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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. Simplified numerical simulation of hot channel in sodium cooled reactor

    International Nuclear Information System (INIS)

    The thermal-hydraulic parameter values that restrict the operation of a liquid sodium cooled reactor are not established by the average conditions of the coolant in the reactor core but by the extreme conditions of the hot channel. The present work was developed to analysis of hot channel of a sodium cooled reactor, adapting to this reactor an existent simplified model for hot channel of pressurized water reactor. The model was applied for a standard sodium reactor and the results are considered satisfatory. (author)

  3. Corrosion and hydridation features of RBMK type reactor technological channels

    International Nuclear Information System (INIS)

    Generalization results, obtained in the course of monitoring the corrosion state and hydridation of RBMK-1000 and RBMK-1500 reactor technological channels (TC) are presented. It is shown, that the corrosion behaviour of TC tube metal in reactors differs notably. Comparison of data on hybridization of RBMK-100 and RBMK-1500 reactor technological tubes allows one to suppose a possibly higher tendency to hydrogen absorption in Zr - 2.5% of Nb alloy under TMT-1 and TMT-2 states

  4. 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

  5. Pu-breeding feasibility in irradiation channels of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tomanin, A., E-mail: alice.tomanin@jrc.ec.europa.e [Institute for the Protection and the Security of the Citizen, Joint Research Centre Ispra, Via E. Fermi, 2749, I-21027 ISPRA, Varese (Italy); University Ghent, Engineering Faculty, B-9052 Gent-Zwijnaarde (Belgium); Peerani, P. [Institute for the Protection and the Security of the Citizen, Joint Research Centre Ispra, Via E. Fermi, 2749, I-21027 ISPRA, Varese (Italy); Janssens-Maenhout, G. [Institute for Environment and Sustainability, Joint Research Centre Ispra, Via E. Fermi, 2749, I-21027 ISPRA (Italy); University Ghent, Engineering Faculty, B-9052 Gent-Zwijnaarde (Belgium)

    2011-02-15

    Research highlights: Clandestine plutonium production in irradiation channels of research reactors is a safeguard concern. IAEA concentrates safeguard measures on research reactors with thermal power greater that 25 MWth. The breeding potential in irradiation channels scales with reactor power and available space for irradiation samples. From about 10 MWth and 0.05 m{sup 3} onwards the proliferation concern raises with more than 2 kg of yearly plutonium breeding capability. - Abstract: The breeding potential in the irradiation channels of research reactors is of safeguards concern, because of lacking continuous supervision on the type of experiments in all the irradiation channels. Moreover, the irradiation time can be optimized in order to breed high quality weapon grade plutonium. With regard to the safeguards measures currently adopted, IAEA concentrates its efforts on those reactors whose thermal power is greater than 25 MWth, because it was calculated that a 25 MWth LEU-fuelled reactor produces not more than one Significant Quantity of Pu (8 kg)/year in its spent fuel and a HEU-fuelled reactor of this power would require an annual reload of not more than one Significant Quantity of U{sub 235} (25 kg). In order to investigate whether it would be possible to determine an analogous power level threshold to estimate the clandestine plutonium production capability of different research reactors, the Monte Carlo method was used to determine the neutron flux in the irradiation channels and to calculate the plutonium breeding potential for three different reactor types: (1) a Triga Mark II with 250 kWth, representative for a small size research reactor; (2) a Material Test Reactor (MTR) with 5 MWth, representative for a medium size research reactor; (3) a High Flux Reactor (HFR) with 45 MWth, representative for a large size research reactor. It was observed that the most important factors for plutonium breeding are the neutron flux (to which reaction rates are

  6. Shutdown channels and fitted interlocks in atomic reactors

    International Nuclear Information System (INIS)

    This catalogue consists of tables (one per reactor) giving the following information: number and type of detectors, range of the shutdown channels, nature of the associated electronics, thresholds setting off the alarms, fitted interlocks. These cards have been drawn up with a view to an examination of the reactors safety by the 'Reactor Safety Sub-Commission', they take into account the latest decisions. The reactors involved in this review are: Azur, Cabri, Castor-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, and Ulysse. (authors)

  7. A new log-linear safety channel for reactor power measurement

    International Nuclear Information System (INIS)

    At Pakistan Research Reactor 1(PARR-1) the safety channels measure reactor power from few kW to 150% of full reactor power and provide reactor scram in case of high neutron flux or malfunctioning in the electronics of the channels. The channels are linear and they cover only three decades of reactor power level. So, the channels can not provide any information about the reactor power and channel operating condition at low reactor power up to few kW. A new log-linear safety channel has been developed for reactor power measurements in logarithmic and linear mode. The channel also measures reactor period. Due to the logarithmic mode of operation the channel can cover from startup range up to 200% of full reactor power. The new channel with one ion chamber has complete functionality of existing safety channels and intermediate power range logarithmic channels. So, on the replacement of existing safety channels with the new log-linear channels the existing logarithmic dc channels will not be required. The new channel will reduce the number of channels, improve fault capabilities and, hence, improve safety and availability of the system. The new channel will also be used in nuclear power plant. The log-linear safety channel has been tested at PARR-1 and the results are found in very good agreement with the designed specifications. This paper presents design and construction of the channel and field test results. (author)

  8. Advances in fuel channel technology for CANDU reactors

    International Nuclear Information System (INIS)

    The components of the CANDU fuel channels are being developed to have service lives of over 30 years with large margins of safety. Information from research programs and the examination of components removed from reactors has enable improvements to be made to pressure tubes, spacers, calandria tubes and end fittings. Improvements have also been made to the channel design to facilitate planned retubing. (author). 22 refs., 5 tabs., 31 figs

  9. Control of helium activity in the fuel reactor channels

    International Nuclear Information System (INIS)

    The objective of this task was to study the possibility of detecting a damaged fuel channel, and to introduce automated procedure for continuous control of reactor channels during operation. The existing control systems at the RA reactor (permanent control of heavy water and helium activity, radiation monitoring of heavy water and helium system, measurements of fire damp gas percent) are not sufficient for fast detection of fuel element failures. Since a 'hot' fuel channel cannot be removed from the core because it should be cooled in the core by heavy water circulation, it is not possible to prevent contamination of heavy water by fission products. It is concluded that it is not indispensable to detect the failed fuel element promptly, i.e. that tome is not a critical issue

  10. SCW pressure-channel nuclear reactor some design features

    International Nuclear Information System (INIS)

    Concepts of nuclear reactors cooled with water at supercritical pressure were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (∼$1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 Mpa and outlet temperature up to 625degC), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems. (author)

  11. Exploitation questions regarding channel type reactors: water graphite channel reactors (operation, reconstruction, advantages and disadvantages)

    International Nuclear Information System (INIS)

    An overview of up-grade of the RBMK-type reactors is given. I this paper the core design and core monitoring, pressure boundary integrity, RBMK basic design and safety improvements emergency core cooling system (ECCS) as well as reactor cavity overpressure protection system (RCOPS) are discussed

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

    International Nuclear Information System (INIS)

    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

  13. 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

  14. 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

  15. THESEE-3, Orgel Reactor Performance and Statistic Hot Channel Factors

    International Nuclear Information System (INIS)

    1 - Nature of physical problem solved: The code applies to a heavy-water moderated organic-cooled reactor channel. Different fuel cluster models can be used (circular or hexagonal patterns). The code gives coolant temperatures and velocities and cladding temperatures throughout the channel and also channel performances, such as power, outlet temperature, boiling and burn-out safety margins (see THESEE-1). In a further step, calculations are performed with statistical values obtained by random retrieval of geometrical in- put data and taking into account construction tolerances, vibrations, etc. The code evaluates the mean value and standard deviation for the more important thermal and hydraulic parameters. 2 - Method of solution: First step calculations are performed for nominal values of parameters by solving iteratively the non-linear system of equations which give the pressure drops in subchannels of the current zone (see THESEE-1). Then a Gaussian probability distribution of possible statistical values of the geometrical input data is assumed. A random number generation routine determines the statistical case. Calculations are performed in the same way as for the nominal case. In the case of several channels, statistical performances must be adjusted to equalize the normal pressure drop. A special subroutine (AVERAGE) then determines the mean value and standard deviation, and thus probability functions of the most significant thermal and hydraulic results. 3 - Restrictions on the complexity of the problem: Maximum 7 fuel clusters, each divided into 10 axial zones. Fuel bundle geometries are restricted to the following models - circular pattern 6/7, 18/19, 36/67 rods, with or without fillers. The fuel temperature distribution is not studied. The probability distribution of the statistical input is assumed to be a Gaussian function. The principle of random retrieval of statistical values is correct, but some additional correlations could be found from a more

  16. 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

  17. Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors

  18. 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

  19. Visual inspections of N Reactor horizontal control rod channels

    International Nuclear Information System (INIS)

    Safety surveillance is performed in horizontal control rod (HCR) channels to locate conditions which could slow or block rod travel. The findings guide the application of preventive measures to assure eventual rod motion impairment will not occur. Borescopes and, more recently, miniaturized closed circuit television (CCTV) cameras have been used for these examinations. Inspections and measurement results are documented in annual surveillance reports, however reported CCTV observations have been limited to highlights. The objective of this report is to catalogue the CCTV recordings in a format suitable for analysis and interpretation and to ease the access to any desired location by noting tape counter readings corresponding with each tube block in view. Searching file tapes for conditions in a specific areas in the past required counting blocks as they passed the camera to determine the distance from a feature like the edge of the reflector or a steam vent gap. This report adds the observations from recent rod channel inspections (1987 and 1988) to a comprehensive survey of graphite conditions in the moderator and reflector regions of the N Reactor core. When completed, the stand-by status of graphite components will be available for use in restart or decommissioning deliberations

  20. Two-channel model for dynamic analysis of GCR type reactor, Mathematical model

    International Nuclear Information System (INIS)

    A two-channel model for reactor dynamic analysis was developed. It enables representation of time dependent behaviour of a reactor as a whole and to obtain time and space dependent changes of temperature in any of the reactor channel. Model is suitable for follow-up of phenomena in limited time intervals up to few tens of minutes, since long term variations caused by fuel burnup and fission products are not taken into account in the model. Parameters are defined to cover the reactor power range from minimum to maximum. Model describes two main processes in the reactor: power generation dependent on the neutron flux and cooling

  1. New start-up channels and multichannel analyzer at the RB reactor

    International Nuclear Information System (INIS)

    New start-up channels and a multichannel analyzer were purchased in 1977 for the RB reactor. Both start-up channels contain BF3 neutron detectors, preamplifier, amplifier, single-channel analyzer, scaler, ratemeter, control unit, recording instrument. This document contains detailed technical description of these devices as well as characteristics of the multichannel analyzer which is being tested and will be used for measuring irradiation in the vicinity of the reactor

  2. 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

  3. 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

  4. 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

  5. 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

  6. Post-reactor research of fuel pins, radiated in nitrogen-cooling technological channels ETC

    International Nuclear Information System (INIS)

    The state of ceramics fuel pins after long-term (about 10 hours) tests in flowing nitrogen-cooling technological channels ETC of IVG.1 reactor was researched. Rather high radiation-chemical stability of the fuel pins under definite conditions of reactor tests was determined. (author)

  7. Specific features of the WWR-K reactor horizontal channel as applied to BNCT

    International Nuclear Information System (INIS)

    For several years the studies related to adaptation of one of the horizontal channels of the WWR-K reactor in view of treatment of malignant tumors by neutron capture therapy techniques are carried out at the Institute of Nuclear Physics in the frame of the republican research program 'Development of Nuclear Power in Kazakhstan'. The need in NCT in Kazakhstan is rather urgent, because many people suffer from cancer. The neutron capture therapy (NCT) is widely used over the world but in special medical reactors. An idea to use the research nuclear reactor for (medical purposes seems to be attractive, because financial expenditures for reactor operation in this case will be considerably lower. The reactor experimental horizontal channel GK-1, transporting the neutron beam from the reactor core via biological shield to the reactor hall, was chosen for NCT goals. The neutron beam neutronic parameters over the channel length have been estimated with and without the neutron guide installed in the reactor core. The neutron guide is to increase the neutron flux density at the channel exit and to reduce the beam gamma component. Prior to experimenting, the corresponding calculations were carried out by means of the computer codes as well as modeling experiments in the critical assembly (zero-power reactor). The following outcomes have been obtained: The patterns of the thermal/fast neutron flux density distribution over the length of the horizontal channels have been found for various reactor power levels and various versions of neutron guide. The thermal/fast neutron flux density at the channel exit comprise 1.2·109/5·108 neutrons cm-2c-1. The lead neutron guide, installed in the reactor core, is found to double the thermal neutron flux density at the channel exit. Further studies imply participation of a multi-discipline team of researchers (physicists, biologists, radiologists, oncologists, etc.), and it seems that the appropriate international NCT project will be the

  8. Hot channel factors determination and evaluation for water-cooled nuclear research reactors performance

    International Nuclear Information System (INIS)

    This paper presents the general outlines of the evaluation, selecting, determination, and applying peaking hot channel factors for nuclear research reactors. As an example, the uncertainty in the heat transfer coefficient is a major contributor to the reduction in nuclear research reactor performances and thermal hydraulic safety margins, where the uncertainties are due to the reactor fuel coolant channel design fabrication defects (fuel meat and clad thickness uncertainties), and the fuel element's heterogeneity. Both the uncertainties are important contributors and an area where more information may be useful in reducing this uncertainty. In this case we take the WWR-M2 nuclear reactor (Russian type) fuel coolant channel (fuel Assembly) as sample problem for applying and determination the peaking factors, and understanding this method. (author)

  9. Design and testing of integrated circuits for reactor protection channels

    International Nuclear Information System (INIS)

    Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. Purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing

  10. Automatic control of neutron flux in experimental channels of the WWR-M type reactors

    International Nuclear Information System (INIS)

    The flowsheet of the neutron flux local regulator intended for maintaining the given level of neutron flux distribution in experimental channels of the WWR-M type reactor under stationary and transition modes, is suggested. The functional diagram of the electron regulation block (ERB) in considered. The regulator is tested when the reactor operates with the capacity of 13 MWt along with the staff system of automated regulation and without it. The experiments carried out demonstrate the stable operation of the entire control system and good performance characteristics of the ERR block. The conclusion is made that the suggested method of neutron flux automated regulation in experimental channels can be successfully extended to a higher number of experimental channels and applied at other research reactors. Small size fission chambers and direct charging detectors can be used in local systems as sensors

  11. 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

  12. 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)

  13. Dismantling of Loop-Type Channel Equipment of MR Reactor in NRC 'Kurchatov Institute' - 13040

    International Nuclear Information System (INIS)

    In 2009 the project of decommissioning of MR and RTF reactors was developed and approved by the Expert Authority of the Russian Federation (Gosexpertiza). The main objective of the decommissioning works identified in this project: - complete dismantling of reactor equipment and systems; - decontamination of reactor premises and site in accordance with the established sanitary and hygienic standards. At the preparatory stage (2008-2010) of the project the following works were executed: loop-type channels' dismantling in the storage pool; experimental fuel assemblies' removal from spent fuel repositories in the central hall; spent fuel assembly removal from the liquid-metal-cooled loop-type channel of the reactor core and its placement into the SNF repository; and reconstruction of engineering support systems to the extent necessary for reactor decommissioning. The project assumes three main phases of dismantling and decontamination: - dismantling of equipment/pipelines of cooling circuits and loop-type channels, and auxiliary reactor equipment (2011-2012); - dismantling of equipment in underground reactor premises and of both MR and RTF in-vessel devices (2013-2014); - decontamination of reactor premises; rehabilitation of the reactor site; final radiation survey of reactor premises, loop-type channels and site; and issuance of the regulatory authorities' de-registration statement (2015). In 2011 the decommissioning license for the two reactors was received and direct MR decommissioning activities started. MR primary pipelines and loop-type facilities situated in the underground reactor hall were dismantled. Works were also launched to dismantle the loop-type channels' equipment in underground reactor premises; reactor buildings were reconstructed to allow removal of dismantled equipment; and the MR/RTF decommissioning sequence was identified. In autumn 2011 - spring 2012 results of dismantling activities performed are: - equipment from underground rooms (No. 66, 66

  14. 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.)

  15. OPAL REACTOR: Calculation/Experiment comparison of Neutron Flux Mapping in Flux Coolant Channels

    Energy Technology Data Exchange (ETDEWEB)

    Barbot, L.; Domergue, C.; Villard, J. F.; Destouches, C. [CEA, Paris (France); Braoudakis, G.; Wassink, D.; Sinclair, B.; Osborn, J. C.; Huayou, Wu [ANSTO, Syeney (Australia)

    2013-07-01

    The measurement and calculation of the neutron flux mapping of the OPAL research reactor are presented. Following an investigation of fuel coolant channels using sub-miniature fission chambers to measure thermal neutron flux profiles, neutronic calculations were performed. Comparison between calculation and measurement shows very good agreement.

  16. Analysis of channel with special heat transfer characteristics for cooling nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Improvement of heat transfer characteristics of nuclear reactor cooling channels is analysed. A generalization of the Ambrok solution for the energy integral equation is proposed. A new improved passage shape is determined. Theoretical results are compared with experimental data. The agreement is excellent. (author)

  17. SCW Pressure-Channel Nuclear Reactors: Some Design Features and Concepts

    International Nuclear Information System (INIS)

    Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950's and 1960's in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with supercritical water (SCW) became attractive again as the ultimate development path for water-cooling. The main objectives of using SCW in nuclear reactors are 1) to increase the thermal efficiency of modern nuclear power plants (NPPs) from 33 -- 35% to about 40 -- 45%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (∼$ 1000 US/kW). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625 deg. C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia. Design features related to both channels and fuel bundles are discussed in this paper. Also, Russian experience with operating supercritical steam heaters at NPP is presented. The main conclusion is that development of SCW pressure-channel nuclear reactors is feasible and significant benefits can be expected over other thermal energy systems. (authors)

  18. Channel blockage accident analysis for research reactors with MTR- type fuel elements

    International Nuclear Information System (INIS)

    It is the purpose of this study to investigate the feasibility of removing the residual decay heat from core of TR-2 ,which is a pool-type research reactor, after a channel blockage accident event and to identify the principal factors involved in cooling process. To analyze this accident scenery, THEAP-I computer code, which is a single phase transient 3-D structure/1-D flow thermal hydraulics code developed with the aim to contribute mainly to the safety analysis of the open pool research reactors, was modified and used. All of the analysis results figured out the fact that the core melting was inevitable in case of an uninterrupted operation (continuous operation) preceding a channel blockage accident of the TR-2 Reactor. Such a result will even be met if the blockage occurs only in a single fuel element. The results of analysis are expressed in terms of temperature field distribution as a function of time

  19. 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.

