-07-01

    The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported.

  11. Morphodynamics of Floodplain Chute Channels

    Science.gov (United States)

    David, S. R.; Edmonds, D. A.

    2015-12-01

    Floodplain chute channel formation is a key process that can enable rivers to transition from single-thread to multi-thread planform geometries. Floodplain chute channels are usually incisional channels connecting topographic lows across point bars and in the floodplain. Surprisingly, it is still not clear what conditions promote chute channel formation and what governs their morphodynamic behavior. Towards this end we have initiated an empirical and theoretical study of floodplain chute channels in Indiana, USA. Using elevation models and satellite imagery we mapped 3064 km2 of floodplain in Indiana, and find that 37.3% of mapped floodplains in Indiana have extensive chute channel networks. These chute channel networks consist of two types of channel segments: meander cutoffs of the main channel and chute channels linking the cutoffs together. To understand how these chute channels link meander cutoffs together and eventually create floodplain channel networks we use Delft3D to explore floodplain morphodynamics. Our first modeling experiment starts from a generic floodplain prepopulated with meander cutoffs to test under what conditions chute channels form.We find that chute channel formation is optimized at an intermediate flood discharge. If the flood discharge is too large the meander cutoffs erosively diffuse, whereas if the floodwave is too small the cutoffs fill with sediment. A moderately sized floodwave reworks the sediment surrounding the topographic lows, enhancing the development of floodplain chute channels. Our second modeling experiments explore how floodplain chute channels evolve on the West Fork of the White River, Indiana, USA. We find that the floodplain chute channels are capable of conveying the entire 10 yr floodwave (Q=1330m3/s) leaving the inter-channel areas dry. Moreover, the chute channels can incise into the floodplain while the margins of channels are aggrading, creating levees. Our results suggest that under the right conditions

  12. Mimic of OSU research reactor

    International Nuclear Information System (INIS)

    The Ohio State University research reactor (OSURR) is undergoing improvements in its research and educational capabilities. A computer-based digital data acquisition system, including a reactor system mimic, will be installed as part of these improvements. The system will monitor the reactor system parameters available to the reactor operator either in digital parameters available to the reactor operator either in digital or analog form. The system includes two computers. All the signals are sent to computer 1, which processes the data and sends the data through a serial port to computer 2 with a video graphics array VGA monitor, which is utilized to display the mimic system of the reactor

  13. Methanation assembly using multiple reactors

    Science.gov (United States)

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  14. MINT research reactor safety program

    Energy Technology Data Exchange (ETDEWEB)

    Mohamad Idris bin Taib [Division of Special Project, Malaysian Institute for Nuclear Technology Research (MINT), Bangi (Malaysia)

    2000-11-01

    Malaysian Institute for Nuclear Technology Research (MINT) Research Reactor Safety Program has been done along with Reactor Power Upgrading Project, Reactor Safety Upgrading Project and Development of Expert System for On-Line Nuclear Process Control Project. From 1993 up to date, Neutronic and Thermal-hydraulics analysis, Probabilistic Safety Assessment as well as installation of New 2 MW Secondary Cooling System were done. Installations of New Reactor Building Ventilation System, Reactor Monitoring System, Updating of Safety Analysis Report and Upgrading Primary Cooling System are in progress. For future activities, Reactor Modeling will be included to add present activities. (author)

  15. Thermionic Reactor Design Studies

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1994-08-01

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  16. International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

  17. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  18. Licensed operating reactors

    International Nuclear Information System (INIS)

    THE OPERATING UNITS STATUS REPORT - LICENSED OPERATING REACTORS provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US

  19. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  20. Reactor safety equipments

    International Nuclear Information System (INIS)

    Purpose: To positively recover radioactive substances discharged in a dry well at the time of failure of a reactor. Constitution: In addition to the emergency gas treating system fitted to a reactor building, a purification system connected through a pipeline to the dry well is arranged in the reactor building. This purification system is connected through pipes fitted to the dry well to forced circulation device, heat exchanger, and purification device. The atmosphere of high pressure steam gases in the dry well is derived to the heat exchanger for cooling, and then radioactive substances which are contained in the gases are removed by filter sets charged with the HEPA filters and the HECA filters. At last, there gases are returned to dry well by circulation pump, repeat this process. (Kamimura, M.)

  1. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  2. Welding and reactor safety

    International Nuclear Information System (INIS)

    The high safety requirements which must be demanded of the quality of the welded joints in reactor technique have so far not been fulfilled in all cases. The errors occuring have caused considerable loss of availability and high material costs. They were not, however, so serious that one need have feared any immediate danger to the personnel or to the environment. The safety devices of reactor plants were only called upon in a few cases and to these they responded perfectly. The intensive efforts to complete and improve the specifications are to contribute to that in future, the reactor plants can be counted even more so as one of the safest technical plants ever. (orig./LH)

  3. Backfitting swimming pool reactors

    International Nuclear Information System (INIS)

    Calculations based on measurements in a critical assembly, and experiments to disclose fuel element surface temperatures in case of accidents like stopping of primary coolant flow during full power operation, have shown that the power of the swimming pool type research reactor FRG-2 (15 MW, operating since 1967) might be raised to 21 MW within the present rules of science and technology, without major alterations of the pool buildings and the cooling systems. A backfitting program is carried through to adjust the reactor control systems of FRG-2 and FRG-1 (5 MW, housed in the same reactor hall) to the present safety rules and recommendations, to ensure FRG-2 operation at 21 MW for the next decade. (author)

  4. Reactor operation experience

    International Nuclear Information System (INIS)

    Since the TRIGA Users Conference in Helsinki 1970 the TRIGA reactor Vienna was in operation without any larger undesired shutdown. The integrated thermal power production by August 15 1972 accumulated to 110 MWd. The TRIGA reactor is manly used for training of students, for scientific courses and research work. Cooperation with industry increased in the last two years either in form of research or in performing training courses. Close cooperation is also maintained with the IAEA, samples are irradiated and courses on various fields are arranged. Maintenance work was performed on the heat exchanger and to replace the shim rod magnet. With the view on the future power upgrading nine fuel elements type 110 have been ordered recently. Experiments, performed currently on the reactor are presented in details

  5. The MNSR reactor

    International Nuclear Information System (INIS)

    This tank-in-pool reactor is based on the same design concept as the Canadian Slowpoke. The core is a right circular cylinder, 24 cm diameter by 25 cm long, containing 411 fuel pin positions. The pins are HEU-Aluminium alloy, 0.5 cm in diameter. Critical mass is about 900 g. The reactor has a single cadmium control rod. The back-up shutdown system is the insertion of a cadmium capsule in a core position. Excess reactivity is limited to 3.5mk. In both the MNSR and Slowpoke, the insertion of the maximum excess reactivity results in a power transient limited by the coolant/moderator temperature to safe values, independent of any operator action. This reactor is used primarily in training and neutron activation analysis. Up to 64 elements have been analyzed in a great variety of different disciplines. (author)

  6. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  7. Fusion reactor materials

    International Nuclear Information System (INIS)

    This is the fifteenth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: Alloy Development for Irradiation Performance; Damage Analysis and Fundamental Studies; Special purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the U.S. Department of Energy. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide

  8. Safety systems of heavy water reactors and small power reactors

    International Nuclear Information System (INIS)

    After introductional descriptions of heavy water reactors and natural circulation boiling water reactors the safety philosophy and safety systems like ECCS, residual heat removal, protection systems etc., are described. (RW)

  9. Operation and maintenance of the RA reactor, RA Research reactor. Annual report 1976

    International Nuclear Information System (INIS)

    During 1976 the Ra reactor was operating for about 30% shorter period than usual. The reason were extraordinary repair activities within regular and investment maintenance as well as repair of failures caused by neglected maintenance during previous 6 years. Delay was caused by unavailability of fuel (2% enriched fuel elements are spent) and the new 80% enriched fuel demanded experimental and theoretical analyses before being introduced into the core. Safety analyses concerned with using 80% enriched fuel both experimental and theoretical were successfully fulfilled. The December 1976 successful experimental campaign can be marked as end of the 17 years period of using 2% enriched fuel and start of the new period of using highly enriched fuel. This is significant not only for the reactor itself but for the users, because it would result in increase of neutron flux by 50% with the increase of costs by only 4%. Demand was submitted for obtaining the final license for transition operating regime with highly enriched fuel which would save at least 2 200 000 dinars. This will enable reactor operation in 1977 and later on, without interruption by 'critical' and other experiments related to new highly enriched fuel. A high number of repair and other urgent activities were fulfilled in order to enable safe operation. Some of these activities were done never before and some were neglected during past 6 years. The most important tasks were: purchase of Al tubes made of special alloy, fabrication and mounting of the fuel channel; overall investigation of reactor vessel leakage; repair of the heavy water pump; exchange of two vertical channels. basic equipment for construction of emergency cooling system was purchased. Hot cells are equipped for independent utilisation

  10. Zirconium ignition in exposed fuel channel

    Energy Technology Data Exchange (ETDEWEB)

    Elias, E., E-mail: merezra@technion.ac.il; Hasan, D.; Nekhamkin, Y.

    2015-05-15

    Highlights: • We demonstrate the idea of runaway zirconium–steam reactions in severe accidents in today's LWRs. • We predict the thermal-hydraulics conditions relevant to cladding oxidation in an exposed fuel channel of a partially uncovered core. • The Semenov theory of metal combustion is extended to define a criterion for runaway oxidation reaction in fuel cladding. - Abstract: A theoretical model based on simultaneous solution of the heat and mass transfer equations is developed for predicting the rate of thermo-chemical reaction between zirconium cladding and a hot steam environment. Ignition conditions relevant to cladding oxidation in an exposed fuel channel of a partially uncovered core are predicted based on the theory of metal combustion. A range of decay power, convective heat transfer coefficients, and initial temperatures leading to uncontrolled runaway cladding oxidation is identified. The model could be readily integrated as part of a fuel channel analysis code for predicting possible outcomes of different accident mitigation procedures in light water nuclear reactors under LOCA conditions.

  11. AREVA's nuclear reactors portfolio

    International Nuclear Information System (INIS)

    A reasonable assumption for the estimated new build market for the next 25 years is over 340 GWe net. The number of prospect countries is growing almost each day. To address this new build market, AREVA is developing a comprehensive portfolio of reactors intended to meet a wide range of power requirements and of technology choices. The EPR reactor is the flagship of the fleet. Intended for large power requirements, the four first EPRs are being built in Finland, France and China. Other countries and customers are in view, citing just two examples: the Usa where the U.S. EPR has been selected as the technology of choice by several U.S utilities; and the United Kingdom where the Generic Design Acceptance process of the EPR design submitted by AREVA and EDF is well under way, and where there is a strong will to have a plant on line in 2017. For medium power ranges, the AREVA portfolio includes a boiling water reactor and a pressurized water reactor which both offer all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation cost: -) KERENA (1250+ MWe), developed in collaboration with several European utilities, and in particular with Eon; -) ATMEA 1 (1100+ MWe), a 3-loop evolutionary PWR which is being developed by AREVA and Mitsubishi. AREVA is also preparing the future and is deeply involved into Gen IV concepts. It has developed the ANTARES modular HTR reactor (pre-conceptual design completed) and is building upon its vast Sodium Fast Reactor experience to take part into the development of the next prototype. (author)

  12. Oscillatory flow chemical reactors

    Directory of Open Access Journals (Sweden)

    Slavnić Danijela S.