  20. 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...

  1. 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)

  2. Verification of performance of the power percentage channel for the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    It was found that the response that gives the power percent channel is correct, given the positive results of the independent tests that were carried out to the gamma ionization chamber and the electronics associated to this channel. Regarding the gamma chamber, it was verified that the appropriate operation voltage is 800 V, and that for operations in stationary state to 1 MW during 2 h, presented maximum variations of 3%. Also it was determined that the degradation percentage in the sensitivity to the gamma radiation is 10.24%, because this chamber has not been changed since the reactor enters in operation at November 8, 1968 by what will be considered to short term the substitution of the same one due to the burnt that it presents. In connection with the electronics of the channel, it was simulated the response of the chamber for intervals of 6 h and in the 4 analyzed cases the response of the channel was lineal. (Author)

  3. 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)

  4. 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

  5. In-pile creep test technique for zirconium alloys examination in BR-10 reactor channels

    International Nuclear Information System (INIS)

    The irradiation enhanced creep phenomenon was discovered in stainless steels as a specific physical process accompanying high-intensity neutron flux irradiation in fast reactors. IPPE is also experienced in irradiation creep test activities, studying different types of materials under irradiation in BR-10 fast reactor. Series of in-channel type test facilities were constructed and tested in BR-10 reactor's 'dry' channels in order to carry out full-scale instrumented examination regarded to in-pile creep behaviour of different reactor materials. As a result, a specific test technique, named 'Tensometric method', has been developed and experimentally proved to be power enough in order to investigate irradiation creep of materials right in situ under neutron irradiation. The main peculiarity of test facility, which is constructed to apply the tensometric method, consists in absence of any special deformation-measurement cell at all. The in-pile creep strain measurement technique developed at IPPE is based on the non-direct measurement of specimen's deformation (either linear tensile strain or angular twisting one), which directly affects the loaded draws' tension parameters. Starting from 1993, in-pile creep experiments to investigate in-reactor creep behaviour of E110 and E635 zirconium alloys were carried out in BR-10. Experimental results and data collected during more than 20-year of BR-10 in-reactor creep test experience can be assumed as a strong evidence that the tensometric technique is a powerful instrument, which can give a chance to study different irradiation effects on reactor materials directly under irradiation. (author)

  6. Tasks related to increase of RA reactor exploitation and experimental potential, 03. Crane for handling the vertical experimental channels of the RA reactor - design project

    International Nuclear Information System (INIS)

    Within the work related to improvement of experimental potential of the RA reactor, this document describes the design project of the new crane for handling the vertical experimental channels of the RA reactor, engineering drawings of the crane main elements, mechanical part, design project of the electrical part of the crane and cost estimation

  7. Experimental and MCNP calculations of neutron flux parameters in irradiation channel at Es-Salam reactor

    International Nuclear Information System (INIS)

    The Algerian research reactor (Es-Salam) is a 15 MW heavy water reactor type, operating since 1992. It became essential to characterize the neutron field in the most useful irradiation positions, in order to guarantee the accuracy in the application of k0-neutron activation analysis (k0-NAA). Experimental value of the thermal to epithermal neutron flux ratio (f) and of the deviation of the epithermal neutron spectrum from 1/E shape (α) were determined using different methods. This work focuses the verification of Monte Carlo neutron flux calculation in typical irradiation channel. Comparison of the results for parameter f obtained experimentally and by Monte Carlo simulations shows good agreement in the irradiation channel studied. The difference between both results is about 2.08%. (author)

  8. Biological samples positioning device for irradiations on a radial channel at the nuclear research reactor

    International Nuclear Information System (INIS)

    For the demand of an experimental device for biological samples positioning system for irradiations on a radial channel at the nuclear research reactor in operation was constructed and started up a device for the place and remove of the biological samples from the irradiation channels without interrupting the operation of the reactor. The economical valuations are effected comparing with another type of device with the same functions. This work formed part of an international project between Cuba and Brazil that undertook the study of the induced damages by various types of ionizing radiation in DNA molecules. Was experimentally tested the proposed solution, which demonstrates the practical validity of the device. As a result of the work, the experimental device for biological samples irradiations are installed and operating in the radial beam hole No3(BH3) for more than five years at the IEA-R1 Brazilian research reactor according to the solicited requirements the device. The designed device increases considerably the type of studies can be conducted in this reactor. Its practical application in research taking place in that facility, in the field of radiobiology and dosimetry, and so on is immediate

  9. Independent loop channel for testing the fuel elements for the BREST-OD-300 reactor

    International Nuclear Information System (INIS)

    The problems connected with irradiation tests of the fuel element simulators for the BREST-OD-300 fast lead cooled reactor in the BOR-60 reactor are discussed. The mononitride uranium-plutonium fuel in the container-type fuel elements is used in the BREST-OD-300 pilot reactor being under design. An independent loop channel with lead coolant is created in the BOR-60 reactor cell D-23 for testing the fuel element simulators. The channel thermal power is of the order of 45 kw, its power service life time amounts to 11000 hours, the fuel assembly inlet/outlet temperatures are up to 480/540 deg C, the lead coolant flow rate is up to 4.5 kg/s. The tests will give an opportunity to refine the information on the fuel properties in low-temperature range, as well as to determine the fuel can radiation and corrosion resistance under simultaneous effects of coolant, irradiation and spacing grids

  10. 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)

  11. 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

  12. 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.

  13. SUBCHANFLOW: a thermal hydraulic sub-channel program to analyse fuel rod bundles and reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, V.; Imke, U.; Ivanov, A. [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Gomez, R., E-mail: Victor.Sanchez@kit.ed [University of Applied Sciences Offenburg, Badstr. 24, 77652 Offenburg (Germany)

    2010-10-15

    The improvement of numerical analysis tools for the design and safety evaluation of reactor cores in a continuous effort in the nuclear community not only to improve the plant efficiency but also to demonstrate high degree of safety. Investigations at the Institute of Neutron Physics and Reactor Technology of the Karlsruhe Institute of Technology (KIT) are focused on the further development and qualification of subchannel and system codes using experimental data. The majority of sub-channel codes in use likes Thesis, Bacchus, Cobra and Matra, were developed in the seventies and eighties. The programming style is rather obsolete and most of these codes are working internally with British Units instead of Si-Units. In the case of water, outdated steam tables are used. Both the trends to improve the efficiency of light water reactors (LWR) and the involvement of KIT in European projects related to the study of the technical feasibility of different fast reactors systems reinforced the need for the development and improvement of sub-channel codes, since they will play a key role in performing better designs as stand-alone tools or coupled to neutron physical codes (deterministic or stochastic). Hence, KIT started the development of a new sub-channel code SUBCHANFLOW based on the Cobra-family. SUBCHANFLOW is a modular code programmed in Fortran-95 with dynamic memory allocation using Si-units. Different fluids like liquid metals and water are available as coolant. In addition some models were improved or replaced by new ones. In this paper the structure, the physical models and the current validation status will be presented and discussed. (Author)

  14. Effect of Reactor Channel Modelling On Rewetting for AHWR Fuel Cluster

    International Nuclear Information System (INIS)

    Effect of reactor channel modeling on the rewetting pattern has been studied for the proposed pressure tube type, natural circulation cooled Advanced Heavy Water Reactor (AHWR). A direct quenching of the nuclear fuel pins with cold water from Emergency Core Cooling System (ECCS) has been proposed to limit the consequences of Loss of Coolant Accident (LOCA). As the injection is designed to take place from the central position of the fuel cluster, the water injection experiences several path resistances during quenching process due to the compact cluster geometry, steam condensation and cross current two phase flow. Assessment of rewetting of the fuel cluster during maximum credible accident condition is required to prove the capability of mode of ECCS injection. Analytical assessment has been carried out in detail considering the effect of space discretization, radial power profile in fuel cluster, subchannel effect and Counter Current Flow Limitation (CCFL) effect. Thermal-hydraulic computer codes RELAP5/MOD3.2 has been used for this study. For each study a RELAP5 specific model has been developed from a reference model. The study shows sensitivity of reactor channel modelling on fuel heatup and the rewetting period. The injection performance is found to be satisfactory with all the models as all fuel pins get re-wetted with varied rewetting period. (authors)

  15. 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.

  16. 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)

  17. 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

  18. Analysis of systematic error deviation of water temperature measurement at the fuel channel outlet of the reactor Maria

    International Nuclear Information System (INIS)

    The reactor Maria has two primary cooling circuits; fuel channels cooling circuit and reactor pool cooling circuit. Fuel elements are placed inside the fuel channels which are parallely linked in parallel, between the collectors. In the course of reactor operation the following measurements are performed: continuous measurement of water temperature at the fuel channels inlet, continuous measurement of water temperature at the outlet of each fuel channel and continuous measurement of water flow rate through each fuel channel. Based on those thermal-hydraulic parameters the instantaneous thermal power generated in each fuel channel is determined and by use of that value the thermal balance and the degree of fuel burnup is assessed. The work contains an analysis concerning estimate of the systematic error of temperature measurement at outlet of each fuel channel and so the erroneous assessment of thermal power extracted in each fuel channel and the burnup degree for the individual fuel element. The results of measurements of separate factors of deviations for the fuel channels are enclosed. (author)

  19. Diagnostic technology for degradation of feeder pipes and fuel channels in CANDU reactor

    International Nuclear Information System (INIS)

    Degradation of critical components of nuclear power plants has become important as the operating years of plants increase. The necessity of degradation study including detection and monitoring technology has raised its head. Because the feeder pipes and the fuel channels are particularly one of the critical components in CANDU nuclear plant, they are treated as a major research target in order to counteract the possible problems and establish the counterplan for the improvement of CANDU reactor safety. To ensure the integrity of feeder pipes and fuel channels in CANDU nuclear plant, the following 3 research tasks were performed in the first stage. - Development of a model for prediction of feeder wall thinning - Development of RFEC detection technology - Development of ICFD noise signal analysis. The technologies developed in this study could contribute to the nuclear safety and estimation of the remaining life of operating CANDU nuclear power plants

  20. 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

  1. Thermo-mechanical behaviour of coolant channels for heavy water reactors under accident conditions

    International Nuclear Information System (INIS)

    The objective of nuclear safety research programme is to develop and verify computer models to accurately predict the behavior of reactor structural components under operating and off normal conditions. Indian Pressurised Heavy Water Reactors (PHWRs) are tube type of reactors. The coolant channel assemblies, being one of the most important components, need detailed analysis under all operating conditions as well as during postulated conditions of accidents for its thermo-mechanical behaviour. One of the postulated accident scenarios for heavy water moderated pressure tube type of reactors i.e. PHWRs is Loss Of Coolant Accident (LOCA) coincident with Loss Of Emergency Core Cooling System (LOECCS). In this case, even though the reactor is tripped, the decay heat may not be removed adequately due to low or no flow condition and inventory depletion of primary side. Since the emergency core cooling system is presumed to be not available, the cooling of the fuel pins and the coolant channel assembly depends on the moderator cooling system, which is assumed to be available. Moderator cooling system is a separate system in PHWRs. In PHWRs, the fuel assembly is surrounded by pressure tube, an annulus insulating environment and a concentric calandria tube. In this postulated accident scenario, a structural integrity evaluation has been carried out to assess the modes of deformation of pressure tube-calandria tube assembly in a tube type nuclear reactor. The loading of pressure and temperature causes the pressure tube to sag/balloon and come in contact with the outer cooler calandria tube. The resulting heat transfer could cool and thus control the deformation of the pressure tube thus introducing inter-dependency between thermal and mechanical contact behaviour. The amount of heat thus expelled significantly depends on the thermal contact conductance and the nature of contact between the two tubes. Deformation of pressure tube creates a heat removal path to the relatively

  2. 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

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

    International Nuclear Information System (INIS)

    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 1.5245x10-4 s was recorded for the new boron carbide designed model while a value of 1.5571x10-7 s was recorded for the original MCNP design of the GHARR-1.

  4. 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. PMID:20637646

  5. Supercritical water-cooled nuclear reactors: NPP layouts and thermal design options of pressure channels

    International Nuclear Information System (INIS)

    Research activities are currently underway worldwide to develop generation IV nuclear reactor concepts with the objective of improving thermal efficiency and increasing economic competitiveness of generation IV nuclear power plants (NPPs) compared to modern NPPs. The supercritical water-cooled reactor (SCWR) concept is one of six generation-IV options chosen for further investigation and development in many countries worldwide, including Canada. Water-cooled reactors operating at subcritical pressures (7 - 16 MPa) have provided significant electricity production for the past 50 years. However, the thermal efficiency of current NPPs is not very high (30 - 35%). As such, more competitive designs with higher thermal efficiencies, close to those of modern supercritical (Sc) thermal power plants (45 - 50%), need to be developed and implemented. Previous studies have shown that direct cycles with single-reheat and no-reheat configurations are the best options for an SCWR concept. There are a few technical challenges associated with the single-reheat and no-reheat supercritical water (SCW) NPP configurations. The single-reheat cycle requires nuclear steam-reheat, thus increasing the complexity of the reactor core design. Conversely, the major technical challenge associated with an Sc no-reheat turbine is high moisture content in the low-pressure-turbine exhaust. The SCWR-core concept investigated in this paper is based on a generic pressure-tube (pressure-channel) reactor with a 43-element bundle string cooled with supercritical water. The considered 1200-MW el reactor has the following operating parameters: pressure of 25 MPa and reactor inlet/outlet temperatures of 350/625 C. Previous studies have shown that is uranium dioxide (UO2) is used, the fuel centerline temperature might exceed the industry accepted limit of 1850 C. Therefore, this paper investigates a possibility of using uranium carbide (UC), uranium nitride (UN), uranium dicarbide (UC2), uranium dioxide plus

  6. Hydrodynamic characterization and evaluation of an open channel reactor for the degradation of paracetamol

    International Nuclear Information System (INIS)

    The conventional wastewater treatment plants do not guarantee the degradation of Persistent Organic Pollutants (POPs). Advanced oxidation processes, like photodegradation that use artificial ultraviolet and solar radiation, are proposed as an alternative for the treatment of contaminated water with POPs. In the present work, the hydrodynamic characterization and evaluation of an open channel reactor for the degradation of paracetamol are presented. The hydrodynamic characterization was performed through the analysis of the residence time distribution using a radioisotope 99mTc. This process was done in two steps. First, the open channel reactor was evaluated in continuous mode operation. To study the influence of the fluid volume in the reactor and the diameter of the flow distributor's orifices on the flow pattern, an experimental 32 design with two replicas in the center was used. The dependent variables were the number of perfectly mixed tanks (J), the mean residence time of the model (τ) and the experimental mean residence time (Trm). The model of perfectly mixed tanks in series exchanging with stagnant zones was assumed as the best model. In a second moment, the mixing time of the system operating in close loop mode was determined. Finally, the degradation of paracetamol in aqueous dissolution trough photolysis, photolysis intensified with H2O2, photo-Fenton with artificial ultraviolet radiation and photo-Fenton with solar radiation was evaluated. The results show that the photo-Fenton processes employing artificial ultraviolet and solar radiation warranty the total degradation of the pharmaceutical after 15 minutes of reaction. (Author)

  7. Radiological consequences of a postulated cooling channel blockage incident at a pool-type research reactor

    International Nuclear Information System (INIS)

    An assessment of the radiological consequences of a postulated coolant flow blockage incident at the Hoger Onderwijs Reactor (HOR) is being presented. The HOR is a swimming-pool type research reactor with a maximum licensed power of 3 MW. Assuming a sudden blockage of cooling channels in the high power density region of the core, the source term for the release of radioactivity into the environment was calculated. The magnitude of this source term is required by actions of the HOR protection system as well as by physical processes acting on the fission products. Hence, almost 99% of the calculated release from the containment consists of noble gases; most of the aerosol-type activity set free in the environment results from the decay of these noble gases. The deposit of long-living radionuclides outside the reactor building is very low. Radio-iodine will be the main contributor to the environmental radiation dose, ingestion of contaminated food being the critical pathway. Despite the conservativeness of most assumptions used, the calculated thyroid dose for critical individuals at all distances from the site boundary remains well below the emergency reference levels recommended by national and international organizations and the national dose limit for members of the public

  8. 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)

  9. Application of Shuttle Remote Manipulator System technology to the replacement of fuel channels in the Pickering CANDU reactor

    International Nuclear Information System (INIS)

    Spar Aerospace Limited of Toronto was the prime contractor to the National Research Council of Canada for the design and development of the Shuttle Remote Manipulator (SRMS). Spar is presently under contract to Ontario Hydro to design and build a Remote Manipulation Control System to replace the fuel channels in the Pickering A Nuclear Generating Station. The equipment may be used to replace the fuel channels in six other early generation CANDU reactors

  10. On adaptation of the WWR-K reactor horizontal channel to the neutron-capture therapy tasks

    International Nuclear Information System (INIS)

    Full text: At the Institute of Nuclear Physics of the National Nuclear Center of the Republic of Kazakhstan water-water research reactor the studies related to adaptation of one of the reactor horizontal channel to treatment of malignant tumors by techniques of neutron-capture therapy (NCT) are carried out for several years. The studies are implemented in the frame of the republican research program 'Development of Nuclear Power in Kazakhstan'. The need in NCT in Kazakhstan is rather urgent, because the number of people suffered from cancer diseases is large. The NCT technique is widely used over the world but in special medical reactors. An idea of utilization of the research nuclear reactor for medical purposes seems to be rather attractive, because in this case financial expenditures will be considerably lower. In view of NCT, the reactor horizontal channel GK-1, transporting the neutron beam from the reactor core via the biological shield to the reactor hall, was chosen. Evaluation of the neutronic parameters of the neutron beam along the channel length and the channel exit was performed with/without the lead neutron guide installed in the reactor core, in order to increase the neutron flux density at the channel exit and to reduce the photon component of the beam. Before in-reactor experiments, the appropriate calculations by means of special computer codes and the modeling experiments at the critical assembly were performed. As a whole, the following has been done in the frame of the studies in question: the distributions of the thermal/fast neutron flux density over the length of the horizontal channel have been measured with/without the neutron guide at various levels of the reactor power; the corresponding indices (cadmium ratios) have been found; it has been found that the lead-made neutron guide, installed in the reactor core, doubles the thermal neutron flux density at the beam exit from the channel; the gamma-radiation dose rates in the channel and in

  11. 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)

  12. Eddy current detection of spacers in the fuel channels of CANDU nuclear reactors

    International Nuclear Information System (INIS)

    Garter Spring (GS) spacers in the fuel channels of CANDU nuclear reactors maintain separation between the hot pressure tube and surrounding moderator cooled calandria tube. Eddy current detection of the four GSs provides assurance that spacers are at or close to design positions and are performing their intended function of maintaining a non-zero gap between pressure tube and calandria tube. Pressure tube constrictions, resulting from relatively less diametral creep at end-of-fuel bundle locations, also produce large eddy current signals. Large constrictions, present in higher service pressure tubes, can produce signals that are 10 times larger than GS signals, reducing GS detectability to 30% in standard GS-detect probes. The introduction of field-focussing elements into the design of the standard GS detection eddy current probe has been used to recover the detectability of GS spacers by increasing the signal amplitude obtained from GSs relative to that from constrictions by a factor of 10. The work presented here compares laboratory, modelling and in-reactor measurements of GS and constriction signals obtained from the standard probe with that obtained from field-focussed eddy current probe designs. (author)

  13. 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)

  14. Monitoring of coolant flow rate and velocity in the hot channel of the IPR-R1 TRIGA reactor core

    International Nuclear Information System (INIS)

    The 250 kW IPR-R1 TRIGA research reactor of the Centre of Nuclear Technology Development (CDTN) at Belo Horizonte is a pool type reactor cooled by light water under natural circulation. The fuel is an alloy of zirconium hydride and uranium enriched at 20% in 235U. The core channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. In the natural convection the driving force is supplied by the buoyancy of the heated water in the core channels. A forced heat removal system is provided for removing heat from the reactor pool water. The water is pumped through a heat exchanger, where heat is transferred from the primary to the secondary loop. The forced cooling system acts in opposition to the natural circulation, and its main purpose is to create a standing water volume at the pool top in order to improve the biological shield. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermohydraulic parameters like the coolant velocity, flow rate, mass flow rate and Reynolds's number for the hot channel, with the forced cooling system switched off and on. (authors)

  15. 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

  16. Development of multi-channel amplifier-discriminator based on measurement system of neutron fluence rate relative distribution in reactor

    International Nuclear Information System (INIS)

    For the measurement of neutron fluence rate in reactor and reliable assurance of γ count measurement of activated 55Mn-58Ni 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. (authors)

  17. Development of advanced techniques for life management and inspection of advanced heavy water reactor (AWHR) coolant channel components

    International Nuclear Information System (INIS)

    Operating life of pressure tubes of Pressurized Heavy Water Reactor (PHWR) is limited due to the presence of various issues associated with the material like hydrogen pick up, delayed hydride cracking, axial elongation and increase in diameter due to irradiation creep and growth. Periodic monitoring of the health of the pressure tube under in-situ conditions is essential to ensure the safe operation of the reactor. New designs of reactor call for innovative design philosophy, modification in fabrication route of pressure tube, development of reactor specific tools, both analytical and hardware for assessing the fitness for service of the pressure tube. Feedback from existing reactors has enhanced the understanding about life limiting parameters. This paper gives an insight into the life limiting issues associated with pressure tube and the efforts pursued for development of life management techniques for coolant channel of Advanced Heavy Water Reactor (AHWR) designed in India. The tools and techniques for in-situ property/hydrogen measurement, pulsed eddy current technique for zirconium alloy in-homogeneity characterization, horizontal shear wave EMAT system for dissimilar metal weld inspection, sliver sampling of vertical channel etc. are elaborated in the paper. (author)

  18. 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 ...

  19. A new DNB correlation proposed for vertical annuli and rectangular channels heated from one or both sides in water-cooled nuclear research reactors

    International Nuclear Information System (INIS)

    An empirical correlation has been developed for predicting departure from nucleate boiling (DNB) using data for an axially heat flux distribution. This correlation is recommended for use in predicting the DNB heat flux in a nuclear research reactor fuel coolant channel (annuli) or any rectangular channel. In this paper, an investigation was carried out on identify the important parameters affecting DNB heat flux in annuli and rectangular channels, focusing on the effects of channel operation conditions (inlet temperature, pressure, coolant volume flow rate). The predicted DNB results for nuclear reactor fuel coolant channel results have been compared and agree closely with the exiting international DNB heat flux data.(author)

  20. 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.

  1. Thermohydraulic calculations in rectangular channels for RA-6 type reactors with transition regime

    International Nuclear Information System (INIS)

    In August 2000 and within the framework of the RA-6 core conversion from high to low enrichment (20%), a preliminary analysis was performed to evaluate the maximum power that the reactor could operate with the new kernel without makeing substantial changes. This meant keeping intact, for example, the concrete shield of the pool and the nucleus inlet and outlet pipes embedded in the walls. Preliminary results indicated that for these boundary conditions a maximum power of about 3 MWt could be achieved. In August 2005 the project was resumed and new calculations performed taking as a starting point the ECBE plate fuel element(U3O8-Al). A core was developed with cooling channle widths of 2.6 mm for the control fuel elements and 2.7 mm for standard fuel elements. The thermo-hydraulic calculation puts in evidence that coolant flow into the core was in the transitional regime for the vast majority of configurations. While TERMIC code, used for thermo-hydraulic design, has been extensively tested and validated for use in research reactors under turbulent and laminar flows, this is not so for transition conditions. The transition regime is strongly dependent on conditions such as flow inlet characteristics, channel geometry, etc.. and therefore there are no reliable correlations for general use. For this reason we found it convenient to carry out experiments simulating the working conditions in order to adjust the code results with experimental data. In the present work we show the experimental results, the simulation of the experiences using the TERMIC code, and the adjustments made to the correlations used by the code so that it can be applied to the thermo-hydraulic design of the new core.

  2. Theoretical and experimental modeling of processes accompanying single fuel channel (pressure tube) rupture for RBMK Reactor. Part A

    International Nuclear Information System (INIS)

    The analysis of the MPTR (Multiple Pressure Tube Rupture) problem requires a series of theoretical and experimental studies of separate physical processes in the RBMK reactor, development of mathematical models and their physical equivalents. The experimental rigs concerned with MPTR problem were designed and constructed at Electrogorsk Research and Engineering Center, Russia. An investigation of the circumstances and mechanism of a rupture in a single channel at different conditions and scenarios is one of the main stages of the aforesaid analysis. Theoretical models of the single channel rupture under thermal and mechanical loading was developed including a channel constrained graphite block. Computer program based on this models enable to describe thermomechanical deformation process of the single channel and to predict rupture moment. Theoretical investigations supplements with experimental modelling single channel rupture by means of series experimental examinations at TKR-F (Model of an Accidental Channel) test rig. One represents a model of the single ruptured fuel channel in a surrounding graphite column. Experimental examinations enable to develop and verify theoretical models, conditions and mechanism of a rupture in a single channel. The flow process of steam or steam-water mixture through narrow graphite gaps is another important process should be modelled in frame of the MPTR problem analysis. The TKR-F (graphite) facility represents two test sections: for high speeds of medium flow near the ruptured channel and for low speeds far from the rupture. In the first case dynamical interaction escaping steam from the ruptured channel with (moving) graphite columns is modelled. The second case is attended to studying of hydraulical resistances and heat transfer of the steady-state steam-water flow within graphite gaps with different cross-section. Theoretical and experimental modelling consolidation sets out technique of the authentic analysis of the

  3. Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

    International Nuclear Information System (INIS)

    Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation

  4. Transfer coefficients in a four-cusp duct simulating a typical nuclear reactor channel degraded by accident

    International Nuclear Information System (INIS)

    An experimental study on forced convection in a four-cusp duct simulating a typical nuclear reactor channel degraded by accident is presented. Transfer coefficients were obtained by using the analogy between heat and mass tranfer, with the naphtalene sublimation technique. The experiment consisted in forcing air past a four-cusp naphthalene moulded duct. Mass transfer coefficients were determined in nondimensional form as Sherwood number. Experimental curves correlating the Sherwood number with a nondimensional length, x+, were obtained for Reynolds number varying from 891 to 30.374. This range covers typical flow rates that are expected to exist in a degraded nuclear reactor core. (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 the zirconium bare triple method. Results obtained for f and α in two irradiation channels show that f parameter determined in this way is different in the carousel facility (CR1) 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 neutron spectrum is relatively well thermalized. Parameter α in the PT has a negative value, which is very small and cannot significantly influence the final results 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. (author)

  6. 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.

  7. 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)

  8. Measurement of the power and of the period of a nuclear reactor using a simultaneously linear and logarithmic channel

    International Nuclear Information System (INIS)

    A simultaneously logarithmic and linear channel for measuring the power of a nuclear reactor using a single detector has been developed. It consists of a linear element and of a logarithmic element in the counter-reaction circuit of the measurement amplifier. The operation of this channel appears to be perfectly satisfactory. It can be connected to a period meter. The range of the logarithmic measurement extends from 10-10 to 1O-4 A and that of the linear measurement from 10-9 to 10-3 A. It will be possible to extend the linear measurement to 10-10 A. (author)

  9. 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)

  10. Remote repairs of the fixed channel gas outlet thermocouple pockets on the Dungeness 'A' and oldbury Magnox reactors

    International Nuclear Information System (INIS)

    This article describes a reactor remote repair project undertaken by NNC on behalf of Nuclear Electric plc. During routine biennial inspection of the Magnox reactors at Dungeness A and Oldbury in the United Kingdom, some of the fixed channel gas outlet thermocouple pocket and sleeve details were observed to be showing signs of oxidation damage. NNC were awarded the contract for design, development and supply of a scheme to secure the details against the effects of further oxidation damage. The article outlines the problem and the evolution of the project technical strategy from the many objectives and constraints, and covers the development of both the installed repair scheme and the remote installation techniques/equipment. A few of the more interesting or innovative items of equipment/engineering schemes are described, and the reactor 'pilot' repairs at Oldbury Power Station are briefly referenced. (author)

  11. 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)

  12. Pre-service ultrasonic inspection of coolant channel rolled joints and surrounding areas in 540 MWe pressurised heavy water reactors

    International Nuclear Information System (INIS)

    To ensure safety of coolant channel rolled joints, in-situ, immersion ultrasonic testing was carried out to check presence of any unacceptable flaw. Twenty numbers of channels (≅ Square root of total channels in a reactor) covering central to peripheral region were selected for pre-service inspection. The calibration notches were made by EDM, both in axial and circumferential directions on OD and ID in adjacent and upset region. Sample rolled joint was cut open and freed from end fitting for making 2 % (0.086mm) depth artificial notches. Two numbers of inspection heads, housing two axial and two circumferential line focused ultrasonic probes were fabricated. In an annular Perspex cylindrical rod of 4 inch diameter, holes inclined at 27 deg were machined in axial and circumferential orientations. Immersion line focused probes of 10 MHz frequency were fitted in the drilled holes to generate 45 deg mode converted shear waves in the wall of pressure tube. No reportable flaw in the region adjacent to the rolled joint was observed in any channel of any of the two reactors. The coolant tube 11 to 1 'O' clock and between the two-rolled joint was also free from any discernible defect. (author)

  13. Transient thermal–hydraulic analysis of complete single channel blockage accident of generic 10 MW research reactor

    International Nuclear Information System (INIS)

    Highlights: • Single channel blockage simulation of research reactor fuel is performed. • Effect of different axial power profiles on fuel integrity is tested. • Effect of vapor generation and presence of oxide layer on fuel temperature is seen. • Point kinetics model is used to simulate neutronic feedback during transient. • Varying adaptation scale of feedback yields deviating fuel temperature results. - Abstract: Thermal–hydraulic behavior for a complete blockage of a single fuel channel in a generic 10 MW research reactor is studied by using the system analysis code RELAP5/MOD3.3 which is widely used in the nuclear industry. Fuel assembly geometry is lumped into a 4 channel model to model high and average power cases which are spatially discretized. Various axial power shapes coming from different control rods positions are considered in the analysis, where the minimum wall subcooled margin is found to exist for case with highest peaking for an average powered channel blockage transient. Vapor generation is observed from first and second highest peaking cases where cyclic variation of vapor inventory inside a blocked channel resulted in oscillatory behavior of the fuel temperature. Effect of a presence of an oxide layer is also tested which showed a slight increase in structure temperatures and vapor generation. Point kinetics model is utilized in the analysis code to observe the effect of reactivity feedback and consequences from different application ranges are compared. Analysis shows a consideration of assembly wise feedback results in increased feedback effect and decreased boiling which deviate from single channel wise feedback case. This calls for a detailed multi-dimensional simulation with neutronics and thermal–hydraulics simultaneously considered. Analyses results show that the consideration of feedback improves the outcome in terms of fuel temperature, and its integrity is conserved for all test cases

  14. Experimental studies on coolant technology and fuel elements of high temperature reactors in the helium loop PG100 and in irradiation channels

    International Nuclear Information System (INIS)

    An important part of experimental researches in SSSR on fuel element fabrication and equipment of the primary coolant circuit of high temperature reactors are made on the following test benches: an installation for the study of leak-tight fuel elements (OSA installation), an helium loop (PG100) in the MR reactor, tube channels for testing ''Kashtan'' pebble fuels, tube channels for testing ''Karat'' fuel microelement. After testing fuel elements and equipment are examined in a hot laboratory

  15. Influence of the control bars pattern on the response of the operation channels of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    The local flow perturbations not generated by movements of bars not planned adequately to operate the reactor to 1 MW of thermal power, are reflected in the independent responses of the operation channels of the same one, find variations average from 17% to 30% for the channel of the power percent and of until 10% for the logarithmic channel. For the case of the lineal and percent power channels, these are between 14% and 46% as maximum when moving some of the bars. These variations can diminish until 5% in the channel of the power percent and until 3% on the average for the logarithmic one, all times when the calculated bars pattern for that irradiation considers that all the bars operate inside the lineal region of its calibration curve with approximately the same reactivity value each one and that during the operation the required reactivity compensations are carried out with the diametrically opposed bar to the irradiation installation used in that experiment. (Author)

  16. Parallel channel instabilities in boiling water reactor systems: boundary conditions for out of phase oscillations

    International Nuclear Information System (INIS)

    In this paper we study the boundary conditions during out of phase oscillations, in a system formed by two parallel channels coupled to multimodal neutron kinetics. The fact that the pressure drop can change with time, but remains the same in all the parallel channels, leads us to analytical integration of the time derivative term of the channel momentum equation, in order to get a dynamic equation for the inlet mass flux to each channel. From these inlet mass flux dynamic equations plus the appropriate boundary conditions, we have obtained the expression of the time dependent common pressure drop to all the channels. Finally the oscillations in the pressure drop and in the inlet mass flux to the channels have been investigated when out of phase oscillations take place

  17. 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

  18. 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.