    2014-01-01

    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  13. Fluid flow characteristics of string reactors packed with spherical particles

    Energy Technology Data Exchange (ETDEWEB)

    Vonortas, A.; Papayannakos, N. [National Technical University of Athens, School of Chemical Engineering, Zografos, Athens (Greece); Hipolito, A.; Rolland, M.; Boyer, C. [Institut Francais du Petrole, 69360 Solaize (France)

    2011-02-15

    A hydrodynamic investigation of three geometries of string pellet reactors filled with spheres was conducted. Two geometries were circular spiral channels, while the third was a straight horizontal square channel. Stimulus-response experiments provided data for residence time distribution analysis from which Pe numbers and liquid holdup were deduced. Flow regimes and transitions were determined from visual observations through the transparent tube wall. For the whole range of the experimental conditions applied in this work and for all reactors, the ratio of gas to liquid velocities, V{sub g}/V{sub l}, is a controlling parameter for Pe number, holdup, and pressure drop. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  14. Detecting, locating and identifying failed fuel in Canadian power reactors

    International Nuclear Information System (INIS)

    This document summarizes how defected fuel elements are detected, located and identified in Canadian CANDU power reactors. Fuel defects are detected by monitoring the primary coolant for gaseous fission products and radioiodines, while location in core is usually performed on-power by delayed neutron monitoring of coolant samples from individual fuel channels or off-power by gamma-ray monitoring of the channel feeder pipes. The systems and techniques used to detect and locate defected fuel in both Ontario Hydro and CANDU 6 power stations are described, along with examples provided by station experience. The ability to detect and locate defected fuel in power stations was greatly enhanced by a fundamental R and D program, which provided an understanding and models of fission-product release and transport, and the post-defect deterioration of failed fuel. Techniques and equipment used to identify and store defected fuel after it has been discharged from the reactor are briefly reviewed

  15. Reactor Materials Research

    International Nuclear Information System (INIS)

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel

  16. Nuclear reactor simulator

    International Nuclear Information System (INIS)

    The Nuclear Reactor Simulator was projected to help the basic training in the formation of the Nuclear Power Plants operators. It gives the trainee the opportunity to see the nuclear reactor dynamics. It's specially indicated to be used as the support tool to NPPT (Nuclear Power Preparatory Training) from NUS Corporation. The software was developed to Intel platform (80 x 86, Pentium and compatible ones) working under the Windows operational system from Microsoft. The program language used in development was Object Pascal and the compiler used was Delphi from Borland. During the development, computer algorithms were used, based in numeric methods, to the resolution of the differential equations involved in the process. (author)

  17. Diagnostics for hybrid reactors

    International Nuclear Information System (INIS)

    The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

  18. Small mirror fusion reactors

    International Nuclear Information System (INIS)

    Basic requirements for the pilot plants are that they produce a net product and that they have a potential for commercial upgrade. We have investigated a small standard mirror fusion-fission hybrid, a two-component tandem mirror hybrid, and two versions of a field-reversed mirror fusion reactor--one a steady state, single cell reactor with a neutral beam-sustained plasma, the other a moving ring field-reversed mirror where the plasma passes through a reaction chamber with no energy addition

  19. Reactor neutron dosimetry

    International Nuclear Information System (INIS)

    An analysis of requirements and possibilities for experimental neutron spectrum determination during the reactor pressure vessel surveil lance programme is given. Fast neutron spectrum and neutron dose rate were measured in the Fast neutron irradiation facility of our TRIGA reactor. It was shown that the facility can be used for calibration of neutron dosimeters and for irradiation of samples sensitive to neutron radiation. The investigation of the unfolding algorithm ITER was continued. Based on this investigations are two specialized unfolding program packages ITERAD and ITERGS written this year. They are able to unfold data from activation detectors and NaI(T1) gamma spectrometer respectively

  20. Perspectives on reactor safety

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

  1. Perspectives on reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  2. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    A nuclear reactor construction comprising a reactor core submerged in a pool of liquid metal coolant in a primary vessel which is suspended from the roof structure of a containment vault. Control rods supported from the roof structure are insertable in the core which is carried on a support structure from the wall of the primary vessel. To prevent excessive relaxation of the support structure whereby the control rods would be displaced relative to the core, the support structure incorporates a normally inactive secondary structure designed to become effective in bracing the primary structure against further relaxation beyond a predetermined limit. (author)

  3. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2000-07-01

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised.

  4. Reactor gamma spectrometry: status

    International Nuclear Information System (INIS)

    Current work is described for Compton Recoil Gamma-Ray Spectrometry including developments in experimental technique as well as recent reactor spectrometry measurements. The current status of the method is described concerning gamma spectromoetry probe design and response characteristics. Emphasis is given to gamma spectrometry work in US LWR and BR programs. Gamma spectrometry in BR environments are outlined by focussing on start-up plans for the Fast Test Reactor (FTR). Gamma spectrometry results are presented for a LWR pressure vessel mockup in the Poolside Critical Assembly (PCA) at Oak Ridge National Laboratory

  5. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  6. Experimental study of prompt neutron decay constant α for 300# pool reactor under mixed core

    International Nuclear Information System (INIS)

    The experimental study of prompt neutron decay constant α for 300# pool reactor under mixed core was carried out through a suit of reactor power spectral density measurement system. The two channel continuous current signals of neutron in the reactor were acquired by ionization chamber DL129 which was symmetrically putted in reactor core. The power spectral density, for two channel signals, was computed using the application program of data acquirement and data process analysis. Finally, by using the non-linear least squares method, the prompt neutron decay constant α was fitted. By comparison, the experimental results well accord to the theory calculation within the error range. The deviation can meet the actual need of project. (authors)

  7. MEMS in microfluidic channels.

    Energy Technology Data Exchange (ETDEWEB)

    Ashby, Carol Iris Hill; Okandan, Murat; Michalske, Terry A.; Sounart, Thomas L.; Matzke, Carolyn M.

    2004-03-01

    Microelectromechanical systems (MEMS) comprise a new class of devices that include various forms of sensors and actuators. Recent studies have shown that microscale cantilever structures are able to detect a wide range of chemicals, biomolecules or even single bacterial cells. In this approach, cantilever deflection replaces optical fluorescence detection thereby eliminating complex chemical tagging steps that are difficult to achieve with chip-based architectures. A key challenge to utilizing this new detection scheme is the incorporation of functionalized MEMS structures within complex microfluidic channel architectures. The ability to accomplish this integration is currently limited by the processing approaches used to seal lids on pre-etched microfluidic channels. This report describes Sandia's first construction of MEMS instrumented microfluidic chips, which were fabricated by combining our leading capabilities in MEMS processing with our low-temperature photolithographic method for fabricating microfluidic channels. We have explored in-situ cantilevers and other similar passive MEMS devices as a new approach to directly sense fluid transport, and have successfully monitored local flow rates and viscosities within microfluidic channels. Actuated MEMS structures have also been incorporated into microfluidic channels, and the electrical requirements for actuation in liquids have been quantified with an elegant theory. Electrostatic actuation in water has been accomplished, and a novel technique for monitoring local electrical conductivities has been invented.

  8. Mitochondrial Ion Channels

    Science.gov (United States)

    O’Rourke, Brian

    2009-01-01

    In work spanning more than a century, mitochondria have been recognized for their multifunctional roles in metabolism, energy transduction, ion transport, inheritance, signaling, and cell death. Foremost among these tasks is the continuous production of ATP through oxidative phosphorylation, which requires a large electrochemical driving force for protons across the mitochondrial inner membrane. This process requires a membrane with relatively low permeability to ions to minimize energy dissipation. However, a wealth of evidence now indicates that both selective and nonselective ion channels are present in the mitochondrial inner membrane, along with several known channels on the outer membrane. Some of these channels are active under physiological conditions, and others may be activated under pathophysiological conditions to act as the major determinants of cell life and death. This review summarizes research on mitochondrial ion channels and efforts to identify their molecular correlates. Except in a few cases, our understanding of the structure of mitochondrial ion channels is limited, indicating the need for focused discovery in this area. PMID:17059356

  9. Channel Identification Machines

    Directory of Open Access Journals (Sweden)

    Aurel A. Lazar

    2012-01-01

    Full Text Available We present a formal methodology for identifying a channel in a system consisting of a communication channel in cascade with an asynchronous sampler. The channel is modeled as a multidimensional filter, while models of asynchronous samplers are taken from neuroscience and communications and include integrate-and-fire neurons, asynchronous sigma/delta modulators and general oscillators in cascade with zero-crossing detectors. We devise channel identification algorithms that recover a projection of the filter(s onto a space of input signals loss-free for both scalar and vector-valued test signals. The test signals are modeled as elements of a reproducing kernel Hilbert space (RKHS with a Dirichlet kernel. Under appropriate limiting conditions on the bandwidth and the order of the test signal space, the filter projection converges to the impulse response of the filter. We show that our results hold for a wide class of RKHSs, including the space of finite-energy bandlimited signals. We also extend our channel identification results to noisy circuits.

  10. Risk prevention during reactor shutdown

    International Nuclear Information System (INIS)

    During reactor shutdown potential risks are issued of a number of maintenance operations. In this text we analyse these operations and give the modifications of technical specifications to ameliorate the reactor safety. 4 figs

  11. Power calibrations for TRIGA reactors

    International Nuclear Information System (INIS)

    The purpose of this paper is to establish a framework for the calorimetric power calibration of TRIGA reactors so that reliable results can be obtained with a precision better than ± 5%. Careful application of the same procedures has produced power calibration results that have been reproducible to ± 1.5%. The procedures are equally applicable to the Mark I, Mark II and Mark III reactors as well as to reactors having much larger reactor tanks and to TRIGA reactors capable of forced cooling up to 3 MW in some cases and 15 MW in another case. In the case of forced cooled TRIGA reactors, the calorimetric power calibration is applicable in the natural convection mode for these reactors using exactly the same procedures as are discussed below for the smaller TRIGA reactors (< 2 MW)

  12. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1982 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Since fiscal 1982, Systematic research and development work on safeguards technology has been added to the activities of the Department. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  13. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

  14. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  15. High Flux Isotope Reactor (HFIR)

    Data.gov (United States)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  16. Shielding design for research and education reactor

    International Nuclear Information System (INIS)

    For the purpose of education and research at the University, 20-KW powered SLOWPOKE-2 research reactor has been chosen as a prototype reactor. In order to study the safety characteristics of the reactor, exposure rate has been estimated at the pool boundary. Reactor core as a radiation source is assumed to be cylindrical volume source. Thus point kernel integration method can be applied to determine the exposure rate. For the sake of simplicity, calculation was done only for the prompt fission gamma rays and fission product gamma rays. As a result, the maximum exposure rate at the pool boundary was estimated to be 18R/min at the same height of the center of the core. In order to examine the accuracy for the point kernel integration method, two shielding experiments were carried out: one for the water tank only and the other for with concrete blocks outside the water tank. Water tank was made of wood pieces which is 13.4cm wide, 1.5cm thick and 2.15m long. Thus the water tank has the total dimension of 1 m radius and 2.1 m height. The experiment was carried out for the radiation source of 0.968 mCi Co-60 at the center of the water tank and the penetrated gamma rays were measured at 5 different detector positions. For the measurement and analysis of the responses, NaI(T1) 3''x3'' detector and 256 channel multichannel analyzer was utilized. To convert pulse height distribution to the exposure rate, Moriuchi conversion factor was adopted. Data from the calculations by point kernel method were well agreed within 10% band with the data from the the experiments. (Author)

  17. Computational analysis of irradiation facilities at the JSI TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Snoj, Luka, E-mail: luka.snoj@ijs.si [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Zerovnik, Gasper, E-mail: gasper.zerovnik@ijs.si [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Trkov, Andrej, E-mail: andrej.trkov@ijs.si [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2012-03-15

    Characterization and optimization of irradiation facilities in a research reactor is important for optimal performance. Nowadays this is commonly done with advanced Monte Carlo neutron transport computer codes such as MCNP. However, the computational model in such calculations should be verified and validated with experiments. In the paper we describe the irradiation facilities at the JSI TRIGA reactor and demonstrate their computational characterization to support experimental campaigns by providing information on the characteristics of the irradiation facilities. - Highlights: Black-Right-Pointing-Pointer TRIGA reactor at JSI suitable for irradiation under well defined conditions. Black-Right-Pointing-Pointer It features irradiation channels of different fluxes, spectra, and dimensions. Black-Right-Pointing-Pointer Computational model has been developed and experimentally verified. Black-Right-Pointing-Pointer The model used for optimization of experiments and evaluation of uncertainties.

  18. Computational intelligent systems for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Nearly 15000 process signals are digitized by physically and functionally distributed embedded systems in Prototype Fast Breeder Reactor (PFBR). Digitized signals are processed and relevant information is displayed through Large video display systems at Control Room. It is necessary that correct and reliable information need to be provided to the plant operator. Computational intelligent systems play a major role in enhancing the safe operation of the Nuclear reactor. The paper explains the features of three such systems, one for on-line validation of neutronic power channel through on-line thermal balance calculation and another for detection of anomalous reactivity addition through on-line reactivity balance computation and third for on-line computation of Reactor power from fluctuations of core thermocouple signals. (author)

  19. QKD Quantum Channel Authentication

    CERN Document Server

    Kosloski, J T

    2006-01-01

    Several simple yet secure protocols to authenticate the quantum channel of various QKD schemes, by coupling the photon sender's knowledge of a shared secret and the QBER Bob observes, are presented. It is shown that Alice can encrypt certain portions of the information needed for the QKD protocols, using a sequence whose security is based on computational-complexity, without compromising all of the sequence's entropy. It is then shown that after a Man-in-the-Middle attack on the quantum and classical channels, there is still enough entropy left in the sequence for Bob to detect the presence of Eve by monitoring the QBER. Finally, it is shown that the principles presented can be implemented to authenticate the quantum channel associated with any type of QKD scheme, and they can also be used for Alice to authenticate Bob.

  20. Dequantization Via Quantum Channels

    Science.gov (United States)

    Andersson, Andreas

    2016-10-01

    For a unital completely positive map {Φ} ("quantum channel") governing the time propagation of a quantum system, the Stinespring representation gives an enlarged system evolving unitarily. We argue that the Stinespring representations of each power {Φ^m} of the single map together encode the structure of the original quantum channel and provide an interaction-dependent model for the bath. The same bath model gives a "classical limit" at infinite time {mto∞} in the form of a noncommutative "manifold" determined by the channel. In this way, a simplified analysis of the system can be performed by making the large- m approximation. These constructions are based on a noncommutative generalization of Berezin quantization. The latter is shown to involve very fundamental aspects of quantum-information theory, which are thereby put in a completely new light.

  1. Dequantization Via Quantum Channels

    Science.gov (United States)

    Andersson, Andreas

    2016-08-01

    For a unital completely positive map {Φ} ("quantum channel") governing the time propagation of a quantum system, the Stinespring representation gives an enlarged system evolving unitarily. We argue that the Stinespring representations of each power {Φ^m} of the single map together encode the structure of the original quantum channel and provide an interaction-dependent model for the bath. The same bath model gives a "classical limit" at infinite time {mto∞} in the form of a noncommutative "manifold" determined by the channel. In this way, a simplified analysis of the system can be performed by making the large-m approximation. These constructions are based on a noncommutative generalization of Berezin quantization. The latter is shown to involve very fundamental aspects of quantum-information theory, which are thereby put in a completely new light.

  2. Entanglement-saving channels

    Science.gov (United States)

    Lami, L.; Giovannetti, V.

    2016-03-01

    The set of Entanglement Saving (ES) quantum channels is introduced and characterized. These are completely positive, trace preserving transformations which when acting locally on a bipartite quantum system initially prepared into a maximally entangled configuration, preserve its entanglement even when applied an arbitrary number of times. In other words, a quantum channel ψ is said to be ES if its powers ψn are not entanglement-breaking for all integers n. We also characterize the properties of the Asymptotic Entanglement Saving (AES) maps. These form a proper subset of the ES channels that is constituted by those maps that not only preserve entanglement for all finite n but which also sustain an explicitly not null level of entanglement in the asymptotic limit n → ∞. Structure theorems are provided for ES and for AES maps which yield an almost complete characterization of the former and a full characterization of the latter.

  3. Chaos in quantum channels

    CERN Document Server

    Hosur, Pavan; Roberts, Daniel A; Yoshida, Beni

    2015-01-01

    We study chaos and scrambling in unitary channels by considering their entanglement properties as states. Using out-of-time-order correlation functions to diagnose chaos, we characterize the ability of a channel to process quantum information. We show that the generic decay of such correlators implies that any input subsystem must have near vanishing mutual information with almost all partitions of the output. Additionally, we propose the negativity of the tripartite information of the channel as a general diagnostic of scrambling. This measures the delocalization of information and is closely related to the decay of out-of-time-order correlators. We back up our results with numerics in two non-integrable models and analytic results in a perfect tensor network model of chaotic time evolution. These results show that the butterfly effect in quantum systems implies the information-theoretic definition of scrambling.

  4. Voltage-Gated Proton Channels

    OpenAIRE

    DeCoursey, Thomas E.

    2008-01-01

    Voltage-gated proton channels, HV1, have vaulted from the realm of the esoteric into the forefront of a central question facing ion channel biophysicists, namely the mechanism by which voltage-dependent gating occurs. This transformation is the result of several factors. Identification of the gene in 2006 revealed that proton channels are homologues of the voltage-sensing domain of most other voltage-gated ion channels. Unique, or at least eccentric, properties of proton channels include dime...

  5. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  6. Reactor safety in Eastern Europe

    International Nuclear Information System (INIS)

    The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. All papers are indexed separately in report GRS--117. (HP)

  7. Some questions on nuclear safety of heavy-water power reactor operating in self-sufficient thorium cycle

    Directory of Open Access Journals (Sweden)

    Bergelson Boris R.

    2008-01-01

    Full Text Available In this paper the comparative calculations of the void coefficient have been made for different types of channel reactors for the coolant density interval 0.8-0.01 g/cm3. These results demonstrate the following. In heavy-water channel reactors, the replacement of D2O coolant by H2O, ensuring significant economic advantage, leads to the essential reducing of nuclear safety of an installation. The comparison of different reactors by the void coefficient demonstrates that at the dehydration of channels the reactivity increase is minimal for HWPR(Th, operating in the self-sufficient mode. The reduction of coolant density in channels in most cases is accompanied by the increase of power and temperatures of fuel assemblies. The calculations show that the reduction of reactivity due to Doppler effect can compensate the effect of dehydration of a channel. However, the result depends on the time dependency of heat-hydraulic processes, occurring in reactor channels in the specific accident. The result obtained in the paper confirms that nuclear safety of HWPR(Th lies on the same level as nuclear safety of CANDU type reactors approved in practice.

  8. The neutron channeling phenomenon.

    Science.gov (United States)

    Khanouchi, A; Sabir, A; Boulkheir, M; Ichaoui, R; Ghassoun, J; Jehouani, A

    1997-01-01

    Shields, used for protection against radiation, are often pierced with vacuum channels for passing cables and other instruments for measurements. The neutron transmission through these shields is an unavoidable phenomenon. In this work we study and discuss the effect of channels on neutron transmission through shields. We consider an infinite homogeneous slab, with a fixed thickness (20 lambda, with lambda the mean free path of the neutron in the slab), which contains a vacuum channel. This slab is irradiated with an infinite source of neutrons on the left side and on the other side (right side) many detectors with windows equal to 2 lambda are placed in order to evaluate the neutron transmission probabilities (Khanouchi, A., Aboubekr, A., Ghassoun, J. and Jehouani, A. (1994) Rencontre Nationale des Jeunes Chercheurs en Physique. Casa Blanca Maroc; Khanouchi, A., Sabir, A., Ghassoun, J. and Jehouani, A. (1995) Premier Congré International des Intéractions Rayonnements Matière. Eljadida Maroc). The neutron history within the slab is simulated by the Monte Carlo method (Booth, T. E. and Hendricks, J. S. (1994) Nuclear Technology 5) and using the exponential biasing technique in order to improve the Monte Carlo calculation (Levitt, L. B. (1968) Nuclear Science and Engineering 31, 500-504; Jehouani, A., Ghassoun, J. and Aboubker, A. (1994) In Proceedings of the 6th International Symposium on Radiation Physics, Rabat, Morocco). Then different geometries of the vacuum channel have been studied. For each geometry we have determined the detector response and calculated the neutron transmission probability for different detector positions. This neutron transmission probability presents a peak for the detectors placed in front of the vacuum channel. This study allowed us to clearly identify the neutron channeling phenomenon. One application of our study is to detect vacuum defects in materials. PMID:9463884

  9. Scenarios for the transmutation of actinides in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, Bronwyn, E-mail: hylandb@aecl.ca [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada); Gihm, Brian, E-mail: gihmb@aecl.ca [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

    2011-12-15

    With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU(R) reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past. The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100-1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.

  10. Fuel condition in Canadian CANDU 6 reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  11. Fuel condition in Canadian CANDU 6 reactors

    International Nuclear Information System (INIS)

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

  12. BLIND CHANNEL ESTIMATION IN DELAY DIVERSITY FOR FREQUENCY SELECTIVE CHANNELS

    Institute of Scientific and Technical Information of China (English)

    Zhao Zheng; Jia Ying; Yin Qinye

    2003-01-01

    Delay diversity is an effective transmit diversity technique to combat adverse ef-fects of fading. Thus far, previous work in delay diversity assumed that perfect estimates ofcurrent channel fading conditions are available at the receiver and training symbols are requiredto estimate the channel from the transmitter to the receiver. However, increasing the number ofthe antennas increases the required training interval and reduces the available time within whichdata may be transmitted. Learning the channel coefficients becomes increasingly difficult for thefrequency selective channels. In this paper, with the subspace method and the delay character ofdelay diversity, a channel estimation method is proposed, which does not use training symbols. Itaddresses the transmit diversity for a frequency selective channel from a single carrier perspectivein the form of a simple equivalent fiat fading model. Monte Carlo simulations give the perfor-mance of channel estimation and the performance comparison of our channel-estimation-baseddetector with decision feedback equalization, which uses the perfect channel information.

  13. K(v)7 channels: function, pharmacology and channel modulators.

    Science.gov (United States)

    Dalby-Brown, William; Hansen, Henrik H; Korsgaard, Mads P G; Mirza, Naheed; Olesen, Søren-P

    2006-01-01

    K(v)7 channels are unique among K(+) channels, since four out of the five channel subtypes have well-documented roles in the development of human diseases. They have distinct physiological functions in the heart and in the nervous system, which can be ascribed to their voltage-gating properties. The K(v)7 channels also lend themselves to pharmacological modulation, and synthetic openers as well as blockers of the channels, regulating neuronal excitability, have existed even before the K(v)7 channels were identified by cloning. In the present review we give an account on the focused efforts to develop selective modulators, openers as well as blockers, of the K(v)7 channel subtypes, which have been undertaken during recent years, along with a discussion of the K(v)7 ion channel physiology and therapeutic indications for modulators of the neuronal K(v)7 channels.

  14. Towards a sustainable future using pressure tube reactor technology

    International Nuclear Information System (INIS)

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

  15. Reactor Cavity in Case of Station Blackout in RBMK-1500

    Directory of Open Access Journals (Sweden)

    Algirdas Kaliatka

    2007-01-01

    Full Text Available Ignalina NPP is equipped with channel-type boiling-water graphite-moderated reactor RBMK-1500. Results of the level-1 probabilistic safety assessment of the Ignalina NPP have shown that in topography of the risk, the transients with failure of long-term core cooling other than LOCA are the main contributors to the core damage frequency. The total loss of off-site power with a failure to start any diesel generator, that is station blackout, is the event which could lead to the loss of long-term core cooling. Such accident could lead to multiple ruptures of fuel channels with severe consequences and should be analyzed in order to estimate the timing of the key events and the possibilities for accident management. This paper presents the results of the analysis of station blackout at Ignalina NPP. Analysis was performed using thermal-hydraulic state-of-the-art RELAP5/MOD3.2 code. The response of reactor cooling system and the processes in the reactor cavity and its venting system in case of a few fuel-channel ruptures due to overheating were demonstrated. The possible measures for prevention of the development of this beyond design basis accident (BDBA to a severe accident are discussed.

  16. A CANDU-type small/medium power reactor

    International Nuclear Information System (INIS)

    The assembly known as a CANDU power reactor consists of a number of standardized fuel channels or 'power modules'. Each of these channels produces about 5 thermal megawatts on average. Within practical limitations on fuel enrichment and ultimately on economics, the number of these channels is variable between about 50 and approximately 700. Small reactors suffer from inevitable disadvantages in terms of specific cost of design/construction as well as operating cost. Their natural 'niche' for application is in remote off-grid locations. At the same time this niche application imposes new and strict requirements for staff complement, power system reliability, and so on. The distinct advantage of small reactors arises if the market requires installation of several units in a coordinated installation program - a feature well suited to power requirements in Canada's far North. This paper examines several of the performance requirements and constraints for installation of these plants and presents means for designers to overcome the consequent negative feasibility factors.