  19. Analysis of some mass and momentum transfer correlations applicability in an adiabatic model of flooding of a reactor channel

    International Nuclear Information System (INIS)

    In this paper an analysis of the applicability of entrainment inception, entrainment rate and interfacial friction correlation in case of counter current annular flow and flooding in a reactor channel has been performed. The thermo-hydraulic model formulation consists of separate treatment of the liquid film and gas droplet mixture in the channel core, while droplet phase of the gas-droplet mixture has been treated with drift-flux model. The results of the analysis have been compared and verified with the correspondingly selected experimental data. In this paper some modifications of the before mentioned correlations in the model have been suggested to achieve better agreement of the results with the experimental data. (author)

  20. New selection criteria for channel refueling of a Candu-6 reactor: introduction to floppy rules

    International Nuclear Information System (INIS)

    A revised set of rules is in use at Gentilly-2 NGS for the selection of channels for refuelling. Traditional hard channel rejection rules (of go/no-go type) have been replaced by a more efficient set of soft evaluation rules based on concepts borrowed to the Fuzzy Logic. New evaluation rules, labelled as 'Floppy Rules', enable to assess and rate the channel suitability for refuelling by using a smooth and natural continuum of values qualifying excellent, good, fair and poor choices. Global channel suitability for refuelling is measured by combining separate ratings obtained from individual evaluation rules. Each evaluation rule is based on a specific control parameter related to local or lumped core properties. Two new software codes (NEWRULES and REFUEL) designed around the concept of Floppy Rules enable to perform a very efficient selection of optimized channel refuelling sequences either in manual and automatic mode. (author)

  1. The BOR-60 loop-channel design for testing the BREST reactor fuel

    International Nuclear Information System (INIS)

    The paper is devoted to development and calculated substantiation of the design of the autonomous lead-cooled loop (AILCL) intended for testing the fuel pins prototypes in the BOR-60 reactor for the BREST-OD-300 reactor. The design features of the loop, its characteristics, instrumentation are considered. The auxiliary systems required to provide the loop operation are described. The main neutron-physical and thermohydraulic characteristics of the loop are presented. The basic operating conditions of the loop are shown. Great consideration is given to analysis of faults and failures affecting the loop serviceability and reactor safety. It is shown that the BOR-60 reactor safety is provided under all normal modes and postulated failures. (author)

  2. Measurements at the RA Reactor related to the VISA-2 project - Part 2, Measurement of neutron flux in VISA-2 channels

    International Nuclear Information System (INIS)

    This report describes the task concerned with measurements of neutron flux in four experimental channels, called VISA-2 channels. All the channels are made of aluminium tubes, one is sealed to prevent contact of foils with heavy water, and are placed in the regular reactor lattice next to the central experimental channel VK-5. Measuring results, i.e. absolute values of neutron flux and flux distribution are needed for realisation of the VISA-2 project. Measurements of neutron flux are done by activation method. Activation foils are placed in cylindrical aluminium tubes specially prepared for this purpose and placed in VISA-2 channels. Foils are irradiated simultaneously for 5 hours at reactor power of 150 kW. Neutron flux distribution is determined by measuring the relative activity of cobalt foils

  3. 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

  4. Evaluation of neutron flux in the WWR-SM reactor channel and in the irradiating zone of U-150 cyclotron

    International Nuclear Information System (INIS)

    Full text: For effective work of a reactor, and correct planning of experiments related to the reactor irradiation of various materials it is required to control a neutron flux in the given irradiation point for a long irradiation period. For realization of research works on topazes ennobling under irradiation by reactor neutrons as well as by secondary neutrons produced in a cyclotron it is necessary to know the total neutron flux and spectra. To resolve the problem a technique for registration of neutrons with different energy and calculation of a neutrons spectrum in the given irradiation points in reactor channels and in cyclotron behind the nickel target has been developed. Neutron flux density and energy spectra were monitored by use of the following nuclear reactions: 59Co(n,γ)60Co, 197Au(n,γ)198Au, 58Ni(n,p)58Co, 24Mg(n,p)24Na, 48Ti(n,p)48Sc, 46Ti(n,p)46Sc, 54Fe(n,p)54Mn, 89Y(n,2n)88Y, 60Ni(np)60Co. Gamma spectrometer composed of HPGe detector (Rel. Eff. - 15%) and Digital Spectra Analyzer DSA-1000 (Canberra Ind., USA) was used to measure gamma activity of irradiated samples. Acquired gamma spectra were processed by means of Genie 2000 standard software package. The σ(E) functions and neutron spectra were calculated by using the least squares method and approximating the tabular and experimental data with power polynomials. The developed technique was applied for the adjustment of the topazes irradiation regimes in the reactor core and under secondary neutrons flux from a nickel target in the cyclotron. The given technique allows to calculate a logarithmic spectrum of neutrons in a energy range from 0,025 eV up to 12 MeV with the uncertainty of about 10 %. (author)

  5. The final calibration of the nuclear power channels of the IPEN/MB-01 reactor by the use of the activation foils and the Monte Carlo method

    International Nuclear Information System (INIS)

    This work aims the final calibration of the nuclear power channels of the IPEN/MB-01 reactor using infinitely dilute gold foils (1% Au - 99% Al), this is, metallic alloy in the concentration levels such that the phenomena of flux disturbance, as the self-shielding factors become worthless. During the irradiations were monitored the nuclear power channels of the reactor used to obtain the neutron flux and consequently the power operation of the reactor. The current values were digitally acquired during each second of operation. Once the foils were irradiated for the analysis of its induced activity it was used a detection system of hyper-pure germanium. Ally to this experimental procedure it was used the computational code MCNP-4C as a tool for theoretical modeling of the core of the IPEN/MB-01 reactor. Thus it was possible to determine the parameters necessary to obtain the power operation of the reactor, such as the inverse of the thermal disadvantage factor and fast fission factor. Thus, using the correlation between average thermal neutron flux, proportional to a power operation and the average of the digital values of current of the nuclear channels, during the irradiations of the foils, it was obtained the calibration of the nuclear power channels, the ionization chambers number 5 and 6 of the IPEN/MB-01 reactor. (author)

  6. Numerical modeling of velocity distribution of coolant flow in fuel channel of a miniature neutron source reactor using navier-stokes equations

    International Nuclear Information System (INIS)

    One major requirement for safe operation of a nuclear reactor is adequate cooling system during normal and emergency conditions. Analysis of the heat and mass transfer, and coolant flow rate variables during operations of nuclear reactors are required for performing risk and hazard management to plan optimum reactor recovery strategies. The velocity distribution of coolant in the channel of the Ghana Research Reactor -1 (GHARR-1) is of major concern because when the velocity of the coolant is too fast it results in pool cooling. On the contrary when the flow rate is also slow, the possibility of boiling occurs and hence poor cooling. This is because the virtually stagnant coolant will take much of the heat generated from the core and would begin to boil which can also lead to core meltdown. These parameters can be determined through experimental setup and computer simulations using models based on the laws of fluid dynamics and thermodynamics. The research took into consideration a computer based model which used Navier Stokes equations of continuity, momentum and energy conservation to simulate flow patterns in the channels. The Navier Stokes equation was then expressed in algorithm using the Marker and Cell (MAC) finite different technique. The algorithm equations were then developed into matrix form (algebraic equations) by discretization. MATLAB has been used to code and solve the resulting finite different equations numerically to determine the velocity fields of coolant in the GHARR-1 reactor core channel. The velocity distribution in the Ghana research reactor (GHARR-1) was determined to be in the range of 0.9 m/s to 1.9 m/s. It is observed that the flow in the hottest channel was faster than the channel which were far from the center of the core and hence helped in removing much heat from the core of the reactor and ensuring reactor safety. (au)

  7. 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)

  8. Safety analysis of neutron flux optimization in irradiation channels at the NUR research reactor

    Directory of Open Access Journals (Sweden)

    Zergoug Tewfik

    2006-01-01

    Full Text Available Prior to core reloading, planned power upgrading, or as a part of required analyses of past events, accurate safety evaluations should be carried out. Generally speaking, the content of a safety report has to be modified whenever a new type or design of fuel is to be used in a reactor core. As the existing plants have well established licensing procedures, including well founded analysis methods, the application of new analysis methods has to be thoroughly evaluated, with specific emphasis on their capability of producing results beneficial to reactor operation. The detailed study presented here was carried out so as to insure that the allowed operational safety limits of the NUR research reactor are not exceeded under any circumstances.

  9. The research of the ceramic fuel pins, irradiated in the technological channels of the ATS of IVG.1 reactor

    International Nuclear Information System (INIS)

    The basic objective of the work is to install the qualitative and quantitative indexes of the serviceability of the rod-type carbide fuel pins as applied to the exploitation conditions in the high temperature gas cooled reactor, where the nitrogen will be used as the coolant. For that purpose the state of the fuel pins, tested in the nitrogen technological channels of the ATS of IVG.1 (EWG-1) reactor in the series of four research nitrogen start-ups , including the energetic start-up with the duration of 160 s. and three standard start-ups with the duration of 510 s., was researched. On the base of the results of the post-reactor research of the fuel pins of ten channels of ATS it is determined that the ceramic fuel pins of (U, Zr)C+C, (U, Nb)C+C and (U, Zr,Nb)C are enough serviceable in the severe conditions of the high temperature tests in the flow path of the chemically aggressive coolant. The lack of the surface cracks in the fuel pins, lack of the fuel pins failures and lack of overweight and thickening of the fuel pins are revealed. It is observed the oxy-nitration of the fuel pins surfaces (at appearance of the characteristic color tones and presence of the slight burning of the fuel rods to each other), however, the depth of the oxy-nitration , even of the fuel pins of the output heating sections, tested at 2800 K, did not exceed 10 μm. It is found out that the levels of the radioactive change of the fuel pins parameters are the same as of the fuel pins of the hydrogen technological channels, tested at the same temperatures and up to the same neutrons fluences. The low change of the fuel pins strength is observed; the strengthening of the fuel pins on the output heating sections for ∼ 20 % (due to the appearance of the residual radiation macro stresses) and weakening of the fuel pins in the output sections for ∼30 % (due to oxy-nitration and erosion of the surface, and also non-congruent evaporating of the surface material). The prognostic analysis of

  10. Flow with boiling in four-cusp channels simulating damaged core in PWR type reactors

    International Nuclear Information System (INIS)

    The study of subcooled nucleate flow boiling in non-circular channels is of great importance to engineering applications in particular to Nuclear Engineering. In the present work, an experimental apparatus, consisting basically of a refrigeration system, running on refrigerant-12, has been developed. Preliminary tests were made with a circular tube. The main objective has been to analyse subcooled flow boiling in four-cusp channels simulating the flow conditions in a PWR core degraded by accident. Correlations were developed for the forced convection film coefficient for both single-phase and subcooled flow boiling. The incipience of boiling in such geometry has also been studied. (author)

  11. Transversal heat exchange in a rectangular channel heated non-uniformly. Application to the calculation of temperatures in a cell of the reactor Siloe

    International Nuclear Information System (INIS)

    In the techniques used for cooling nuclear reactors, the existence of flux peaks is a great nuisance. The case is considered here of a fuel lattice made up of plates and thus giving rectangular cooling channels. A theoretical and practical study has been made of transversal heat exchange in this structure. An example is given of an application which is in the form of a calculation of the temperature distributions at the exit of a cooling channel of the swimming-pool type reactor Siloe operating at 10 MW. (author)

  12. Large scale replacement of fuel channels in the Pickering CANDU reactor using a man-in-the-loop remote control system

    International Nuclear Information System (INIS)

    Spar Aerospace Limited of Toronto is presently under contract to Ontario Hydro to design a Remote Manipulation and Control System (RMCS) to be used during the large scale replacement of the fuel channels in the Pickering A Nuclear Generating Station. The system is designed to support the replacement of all 390 fuel channels in each of the four reactors at the Pickering A station in a safe manner that minimizes worker radiation exposure and unit outage time

  13. Inactive commissioning of a micro channel catalytic reactor for highly tritiated water production in the CAPER facility of TLK

    International Nuclear Information System (INIS)

    Highlights: ► In a DT fusion machine several events will generate highly tritiated water (HTW). ► PERMCAT appears a promising process to recover tritium from HTW. ► In order to perform R and D activity on HTW processing with PERMCAT, such water has to be produced on purpose. ► A tritium compatible micro-channel catalytic reactor (μCCR) has been designed and manufactured to produce up to 10 mL min−1 of HTW with very high specific tritium activity. ► The paper presents the inactive commissioning of the μCCR required before the integration in CAPER facility. ► The combination of the μCCR with the O2 sensor represents a reliable system able to produce HTW in a safe way and without radioactive waste. - Abstract: In future DT fusion machines, several events will generate highly tritiated water (HTW). Among potential techniques for HTW processing, isotopic swamping in a catalytic membrane reactor (PERMCAT) appears promising. The experimental demonstration of PERMCAT for HTW processing has started in the CAPER facility at the Tritium Laboratory of Karlsruhe (TLK). Without any HTW source, such water has to be produced on purpose. Catalytic HT oxidation would ensure clean operation but could be critical for operation due to possible occurrence of explosive mixture. A tritium compatible micro-channel catalytic reactor (μCCR) has been designed and manufactured to produce up to 10 mL min−1 of HTW with very high specific tritium activity (stoichiometric DTO: 5.2 × 1016 Bq kg−1). Prior to its integration in CAPER for tritium operation, this reactor has been commissioned at different feed flow rates, gas composition (air or Helium), and temperature. The results demonstrate the good performances of the μCCR in producing water. The combination of the μCCR with the O2 sensor represents a reliable system able to produce HTW in a safe way and without radioactive waste. Accordingly, the CAPER facility can be upgrade in order to continue the R and D activity on

  14. Experience of detecting blisters in irradiated coolant channels of Indian Pressurised Heavy Water Reactors

    International Nuclear Information System (INIS)

    A number of irradiated pressure tubes which were in contact with calandria tube during reactor operation have been subjected to detailed examination. In case of contact, calandria tube/ pressure tube (CT/PT) contact hydrogen absorbed in the pressure tube migrates and keeps accumulating in the contact region cold spot under thermal gradient. Over a length of time, accumulated hydrogen at the contact zone forms localized massive concentration of δ-phase zirconium hydride, which is termed as Blister. Blister grows in size with time in the reactor and reaches a critical size when it can crack. Presence of a cracked blister is a matter of concern for the safety of pressure tubes. Ultrasonic velocity ratio measurement technique has been developed and applied to evaluate formation of hydride blisters in irradiated pressure tube during the course of post irradiation examination. (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. Verification of Monte Carlo calculations of the neutron flux in the carousel channels of the TRIGA Mark II reactor, Ljubljana

    International Nuclear Information System (INIS)

    In this work experimental verification of Monte Carlo neutron flux calculations in the carousel facility (CF) of the 250 kW TRIGA Mark II reactor at the Jozef Stefan Institute is presented. Simulations were carried out using the Monte Carlo radiation-transport code, MCNP4B. The objective of the work was to model and verify experimentally the azimuthal variation of neutron flux in the CF for core No. 176, set up in April 2002. '1'9'8Au activities of Al-Au(0.1%) disks irradiated in 11 channels of the CF covering 180'0 around the perimeter of the core were measured. The comparison between MCNP calculation and measurement shows relatively good agreement and demonstrates the overall accuracy with which the detailed spectral characteristics can be predicted by calculations.(author)

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

    International Nuclear Information System (INIS)

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

  18. 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)

  19. 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.

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

    Czech Academy of Sciences Publication Activity Database

    Štefánik, Milan; Katovsky, K.; Vinš, M.; Šoltéš, J.; Závorka, L.

    2014-01-01

    Roč. 104, NOV (2014), s. 302-305. ISSN 0969-806X. [1st International Conference on Dosimetry and its Applications (ICDA). Prague, 23.6.2013-28.6.2013] R&D Projects: GA MŠk LG14004 Institutional support: RVO:61389005 Keywords : spectral index * neutron spectrometry * dosimetry-foils activation technique * irradiation channel * reaction rate * Gamma-spectroscopy Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.380, year: 2014

  1. 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

  2. Influence of single-phase heat transfer correlations on safety analysis of research reactors with narrow rectangular fuel channels

    International Nuclear Information System (INIS)

    The influence of different single-phase heat transfer correlations on the fuel temperature and minimum critical heat flux ratio (MCHFR) during a typical accident of a 5 MW research reactor is investigated. A reactor uses plate type fuel, of which the cooling channels have a narrow rectangular shape. RELAP5/MOD3.3 tends to over-predict the Nusselt number (Nu) at a low Reynolds number (Re) region, and therefore the correlation set is modified to properly describe the thermal behavior at that region. To demonstrate the effect of Nu at a low-Re region on an accident analysis, a two-pump failure accident was chosen as a sample problem. In the accident, the downward core flow decreases by a pump coast-down, and then reverses upward by natural convection. During the pump coast-down and flow reversal, the flow undergoes a laminar flow regime which has a different Nu with respect to the correlation sets. Compared to the results by the original RELAP5/MOD3.3, the modified correlation set predicts the fuel temperature to be a little higher than the original value, and the MCHFR to be a little lower than the original value. Although the modified correlation set predicts the fuel temperature and the MCHFR to be less conservative than those calculated from the original correlation of RELAP5/MOD3.3, the maximum fuel temperature and the MCHFR still satisfy the safety acceptance criteria

  3. 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)

  4. 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)

  5. 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)

  6. Sub-channel analysis of Pb-Bi cooled fast breeder reactor PEACER fuel assembly using MATRA

    International Nuclear Information System (INIS)

    Full text of publication follows: The nuclear power is the one of the realistic means that can solve the shortage of usable energies due to depletion of fossile fuel and due to the environmental contamination. However, since the nuclear fission yields a kind of fission fragment as a by-product, a radiological hazard of spent fuel is now a major problem. To overcome this difficulty, a number of studies are being performed and planned. One of the key solutions to this problem is to eliminate spent fuel by nuclear transmutation. According to this research, the most significant long-lived isotopes in spent fuels of the current power reactors can be transmuted into short-lived ones by using fast neutron spectrum with localized thermal traps. For this reason, liquid metal cooled fast breeder reactor is widely chosen as the answer to solve the problem. In Korea, PEACER(Proliferation-resistant, Environment-friendly, Accident-tolerant, Continual and Economical Reactor) is under study to work out this issue. PEACER core is designed to produce about 1560 MW of the thermal output with electric output up to 550 MW which efficiency is about 0.35. PEACER uses control rode made of B4C to perform reactivity control and Pb-Bi liquid metal is adopted as a coolant for primary system. In nuclear power plants it is important to keep the temperature of the reactor core structures under certain criteria in order to prevent damage of fuel materials which can advance to severe situations such as radiation leakage, and even meltdown of the fuel. This study was intended to see the liquid metal coolant behavior along the PEACER fuel channels and to find out whether the given heat flux profiles and geometrical arrangement of the fuel rods yields reasonable fluid dynamic distribution under nominal operation by using subchannel approach. The subchannel analysis of the fuel assembly under nominal operation condition was performed using MATRA (Multi-channel Analyzer for Transient and steady

  7. Multi-channel heat exchanger-reactor using arborescent distributors: A characterization study of fluid distribution, heat exchange performance and exothermic reaction

    International Nuclear Information System (INIS)

    A multi-functional heat exchanger-reactor comprising arborescent (tree-like) distributors and collector, 16 mini-channels in parallel and T-mixers is introduced in this paper. Flow distribution property, pressure drop and heat exchange performance of proposed heat exchanger-reactor are tested and discussed. Firstly, flow distribution uniformity is characterized by CFD simulation and then qualitatively confirmed by visualization experiment. Results show that for total flowrates ranging from 5 mL s−1 to 20 mL s−1, good distribution uniformity is obtained, with maximum flowrate deviation less than 10%. Then, experiments of heat exchange between hot and cold water are carried out. High values of overall heat transfer coefficient ranging from 2000 to 5000 W m−2 °C−1 are obtained under our working conditions. The volumetric heat exchange capability (UA/V) is found to be around 200 kW m−3 °C−1, showing a high heat exchange capability with compact design. The roles of end-effect and non-established flow are discussed and are supposed to be responsible for efficient heat transfer. Finally a typical fast exothermic reaction, neutralization between acid and basic solutions, is carried out to test the thermal control capability of the studied heat exchanger-reactor. Results indicate that isothermal condition could be realized by circulating appropriate flowrate of coolant through the heat exchanger. The design of heat exchanger-reactor with arborescent distributor and collector makes possible the application of multi-channel systems. This paper introduces systematically the successful integration of heat exchanger-reactor and its performance evaluation. - Highlights: • A design of mini scale, multichannel heat exchanger-reactor is proposed. • Uniform distribution for parallel channels is obtained with arborescent structure. • High global heat exchange coefficient is found experimentally. • Thermal control capability is verified with an

  8. 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)

  9. 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

  10. Effect of coolant channel width on group constants and multiplication factor of research reactors using MTR type low enriched uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, A.N. E-mail: nasir@pieas.edu.pk

    2002-03-01

    One dimensional transport theory lattice code WIMS-D/4 and three dimensional diffusion theory code CITATION have been used to study the effect of fuel loading on critical cores of low enriched uranium (LEU) fuelled material testing reactors (MTRs). The fuel loading in a fuel element was varied by changing the fuel density in the fuel meat. In order to keep the reactor critically moderated, the optimal coolant channel width for a given fuel loading was calculated. For the purpose of optimization, the group constants D, {sigma}{sub a} and {nu}{sigma}{sub f}, and infinite multiplication factor (k{sub {infinity}}) were calculated as a function of coolant channel width using WIMS-D/4. An increase in {sup 235}U loading per fuel plate results in an increase in the optimal coolant channel width and k{sub {infinity}}. The calculated values were found to be in good agreement with the typical design of MTR. CITATION was then used to determine the critical cores for different fuel loading with optimized fuel dimensions. Both critical mass and volume were found to decrease with an increase in the fuel loading. The criticality studies of Pakistan research reactor-1 (PARR-1) are in good agreement with the predictions.