  17. Fast reactor programme

    International Nuclear Information System (INIS)

    This progress report summarizes the fast reactor research carried out by ECN during the period covering the year 1980. This research is mainly concerned with the cores of sodium-cooled breeders, in particular the SNR-300, and its related safety aspects. It comprises six items: A programme to determine relevant nuclear data of fission- and corrosion-products; A fuel performance programme comprising in-pile cladding failure experiments and a study of the consequences of loss-of-cooling and overpower; Basic research on fuel; Investigation of the changes in the mechanical properties of austenitic stainless steel DIN 1.4948 due to fast neutron doses, this material has been used in the manufacture of the reactor vessel and its internal components; Study of aerosols which could be formed at the time of a fast reactor accident and their progressive behaviour on leaking through cracks in the concrete containment; Studies on heat transfer in a sodium-cooled fast reactor core. As fast breeders operate at high power densities, an accurate knowledge of the heat transfer phenomena under single-phase and two-phase conditions is sought. (Auth.)

  18. Thermal Reactor Safety

    International Nuclear Information System (INIS)

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods

  19. NEUTRONIC REACTOR FUEL ELEMENT

    Science.gov (United States)

    Picklesimer, M.L.; Thurber, W.C.

    1961-01-01

    A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

  20. Stabilized Spheromak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2007-04-03

    The U.S. fusion energy program is focused on research with the potential for studying plasmas at thermonuclear temperatures, currently epitomized by the tokamak-based International Thermonuclear Experimental Reactor (ITER) but also continuing exploratory work on other plasma confinement concepts. Among the latter is the spheromak pursued on the SSPX facility at LLNL. Experiments in SSPX using electrostatic current drive by coaxial guns have now demonstrated stable spheromaks with good heat confinement, if the plasma is maintained near a Taylor state, but the anticipated high current amplification by gun injection has not yet been achieved. In future experiments and reactors, creating and maintaining a stable spheromak configuration at high magnetic field strength may require auxiliary current drive using neutral beams or RF power. Here we show that neutral beam current drive soon to be explored on SSPX could yield a compact spheromak reactor with current drive efficiency comparable to that of steady state tokamaks. Thus, while more will be learned about electrostatic current drive in coming months, results already achieved in SSPX could point to a productive parallel development path pursuing auxiliary current drive, consistent with plans to install neutral beams on SSPX in the near future. Among possible outcomes, spheromak research could also yield pulsed fusion reactors at lower capital cost than any fusion concept yet proposed.

  1. Reactors. Nuclear propulsion ships

    International Nuclear Information System (INIS)

    This article has for object the development of nuclear-powered ships and the conception of the nuclear-powered ship. The technology of the naval propulsion P.W.R. type reactor is described in the article B.N.3 141 'Nuclear Boilers ships'. (N.C.)

  2. Pressure tube type reactor

    International Nuclear Information System (INIS)

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

  3. Fusion reactor materials

    International Nuclear Information System (INIS)

    At the Belgian Nuclear Research Centre SCK-CEN, activities related to fusion focus on environmental tolerance of opto-electronic components. The objective of this program is to contribute to the knowledge on the behaviour, during and after neutron irradiation, of fusion-reactor materials and components. The main scientific activities for 1997 are summarized

  4. Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R ampersand D

  5. Studies on reactor physics

    International Nuclear Information System (INIS)

    Most of the peaceful applications of atomic energy are inherently dependent on advances in the science and technology of nuclear reactors, and aspects of this development are part of a major programme of the International Atomic Energy Agency. The most useful role that the Agency can play is as a co-ordinating body or central forum where the trends can be reviewed and the results assessed. Some of the basic studies are carried out by members of the Agency's own scientific staff. The Agency also convenes groups of experts from different countries to examine a particular problem in detail and make any necessary recommendations. Some of the important subjects are discussed at international scientific meetings held by the Agency. One of the subjects covered by such studies is the physics of nuclear reactors and a specific topic recently discussed was Codes for Reactor Computations, on which a seminar was held in Vienna in April this year. Another The members of the Panel described the development of heavy water reactors, the equipment and methods of research currently used, and plans for further development in their respective countries meeting of Panel of Experts on Heavy Water Lattices was held in Vienna in August 1959

  6. Nuclear power reactor physics

    International Nuclear Information System (INIS)

    The purpose of this book is to explain the physical working conditions of nuclear reactors for the benefit of non-specialized engineers and engineering students. One of the leading ideas of this course is to distinguish between two fundamentally different concepts: - a science which could be called neutrodynamics (as distinct from neutron physics which covers the knowledge of the neutron considered as an elementary particle and the study of its interactions with nuclei); the aim of this science is to study the interaction of the neutron gas with real material media; the introduction will however be restricted to its simplified expression, the theory and equation of diffusion; - a special application: reactor physics, which is introduced when the diffusing and absorbing material medium is also multiplying. For this reason the chapter on fission is used to introduce this section. In practice the section on reactor physics is much longer than that devoted to neutrodynamics and it is developed in what seemed to be the most relevant direction: nuclear power reactors. Every effort was made to meet the following three requirements: to define the physical bases of neutron interaction with different materials, to give a correct mathematical treatment within the limit of necessary simplifying hypotheses clearly explained; to propose, whenever possible, numerical applications in order to fix orders of magnitude

  7. Cermet fuel reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  8. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

  9. The Chernobyl reactor accident

    International Nuclear Information System (INIS)

    The documentation abstracted contains a complete survey of the broadcasts transmitted by the Russian wire service of the Deutsche Welle radio station between April 28 and Mai 15, 1986 on the occasion of the Chernobyl reactor accident. Access is given to extracts of the remarkable eastern and western echoes on the broadcasts of the Deutsche Welle. (HP)

  10. SRP reactor safety evolution

    International Nuclear Information System (INIS)

    The Savannah River Plant reactors have operated for over 100 reactor years without an incident of significant consequence to on or off-site personnel. The reactor safety posture incorporates a conservative, failure-tolerant design; extensive administrative controls carried out through detailed operating and emergency written procedures; and multiple engineered safety systems backed by comprehensive safety analyses, adapting through the years as operating experience, changes in reactor operational modes, equipment modernization, and experience in the nuclear power industry suggested. Independent technical reviews and audits as well as a strong organizational structure also contribute to the defense-in-depth safety posture. A complete review of safety history would discuss all of the above contributors and the interplay of roles. This report, however, is limited to evolution of the engineered safety features and some of the supporting analyses. The discussion of safety history is divided into finite periods of operating history for preservation of historical perspective and ease of understanding by the reader. Programs in progress are also included. The accident at Three Mile Island was assessed for its safety implications to SRP operation. Resulting recommendations and their current status are discussed separately at the end of the report. 16 refs., 3 figs

  11. Fusion reactor materials

    International Nuclear Information System (INIS)

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics

  12. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  13. Department of reactor technology

    International Nuclear Information System (INIS)

    The activities of the Department of Reactor Technology at Risoe during 1980 are described. The work is presented in three chapters: General Information on the Department, Summary of the Department's Development during 1980, and Activities of the Department. Lists of staff, publications, computer programs, and test facilities are included. (author)

  14. The AP1000 reactor

    International Nuclear Information System (INIS)

    The design of the AP1000 reactor began 20 years ago when Westinghouse launched the AP600 reactor project. In fact by re-assessing AP600's safety margins Westinghouse realized that the its power output could be raised without putting at risk its safety standard. The AP1000 was born, it yields 1100 MWe. The main AP1000's design features is its passive safety (particularly after the Fukushima accident) and its modularity. The passive safety of the AP1000 implies: -) no humane intervention needed for 72 hours at least after the incident; -) no necessity for redundant complex safety systems. The modularity means that the plant, the reactor and other buildings are constructed from a choice of 300 modular units. These units can be built off-site and fit together on site. The modularity allows more construction activities to be led simultaneously and more chances to cope with the construction schedule. The NRC has approved the operation license for 30 years of the first AP1000 being built in the Usa (Vogtle plant in Georgia). 4 AP1000 are being built in China (Sanmen and Haiyang sites) and 6 others are planned in the Usa. Westinghouse is convinced that the AP1000's passive safety makes it more attractive. Let us not forget that Westinghouse was at the origin of the concept of pressurized water reactors, an idea adopted for half the nuclear power stations in the world and for all the plants now active in France. (A.C.)

  15. Fast reactor programme

    International Nuclear Information System (INIS)

    This progress report summarises the fast reactor research carried out at the Netherlands Energy Research Centre during the year 1981. The neutron and fission product cross sections of various isotopes have been evaluated. In the fuel performance programme, some preliminary results are given and irradiation facilities described. Creep experiments on various stainless steel components are reported

  16. Nuclear rocket engine reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lanin, Anatoly

    2013-07-01

    Covers a new technology of nuclear reactors and the related materials aspects. Integrates physics, materials science and engineering Serves as a basic book for nuclear engineers and nuclear physicists. The development of a nuclear rocket engine reactor (NRER) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  17. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  18. Fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1989-01-01

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

  19. Comparison of transient analysis of LBE-cooled fast reactor and ADS under loss of heat sink accident

    International Nuclear Information System (INIS)

    Highlights: • A natural circulation LBE-cooled research reactor is developed by Institute of Nuclear Energy Safety Technology. • The safety advantages for 10 MWth fast and ADS reactors under loss of heat sink were compared. • A neutronics and thermal-hydraulics coupled simulation code NTC was employed in this study. - Abstract: A natural circulation LBE-cooled research reactor developed by Institute of Nuclear Energy Safety Technology (INEST) is proposed and designed to be 10 MWth. The reactor has two operation modes, which are LBE-cooled critical fast reactor mode and LBE-cooled accelerator-driven sub-critical system (ADS) reactor mode. In this paper, protected loss of heat sink (PLOHS) and unprotected loss of heat sink (ULOHS) transient accidents were simulated for both the critical and ADS reactors above by NTC-2D code, aiming at the investigation of the safety advantages for the two reactor modes under this typical transient condition (loss of heat sink, LOHS). The results showed that during PLOHS for both the two modes, all the key parameters (core power, fuel, cladding and coolant temperatures in the hottest channel) decreased to very small values after the reactor scrammed, which meant the reactors under the two modes were both safe. For ULOHS, the fuel, cladding and coolant temperatures of the fast reactor increased smaller than those of the sub-critical one, which means the fast reactor had a better safety advantage under LOHS transient

  20. Tritium content in atmospheric water vapor inside of the reactor hall (reactor R-A) in Institute Vinca

    International Nuclear Information System (INIS)

    Tritium content in atmospheric water vapour inside of the reactor hall was measured during the regular inspection of the fuel channels in Institute of Nuclear Sciences 'Vinca', in March and May 2006. Tritium content in HTO form varied from 1.56·102 Bqm-3 to 4.05·102 Bqm-3. Tritium concentrations in precipitation collected at Zeleno Brdo and Institute 'Vinca' during the 2006. were (-1 and (3.52 - 13.09) Bql-1, respectively. (author)

  1. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis

  2. Fast reactors: potential for power

    International Nuclear Information System (INIS)

    The subject is discussed as follows: basic facts about conventional and fast reactors; uranium economy; plutonium and fast reactors; cooling systems; sodium coolant; safety engineering; handling and recycling plutonium; safeguards; development of fast reactors in Britain and abroad; future progress. (U.K.)

  3. Reactor physics problems on HCPWR

    International Nuclear Information System (INIS)

    Reactor physics problems on high conversion pressurized water reactors (HCPWRs) are discussed. Described in this report are outline of the HCPWR, expected accuracy for the various reactor physical qualities, and method for K-effective calculation in the resonance energy area. And requested further research problems are shown. The target value of the conversion ratio are also discussed. (author)

  4. An indigenous system for in-service inspection of coolant channels of 235 MWe PHWRs

    International Nuclear Information System (INIS)

    In-service inspection of coolant channels of nuclear power plants is essential to provide information on ageing effects. A channel inspection system (BARCIS- BARC Channel Inspection System) for in-service inspection of coolant channels of 235 MWe PHWRs has been recently developed at BARC. The need for such a system was being felt for quite sometime. The system is designed with the aim to minimise radiation exposure to inspection personnel and to perform the inspection so as to minimise reactor downtime. Sixteen coolant channels of RAPS-2 and fourteen coolant channels of MAPS-2 have been inspected using BARCIS MARK-1 system. Details of the system developed and the results of inspection are presented. (author). 3 refs., 3 tabs., 5 figs

  5. Pickering NGS units 1 and 2 LSFCRP fuel channel reinstallation process

    International Nuclear Information System (INIS)

    Replacement of fuel channels at Ontario Hydros' Pickering reactor units 1 and 2 is currently in progress. In this replacement process, the east and fitting at each lattice site is left in place. The west end fitting, the pressure tube and the garter springs are removed from the west side and replaced from the same end. This paper describes the reinstallation of fuel channel components and the tooling associated with this process

  6. Design of a rotary reactor for chemical-looping combustion. Part 1: Fundamentals and design methodology

    KAUST Repository

    Zhao, Zhenlong

    2014-04-01

    Chemical-looping combustion (CLC) is a novel and promising option for several applications including carbon capture (CC), fuel reforming, H 2 generation, etc. Previous studies demonstrated the feasibility of performing CLC in a novel rotary design with micro-channel structures. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet, and depleted air and product streams at exit. The rotary wheel consists of a large number of micro-channels with oxygen carriers (OC) coated on the inner surface of the channel walls. In the CC application, the OC oxidizes the fuel while the channel is in the fuel zone to generate undiluted CO2, and is regenerated while the channel is in the air zone. In this two-part series, the effect of the reactor design parameters is evaluated and its performance with different OCs is compared. In Part 1, the design objectives and criteria are specified and the key parameters controlling the reactor performance are identified. The fundamental effects of the OC characteristics, the design parameters, and the operating conditions are studied. The design procedures are presented on the basis of the relative importance of each parameter, enabling a systematic methodology of selecting the design parameters and the operating conditions with different OCs. Part 2 presents the application of the methodology to the designs with the three commonly used OCs, i.e., nickel, copper, and iron, and compares the simulated performances of the designs. © 2013 Elsevier Ltd. All rights reserved.

  7. Selection of fuel channels for Thermal Power Measurement in 700 MWe Indian PHWR by evolutionary algorithm

    International Nuclear Information System (INIS)

    Highlights: ► Thermal Power Monitoring System is used for on-line estimation of reactor bulk power and zone powers in PHWRs. ► This requires making best choice of 44 fuel channels out of 392 to be instrumented for measurement of flow and temperature. ► This constrained optimization problem is solved by Estimation of Distribution Algorithm. ► With the designed TPMS, bulk power and zone power can be predicted with an accuracy of 0.5% and 2% respectively. - Abstract: This paper presents studies on the design of Thermal Power Monitoring System (TPMS) for the forthcoming 700 MWe Indian Pressurized Heavy Water Reactor (PHWR). This reactor contains total 392 horizontal fuel channels. Each channel contains clustered natural Uranium fuel along with associated heavy water coolant placed inside a pressure tube. The coolant in different fuel channels is physically and thermally isolated from each other inside the core. It is necessary to select 44 fuel channels (out of 392) for keeping instrumentation to measure flow and temperature of coolant. The reactor is logically divided into 7 radial zones each containing certain number of fuel channels. The selection of instrumented channels is to be made such that power measured by them in terms of per unit basis represents the true zone-wise and global powers fairly accurately. This should be possible for a large number of reactor configurations that can occur because of the movement of reactivity devices in the core. Such a study is useful to make the TPMS more accurate means to measure the reactor bulk power and zone powers. The choice of 44 channels is an optimization problem in which the error in zonal and global power prediction is to be minimized. There are several constraints on the selection of instrumented channels. Therefore, a constrained combinatorial optimization problem has to be solved. An evolutionary technique based on Estimation of Distribution Algorithm (EDA) is used for this purpose. A suitable pattern of

  8. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    Science.gov (United States)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  9. Chemistry in Microfluidic Channels

    Science.gov (United States)

    Chia, Matthew C.; Sweeney, Christina M.; Odom, Teri W.

    2011-01-01

    General chemistry introduces principles such as acid-base chemistry, mixing, and precipitation that are usually demonstrated in bulk solutions. In this laboratory experiment, we describe how chemical reactions can be performed in a microfluidic channel to show advanced concepts such as laminar fluid flow and controlled precipitation. Three sets of…

  10. Beyond the Manual Channel

    DEFF Research Database (Denmark)

    the manual channel. Not surprisingly, most papers deal with non-manuals on the face. Once again, the papers at this workshop clearly identify the potentials of even closer cooperation between sign linguists and sign language engineers, and we think it is events like this that contribute a lot to a better...

  11. TRP channels: an overview

    DEFF Research Database (Denmark)

    Pedersen, Stine Falsig; Owsianik, Grzegorz; Nilius, Bernd

    2005-01-01

    to a plethora of data on the roles of TRPs in a variety of tissues and species, including mammals, insects, and yeast. The present review summarizes the most pertinent recent evidence regarding the structural and functional properties of TRP channels, focusing on the regulation and physiology of mammalian TRPs....

  12. All channels open

    NARCIS (Netherlands)

    Frank Huysmans; Jos de Haan

    2010-01-01

    Original title: Alle kanalen staan open. The rapid changes taking place in the media landscape in the Netherlands - characterised by digitisation and convergence of media technologies - raise the question of how the Dutch are dealing with the many new opportunities that have opened up. All channels

  13. Channeling through Bent Crystals

    Energy Technology Data Exchange (ETDEWEB)

    Mack, Stephanie; /Ottawa U. /SLAC

    2012-09-07

    Bent crystals have demonstrated potential for use in beam collimation. A process called channeling is when accelerated particle beams are trapped by the nuclear potentials in the atomic planes within a crystal lattice. If the crystal is bent then the particles can follow the bending angle of the crystal. There are several different effects that are observed when particles travel through a bent crystal including dechanneling, volume capture, volume reflection and channeling. With a crystal placed at the edge of a particle beam, part of the fringe of the beam can be deflected away towards a detector or beam dump, thus helping collimate the beam. There is currently FORTRAN code by Igor Yazynin that has been used to model the passage of particles through a bent crystal. Using this code, the effects mentioned were explored for beam energy that would be seen at the Facility for Advanced Accelerator Experimental Tests (FACET) at a range of crystal orientations with respect to the incoming beam. After propagating 5 meters in vacuum space past the crystal the channeled particles were observed to separate from most of the beam with some noise due to dechanneled particles. Progressively smaller bending radii, with corresponding shorter crystal lengths, were compared and it was seen that multiple scattering decreases with the length of the crystal therefore allowing for cleaner detection of the channeled particles. The input beam was then modified and only a portion of the beam sent through the crystal. With the majority of the beam not affected by the crystal, most particles were not deflected and after propagation the channeled particles were seen to be deflected approximately 5mm. After a portion of the beam travels through the crystal, the entire beam was then sent through a quadrupole magnet, which increased the separation of the channeled particles from the remainder of the beam to a distance of around 20mm. A different code, which was developed at SLAC, was used to

  14. Reactor monitoring using antineutrino detectors

    Science.gov (United States)

    Bowden, N. S.

    2011-08-01

    Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactor as part of International Atomic Energy Agency (IAEA) and/or other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway worldwide.

  15. Reactor physics activities in Japan

    International Nuclear Information System (INIS)

    This report reviews the research activity in reactor physics field in Japan during July, 1992 - July, 1993. The review was performed in the following fields : nuclear data evaluation, calculational method development, fast reactor physics, thermal reactor physics, advanced core design, fusion reactor neutronics, nuclear criticality safety, shielding, incineration of radioactive nuclear wastes, noise analysis and control and national programs. The main references were taken from journals and reports published during this period. The research committee of reactor physics is responsible for the review work. (author)

  16. Development and application of reactor noise diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Karlsson, Joakim K.H

    1999-04-01

    A number of problems in reactor noise diagnostics have been investigated within the framework of the present thesis. The six papers presented cover three relatively different areas, namely the use of analytical calculations of the neutron noise in simple reactor models, some aspects of boiling water reactor (BWR) stability and diagnostics of core barrel motion in pressurized water reactors (PWRs). The noise induced by small vibrations of a strong absorber has been the subject of several previous investigations. For a conventional {delta}-function source model, the equations can not be linearized in the traditional manner. Thus, a new source model, which is called the {epsilon}/d model, was developed. The correct solution has been derived in the {epsilon}/d model for both 1-D and 2-D reactor models. Recently, several reactor diagnostic problems have occurred which include a control rod partially inserted into the reactor core. In order to study such problems, we have developed an analytically solvable, axially non-homogenous, 2-D reactor model. This model has also been used to study the noise induced by a rod maneuvering experiment. Comparisons of the noise with the results of different reactor kinetic approximations have yielded information on the validity of the approximations in this relatively realistic model. In case of an instability event in a BWR, the noise may consist of one or several co-existing modes of oscillation and besides the fundamental mode, a regional first azimuthal mode has been observed in e.g. the Swedish BWR Ringhals-1. In order to determine the different stability characteristics of the different modes separately, it is important to be able to decompose the noise into its mode constituents. A separation method based on factorisation of the flux has been attempted previously, but without success. The reason for the failure of the factorisation method is the presence of the local component of the noise and its axial correlation properties. In

  17. Defueled channel experiments in ZED-2 in support of ACR-1000 ROP analysis

    International Nuclear Information System (INIS)

    Defueled channel experiments were performed in ZED-2 to help resolve discrepancies between calculated flux detector response during refueling in ACR-1000 according the reactor codes RFSP and MCNP. The data produced from these experiments was later used in a separate Regional-Over-Power (ROP) analysis to verify MCNP and RFSP neutron response predictions during refueling. These experiments provided information on thermal flux distributions interior and exterior to a fueled and defueled channel; and on epithermal absolute flux distributions exterior to the same channel. Critical height and moderator temperature data for fueled and defueled channel conditions were also measured. In addition, standard platinum-clad Inconel Self-Powered Detector (SPD) performance data was obtained. The following reactor physics and SPD parameters were measured in these experiments: C Radial flux distribution inside the channel of interest (fueled and defueled), C Radial flux distribution outside the channel of interest (fueled and defueled), C Epithermal radial flux distribution outside the channel of interest (fueled and defueled), and C SPD response parallel to and normal to the channel of interest (fueled and defueled).

  18. Estimation of random deviation effect of pipeline geometry on hydraulic nonuniformity at reactor inlet

    International Nuclear Information System (INIS)

    Approximate formula enabling to evaluate hydraulic nonuniformity at reactor core inlet is given. The mentioned nonuniformity depends on 13 structural parameters of reactor hydraulic pipeline - from branch pipes up to the header. Dispersion technique for evaluation of the effect of structural parameter deviation on the spread of resulting nonuniformity is applied to the dependence. Coefficients of the effect have been obtained. Calculation has been conducted for the example, the minimal flow rate in reactor channel has been evaluated as 0.65±0.14

  19. Monte Carlo Model of TRIGA Reactor to Support Neutron Activation Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Zerovnik, G.; Snoj, L.; Trkov, A. [Reactor Physics Department, Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2011-07-01

    The TRIGA reactor at Jozef Stefan Institute is used as a neutron source for neutron activation analysis. The accuracy of the method depends on the accuracy of the neutron spectrum characterization. Therefore, computational models on different scales have been developed: Monte Carlo full reactor model, model of an irradiation channel and deterministic code for self-shielding factor calculations. The models have been validated by comparing against experiment and thus provide a very strong support for neutron activation analysis of samples irradiated at the TRIGA reactor. (author)

  20. Slow neutron flux extrapolation distances in R-5 and CIRUS reactors

    International Nuclear Information System (INIS)

    In order to calculate the core reactivity, fuel channel power outputs and neutron flux levels in the R-5 reactor at Trombay, axial flux extrapolation distances are required. For this, an analysis is carried out considering the reactor core as a two region neutron multiplying system in axial direction. The slow neutron diffusion equations for both the regions are solved analytically by applying suitable boundary conditions. Application of this method for the estimation of top extrapolation distances in CIRUS, has given results which agree well with accepted values for the reactor. (author)

  1. Sensitivity analysis of the RELAP5 nodalization to IPR-R1 TRIGA research reactor

    International Nuclear Information System (INIS)

    The main aim of this work is to identify how much the code results are affected by code user in the choice of, for example, the number of thermal-hydraulic channels in a nuclear reactor nodalization. To perform this, two essential modifications were made on a previous validated nodalization for analysis of steady state and forced recirculation off transient in the IPR-R1 TRIGA research reactor. Experimental data were taken as reference to compare the behavior of the reactor for two different types of model. The results found highlight the necessity of sensitivity analysis to obtain the ideal simulation model of a system. (author)

  2. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  3. Continuous thermal balance monitoring for IEA-R1 nuclear research reactor power determination

    International Nuclear Information System (INIS)

    This research deals with thermal balance calculation for real time power level determination of IEA-R1 nuclear research reactor. It is also shown the development of a supervision software (Visual Basic) of operation parameters. The assembled data acquisition system allows data analysis during reactor operation, giving a reliable measurement of reactor power, and the organization of a data base allows a back-up surveillance of reactor operation whenever necessary. Results obtained from temperature and primary flow are shown in a continuous form and also the Data Base implementation for further studies and analysis of energy balance behavior of the many reactor components. Besides it is planned to manage N-16 activity measurement channel (monitoring) for comparison of acquired data results for thermal calculations. The results of this acquisition and related thermal balance calculations are shown in a continuous shape (On-Line) by means of windows operational system using Visual Basic VB6 software for development. (author)

  4. Reactivity worths of two perpendicular control mechanisms in a large graphite-moderated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lan, J.S.; Toffer, H.; Carter, L.L.; Omberg, R.P.