  11. Effect of coolant channel width on group constants and multiplication factor of research reactors using MTR type low enriched uranium fuel

    International Nuclear Information System (INIS)

    One dimensional transport theory lattice code WIMS-D/4 and three dimensional diffusion theory code CITATION have been used to study the effect of fuel loading on critical cores of low enriched uranium (LEU) fuelled material testing reactors (MTRs). The fuel loading in a fuel element was varied by changing the fuel density in the fuel meat. In order to keep the reactor critically moderated, the optimal coolant channel width for a given fuel loading was calculated. For the purpose of optimization, the group constants D, Σa and νΣf, and infinite multiplication factor (k∞) were calculated as a function of coolant channel width using WIMS-D/4. An increase in 235U loading per fuel plate results in an increase in the optimal coolant channel width and k∞. The calculated values were found to be in good agreement with the typical design of MTR. CITATION was then used to determine the critical cores for different fuel loading with optimized fuel dimensions. Both critical mass and volume were found to decrease with an increase in the fuel loading. The criticality studies of Pakistan research reactor-1 (PARR-1) are in good agreement with the predictions

  12. 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.

  13. 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

  14. Reactors

    International Nuclear Information System (INIS)

    Purpose: To provide a spray cooling structure wherein the steam phase in a bwr reactor vessel can sufficiently be cooled and the upper cap and flanges in the vessel can be cooled rapidly which kept from direct contaction with cold water. Constitution: An apertured shielding is provided in parallel spaced apart from the inner wall surface at the upper portion of a reactor vessel equipped with a spray nozzle, and the lower end of the shielding and the inner wall of the vessel are closed to each other so as to store the cooling water. Upon spray cooling, cooling water jetting out from the nozzle cools the vapor phase in the vessel and then hits against the shielding. Then the cooling water mostly falls as it is, while partially enters through the apertures to the back of the shielding plate, abuts against stoppers and falls down. The stoppers are formed in an inverted L shape so that the spray water may not in direct contaction with the inner wall of the vessel. (Horiuchi, T.)

  15. COBRA4i-MIT: an updated sub-channel analysis code for sodium fast reactor design

    International Nuclear Information System (INIS)

    Proper modeling of the coolant behavior in Sodium Fast Reactors (SFR) is necessary for design and safety reasons. Fuel performance, for example, can only be accurately understood by knowing the full history of local coolant conditions. Computational Fluid Dynamics (CFD) is a powerful tool for fluid modeling; however it is still too computationally expensive for parametric studies and/or transient analysis of whole assemblies. The sub-channel analysis approach is better suited for the task, trading in a luxurious level of detail for a necessary boost in speed. Most existing sub-channel analysis codes for sodium, including SABRE, SLTHEN, COBRA, and MATRA, are capable of producing reasonably accurate results, however are limited in availability or lack the most current empirical correlations. COBRA4i, which provides robust implicit and explicit solutions schemes, suffers only from the latter malady. COBRA4i produced good results when previously tested with experimental data and with its multiple solution schemes is viable for a large spectrum of operating conditions and transients. The main shortcoming of the code is its archaic nature, in its programming language (FORTRAN66) and its correlations, both of which can be remedied as described. COBRA4i was brought up to date so it could be interpreted by modern compilers. A through literature search determined the most accurate and up to date correlations for pressure drop, mixing, and heat transfer. These correlations were added to the code, which was then run parametrically to determine how different combinations of old and new correlations affected code performance. All flow (laminar, transition and turbulent) and convection (natural, mixed and forced) regimes were included in the update. A recommended set of correlations was determined. Experimental benchmarks were preformed on data from the ORNL 19-rod test assembly, Toshiba 37-rod bundle and WARD 61-rod bundle, along with a code-to-code benchmark on results from the

  16. Multiple pressure tube rupture in channel type reactors. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    Accident scenarious which could potentially lead to multiple pressure tube ruptures have been recognized as a major safety issue for the RBMK. Therefore, a topical meeting on multiple pressure tube rupture analysis in channel type reactors was convened by the IAEA at the RDIPE in Moscow from 31 January to 4 February 1994 in the framework of the IAEA Extrabudgetary Programme on the Safety of RBMK Nuclear Power Plants. The objective of the meeting was to exchange experience on approaches adopted in Member States operating channel type reactors and review analysis methodology, criteria and results obtained related to multiple pressure tube rupture in RBMK type reactors including related regulatory requirements. The review was carried out in two broader technical areas: Pressure tube integrity and potential for failure propagation; Multiple pressure tube rupture scenarios. The following conclusions have been derived: Propagation of a single tube rupture is unlikely; The analysis of accident scenarios which could lead to multiple pressure tube ruptures in RBMKs is among the highest priority safety issues. International co-operation including independent analyses by experts from OECD countries, is required to resolve this issue; Further experimental information is needed both to validate the computer codes and to obtain a better understanding of some of the physical phenomena involved; No specific scenario for multiple failures following a localized power excursion was identified. 2 refs, 12 figs, 4 tabs

  17. Neutron-photon energy deposition in CANDU reactor fuel channels: a comparison of modelling techniques using ANISN and MCNP computer codes

    International Nuclear Information System (INIS)

    In order to assess irradiation-induced corrosion effects, coolant radiolysis and the degradation of the physical properties of reactor materials and components, it is necessary to determine the neutron, photon, and electron energy deposition profiles in the fuel channels of the reactor core. At present, several different computer codes must be used to do this. The most recent, advanced and versatile of these is the latest version of MCNP, which may be capable of replacing all the others. Different codes have different assumptions and different restrictions on the way they can model the core physics and geometry. This report presents the results of ANISN and MCNP models of neutron and photon energy deposition. The results validate the use of MCNP for simplified geometrical modelling of energy deposition by neutrons and photons in the complex geometry of the CANDU reactor fuel channel. Discrete ordinates codes such as ANISN were the benchmark codes used in previous work. The results of calculations using various models are presented, and they show very good agreement for fast-neutron energy deposition. In the case of photon energy deposition, however, some modifications to the modelling procedures had to be incorporated. Problems with the use of reflective boundaries were solved by either including the eight surrounding fuel channels in the model, or using a boundary source at the bounding surface of the problem. Once these modifications were incorporated, consistent results between the computer codes were achieved. Historically, simple annular representations of the core were used, because of the difficulty of doing detailed modelling with older codes. It is demonstrated that modelling by MCNP, using more accurate and more detailed geometry, gives significantly different and improved results. (author). 9 refs., 12 tabs., 20 figs

  18. The complex nonlinear dynamics in the multiple boiling channels coupling with multi-point reactors with constant total flow rate

    International Nuclear Information System (INIS)

    Highlights: • The nuclear-coupled effect has a distinct influence on the system dynamic behaviors and oscillation modes. • The effect of channel-to-channel interaction makes the 5-channel system more unstable than a 3-channel one. • Complex nonlinear bifurcation phenomena may appear in the present system. • The unimaginable types of complex periodic oscillations may exist in such a system. - Abstract: The present study explores the effect of nuclear-coupled feedback on the oscillation modes and nonlinear phenomena of a five nuclear-coupled boiling channel system by a nonlinear dynamic model previously developed by the authors. The results show that the combined effects of stable neutron interaction and unstable void-reactivity feedback generate distinct influence on the system stability, particularly a significant unstable effect as in the 4Cα cases. The effect of channel-to-channel interaction will drive the 5-channel system more unstable than a 3-channel one. Such a nuclear-coupled effect may affect the oscillation modes and nonlinear phenomena among the channels substantially. For the present system with a constant total flow rate, the superimposition of the dominant single-phase frictional pressure drop and strengthening void-reactivity feedback may result in the departure from the out-of-phase mode oscillations at some system states. The results demonstrate the appearance of different bifurcation phenomena in the unstable region and complex nonlinear phenomena, i.e. various periodic oscillations and complex Rossler type of chaotic oscillations, in such a system subject to certain nuclear-coupled feedbacks. A special type of complex P-3 oscillations is identified in this system. It suggests that there may be immeasurable types of the periodic nonlinear oscillations in the limited unstable space of this five nuclear-coupled boiling channel system

  19. 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)

  20. Calibration of the nuclear power channels for the cylindrical configuration of the IPEN/MB-01 reactor obtained from the measurements of the spatial neutron flux distribution in the reactor core through the irradiation of gold foils

    International Nuclear Information System (INIS)

    The activation foil is one of the most used techniques to obtain and compare nuclear parameters from the nuclear data libraries, given by a gamma spectrometry system. Through the measurements of activity induced in the foils, it is possible to determine the neutron flux profile exactly where it has been irradiated. The power level operation of the reactor is a parameter directly proportional to the average neutron flux in the core. The objective of this work is to obtain, for a cylindrical configuration, the power generation through a spatial thermal neutron flux distribution in the core of IPEN/MB-01 Reactor, by irradiating gold foils positioned symmetrically into the core. They are put in a Lucite plate which will not interfere in the analysis of the neutron flux, because of its low microscopic absorption cross section for the analyzed neutrons. The foils are irradiated with and without cadmium covered small plates, to obtain the thermal and epithermal neutron flux, through specific equations. The correlation between the average power neutron flux, as a result of the foil's irradiation, and the average power digital neutron flux of the nuclear power channels, allows the calibration of the nuclear channels of the reactor. This same correlation was done in 2008 with the reactor in a rectangular configuration, which resulted in a specific calibration of the power level operation. This calibration cannot be used in the cylindrical configuration, because the nuclear parameters could change, which may lead to a different neutron profile. Furthermore, the precise knowledge of the power neutron flux in the core also validates the mathematics used to calculate the power neutron flux. (author)

  1. Experimental operation of the RA reactor with 4 fuel channels containing 80% enriched dispersion fuel - Operational Report

    International Nuclear Information System (INIS)

    Start of utilization of the new 80% enriched dispersion nuclear fuel is underway in the RA reactor core. Both economic and technical analyses were in favor of introducing the new fuel elements gradually into the RA reactor core. Thus overall theoretical and experimental analyses as well as other preparations are directed to transition regime based on gradual introducing of new fuel into the core, i.e. reactor core with two types of fuel. The objective of these analyses and preparation is establishment of conditions for safe reactor operation during transition period. The analyses and preparations are almost completed. The experimental data about fuel burnup during a time period of operation at nominal power i.e. daily decrease of excess reactivity is missing. This data is needed for planning the refueling (quantity of fresh fuel and frequency of refueling) during the transient period. This data can be obtained only by normal operation of the reactor during a period of time significantly longer than the period of attaining equilibrium poisoning, as time between two D2O condensate overflows into the RA reactor core. Thus a ten day experimental campaign was planned to be done in December 1976. This report presents the most important results of safety analyses and preparation which show that, during this experimental period, the reactor operation is absolutely safe taking into account the most important parameters influencing reactor safety, as reactivity, thermal and temperature limits for fuel and the reactor, etc. Data to be obtained during this experimental campaign are significant because they would enable definition of future supply of fresh fuel during the transition period

  2. Calculation and measurement of uranium temperature during irradiation in the experimental channel in the RA reactor reflector

    International Nuclear Information System (INIS)

    Irradiation of uranium samples was not done at the RA reactor. Irradiation of fissile materials demands special study as well as safety analysis and safety report. Theoretical studies showed that the temperature of samples could be up to 300 deg C at 6.5 MW reactor power. For that reason it was decided to measure the temperature of capsules containing samples during irradiation which causes problems since they are placed in the core. Temperature dependent on the reactor power was measured by thermocouples and showed that the temperature of the irradiation capsule was 200 deg C at 6.5 MW

  3. Supercritical water-cooled nuclear reactors: thermodynamic-cycles plant layouts and thermal aspects of pressure-channel design

    International Nuclear Information System (INIS)

    Research activities are currently conducted worldwide to develop Generation IV nuclear reactor concepts with the objective of improving thermal efficiency and increasing economic competitiveness of Generation IV Nuclear Power Plants (NPPs) compared to modern thermal power plants. The Super-Critical Water-cooled Reactor (SCWR) concept is one of the six Generation IV options chosen for further investigation and development in several countries including Canada. Water-cooled reactors operating at subcritical pressures (7-16 MPa) have provided a significant amount of electricity production for the past 50 years. However, the thermal efficiency of the current NPPs is not very high (30-35%). As such, more competitive designs, with higher thermal efficiencies, which will be close to that of modern coal-fired thermal power plants (45-50%), need to be developed and implemented. Previous studies have shown that direct cycles with no-reheat and single-reheat configurations are the best choices for SCWR concepts. This paper presents three SCW NPP concepts based on direct single-reheat and no-reheat regenerative cycles, and indirect single-reheat regenerative cycle. In general, there are many technical challenges associated with the single-reheat and no-reheat SCW NPP configurations. The direct single-reheat cycle requires nuclear steam-reheat, thus increasing the complexity of the pressure-tube reactor-core design. The nuclear steam reheat seems to be possible only in a pressure-tube reactor. Conversely, the major technical challenge associated with the no-reheat cycle is the high moisture content in the low-pressure-turbine exhaust. The direct no-reheat cycle can be implemented in NPP with pressure-vessel or pressure-tube reactors. The gross thermal efficiency of the direct cycles was determined to be about 50%. The indirect single-reheat cycles utilize heat exchangers (steam generators) to transfer heat from the reactor coolant to the secondary-loop working fluid. The

  4. 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.

  5. 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)

  6. Reactor utilization

    International Nuclear Information System (INIS)

    In 1962, the RA reactor was operated almost three times more than in 1961, producing total of 25 555 MWh. Diagram containing comparative data about reactor operation for 1960, 1961, and 1962, percent of fuel used and U-235 burnup shows increase in reactor operation. Number of samples irradiated was 659, number of experiments done was 16. mean powered level was 5.93 MW. Fuel was added into the core twice during the reporting year. In fact the core was increased from 56 to 68 fuel channels and later to 84 fuel channels. Fuel was added to the core when the reactivity worth decreased to the minimum operation level due to burnup. In addition to this 5 central fuel channels were exchanged with fresh fuel in february for the purpose of irradiation in the VISA-2 channel

  7. 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

  8. 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

  9. Calculation and measurement of the uranium temperature during irradiation in the experimental channel in the reflector of the RA reactor - Annex 15

    International Nuclear Information System (INIS)

    Upon demand of the Laboratory for fuel reprocessing, six domestic metal uranium pellets were exposed to neutron flux ( 4 - 5 1012 n cm-2 sec -1) in the RA reactor. Irradiation of fuel demanded special analyses for safety reasons. Weight of the fuel pellets was 13 - 20 g, having diameter 20 mm. pellets were placed in leak tight aluminium capsules with helium. The irradiation was dome in the aluminium experimental channel in the graphite reflector. Theoretical study has shown that the expected fuel temperature in the core could be up to 300 deg C at nominal power. For that reason temperature of the capsule with the uranium sample was measured during irradiation by using thermocouples. Results showed the discrepancy between measure and calculated values to be about 30%

  10. 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)

  11. Progress of Filtered Neutron Beams Development and Applications at the Horizontal Channels No.2 and No.4 of Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    The neutron filter technique has been applied to create mono-energetic neutron beams with high intensity, at the horizontal channels No.2 and No.4 of the Dalat nuclear research reactor. The mono-energetic neutron beams that have been developed for researches and applications are thermal (0.025 eV), 24 keV, 54 keV, 59 keV, 133 keV and 148 keV. The relative intensities of main peak in filtered neutron energy spectra and the collimated neutron fluxes at the sample irradiation positions are 90 - 96% and 2.8×105 - 7.8×106 n/cm2.s, respectively. Monte Carlo simulations and transmission calculations were performed to each neutron energy beam for optimal design of geometrical structure and neutron filter materials. These filtered neutron beams have been applied efficiently for experimental researches on neutron total and capture cross sections measurements, and elemental analysis in various kinds of samples based on the prompt gamma neutron activation analysis method. This paper reviews the progress of filtered neutron beams development and its applications for past many years at the Dalat nuclear research reactor. (author)

  12. Calibration of the B54X channel and implementation of k0-NAA at the RA3-reactor, Ezeiza, Argentina

    International Nuclear Information System (INIS)

    The B54X position of the 8 MW RA3 research reactor at the Ezeiza Atomic Centre of the Argentine National Atomic Energy Commission is currently being used for NAA irradiations. The facility with a nominal average fluence of 5 × 1013 cm-2 s-1 is dedicated to long irradiations of up to 5 h. Samples are being measured after a decay of typically 7 and 30 days. With the aim of implementing the k0-NAA method at the Nuclear Analytical Techniques Laboratory of the Centre, the reactor parameters α and f were estimated applying multi monitor methods using the Kayzero for Windows software. After a careful recalibration of the HPGe detector, SMELS III, NIST SRM 1633b and several other matrix RM's were analyzed using the k0 standardization in order to verify the proper implementation of the k0-NAA approach. The found accuracy and associated uncertainties are discussed. In general, good agreement was obtained between results of this work and the reference values of the individual reference materials, thus proving successful first implementation of the above method and trueness of the results achieved. The obtained detection limits for several elements were evaluated. (author)

  13. Nuclear reactor

    International Nuclear Information System (INIS)

    In an improved reactor core for a high conversion BWR reactor, Pu-breeding type BWR type reactor, Pu-breeding type BWR type rector, FEBR type reactor, etc., two types of fuel assemblies are loaded such that fuel assemblies using a channel box of a smaller irradiation deformation ratio are loaded in a high conversion region, while other fuel assemblies are loaded in a burner region. This enables to suppress the irradiation deformation within an allowable limit in the high conversion region where the fast neutron flux is high and the load weight from the inside of the channel box due to the pressure loss is large. At the same time, the irradiation deformation can be restricted within an allowable limit without deteriorating the neutron economy in the burner region in which fast neutron flux is low and the load weight from the inside of the channel box is small since a channel box with smaller neutron absorption cross section or reduced wall thickness is charged. As a result, it is possible to prevent structural deformations such as swelling of the channel box, bending of the entire assemblies, bending of fuel rods, etc. (K.M.)

  14. Aerodynamic study of the fluid flow in the channel of a reactor filled with internally and externally cooled fuel elements

    International Nuclear Information System (INIS)

    A study is made of the problem of the flow-rate and pressure distributions along the length of two volumes, internal and external, bounded by a series of non-continuous annular elements placed along the channel axis. It is observed that the phenomenon can easily be represented by equations. The theoretical expressions observed are particularly simple when the distances between the elements are above a certain minimum value. The experimental work has made it possible to show that the theoretical formulation derived is valid with a very great accuracy. The experimental study has also been carried out in the case of a very small spacing between the elements. It has been possible to show in this case that the hypothesis made for deriving the theoretical expressions was perfectly justified. In the last part finally, we consider the practical problem of evaluating the pressure-drops between the ends of a series of annular elements. (author)

  15. 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

  16. Re-determination and re-evaluation of the f and α parameters in channels Y4 and S84 of the BR1 reactor, for use in k0-NAA at DSM Research

    Science.gov (United States)

    De Wispelaere, A.; De Corte, F.; Bossus, D. A. W.; Swagten, J. J. M. G.; Vermaercke, P.

    2006-08-01

    Since the introduction of k0-based NAA as an analytical tool in 1989, all irradiations by DSM Research are done in channels Y4 and S84 of the BR1 reactor in Mol (Belgium). The last determination of f and α-values for these channels was performed in 1993. Although the configuration of the reactor did not change over all these years and therefore no change in f and α was to be expected, DSM Research decided to re-determine both parameters in both channels. Having much experience in this field, the INW k0-group was asked by DSM Research to perform this re-determination, in co-operation with the SCK, Mol. As the flux in channel Y4 is not constant during the start up and the scramming of the reactor, a numerical integration method was applied. This is a new approach in comparison with all previous reported data from DSM Research, where this change in flux was not taken into account. For the work presented here, use was made of the most recent nuclear data available in the literature.

  17. Design and the construction of some functional electronics modular for (n,2γ) spectrometer at a horizontal channel in the Dalat research reactor

    International Nuclear Information System (INIS)

    As the nuclear disintegration is characteristic for a given isotope, specific measurements can be performed by means of coincidence techniques, whereby correlated phenomena must be simultaneously detected in order to be counted. As well bete-gamma as gamma-gamma cascades of the disintegration, which occur within very short time intervals, are suitable for these purposes. Also both annihilation gamma rays can be measured in coincidence. The pulses coming from the components of the cascades can be selected in energy by means of a pulse height analyser, and are fed into the coincidence circuit. In order to be counted, two pulses must arrive within the resolving time τ of the coincidence unit. Typical values of τ are of the order of the as for 'slow' coincidence and down to the ns for 'fast' coincidence. Actually, Coincidence and Linear amplifier units are two important pieces of the measuring system. The main task of the interbal sub-project is to study on and to design these NIM-standard blocks those are able to combine with other needed electronics modulars for the performance of a gamma-gamma coincidence system with the sake of nuclear structure research at a horizontal channel in the research reactor Dalat. (author)

  18. Flow visualization study of two-phase flow in the horizontal annulus of the fuel-channel outlet end-fitting of a CANDU reactor

    International Nuclear Information System (INIS)

    In CANDU-6 reactors, the pressurized hightemperature coolant flows through 380 fuel channels passing horizontally through the core. In 1996, higher than expected rates of wall thinning of the outlet feeders were ascribed to flow-accelerated corrosion (FAC). Such corrosion is strongly influenced by the hydrodynamics of the coolant. Results of preliminary flow visualization and modelling studies have suggested that flow conditions in the end-fitting annulus upstream of the outlet feeder may influence the pattern of FAC. For a full-scale flow visualization, an acrylic test section was built to simulate the cylindrical end-fitting with its annulus flow path. The tests were performed with water and air at atmospheric pressure and room temperature. The phase distribution along the length of the annulus was recorded with a digital video recorder. Size, concentration and velocity of the air bubbles at particular locations were studied with a high-speed digital still camera and a high-speed digital video camera. Phase distributions and variations in bubble size with velocity were determined. Significant effects on the flow patterns of spacer buttons in the annulus were observed. A commercial computational fluid dynamics (CFD) code-Fluent 6.1-was used to model the results. (authors)

  19. Nuclear reactors

    International Nuclear Information System (INIS)

    This draft chart contains graphical symbols from which the type of (nuclear) reactor can be seen. They will serve as illustrations for graphical sketches. Important features of the individual reactor types are marked out graphically. The user can combine these symbols to characterize a specific reactor type. The basic graphical symbol is a square with a point in the centre. Functional groups can be depicted for closer specification. If two functional groups are not clearly separated, this is symbolized by a dotted line or a channel. Supply and discharge lines for coolant, moderator and fuel are specified in accordance with DIN 2481 and can be further specified by additional symbols if necessary. The examples in the paper show several different reactor types. (orig./AK)

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

    International Nuclear Information System (INIS)

    The objective of the project was to design the head of the CO2 coolant loop for cooling the materials during irradiation in the RA reactor. Six heads of coolant loops will be placed in the RA reactor, two in the region of heavy water in the experimental channels VEK-6 and four in the graphite reflector in the channels VEK-G. Materials for irradiation are metallurgy and chemical samples. In addition to the project objectives, this volume includes technical specifications of the coolant loop head, thermal calculations, calculations of mechanical stress, antireactivity and activation of the construction materials, cost estimation, scheme of the coolant loop head, diagrams of CO2 gas temperature, thermal neutron flux distribution, design specifications of two proposed solutions for head of low temperature coolant loop

  2. 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.

  3. Research nuclear reactors

    International Nuclear Information System (INIS)

    Since the divergence of the first nuclear reactor in 1942, about 600 research or test reactors have been built throughout the world. Today 255 research reactors are operating in 57 countries and about 70% are over 25 years old. Whereas there are very few reactor types for power plants because of rationalization and standardisation, there is a great diversity of research reactors. We can divide them into 2 groups: heavy water cooled reactors and light water moderated reactors. Heavy water cooled reactors are dedicated to the production of high flux of thermal neutrons which are extracted from the core by means of neutronic channels. Light water moderated reactors involved pool reactors and slightly pressurized closed reactors, they are polyvalent but their main purposes are material testing, technological irradiations, radionuclide production and neutron radiography. At the moment 8 research reactors are being built in Canada, Germany, Iran, Japan, Kazakhstan, Morocco, Russia and Slovakia and 8 others are planned in 7 countries (France, Indonesia, Nigeria, Russia, Slovakia, Thailand and Tunisia. Different research reactors are described: Phebus, Masurca, Phenix and Petten HFR. The general principles of nuclear safety applied to test reactors are presented. (A.C.)

  4. Boiling in MTR Fuel Element Channels as Cooling Mechanism During Partial Loss-of-Coolant Accident (LOCA) in Pool Type Reactors

    International Nuclear Information System (INIS)

    The main objective of the present work is to measure the effective wetted zone above the nominal water level in the fuel element channels in various channel power and nominal water levels. These results serve as the basis for a simple model for calculating the maximum temperature in the real case of an MTR fuel element

  5. 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.

  6. 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.

  7. Detail analysis of tritium permeation in the metal liquid channels of the regenerating sheaths of a fusion reactor in presence of helium bubbles; Analisis de detalle de la permeacion de tritio en los caneles de metal liquido de las envolturas regeneradoras de un reactor de fusion en presencia de burbujas de helio

    Energy Technology Data Exchange (ETDEWEB)

    Banet, L.; Mas de les Valls, E.; Sedano, L. A.

    2012-07-01

    Inside the channels of liquid metal of the fusion reactor regenerative wrappers, the possible existence of nucleated helium bubbles is not remote. Helium is formed joined the tritium in the escaped neutrons of plasma with lithium. The accumulation of helium in the contact surfaces, between the structure and ML, lead a reduction of heat transfer, at the same time a reduction in the permeation of tritium. The coexistence of three phases in touch: metal liquid, helium and structural material, makes the transport of heat and tritium in a complex phenomenon. To enrich tritium transport studies conducted in the past, there is now a detail analysis of the helium bubble environment adhered to the channel ML wall of a regenerative wrap. For the study we used a CFD tool development on free code OpenFOAM.

  8. 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)

  9. 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

  10. RB research reactor Safety Report

    International Nuclear Information System (INIS)

    This RB reactor safety report is a revised and improved version of the Safety report written in 1962. It contains descriptions of: reactor building, reactor hall, control room, laboratories, reactor components, reactor control system, heavy water loop, neutron source, safety system, dosimetry system, alarm system, neutron converter, experimental channels. Safety aspects of the reactor operation include analyses of accident causes, errors during operation, measures for preventing uncontrolled activity changes, analysis of the maximum possible accident in case of different core configurations with natural uranium, slightly and highly enriched fuel; influence of possible seismic events

  11. 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.

  12. 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

  13. A trolley mounted magazine for reactor maintenance

    International Nuclear Information System (INIS)

    This paper describes the design of a mechanism incorporating a rotary magazine to be mounted on a fuelling machine transport trolley for use at a Darlington reactor during a feeder replacement or maintenance outage. The magazine stores reactor channel maintenance components, such as channel isolation plugs and vented closure plugs, in twelve available magazine channels. Use of the magazine rather than a fuelling machine reduces the time required to transfer such components between the Central Service Area and reactor channels. Component transfers are accomplished by locking the fuelling machine onto one of the magazine channels and using a local controller to execute commands received from the fuel handling control system. (author)

  14. Radioactive products of corrosion in the experimental channels and in the water of the first loop of the thermal neutron research reactor 2000 in Sofia

    International Nuclear Information System (INIS)

    The activity of the products of corrosion in the horizontal and vertical channels and in the water of the first circulation loop of the IRT-2Ge(Li) detector with a sensitivity of 10 cm3 and a 4000-channel analyser were used. The results show that besides the isotopes with a short life time 24Na (15.05 h) and 65Ni (2.5 h) obtained from the reaction 27Al(n,α)24Na and 65Cu(n,p)65Ni the samples contain the isotopes with a long lifetime: 56Nn (313.5 h), 45Sc (83.9 days), 59Fe (46.5 days) and 60Co (5.26 years). The application of a Ge(Li) detector makes possible a fast control with an accurate determination of the isotope content, which is of great importance for the Nuclear Power Stations. (author)

  15. A BWR fuel channel tracking system

    International Nuclear Information System (INIS)

    A relational database management system with a query language, Reference 1, has been used to develop a Boiling Water Reactor (BWR) fuel channel tracking system on a microcomputer. The software system developed implements channel vendor and Nuclear Regulatory Commission recommendations for in-core channel movements between reactor operating cycles. A BWR Fuel channel encloses the fuel bundle and is typically fabricated using Ziracoly-4. The channel serves three functions: (1) it provides a barrier to separate two parallel flow paths, one inside the fuel assembly and the other in the bypass region outside the fuel assembly and between channels; (2) it guides the control rod as it moves between fuel assemblies and provides a bearing surface for the blades; and (3) it provides rigidity for the fuel bundle. All of these functions are necessary in typical BWR core designs. Fuel channels are not part of typical Pressurized Water Reactor (PWR) core designs

  16. Study of instabilities in phase by using the tool Dynamics: analysis of the evolution space temporary of the waves of density in channels of reactors BWR

    International Nuclear Information System (INIS)

    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.

  17. 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

  18. 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.

  19. Cesium Sorption Rate on Non-Crushed Rock Measured by a New Apparatus Based on a Micro-Channel-Reactor Concept

    International Nuclear Information System (INIS)

    Since nuclide migration through rock mediums is an extremely slow process, experimental effort to evaluate the barrier performance of geologic disposal such as the diffusion coefficient (De) and the distribution coefficient (Kd) requires relatively long testing periods and chemically stable conditions. We have developed a fast method to determine both De and Kd by using a non-crushed rock sample. In this method, a fluoro-plastic plate with a micro channel (10- 200-μm depth) is placed just beneath a rock-sample plate, and a radionuclide solution is injected into the channel at constant rate. A part of radionuclide diffuses into the rock matrix and/or adsorbs on the rock surface. The difference between the inlet and outlet radionuclide flux is simply related to the apparent diffusion coefficient (Da) of the rock sample. In this study, we estimated Kd of Cs for granite by using the equilibrium model, finding that Kd decreased with increasing flow rate. This dependence of Kd on flow rate implies the state of sorption equilibrium. The adsorption and desorption curves of 134Cs were thus measured, and the rate constants for both processes were obtained by adopting a first-order rate law. The rate constants of sorption (k+) and desorption (k-) were obtained as a function of flow velocity; constant values of both were observed. Kd was calculated from k+/ k- and then compared with that determined by conventional batch sorption method using a crushed rock sample. The Kd values determined by the present and conventional methods are in good accordance; however, the testing periods for each method are very different; namely, 1 day and 7 days, respectively. (authors)

  20. 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

  1. 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

  2. On the boundary heating conditions difference in the DNB channels

    International Nuclear Information System (INIS)

    Theoretical and experimental work on DNB in reactor channels is reviewed. The effects of channel characteristics, temperature, heat flux and flow rate on DNB correlations are discussed and relations to burnout are considered

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Cover gas spaces for primary coolant vessel, such as a reactor container, a pump vessel and an intermediate heat exchanger vessel are in communication with each other by an inverted U-shaped pressure conduit. A transmitter and a receiver are disposed to the pressure conduit at appropriate positions. If vibration frequencies (pressure vibration) from low frequency to high frequency are generated continuously from the transmitter to the inside of the communication pipe, a resonance phenomenon (air-column resonance oscillation) is caused by the inherent frequency or the like of the communication pipe. The frequency of the air-column resonance oscillation is changed by the inner diameter and the clogged state of the pipelines. Accordingly, by detecting the change of the air-column oscillation characteristics by the receiver, the clogged state of the flow channels in the pipelines can be detected even during the reactor operation. With such procedures, steams of coolants flowing entrained by the cover gases can be prevented from condensation and coagulation at a low temperature portion of the pipelines, otherwise it would lead clogging in the pipelines. (I.N.)

  4. 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

  5. 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

  6. 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)

  7. 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...

  8. Hydrodynamics of a Monolithic Stirrer Reactor

    OpenAIRE

    Kritzinger, H.P.

    2011-01-01

    The Monolithic Stirrer Reactor (MSR) is a novel concept for heterogeneously catalyzed reactors and is presented as an alternative device to slurry reactors. It uses a modified stirrer on which structured catalyst supports (monoliths) are fixed to form permeable blades. The monoliths consist of small square parallel channels on which a layer of catalytic material can be applied. The stirrer now has both a catalytic and a mixing function. The main advantage of this reactor type is the ease of t...

  9. Measurements of neutron flux in the RA reactor

    International Nuclear Information System (INIS)

    This report includes results of the following measurements performed at the RA reactor: thermal neutron flux in the experimental channels, epithermal and fast neutron flux, neutron flux in the biological shield, neutron flux distribution in the reactor cell

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

    OpenAIRE

    Khater Hany; Abu-El-Maty Talal; El-Morshdy El-Din Salah

    2006-01-01

    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...