    1989-01-01

    In a nuclear reactor, one of the more difficult tasks is evaluating control system worth where such systems are orthogonal to the fuel charges. Only complex, multidimensional analysis tools such as Monte Carlo techniques can provide answers for such complex reactor geometries. This paper discusses the successful application of the Monte Carlo neutron-proton (MCNP) code to control system worth evaluations for the N Reactor. The graphite-moderated, light-water-cooled Hanford site N reactor, operated by Westinghouse Hanford Company for the US Department of Energy, has a very complex geometric arrangement of horizontal pressure tubes, control rods, and ball backup safety system channels. The availability of the Hanford site version of the MCNP code enabled the set-up of numerical experiments for the N reactor lattice. Explicit representation of the entire N reactor core by Monte Carlo methods requires even larger and faster computers than are currently in use. This paper presents newly developed meter-cube model or hypercell model for MCNP analyses. The degree of representation that this model renders for the whole reactor core and the extent of the capability that this model could simulate under a variety of operational conditions are the subject of the discussion here. The reactivity worths of the two perpendicular control systems (control rods and safety ball channels) are specifically considered, neither of which has been explicitly calculated before with one code.

  5. Reactor simulator development. Workshop material

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in reactor operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. This publication consists of course material for workshops on development of such reactor simulators. Participants in the workshops are provided with instruction and practice in the development of reactor simulation computer codes using a model development system that assembles integrated codes from a selection of pre-programmed and tested sub-components. This provides insight and understanding into the construction and assumptions of the codes that model the design and operational characteristics of various power reactor systems. The main objective is to demonstrate simple nuclear reactor dynamics with hands-on simulation experience. Using one of the modular development systems, CASSIMtm , a simple point kinetic reactor model is developed, followed by a model that simulates the Xenon/Iodine concentration on changes in reactor power. Lastly, an absorber and adjuster control rod, and a liquid zone model are developed to control reactivity. The built model is used to demonstrate reactor behavior in sub-critical, critical and supercritical states, and to observe the impact of malfunctions of various reactivity control mechanisms on reactor dynamics. Using a PHWR simulator, participants practice typical procedures for a reactor startup and approach to criticality. This workshop material consists of an introduction to systems used for developing reactor simulators, an overview of the dynamic simulation

  6. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  7. Improving Virtual Channel Discrimination in a Multi-Channel Context

    OpenAIRE

    Srinivasan, Arthi G.; Shannon, Robert V.; Landsberger, David M.

    2012-01-01

    Improving spectral resolution in cochlear implants is key to improving performance in difficult listening conditions (e.g. speech in noise, music, etc.). Current focusing might reduce channel interaction, thereby increasing spectral resolution. Previous studies have shown that combining current steering and current focusing reduces spread of excitation and improves virtual channel discrimination in a single-channel context. It is unclear whether the single-channel benefits from current focusi...

  8. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    In 1998, the NNSA organized to complete the nuclear safety review on the test loop in-reactor operation of the High-flux Engineering Experimental Reactor (HFEER) and the re-operation of the China Pulsed Reactor and the Uranium-water Criticality Facility. The NNSA conducted the nuclear safety review on the CP application of the China Experimental Fast Reactor (CEFR) and the siting of China Advanced Research Reactor (CARR), and carried out the construction supervision on HTR-10, and dealt with the event about the technological tube breakage of HWRR and other events

  9. Compact fusion reactors

    CERN Document Server

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

  10. Elk River Reactor dismantling

    International Nuclear Information System (INIS)

    The dismantling program was carried out in three overlapping phases: the planning phase which included the preliminary planning and selection of the dismantling approach, the dismantling phase which included all work performed to remove the reactor facility and restore the site to its pre-reactor condition, and the closeout phase which included the final site survey and efforts necessary to terminate the AEC license and contract. Of particular interest was the use of a remotely operated plasma cutting torch to section the pressure vessel internals, the pressure vessel and the outer thermal shield, the use of explosives in removal of the biological shield and the method of establishment of the criteria for material disposal

  11. Perturbation analysis of the TRIGA Mark II reactor Vienna

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R. [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad (Pakistan); Villa, M.; Stummer, T.; Boeck, H. [Vienna Univ. of Technology (Austria). Atominstitut; Saeedbadshah [International Islamic Univ., Islamabad (Pakistan)

    2013-04-15

    The safety design of a nuclear reactor needs to maintain the steady state operation at desired power level. The safe and reliable reactor operation demands the complete knowledge of the core multiplication and its changes during the reactor operation. Therefore it is frequently of interest to compute the changes in core multiplication caused by small disturbances in the field of reactor physics. These disturbances can be created either by geometry or composition changes of the core. Fortunately if these changes (or perturbations) are very small, one does not have to repeat the reactivity calculations. This article focuses the study of small perturbations created in the Central Irradiation Channel (CIC) of the TRIGA mark II core to investigate their reactivity influences on the core reactivity. For this purpose, 3 different kinds of perturbations are created by inserting 3 different samples in the CIC. The cylindrical void (air), heavy water (D2O) and Cadmium (Cd) samples are inserted into the CIC separately to determine their neutronics behavior along the length of the core. The Monte Carlo N-Particle radiation transport code (MCNP) is applied to simulate these perturbations in the CIC. The MCNP theoretical predictions are verified by the experiments performed on the current reactor core. The behavior of void in the whole core and its dependence on position and water fraction is also presented in this article. (orig.)

  12. REACTOR MODERATOR STRUCTURE

    Science.gov (United States)

    Greenstreet, B.L.

    1963-12-31

    A system for maintaining the alignment of moderator block structures in reactors is presented. Integral restraining grids are placed between each layer of blocks in the moderator structure, at the top of the uppermost layer, and at the bottom of the lowermost layer. Slots are provided in the top and bottom surfaces of the moderator blocks so as to provide a keying action with the grids. The grids are maintained in alignment by vertical guiding members disposed about their peripheries. (AEC)

  13. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    SCK-CEN's programme on fusion reactor materials includes studies (1) to investigate fracture mechanics of neutron-irradiated beryllium; (2) to describe the helium behaviour in irradiated beryllium at atomic scale; (3) to define the kinetics of beryllium reacting with air or steam; (3) to perform a feasibility study for the testing of integrated blanket modules under neutron irradiation. Progress and achievements in 1997 are reported

  14. Decommissioning of research reactors

    International Nuclear Information System (INIS)

    Research reactors of WWR-S type were built in countries under Soviet influence in '60, last century and consequently reached their service life. Decommissioning implies removal of all radioactive components, processing, conditioning and final disposal in full safety of all sources on site of radiological pollution. The WWR-S reactor at Bucuresti-Magurele was put into function in 1957 and operated until 1997 when it was stopped and put into conservation in view of decommissioning. Presented are three decommissioning variants: 1. Reactor shut-down for a long period (30-50 years) what would entail a substantial decrease of contamination with lower costs in dismantling, mechanical, chemical and physical processing followed by final disposal of the radioactive wastes. The drawback of this solution is the life prolongation of a non-productive nuclear unit requiring funds for personnel, control, maintenance, etc; 2. Decommissioning in a single stage what implies large funds for a immediate investment; 3. Extending the operation on a series of stages rather phased in time to allow a more convenient flow of funds and also to gather technical solutions, better than the present ones. This latter option seems to be optimal for the case of the WWR-S Research at Bucharest-Magurele Reactor. Equipment and technologies should be developed in order to ensure the technical background of the first operations of decommissioning: equipment for scarification, dismantling, dismemberment in a highly radioactive environment; cutting-to-pieces and disassembling technologies; decontamination modern technologies. Concomitantly, nuclear safety and quality assurance regulations and programmes, specific to decommissioning projects should be implemented, as well as a modern, coherent and reliable system of data acquisition, recording and storing. Also the impact of decommissioning must be thoroughly evaluated. The national team of specialists will be assisted by IAEA experts to ensure the

  15. Measurement channel of neutron flow based on software; Canal de medicion de flujo neutronico basado en software

    Energy Technology Data Exchange (ETDEWEB)

    Rivero G, T.; Benitez R, J. S. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: trg@nuclear.inin.mx

    2008-07-01

    The measurement of the thermal power in nuclear reactors is based mainly on the measurement of the neutron flow. The presence of these in the reactor core is associated to neutrons released by the fission reaction of the uranium-235. Once moderate, these neutrons are precursors of new fissions. This process it is known like chain reaction. Thus, the power to which works a nuclear reactor, he is proportional to the number of produced fissions and as these depend on released neutrons, also the power is proportional to the number of present neutrons. The measurement of the thermal power in a reactor is realized with called instruments nuclear channels. To low power (level source), these channels measure the individual counts of detected neutrons, whereas to a medium and high power, they measure the electrical current or fluctuation of the same one that generate the fission neutrons in ionization chambers especially designed to detect neutrons. For the case of TRIGA reactors, the measurement channels of neutron flow use discreet digital electronic technology makes some decades already. Recently new technological tools have arisen that allow developing new versions of nuclear channels of simple form and compacts. The present work consists of the development of a nuclear channel for TRIGA reactors based on the use of the correlated signal of a fission chamber for ample interval. This new measurement channel uses a data acquisition card of high speed and the data processing by software that to the being installed in a computer is created a virtual instrument, with what spreads in real time, in graphic and understandable form for the operator, the power indication to which it operates the nuclear reactor. This system when being based on software, offers a major versatility to realize changes in the signal processing and power monitoring algorithms. The experimental tests of neutronic power measurement show a reliable performance through seven decades of power, with a

  16. Nuclear reactor container

    International Nuclear Information System (INIS)

    Upon reactor accident, hydrogen and oxygen are generated by water-zirconium reaction and radiolysis of water, which are accumulated in the reactor. If the concentration of hydrogen and oxygen exceeds a burning limit, there is a possibility of hydrogen burning to cause a danger of deteriorating the integrity of the reactor container and the equipments therein. The limit for the occurrence of the detonation is determined by a relationship between the scale of a detonation cell and the size of the container, and if the scale is greater than the container, the detonation does not occur. The scale of the cell is determined by a gas combustion rate and, if the combustion reaction is suppressed, detonation does not occur even in a large container. Then, an appropriate diluent is added to increase heat capacity of a gas mixture to thereby suppress the temperature elevation of the gas. Incombustible gases having a great heat capacity are preferred for the diluent, and CO2 is used. As the concentration of the CO2 gas to be added is increased, the detonation cell is made greater. Thus, occurrence of detonation due to combustion of the accumulated hydrogen can be prevented. (N.H.)

  17. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    In the middle of 21st century, the population on the earth is expected to double, and the energy that mankind consumes to triple. The nuclear fusion which is said the ultimate energy source for mankind is expected to solve this energy problem. As for fusion reactors, fuel materials exist inexhaustibly, distributing evenly, they have high safety in principle, the product of burning is harmless nonradioactive substance that does not require the treatment and disposal, and the attenuation of induced radioactivity due to neutrons is quick and the effect to global environment is little. The basic plan of second stage nuclear fusion research and development was decided in 1975, aiming at attaining the critical plasma condition. JT-60 has attained it in 1987. The project of international thermonuclear fusion experimental reactor (ITER) was started, and the conceptual design was carried out. Under such background, the third stage basic plan was decided in 1992, and its objective is self ignition condition, long time burning and the basis of the reactor engineering technology. The engineering design of the ITER is investigated. (K.I.)

  18. High temperature gas reactor

    International Nuclear Information System (INIS)

    The present invention provides a reflector block structure of a high temperature gas reactor in which graphite blocks are not failed even a containing cylinder loaded to a fuel exchanger collides against to secured reflectors upon loading and withdrawing fuel constitutional elements. Namely, a protection plate made of a metal material such as stainless steel is covered on the secured reflector blocks disposed to the upper most step among secured graphite reflector blocks constituting the reactor core. In addition, positioning guide grooves are formed on the protection plate for guiding the containing cylinder loaded to the fuel exchanger to the column of the reactor core constitutional elements. With such a constitution, even if the containing cylinder of fuel exchanger is hoisted down and collided against the inner circumferential edge of the secured reflector blocks due to deviation of the position and the direction upon exchange of fuels, the reflector blocks are not failed since the above-mentioned portion is covered with the metal protection plate. In addition, the positioning guide grooves lead the fuel exchanger to a predetermined column correctly. (I.S.)

  19. BWR type reactor

    International Nuclear Information System (INIS)

    In a coolant circulation in BWR type reactors, since the mixed stream of steam fluid undergoes a great resistance, the pressure loss due to the flow rate distribution when the coolants flow from the upper plenum into the stand pipe is increased upon passing stand pipe. Also in the spontaneous recycling reactor, pressure loss is still left upon passing the swirling blade of a gas-liquid separator. In view of the above, a plurality of vertical members each having a lower end opened to a gas-liquid two phase boundary and an upper end directly suspended from a steam dryer to the gas-liquid separator. The liquid droplets from the 2-phase boundary heated in the reactor core and formed into a mixed gas-liquid 2-phase stream is directed in the vertical direction accompanied with the steam. The liquid droplets spontaneously fallen by gravity from greater ones successively and the droplets in the steam abutted against the vertical member are fallen as a liquid membrane. Thus, the gas-liquid separation is conducted, the dry steam is directly flown into the steam dryer, thereby capable of providing a gas-liquid separator having gas-liquid separation performance with lower loss than usual. (N.H.)

  20. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  1. Fusion reactor safety

    International Nuclear Information System (INIS)

    Nuclear fusion could soon become a viable energy source. Work in plasma physics, fusion technology and fusion safety is progressing rapidly in a number of Member States and international collaboration continues on work aiming at the demonstration of fusion power generation. Safety of fusion reactors and technological and radiological aspects of waste management are important aspects in the development and design of fusion machines. In order to provide an international forum to review and discuss the status and the progress made since 1983 in programmes related to operational safety aspects of fusion reactors, their waste management and decommissioning concepts, the IAEA had organized the Technical Committee on ''Fusion Reactor Safety'' in Culham, 3-7 November 1986. All presentations of this meeting were divided into four sessions: 1. Statements on National-International Fusion Safety Programmes (5 papers); 2. Operation and System Safety (15 papers); 3. Waste Management and Decommissioning (5 papers); 4. Environmental Impacts (6 papers). A separate abstract was prepared for each of these 31 papers. Refs, figs, tabs

  2. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units are provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  3. OECD Halden Reactor Project

    International Nuclear Information System (INIS)

    The OECD Halden Reactor project is an agreement between OECD member countries. It was first signed in 1958 and since then regularly renewed every third year. The activities at the Project is centred around the Halden heavy water rector, the HBWR. The reseach programme comprizes studies of fuel performance under various operating conditions, and the application of computers for process control. The HBWR is equipped for exposing fuel rods to temperatures and pressures, and at heat ratings met in modern BWR's and PWR's. A range of in-core instruments are available, permitting detailed measurements of the reactions of the fuel, including mechanical deformations, thermal behaviour, fission gas release, and corrosion. In the area of computer application, the studies of the communication between operator and process, and the surveillance and control of the reactor core, are of particular interst for reactor operation. 1988 represents the 30th year since the Project was started, and this publication is produced to mark this event. It gives and account of the activities and achievements of the Project through the years 1958-1988

  4. Statistical Hot Channel Analysis for the NBSR

    Energy Technology Data Exchange (ETDEWEB)

    Cuadra A.; Baek J.

    2014-05-27

    A statistical analysis of thermal limits has been carried out for the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The objective of this analysis was to update the uncertainties of the hot channel factors with respect to previous analysis for both high-enriched uranium (HEU) and low-enriched uranium (LEU) fuels. Although uncertainties in key parameters which enter into the analysis are not yet known for the LEU core, the current analysis uses reasonable approximations instead of conservative estimates based on HEU values. Cumulative distribution functions (CDFs) were obtained for critical heat flux ratio (CHFR), and onset of flow instability ratio (OFIR). As was done previously, the Sudo-Kaminaga correlation was used for CHF and the Saha-Zuber correlation was used for OFI. Results were obtained for probability levels of 90%, 95%, and 99.9%. As an example of the analysis, the results for both the existing reactor with HEU fuel and the LEU core show that CHFR would have to be above 1.39 to assure with 95% probability that there is no CHF. For the OFIR, the results show that the ratio should be above 1.40 to assure with a 95% probability that OFI is not reached.

  5. Geysering in boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Aritomi, Masanori; Takemoto, Takatoshi [Tokyo Institute of Technology, Tokyo (Japan); Chiang, Jing-Hsien [Japan NUS Corp. Ltd., Toyko (Japan)] [and others

    1995-09-01

    A concept of natural circulation BWRs such as the SBWR has been proposed and seems to be promising in that the primary cooling system can be simplified. The authors have been investigating thermo-hydraulic instabilities which may appear during the start-up in natural circulation BWRs. In our previous works, geysering was investigated in parallel boiling channels for both natural and forced circulations, and its driving mechanism and the effect of system pressure on geysering occurrence were made clear. In this paper, geysering is investigated in a vertical column and a U-shaped vertical column heated in the lower parts. It is clarified from the results that the occurrence mechanism of geysering and the dependence of system pressure on geysering occurrence coincide between parallel boiling channels in circulation systems and vertical columns in non-circulation systems.

  6. Lipid Ion Channels

    CERN Document Server

    Heimburg, Thomas

    2010-01-01

    The interpretation electrical phenomena in biomembranes is usually based on the assumption that the experimentally found discrete ion conduction events are due to a particular class of proteins called ion channels while the lipid membrane is considered being an inert electrical insulator. The particular protein structure is thought to be related to ion specificity, specific recognition of drugs by receptors and to macroscopic phenomena as nerve pulse propagation. However, lipid membranes in their chain melting regime are known to be highly permeable to ions, water and small molecules, and are therefore not always inert. In voltage-clamp experiments one finds quantized conduction events through protein-free membranes in their melting regime similar to or even undistinguishable from those attributed to proteins. This constitutes a conceptual problem for the interpretation of electrophysiological data obtained from biological membrane preparations. Here, we review the experimental evidence for lipid ion channels...

  7. A transparent Pyrex μ-reactor for combined in situ optical characterization and photocatalytic reactivity measurements

    DEFF Research Database (Denmark)

    Dionigi, Fabio; Nielsen, Morten Godtfred; Pedersen, Thomas;

    2013-01-01

    A new Pyrex-based μ-reactor for photocatalytic and optical characterization experiments is presented. The reactor chamber and gas channels are microfabricated in a thin poly-silicon coated Pyrex chip that is sealed with a Pyrex lid by anodic bonding. The device is transparent to light in the UV...

  8. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  9. Chernobyl reactor transient simulation study

    International Nuclear Information System (INIS)

    This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

  10. Ion Channels, Natural Nanovalves

    OpenAIRE

    Eisenberg, Bob

    2012-01-01

    Ion channels are proteins with holes down their middle that control the flow of ions and electric current across otherwise impermeable biological membranes. The flow of sodium, potassium, calcium (divalent), and chloride ions have been central issues in biology for more than a century. The flow of current is responsible for the signals of the nervous system that propagate over long distances (meters). The concentration of divalent calcium ions is a 'universal' signal that controls many differ...

  11. Proton channel models

    OpenAIRE

    Pupo, Amaury; Baez-Nieto, David; Martínez, Agustín; Latorre, Ramón; González, Carlos

    2014-01-01

    Voltage-gated proton channels are integral membrane proteins with the capacity to permeate elementary particles in a voltage and pH dependent manner. These proteins have been found in several species and are involved in various physiological processes. Although their primary topology is known, lack of details regarding their structures in the open conformation has limited analyses toward a deeper understanding of the molecular determinants of their function and regulation. Consequently, the f...

  12. Athermal channeled spectropolarimeter

    Energy Technology Data Exchange (ETDEWEB)

    Jones, Julia Craven

    2015-12-08

    A temperature insensitive (athermal) channeled spectropolarimeter (CSP) is described. The athermal CSP includes a crystal retarder formed of a biaxial crystal. The crystal retarder has three crystal axes, wherein each axis has its own distinct index of refraction. The axes are oriented in a particular manner, causing an amplitude modulating carrier frequency induced by the crystal retarder to be thermally invariant. Accordingly, a calibration beam technique can be used over a relatively wide range of ambient temperatures, with a common calibration data set.

  13. Radar channel balancing with commutation

    Energy Technology Data Exchange (ETDEWEB)

    Doerry, Armin Walter

    2014-02-01

    When multiple channels are employed in a pulse-Doppler radar, achieving and maintaining balance between the channels is problematic. In some circumstances the channels may be commutated to achieve adequate balance. Commutation is the switching, trading, toggling, or multiplexing of the channels between signal paths. Commutation allows modulating the imbalance energy away from the balanced energy in Doppler, where it can be mitigated with filtering.

  14. Safety analysis code 'COOLTMP' for assessment of PHT cooling under reactor shutdown conditions

    International Nuclear Information System (INIS)

    The thermal energy generated by the reactor core is removed by the Primary Heat Transport (PHT) System when the reactor is under normal operation, by operation of the primary circulation pumps and steam generators. However, when the reactor is shutdown, the decay heat removal is done by the Shut Down (S/D) Cooling heat exchangers and pumps of lower capacity. In the event of loss/stoppage of circulation of PHT under such a situation, the bulk of the decay heat generated will be distributed to the moderator system, end shield system and through the feeders to the feeder cabinet/FM vault environment. However, the PHT inventory in the channel will be heated up because of loss of flow in the channel. The code COOLTMP has been developed to estimate the temperature of PHT following a loss/stoppage of circulation, when the reactor is under shutdown condition. It predicts the increase in the PHT temperature with time for hot channel, average channel or a specific channel under such a condition. It also calculates the apportionment of the decay heat to different heat sinks, viz. moderator, end shield and FM Vault. This computation is required when the plant is required to be under shutdown for doing some maintenance job on the PHT system, feeders or channels where the S/D cooling system has to be stopped and in some cases the headers have to be drained. At that time such a calculation gives whether the peak PHT temperature, or the time available to reach such a temperature, as obtained, is acceptable to carry out such a job. Hence, the schedule of the maintenance job can be decided. This code has been validated for RAPS and MAPS and used extensively for predicting PHT temperature after reactor shutdown to obtain regulatory clearances to stop forced circulation with and without header filled. (author)

  15. Ion channeling revisited

    Energy Technology Data Exchange (ETDEWEB)

    Doyle, Barney Lee [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Corona, Aldo [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Nguyen, Anh [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-09-01

    A MS Excel program has been written that calculates accidental, or unintentional, ion channeling in cubic bcc, fcc and diamond lattice crystals or polycrystalline materials. This becomes an important issue when simulating the creation by energetic neutrons of point displacement damage and extended defects using beams of ions. All of the tables and graphs in the three Ion Beam Analysis Handbooks that previously had to be manually looked up and read from were programed into Excel in handy lookup tables, or parameterized, for the case of the graphs, using rather simple exponential functions with different powers of the argument. The program then offers an extremely convenient way to calculate axial and planar half-angles and minimum yield or dechanneling probabilities, effects on half-angles of amorphous overlayers, accidental channeling probabilities for randomly oriented crystals or crystallites, and finally a way to automatically generate stereographic projections of axial and planar channeling half-angles. The program can generate these projections and calculate these probabilities for axes and [hkl] planes up to (555).