  11. Monitoring parameters for fuel channel safety in AHWR - an overview

    International Nuclear Information System (INIS)

    Full text: Advanced heavy water reactor (AHWR) is a vertical pressure tube type reactor having boiling light water coolant circulating by thermosyphon in reactor coolant channels. The fuel channel monitoring system instrumentation becomes very important for maintaining channel safety and integrity. The various monitoring parameters for fuel channel monitoring system can be classified as (a) continuous monitoring and (b) sample monitoring. The continuous monitoring parameters for fuel channel safety and operation are i) channel inlet pressure ii) channel temperature iii) channel flow and iv) channel power. Coolant activity and Annulus gas monitoring are sample-monitoring parameters for ensuring channel integrity. The inlet pressure to the core being same for all the channels and AHWR being a pressurized boiling water system, pressure and temperature measurement cannot be considered as monitoring parameter for channel safety. Hence the continuously monitoring parameter for fuel channel safety becomes flow and power monitoring only. Since channel DT is negligible, the void fraction measurement is required for getting the channel power. Hence in AHWR the measurement of flow at the single-phase inlet section and measurement of two-phase flow/void fraction at the tail pipe section becomes important for estimating channel flow and channel power for reactor safety and operation. Measurement of two-phase flow in natural circulation system is complex and various attempts were made to understand and develop instrumentation so that an effective channel monitoring can be carried out. The channel integrity is monitored by sampling methods namely annulus gas monitoring and activity monitoring. Since the annulus gap is partially open, the moisture/activity in the annulus air need to be measured accurately to identify and quantify the leak rate and in turn the pressure tube/calandria tube integrity. This paper describes in detail the various aspects of fuel channel monitoring

  12. Measuring vibrations in fuel channels CNE

    International Nuclear Information System (INIS)

    This paper present a description of implementation and execution of vibration measurements made at the request of NUCLEOELECTRICA ARGENTINA S.A. on the ends of the reactor fuel channels of Embalse Nuclear Power Plant to explore possible differences between the dynamic behavior of empty fuel channel and with full charge of fuel elements inside. (author)

  13. Digital instrumentation system for nuclear research reactors

    International Nuclear Information System (INIS)

    This work describes a proposal for a system of nuclear instrumentation and safety totally digital for the Argonauta Reactor. The system divides in the subsystems: channel of pulses, channel of current, conventional instrumentation and safety system. The connection of the subsystems is made through redundant double local net, using the protocol modbus/rtu. So much the channel of pulses, the current channel and safety's system use modules operating in triple redundancy. (author)

  14. BWR type reactor

    International Nuclear Information System (INIS)

    No channel box is mounted to a fuel assembly, but a partition plate for separating coolant flow channels between each of fuel bundles is disposed between each of fuel bundles along the direction of height for the reactor core instead of the channel box. The partition plate has a shape surrounding the fuel bundles only in a specific region, or so that coolant flow channels for a plurality of fuel bundles of identical output are integrated. As a result, cross-flow of coolants can be prevent without channel box and, further, radial expansion of the channel box can be eliminated. As the same time, the bending for the entire assembly due to the irradiation growth of the channel box is also eliminated and structural stability can be attained without using upper grid plates. Further, it is possible to minimize the pressure loss caused between the upper and lower portions of the assembly and it is possible to adjsut with respective thermohydrodynamic properties of the high conversion region and the burner region. (K.M.)

  15. Device for nuclear reactor control

    International Nuclear Information System (INIS)

    The device for power height distribution control in channel-type uranium-graphite reactor cores is described. The device is a water filled vertical channel positioned in the reactor core. The device consists of a controlling rod, displacer in a form of a throttle and gas cavity and discharge throttle. The rod is fixed in upper position with an electromagnet. By shifting a displacer and changing flow rate established are the required height and position of a controlling liquid column. In the emergency protection, a drive shifts the displacer under core space or the displacer drops under the action of its own weight at electromagnet clutch doenergyzation whereas the channel is filled by liquid. The application of the suggested device permits to improve economic and operating characteristics of reactors

  16. Reactor Physics

    International Nuclear Information System (INIS)

    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

  17. Reactor Physics

    International Nuclear Information System (INIS)

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. 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 materials testing 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 2001 are summarised

  18. 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.

  19. Calibration of RB reactor power

    International Nuclear Information System (INIS)

    The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8√2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation

  20. 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

  1. 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

  2. Thermal-hydraulic interfacing code modules for CANDU reactors

    International Nuclear Information System (INIS)

    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

  3. 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.

  4. Innovative reactor technologies - Enabling success

    International Nuclear Information System (INIS)

    Full text: 1. Introduction. AECL has been pursuing innovation in reactor design using an evolutionary approach. The Advanced CANDU Reactor, or ACR design is the logical next step in the CANDU fuel-channel reactor design process, and achieves major improvements in economics while expanding safety margins. ACR technology is also the start of the long-term development path for CANDU fuel channel reactors. This path fits in with the long-term development directions identified in initiatives such as the Generation IV International Forum (GIF) and INPRO. AECL sees the product evolution from the ACR to the Supercritical Water-costed Reactor (SCWR), one of the concepts identified by GIF for further development. One of the prominent characteristics of the heavy-water moderated fuel channel reactor approach is the high potential for innovation. 2. Background to Development Strategy. As part of the pressurized water reactor family, CANDU's share many characteristics with other light water reactors, while retaining a set of distinctive features: High pressure water coolant in individual fuel channels, with low-pressure, low temperature moderator; Horizontal fuel channel design with on-power refuelling; Simple, easily-fabricated fuel bundle design; Use of heavy water to improve neutron efficiency. Since the original CANDU reactors were first built, the advent of deregulated, competitive energy markets has strongly emphasized the importance of low capital cost and short construction time. The ACR design innovations are chosen to respond to this evolution of energy markets, while retaining the proven features of the CANDU line. 3. Innovative CANDU Product Designs. The ACR-700 design is an evolutionary development of familiar CANDU technology, adding a carefully chosen set of innovations to the major improvements in economics, operations and safety margins: Slightly enriched uranium fuel contained in CANFLEX bundles; Light water replacing heavy water as the reactor coolant; More

  5. Gas conduction in a high temperature reactor

    International Nuclear Information System (INIS)

    The hot gas is distributed and mixed by polygonal blocks in the reactor floor. Radial and an annular channel are used for this purpose. This annular channel also carried circular ducts distributed evenly over the circumference, which lead to a heat exchanger. Temperature differences across the crossection of the reactor floor are evened out by multiple deflection, combination and renewed splitting of the gas flows. (DG)

  6. 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.

  7. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  8. Reactor core monitoring device

    International Nuclear Information System (INIS)

    The device of the present invention reliably and conveniently detects an event of rapid increase of a coolant void coefficient at a portion of a channel by flow channel clogging event in a PWR-type reactor. Namely, upon flow channel clogging event, the coolant void coefficient is increased, an effective density is lowered, and a coolant shielding effect is lowered. Therefore, fast neutron fluxes at the periphery of a pressure tube are increased. The increase of the fast neutron fluxes is detected by a fast neutron flux detector disposed in a guide tube of an existent neutron flux detector. Based on the result, increase of coolant void coefficient can be detected. When an average void coefficient reaches from 30% to 100%, for example, the fast neutron fluxes are increased by about twice at a neutron permeation distance of coolants of about 10cm, thereby enabling to perform effective detection. (I.S.)

  9. Practical course on reactor instrumentation

    International Nuclear Information System (INIS)

    This course is based on the description of the instrumentation of the TRIGA-reactor Vienna, which is used for training research and isotope production. It comprises the following chapters: 1. instrumentation, 2. calibration of the nuclear channels, 3. rod drop time of the control rods, 4. neutron flux density measurements using compensated ionization, 5. neutron flux density measurement with fission chambers (FC), 6. neutron flux density measurement with self-powered neutron detectors (SPND), 7. pressurized water reactor simulator, 8. verification of the radiation level during reactor operation. There is one appendix about neutron-sensitive thermocouples. (nevyjel)

  10. fuel management in candu reactors: RFSP code

    International Nuclear Information System (INIS)

    The objective of in-core fuel management is to determine the required refuelling strategies for safe and reliable operation of the reactor with minimum total energy cost. CANDU reactors use natural uranium fuel and rely on semi-continuous on-power refuelling. For the purpose of fuel management, the CANDU core with 380 fuel channels is modelled dividing into inner-and outer core. Refuelling rate in the CANDU reactors is evaluated in three periods for the whole operating life: 1)From the initial core to refuelling onset (100-150 EFPD), 2) the intermediate period (400-500 EFPD), and 3)the equilibrium period (approximately 30 years). A channel in the CANDU-6 reactor contains 12 bundles, in the refuelling operation some bundles do not discharged, but are shifted to other place in the same channel. One of the methods used for selection the channel and determination the bundles to be discharged is simulation method one of which is the RFSP (reactor fuelling simulating program). RFSP is a computer programme to do neutronic calculations for CANDU reactors. It can calculate both static and time-dependent neutron flux and power distributions in the core. It is a modular program containing a lot of modules. RFSP can perform fuel-management calculations and simulate a reactor operating history at specified intervals, taking burnup steps and channel refuelling into account

  11. 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)

  12. 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

  13. Heat and mass transfer intensification in coaxial reactor

    Science.gov (United States)

    Ananyev, D. V.; Halitova, G. R.

    2014-04-01

    The work considers heat and mass transfer in the homophasic polymerization reactor. The reactor is a coaxial channel with internal tube in the form of a channel of confusor-diffuser type. The authors compared the degree of polymer transformation in the intensified coaxial reactor with internal tube of confusor-diffuser type and the reactor with constant rectilinear longitudinal section. It was found that in coaxial channels with internal tube of confusor-diffuser type, it is possible to reach high values of the transformation degree and to improve the quality of the obtained polymer.

  14. Research reactors

    International Nuclear Information System (INIS)

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world's research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted

  15. Reactor container

    International Nuclear Information System (INIS)

    Object: To provide a jet and missile protective wall of a configuration being inflated toward the center of a reactor container on the inside of a body of the reactor container disposed within a biological shield wall to thereby increase safety of the reactor container. Structure: A jet and missile protective wall comprised of curved surfaces internally formed with a plurality of arch inflations filled with concrete between inner and outer iron plates and shape steel beam is provided between a reactor container surrounded by a biological shield wall and a thermal shield wall surrounding the reactor pressure vessel, and an adiabatic heat insulating material is filled in space therebetween. (Yoshino, Y.)

  16. 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.

  17. Reactor inspection and maintenance machine senses and homes in on reactor end fittings

    International Nuclear Information System (INIS)

    The Universal Delivery Machine (UDM) is a new CANDU reactor maintenance tool that allows safe, timely, and cost-effective inspection and maintenance of fuel channels. The UDM must align precisely with reactor end-fittings in order to clamp onto fuel channels without applying excessive force. This alignment process is called fine homing. This paper describes the UDM instrumentation and control design features used in the fine homing process. (author)

  18. 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)

  19. Heterogeneous reactors

    International Nuclear Information System (INIS)

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author)

  20. 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

  1. 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.

  2. 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.

  3. Assessment of Inner Channel Blockage on the Annular Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Shin, C. H.; In, W. K.; Oh, D. S.; Chun, T. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    A dual-cooled annular fuel for a pressurized water reactor (PWR) has been introduced for a significant amount of reactor power uprate. The Korea Atomic Energy Research Institute (KAERI) has been performing a research to develop a dual-cooled annular fuel for the power uprate of 20% in an optimized PWR in Korea, OPR1000. An inner channel blockage is principal one of technical issues of the annular fuel rod. The inner channel in an annular fuel is isolated from the neighbor channels unlike the outer channels. The inner channel will be faced with a DNB accident by the partial blockage. In this paper, the largest fractional channel blockage was assessed by subchannel analysis code MATRA-AF and an end plug design to complement inlet blockage of inner channel was estimated by CFD code, CFD-ACE

  4. Plasma reactor

    OpenAIRE

    Molina Mansilla, Ricardo; Erra Serrabasa, Pilar; Bertrán Serra, Enric

    2008-01-01

    [EN] A plasma reactor that can operate in a wide pressure range, from vacuum and low pressures to atmospheric pressure and higher pressures. The plasma reactor is also able to regulate other important settings and can be used for processing a wide range of different samples, such as relatively large samples or samples with rough surfaces.

  5. Reactor physics

    International Nuclear Information System (INIS)

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  6. Wear inspection for fuel channels in HANARO

    International Nuclear Information System (INIS)

    It has been observed that fuel assemblies have mechanical damage on some components due to the flow-induced vibration in the fuel channels. The major damage was the fretting-wear on the bottom end plates and spacer plates. Both the components are aluminum but cause the fretting-wear on the corresponding surfaces of the zirconium fuel channels. Also the bottom guide arms and top springs of the fuel are considered to be major parts that cause wear on the fuel channels. Therefore, the inspection of the inner surfaces of the fuel channels is required from a lifetime point of view. It is very difficult and time-consuming work to remove and install the fuel channels because of their inherent characteristics and the physical interference of other components in the reactor core. Thus we developed special tools for the inspection of the fuel channels by using an impression material without the removal of the reactor components. The impression material is a compound to replicate the damage of the fuel channels within a limited working time considering the hardening time as well as the radiation effect. The wear-inspection tool is a mechanical tool to press the impression compound against the inner walls of the fuel channels where the fuel components contact. The inspection tool has 36 molding cups that are operated in radial directions by turning the central rod with a fuel-handling tool from the pool top 12m above the reactor core. The same concept but a more compact design is applied to the cylindrical fuel channel. The wear inspection was successfully accomplished for a few fuel channels. The results show visible wear marks on a hexagonal fuel channel at the positions corresponding to fuel components such as the bottom guide arms and top springs. The wear damage is slight, approximately 0.2mm depth, in comparison with the thickness (1.6mm) of the fuel channel. No visible wear mark has been found in the cylindrical flow tubes so far. The wear inspection is being continued

  7. After heat removing device for FBR type reactor

    International Nuclear Information System (INIS)

    An annular liquid (water) pool is formed radially surrounding a reactor container and a reactor safety container. An annular cavity wall is formed in the liquid pool, and the inside of the cavity wall is formed as a liquid channel. If the temperature of liquid sodium in the reactor container rises by the after heat of the nuclear fuels, the temperature of the reactor safety container also rises to a high temperature, and the amount of heat radiated from the surface is increased. Water in the liquid channel heated by undergoing the radiation heat forms upward streams in the liquid channel by an air lift-effect caused by rising of boiling air bubbles. Namely, the water in the liquid pool rises the liquid channel while boiling to cool the reactor safety container. With such a constitution, after heat can be removed continuously by the spontaneously circulating water. (I.N.)

  8. CFD (computational fluid dynamics) analysis of a novel reactor design using ion transport membranes for oxy-fuel combustion

    International Nuclear Information System (INIS)

    Conventional two-channel ITM (ion transport membrane) reactors applied to oxy-combustion, face the potential drawback of high thermal gradients and high local temperatures, which can result in membrane damage. In such reactors, air flows on the feed side and fuel are introduced on the permeate side, where it reacts with the permeated oxygen. In this work, we propose to use a three-channel configuration in which a porous plate is used to separate the permeate stream from the fuel stream, allowing the fuel to diffuse gradually into the permeate side. The gradual combustion of the fuel results in a slow temperature rise and a more spatially uniform temperature distribution along the membrane. We model this three-channel reactor using computational fluid dynamics and compare its performance to a conventional two-channel reactor. It is shown that, indeed, in case of a two-channel reactor, a high temperature zone is concentrated near the inlet, whereas the three-channel reactor produces a milder temperature gradient along the reactor length. The more-uniform heat flux associated with the latter results in a moderate temperature distribution and reduction in the wall shear stress along the channels and the associated pressure drop. The more uniform temperature distribution should be less damaging to the membrane. The reaction zone associated with the gradual fuel diffusion into the sweep side improves the membrane performance by maintaining a more uniform oxygen flux. - Highlights: • A novel concept of 3-channel reactor using ITM (ion transport membrane) is presented. • Comparison of present 3-channel reactor with the conventional 2-channel reactor. • A more spatially uniform temperature is obtained using the 3-channel reactor. • 3-channel reactor produces a milder temperature gradient along the reactor length. • The reaction zone in 3-channel reactor improves the membrane performance

  9. 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.

  10. 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

  11. 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

  12. 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)

  13. 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.

  14. 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.

  15. Multifunctional reactors

    OpenAIRE

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much emphasis in research in the last decade. A survey is given of modern developments and the first successful applications on a large scale. It is explained why their application in many instances is ...

  16. 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)

  17. 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)

  18. Breeder reactors

    International Nuclear Information System (INIS)

    The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components

  19. Innovative reactor technologies - Enabling success

    International Nuclear Information System (INIS)

    Many innovative reactors are being discussed, offering advantages in economics, sustainability, environmental impact, versatility and efficiency. To be successful, however, innovative reactors must meet the requirements for a successful build project. This requires achieving the mixture of innovation and proveness required to meet the first-of-a-kind hurdle. Based on the successful CANDU 6 reactor, a design still being built today, the ACR adds specific innovations in key areas chosen to achieve a balanced design. Capital cost has been significantly reduced by optimising the reactor-core design and simplifying systems. Key changes in this area include a move from a heavy water coolant to a light water coolant, and the adoption of SEU fuel. Construction times have also been reduced by using a modular design that takes advantage of modern construction techniques. Operating performance has been enhanced through improvements in system materials, equipment layout and component specifications. In parallel with these priorities, design adaptations have been applied so as to increase safety margins and defence-in-depth, again adding to the confidence in ACR licensability. The ACR development plan includes early review by regulators to reduce licensing risk, with international regulatory review having commenced. Overall, this places the ACR in a good position to meet the first-of-a-kind challenge, a necessary condition to enabling the success of an innovative reactor. AECL sees a logical evolution from the ACR, via increasing temperature and pressure capability, to the SCWR (Supercritical Water Reactor). AECL's CANDU-X program is already looking at designs for this concept. Inherent features of both ACR and the fuel channel SCWR lend themselves to different fuel cycles for the future. One of the prominent characteristics of the heavy-water moderated fuel channel reactor approach is the high potential for innovation. The evolutionary path allows innovation in practical steps

  20. 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.

  1. 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)

  2. Development of Thermal Margin Method using Double Hot Channel

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyuk; Seo, K. W.; Kim, S. J.; Cha, J. H.; Kim, T. W.; Hwang, D. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    FAST code was developed to calculate the DNBR value and thermal margin in the reactor core of SMART. In the thermal margin analysis of whole core of SMART using FAST, 4-channel lumping model is applied to evaluate the minimum DNBR in the hot channel. There are two kinds of hot channel in SMART which is typical and thimble type. These hot channels have a similar hydraulic diameter in contrast with 16 by 16 type hot channel(OPR-1000 reactor) which is the thimble channel with a large guide tube. In this reason, a separate calculation according to the hot channel type assembly should be required on thermal margin analysis of SMART fuel. To reduce the time consuming process to look for the minimum DNBR value for a typical channel or a thimble channel, double channel method is suggested. The method is available to simultaneously calculate the minimum DNBR for two hot channels. As a result, single stage thermal margin calculation is available. Present study shows the double channel method and comparison results of MATRA-S thermal margin analysis

  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. 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.

  5. Research reactors - an overview

    Energy Technology Data Exchange (ETDEWEB)

    West, C.D.

    1997-03-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

  6. 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.)

  7. Digital instrumentation for nuclear research reactors

    International Nuclear Information System (INIS)

    The problem in measuring the neutron flux in a nuclear reactor start-up pulse channel is related to the statistical fluctuation in a wide measuring range, from a few cps up to 106 cps. This article presents the development of a star up pulse channel with digital filtering of the neutron flux and its rate variation to avoid the statistical fluctuation and obtain stable readings with a response time of 50 ms. Results are shown compared with the traditional analog instrumentation of Argonauta reactor. (author (author)

  8. 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

  9. EL-2 reactor: Thermal neutron flux distribution

    International Nuclear Information System (INIS)

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  10. 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

  11. Reactor Neutrinos

    CERN Document Server

    Lasserre, T; Lasserre, Thierry; Sobel, Henry W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrino oscillation physics in the last years. It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the last undetermined neutrino mixing angle theta13. We conclude by opening on possible use of neutrinos for Society: NonProliferation of Nuclear materials and Geophysics.

  12. 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.)

  13. Nuclear reactors

    International Nuclear Information System (INIS)

    A nuclear reactor has a large prompt negative temperature coefficient of reactivity. A reactor core assembly of a plurality of fluid-tight fuel elements is located within a water-filled tank. Each fuel element contains a solid homogeneous mixture of 50-79 w/o zirconium hydride, 20-50 w/o uranium and 0.5-1.5 W erbium. The uranium is not more than 20 percent enriched, and the ratio of hydrogen atoms to zirconium atoms is between 1.5:1 and 7:1. The core has a long lifetime, E.G., at least about 1200 days

  14. Nuclear reactors

    International Nuclear Information System (INIS)

    In a liquid cooled nuclear reactor, the combination is described for a single-walled vessel containing liquid coolant in which the reactor core is submerged, and a containment structure, primarily of material for shielding against radioactivity, surrounding at least the liquid-containing part of the vessel with clearance therebetween and having that surface thereof which faces the vessel make compatible with the liquid, thereby providing a leak jacket for the vessel. The structure is preferably a metal-lined concrete vault, and cooling means are provided for protecting the concrete against reaching a temperature at which damage would occur. (U.S.)

  15. Pressure Transducer Calibration Flow Speed, Temperature and Water Level on Reactor Protection Instrumentation System

    International Nuclear Information System (INIS)

    This experiment (on RSG-GAS) has calibrated a part of the transducers of reactor protection's measurement channels. The calibration of the transducers is a special program of the RSG-GAS maintenance program. The measurement channels transducers are the transducers of pressure measurement channel, the temperature measurement channel, the flow measurement channel and the water level measurement channel. The calibrations have used the special tools of the pressure and flow test, temperature test and water level test. These calibrations have re adjusted and re standardized all of these mentioned transducers. These work has brought the performance of the 25% of the transducers in reactor protection system back to their base

  16. 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 .

  17. 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

  18. Thorium fuel-cycle studies for CANDU reactors

    International Nuclear Information System (INIS)

    The high neutron economy of the CANDU reactor, its ability to be refuelled while operating at full power, its fuel channel design, and its simple fuel bundle provide an evolutionary path for allowing full exploitation of the energy potential of thorium fuel cycles in existing reactors. AECL has done considerable work on many aspects of thorium fuel cycles, including fuel-cycle analysis, reactor physics measurements and analysis, fuel fabrication, irradiation and PIE studies, and waste management studies. Use of the thorium fuel cycle in CANDU reactors ensures long-term supplies of nuclear fuel, using a proven, reliable reactor technology. (author)

  19. Nuclear reactor fuel rod spacer

    International Nuclear Information System (INIS)

    A spacer for positioning at least the four corner fuel rods in a tubular flow channel of a nuclear reactor is disclosed. The spacer comprises a support member having four side bands interconnected by four corner bands to form a unitary structure. Each of the side bands has a L-shaped lobe adjacent to each of its ends with one leg of each lobe extending to the adjacent end of its side band. Each of the corner bands is narrower than the side bands and is offset so as to be spaced from the lobe. One leg of each lobe is positioned to engage the tubular flow channel to maintain proper spacing between the flow channel and the adjacent corner fuel rod and to improve the thermal-hydraulic performance of the spacer

  20. 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.

  1. 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...

  2. Experimental possibilities of IGR reactor for the researches on the nuclear reactor safety

    International Nuclear Information System (INIS)

    The IGR reactor (National nuclear centre of the Republic of Kazakstan, Kurchatov) with high technical and neutron-physical properties has wide experimental possibilities for the dynamic studies. On this reactor possible curried out two general types of regimes. First regime is a 'flare', non-regular neutron impulse of power of bell form. In this regime the maximum flux density of thermal neutrons. Second regime is a 'impulse', regulated on the given regime (law) power impulse. Profile of power change in this regime has sections of linear ascent and fall, sections of stationary power. IGR reactor has pneumatic hydraulic stand, provided accumulation in the ramps of high pressure. Experimental volume of the reactor are composites central and lateral channels, which are passed through active zone of height equal to 1400 mm. The above mentioned possibilities of IGR reactor are provided unique conditions for studies in the field of nuclear reactor safety

  3. Heat removal by natural convection in a RPR reactor

    International Nuclear Information System (INIS)

    In this paper natural convection in RPR reactor is analysed. The effect of natural convection valves size on cladding temperature is studied. The reactor channel heat transfer problem is solved using finite elements in a two-dimensional analysis. Results show that two valves with Φ = 0.16 m are suited to keep coolant and cladding temperatures below 730C. (author)

  4. 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

  5. Flow and cooling in narrow, vertical rectangular channels

    International Nuclear Information System (INIS)

    Laminar flows and energy transfers in narrow, vertical rectangular channels has gained considerable attention in recent years. the cooling channels of TR-2 reactor of CNAEM research center are same and the width of channels is 2.1 mm. Natural convention cooling in these channels, in case of a loss of forced circulation cooling, as would happen in a accident, has utmost importance. A simple open loop experiment was set up in Nuclear Engineering Dep. for the simulation of TR-2 channels. The dummy fuel plates defining cooling channels were heated electrically and temperature measurements were made by thin wire thermocouples. The fluid used at the moment is air. Constant heat flux case was studied only. For the comparison purposes, simplified forms of Navie-Stokes equations for free convention cooling and incompressible flows were solved also on a variable mesh grid by relaxation technique. Flow and temperature distributions inside the channel nad some integral parameters, such as Nu number, were obtained

  6. The thermalhydraulic behaviour of a CANDU channel during a channel flow blockage accident

    International Nuclear Information System (INIS)

    The aim of this paper is to perform an analysis of the thermal-hydraulic behavior of CANDU channel during a channel flow blockage accident and to present the final results. This event assumes a flow blockage in one of the reactor fuel channels, that leads to a reduction of the flow in the affected channel, and consequently to fuel cladding and fuel temperature increase. The main conclusions of this analysis are as follows: - For the complete range of flow blockage the effective shutdown is achieved and a good cooling is maintained in the unblocked channels of both primary circuit loops; - Behaviour of the blocked channel depends on the size of the blockage; - For the blockage which are not large enough to cause coolant superheating, the channel remains intact and there are no release to the containment; - Coolant superheating occurs only if the channel flow is reduced to less than 20 % of the normal flow; - Cooling in the intact loop is maintained by pumped circulation and heat removal by steam generators; - Thermal-hydraulic results are used in containment, fuel and fuel channel analysis. (authors)

  7. 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.)

  8. Reactor building

    International Nuclear Information System (INIS)

    The present invention concerns a structure of ABWR-type reactor buildings, which can increase the capacity of a spent fuel storage area at a low cost and improved earthquake proofness. In the reactor building, the floor of a spent fuel pool is made flat, and a depth of the pool water satisfying requirement for shielding is ensured. In addition, a depth of pool water is also maintained for a equipment provisionally storing pool for storing spent fuels, and a capacity for a spent fuel storage area is increased by utilizing surplus space of the equipment provisionally storing pool. Since the flattened floor of the spent fuel pool is flushed with the floor of the equipment provisionally storing pool, transfer of horizontal loads applied to the building upon occurrence of earthquakes is made smooth, to improve earthquake proofness of the building. (T.M.)

  9. Nuclear reactors

    International Nuclear Information System (INIS)

    Disclosed is a nuclear reactor cooled by a freezable liquid has a vessel for containing said liquid and comprising a structure shaped as a container, and cooling means in the region of the surface of said structure for effecting freezing of said liquid coolant at and for a finite distance from said surface for providing a layer of frozen coolant on and supported by said surface for containing said liquid coolant. In a specific example, where the reactor is sodium-cooled, the said structure is a metal-lined concrete vault, cooling is effected by closed cooling loops containing NaK, the loops extending over the lined surface of the concrete vault with outward and reverse pipe runs of each loop separated by thermal insulation, and air is flowed through cooling pipes embedded in the concrete behind the metal lining. 7 claims, 3 figures

  10. Nuclear reactor

    International Nuclear Information System (INIS)

    The liquid metal (sodium) cooled fast breeder reactor has got fuel subassemblies which are bundled and enclosed by a common can. In order to reduce bending of the sides of the can because of the load caused by the coolant pressure the can has got a dodecagon-shaped crosssection. The surfaces of the can may be of equal width. One out of two surfaces may also be convex towards the center. (RW)

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    A detector having high sensitivity to fast neutrons and having low sensitivity to thermal neutrons is disposed for reducing influences of neutron detector signals on detection values of neutron fluxes when the upper end of control rod pass in the vicinity of the neutron flux detector. Namely, the change of the neutron fluxes is greater in the thermal neutron energy region while it is smaller in the fast neutron energy region. This is because the neutron absorbing cross section of B-10 used as neutron absorbers of control rods is greater in the thermal neutron region and it is smaller in the fast neutron region. As a result, increase of the neutron detection signals along with the local neutron flux change can be reduced, and detection signals corresponding to the reactor power can be obtained. Even when gang withdrawal of operating a plurality of control rods at the same time is performed, the reactor operation cycle can be measured accurately, thereby enabling to shorten the reactor startup time. (N.H.)

  12. 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.)

  13. 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

  14. Scalable synthesis of ionic liquids: comparison of performances of microstructured and stirred batch reactors

    OpenAIRE

    Iken, Hicham; Guillen, Frédéric; Chaumat, Hélène; Mazières, Marie-Rose; Plaquevent, Jean-Christophe; Tzedakis, Théodore

    2012-01-01

    A range of alkylpyridinium bromide ionic liquids have been synthesized in a stirred reactor at multigram scale and characterized by physical methods (viscosity, conductivity, melting point, electrochemical window, and water content). One ionic liquid, octylpyridinium bromide, was chosen to be synthesized in both macro and reduced scale reactors, in order to compare its performance and to afford evidence of the advantages of a cross channel micro reactor (channel width = 1 mm) compared to a st...

  15. Experimental estimation of the neutron flux density at the reconstructed Rossendorf research reactor

    International Nuclear Information System (INIS)

    The Rossendorf Research Reactor was reconstructed in the years 1986-1989. During start up of the reactor the neutron flux density was investigated in the reactor core and the outer irradiation channels by the help of activation probes and self-powered neutron detectors. The report includes the most important experimental results and a brief description of the measuring techniques. (orig.)

  16. Environmental radiological monitoring at Pakistan research reactor - 1 (PARR-1)

    International Nuclear Information System (INIS)

    The radiological monitoring channels of Pakistan Research Reactor-1 (PARR)1 to monitor the release of radioactive materials into the environment. This paper presents the scope of the radiological monitoring in different areas of reactor facility and describes the detection of various probable hazards and remedial action taken which generally lead to scramming the reactor. This paper also describes a new radiological monitoring channel, which is locally developed and is in use for several years for measurement of nuclear radiation in the environment. (author)

  17. Methods for studying fuel management in advanced gas cooled reactors

    International Nuclear Information System (INIS)

    The methods used for studying fuel and absorber management problems in AGRs are described. The basis of the method is the use of ARGOSY lattice data in reactor calculations performed at successive time steps. These reactor calculations may be quite crude but for advanced design calculations a detailed channel-by-channel representation of the whole core is required. The main emphasis of the paper is in describing such an advanced approach - the ODYSSEUS-6 code. This code evaluates reactor power distributions as a function of time and uses the information to select refuelling moves and determine controller positions. (author)

  18. Actions needed for RA reactor exploitation - I-IV, Part I, Thermotechnical experiments related to RA reactor hot start-up

    International Nuclear Information System (INIS)

    Heavy water coolant loop of the RA reactor includes the reactor, circulation pumps, heat exchangers and pipes. The objective of this task was measuring the thermal parameters of the RA reactor during operation. This report contains the results of the experiment, calculations of thermal regime for the outer and inner tubes, maximum temperature of the fuel element, fluid flow rate in the reactor channels, temperature of the coolant and fuel element cladding

  19. 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.

  20. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    In a BWR type nuclear reactor, the number of first fuel assemblies (uranium) loaded in a reactor core is smaller than that of second fuel assemblies (mixed oxide), the average burnup degree upon take-out of the first fuel assemblies is reduced to less than that of the second fuel assemblies, and the number of the kinds of the fuel rods constituting the first fuel assemblies is made smaller than that of the fuel rods constituting the second fuel assemblies. As a result, the variety of the plutonium enrichment degree is reduced to make the distribution of the axial enrichment degree uniform, thereby enabling to simplify the distribution of the enrichment degree. Then the number of molding fabrication steps for MOX fuel assemblies can be reduced, thereby enabling to reduce the cost for molding and fabrication. (N.H.)

  1. 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)

  2. 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

  3. Experimental study of the heat transfer in a thin vertical rectangular channels

    International Nuclear Information System (INIS)

    Free convection cooling processes are often used in nuclear technologies as well as in vertical channel type structures of some systems, in electronic circuit board cooling and many other fields. Tr-2 reactor of Cekmece Nuclear Research and Training Center (Caiman) is plate type fueled pool type research reactor. The narrow vertical cooling channels of this TR-2 reactor are identical with a width of 2.1 mm. In case of an accident of loss of cooling event, the heat transfer in this channels are supplied by natural convection. An experimental setup was constructed to simulate the TR-2 cooling channel. Dummy fuel plates were heated by direct current and temperature measurement were done by Cu-constant thermocouples in different points. Working fluid is air. At several power and channel widths the temperature has been measured. The average Nu and Ra numbers were calculated for the channel and they are compared with numerical results

  4. PSA Level 2 activities for RBMK reactors

    International Nuclear Information System (INIS)

    Probabilistic safety analyses (PSAs) of the boiling water graphite moderated pressure tube reactors (RBMKs) have been developed only recently and they are limited to Level 1. Activities at the IAEA were first motivated because of the difficulties to characterize core damage for RBMK reactors. Core damage probability is used in documents of the IAEA as a convenient single valued measure, for example for probabilistic safety criteria. The limited number of PSAs that have been completed for the RBMK reactors have shown that several special features of these channel type reactors necessitate revisiting of the characterization of core damage for these reactors. Furthermore, it has become increasingly evident that detailed deterministic analysis of DBAs and beyond design basis accidents reveal considerable insights into RBMK response to various accident conditions. These analyses can also help in better characterizing the outstanding phenomenological uncertainties, improved EOPs and AM strategies, including potential risk-beneficial accident negative backfits. The deterministic efforts should be focused first on elucidating accident progression processes and phenomena, and second on finding, qualifying and implementing procedures to minimize the risk of severe accident states The IAEA PSA procedures were mainly developed in New of vessel type LWRs, and would therefore require extensions to make them directly applicable. to channel type reactors. (author) (author)

  5. Types of Nuclear Reactors

    International Nuclear Information System (INIS)

    The presentation is based on the following areas: Types of Nuclear Reactors, coolant, moderator, neutron spectrum, fuel type, pressurized water reactor (PWR), boiling water reactor (BWR) reactor pressurized heavy water (PHWR), gas-cooled reactor, RBMK , Nuclear Electricity Generation,Challenges in Nuclear Technology Deployment,EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR.

  6. 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.

  7. 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

  8. 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.

  9. 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.

  10. 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.)

  11. 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

  12. Reactor calculations for improving utilization of TRIGA reactor

    International Nuclear Information System (INIS)

    A brief review of our work on reactor calculations of 250 kW TRIGA with mixed core (standard + FLIP fuel) will be presented. The following aspects will be treated: - development of computer programs; - optimization of in-core fuel management with respect to fuel costs and irradiation channels utilization. TRIGAP programme package will be presented as an example of computer programs. It is based on 2-group 1-D diffusion approximation and besides calculations offers possibilities for operational data logging and fuel inventory book-keeping as well. It is developed primarily for the research reactor operators as a tool for analysing reactor operation and fuel management. For this reason it is arranged for a small (PC) computer. Second part will be devoted to reactor physics properties of the mixed cores. Results of depletion calculations will be presented together with measured data to confirm some general guidelines for optimal mixed core fuel management. As the results are obtained using TRIGAP program package results can be also considered as an illustration and qualification for its application. (author)

  13. CANDU-PHW fuel channel replacement experience

    International Nuclear Information System (INIS)

    One of the main characteristics of the CANDU pressurized heavy water reactor is the use of pressure tubes rather than one large pressure vessel to contain the fuel and coolant. This provides an inherent design capability to permit their replacement in an expeditious manner, without seriously affecting the high capacity factors of the reactor units. Of th eight Ontario Hydro commercial nuclear generating units, the lifetime performance places seven of them (including two that have had some of their fuel channels replaced), in the top ten positions in the world's large nuclear-electric unit performance ranking. Pressure tube cracks in the rolled joint region have resulted in 70 fuel channels being replaced in three reactor units, the latest being at the Bruce Nuclear Generating Station 'A', Unit 2 in February 1982. The rolled joint design and rolling procedures have been modified to eliminate this problem on CANDU units subsequent to Bruce 'A'. This paper describes the CANDU pressure tube performance history and expectations, and the tooling and procedures used to carry out the fuel channel replacement

  14. 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.

  15. 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

  16. 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)

  17. Natural and mixed convections in two parallel channels

    International Nuclear Information System (INIS)

    Reactor core reverse flow experiments were made to observe heat transfer and fluid characteristics of forced cooling failure accident of HTGR reactor core. Two channels