  16. DMT of weighted Parallel Channels: Application to Broadcast Channel

    CERN Document Server

    Mroueh, Lina; Othman, Ghaya Rekaya-Ben; Belfiore, Jean-Claude

    2008-01-01

    In a broadcast channel with random packet arrival and transmission queues, the stability of the system is achieved by maximizing a weighted sum rate capacity with suitable weights that depend on the queue size. The weighted sum rate capacity using Dirty Paper Coding (DPC) and Zero Forcing (ZF) is asymptotically equivalent to the weighted sum capacity over parallel single-channels. In this paper, we study the Diversity Multiplexing Tradeoff (DMT) of the fading broadcast channel under a fixed weighted sum rate capacity constraint. The DMT of both identical and different parallel weighted MISO channels is first derived. Finally, we deduce the DMT of a broadcast channel using DPC and ZF precoders.

  17. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    The purpose of this technical paper is to provide status of the United State domestic Research Reactor Infrastructure (RRI) Program at the Idaho National Laboratory. This paper states the purpose of the program, lists the universities operating TRIGA reactors that are supported by the program, identifies anticipated fresh fuel needs for the reactor facilities, discusses spent fuel activities associated with the program, and addresses successes and planned activities for the program. (author)

  18. Thermionic Reactor Design Studies

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1994-06-01

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic

  19. Applications of Research Reactors

    International Nuclear Information System (INIS)

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world.' One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property'. The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. The purpose of the earlier publication, The Application of Research Reactors, IAEA-TECDOC-1234, was to present descriptions of the typical forms of research reactor use. The necessary criteria to enable an application to be performed were outlined for each one, and, in many cases, the minimum as well as the desirable requirements were given. This revision of the publication over a decade later maintains the original purpose and now specifically takes into account the changes in service requirements demanded by the relevant stakeholders. In particular, the significant improvements in

  20. Cascading blockages in channel bundles.

    Science.gov (United States)

    Barré, C; Talbot, J

    2015-11-01

    Flow in channel networks may involve a redistribution of flux following the blockage or failure of an individual link. Here we consider a simplified model consisting of N(c) parallel channels conveying a particulate flux. Particles enter these channels according to a homogeneous Poisson process and an individual channel blocks if more than N particles are simultaneously present. The behavior of the composite system depends strongly on how the flux of entering particles is redistributed following a blockage. We consider two cases. In the first, the intensity on each open channel remains constant while in the second the total intensity is evenly redistributed over the open channels. We obtain exact results for arbitrary N(c) and N for a system of independent channels and for arbitrary N(c) and N=1 for coupled channels. For N>1 we present approximate analytical as well as numerical results. Independent channels block at a decreasing rate due to a simple combinatorial effect, while for coupled channels the interval between successive blockages remains constant for N=1 but decreases for N>1. This accelerating cascade is due to the nonlinear dependence of the mean blocking time of a single channel on the entering particle flux that more than compensates for the decrease in the number of active channels.

  1. Micro-channel plate detector

    Energy Technology Data Exchange (ETDEWEB)

    Elam, Jeffrey W.; Lee, Seon W.; Wang, Hsien -Hau; Pellin, Michael J.; Byrum, Karen; Frisch, Henry J.

    2015-09-22

    A method and system for providing a micro-channel plate detector. An anodized aluminum oxide membrane is provided and includes a plurality of nanopores which have an Al coating and a thin layer of an emissive oxide material responsive to incident radiation, thereby providing a plurality of radiation sensitive channels for the micro-channel plate detector.

  2. Nomenclature for Ion channel Subunits

    OpenAIRE

    Bradley, Jonathan; Frings, Stephan; Yau, King-Wai; Reed, Randall

    2001-01-01

    Presents the nomenclature for ion channel subunits. Role of ion channels in the mediation of visual and olfactory signal transduction; Expression of ion channels in cell types and tissues; Assessment on the nucleotide sensitivity, ion conductance and calcium modulation in heteromers.

  3. Probabilistic assessment of fuel channel bearing travel and end of bearing time

    International Nuclear Information System (INIS)

    Operators in Ontario have been assessing fuel channel elongation in nuclear generating units to ensure fuel channels remain on bearing at least until the next inspection in accordance with CSA N285.4. These assessments have so far been performed deterministically and, among other conservative assumptions, use the worst stack-up of tolerances for all relevant fuel channel components to calculate the end-of-bearing (EOB) times. Since the probability of having the worst stack-up of tolerances occurring is extremely small, a deterministic assessment provides a very conservative representation of the unit condition. The probabilistic assessment presented in this paper provides a more realistic representation of elongation and channel end-of-bearing times. For a given reactor, the probabilistic assessment used Monte Carlo techniques to generate values for fuel channel component dimensions, current fuel channel elongations, future elongations and elongation measurement errors from specified distributions. The simulated values are then used to obtain a distribution of EOB times at each lattice position in the reactor. The expected number of fuel channels reaching EOB in a given operating interval can then be compared to pre-established acceptance criteria to determine if the operating interval is acceptable. The methodology is illustrated with an example reactor. The results from 100,000 simulations at each lattice position show that the EOB times calculated in the deterministic assessment corresponded to a very low percentile of the EOB times calculated in the probabilistic assessment. This indicated that the deterministic assessment was very conservative. For the example reactor, no elongation maintenance would be necessary for approximately one year (8,000 effective full power hours) after the latest outage. (author)

  4. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

  5. CATHENA Models for Wolsong 1 Channels O6 and B10

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y.H.; Lee, K.H.; Choi, H. [Korea Electric Power Research Institute, Taejon (Korea)

    2002-07-01

    This document includes CATHENA model development of slave channel O6 and B10 for Wolsong 1 trip coverage analysis in the project of ''Development of Assessment Technologies for CANDU Reactor Power and Trip Effectivness''. (author). 11 refs., 2 tabs.

  6. Gas-liquid separation mechanism and reactor

    International Nuclear Information System (INIS)

    A space above the reactor core mounted in a pressure vessel of a BWR type reactor is surrounded with a cylindrical shroud. Since vortex and stagnation of coolants are liable to be formed over the entire periphery at the upper portion of the shroud, inlets through which steam bubbles enter are opened to the entire circumference along the upper portion of the shroud. This can attain the gas-liquid separation effect from the entire periphery thereof. Further, since the coolant flowing vector changes abruptly at a portion where the coolant flow turns from the upward to the downward direction, steam bubbles are swirled or stagnated on the side of the downward flow channel in that portion. In view of the above, a coolant inlet submerged in the coolant is disposed to that portion, so that the steam bubbles are captured, involved in the subsequently flowing coolants and urged upwardly into the space above the coolant surface, thereby maintaining extremely high spontaneous recycling force and recycling amount. (T.M.)

  7. The research reactor TRIGA Mainz

    International Nuclear Information System (INIS)

    Paper dwells upon the design and the operation of one of the German test reactors, namely, the TRIGA Mainz one (TRIGA: Training Research Isotope Production General Atomic). The TRIGA reactor is a pool test reactor the core of which contains a graphite reflector and is placed into 2 m diameter and 6.25 m height aluminum vessel. There are 75 fuel elements in the reactor core, and any of them contains about 36 g of 235U. The TRIGA reactors under the stable operation enjoy wide application to ensure tests and irradiation, namely: neutron activation analysis, radioisotope production, application of a neutron beam to ensure the physical, the chemical and the medical research efforts. Paper presents the reactor basic experimental program lines

  8. Spiral-shaped disinfection reactors

    KAUST Repository

    Ghaffour, Noreddine

    2015-08-20

    This disclosure includes disinfection reactors and processes for the disinfection of water. Some disinfection reactors include a body that defines an inlet, an outlet, and a spiral flow path between the inlet and the outlet, in which the body is configured to receive water and a disinfectant at the inlet such that the water is exposed to the disinfectant as the water flows through the spiral flow path. Also disclosed are processes for disinfecting water in such disinfection reactors.

  9. Industrializing the liquid metal reactor

    International Nuclear Information System (INIS)

    Commercial acceptance of the liquid metal reactor had its beginning with the Fermi reactor, over two decades ago. The pattern of industrialization since that time is discussed, contrasting domestic and foreign experience. The recent termination of the Clinch River reactor project marks a watershed in the U.S. approach towards commercialization. The increased emphasis on achieving cost competitive designs reflects an awareness that barriers to industrialization are institutional and financial, and not technological

  10. Acceptability of reactors in space

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.

    1981-04-01

    Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System.

  11. Hydrogen Production in Fusion Reactors

    OpenAIRE

    Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-01-01

    As one of methods of innovative energy production in fusion reactors without having a conventional turbine-type generator, an efficient use of radiation produced in a fusion reactor with utilizing semiconductor and supplying clean fuel in a form of hydrogen gas are studied. Taking the candidates of reactors such as a toroidal system and an open system for application of the new concepts, the expected efficiency and a concept of plant system are investigated.

  12. Fast reactor programme in India

    Indian Academy of Sciences (India)

    P Chellapandi; P R Vasudeva Rao; Prabhat Kumar

    2015-09-01

    Role of fast breeder reactor (FBR) in the Indian context has been discussed with appropriate justification. The FBR programme since 1985 till 2030 is highlighted focussing on the current status and future direction of fast breeder test reactor (FBTR), prototype fast breeder reactor (PFBR) and FBR-1 and 2. Design and technological challenges of PFBR and design and safety targets with means to achieve the same are the major highlights of this paper.

  13. Turning points in reactor design

    International Nuclear Information System (INIS)

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems

  14. Optimal control of nuclear reactors

    International Nuclear Information System (INIS)

    The modern control theory is applied to the design of control systems for experimental nuclear reactors that do not belong to power reactors, the component forms of optimal control systems for nuclear reactors are demonstrated. The adoption of output quadratic integral criterion and incomplete state feedback technique can make these systems both efficient and economical. Moreover, approximate handling methods are given so as to simplify the calculations in design. In addition, the adoptable reference values of parameters are given in the illustration

  15. Turning points in reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Beckjord, E.S.

    1995-09-01

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

  16. Acceptability of reactors in space

    International Nuclear Information System (INIS)

    Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it does not appear that reactors add measurably to the risk associated with the Space Transportation System

  17. INR TRIGA Research Reactors: A Neutron Source for Radioisotopes and Materials Investigation

    International Nuclear Information System (INIS)

    At the INR there are 2 high intensity neutron sources. These sources are in fact the two nuclear TRIGA reactors: TRIGA SSR 14 MW and TRIGA ACPR. TRIGA stationary reactor is provided with several in-core irradiation channels. Other several out-of-core irradiation channels are located in the vertical channels in the beryllium reflector blocks. The maximum value of the thermal neutron flux (E14 cm-2s-1 and of fast neutron flux (E>1 MeV) is 6.89×1013 cm-2s-1. For neutron activation analysis both reactors are used and k0-NAA method has been implemented. At INR Pitesti a prompt gamma ray neutron activation analysis devices has been designed, manufactured ant put into operation. For nuclear materials properties investigation neutron radiography methods was developed in INR. For these purposes two neutron radiography devices were manufacture, one of them underwater and other one dry. The neutron beams are used for investigation of materials properties and components produced or under development for applications in the energy sector (fission and fusion). At TRIGA 14 MW reactor a neutron difractormeter and a SANS devices are available for material residual stress and texture measurements. TRIGA 14 MW reactor is used for medical and industrial radioisotopes production (131I, 125I, 192Ir, etc) and a method for 99Mo-99Tc production from fission is under developing. At INR Pitesti several special programmes for new types of nuclear fuel behavior characterization are under development. (author)

  18. Advanced Fission Reactor Program objectives

    International Nuclear Information System (INIS)

    The objective of an advanced fission reactor program should be to develop an economically attractive, safe, proliferation-resistant fission reactor. To achieve this objective, an aggressive and broad-based research and development program is needed. Preliminary work at Brookhaven National Laboratory shows that a reasonable goal for a research program would be a reactor combining as many as possible of the following features: (1) initial loading of uranium enriched to less than 15% uranium 235, (2) no handling of fuel for the full 30-year nominal core life, (3) inherent safety ensured by core physics, and (4) utilization of natural uranium at least 5 times as efficiently as light water reactors

  19. Reactor safety - an international task

    International Nuclear Information System (INIS)

    The dimensions and the significance of the task of ensuring reactor safety can be defined on the basis of experiences gained from Harrisburg and Chernobyl. The countries that use nuclear energy are tied together to a community by virtue of the risk they share. Therefore the GRS is working in close cooperation with the EC, OECD, IAEO and COMECON. This results in safety examinations of the Greifswald reactor, safety analyses of nuclear reactors in Germany, France and the USA and also considerations on the safety demands to be placed on new reactor concepts. (DG)

  20. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements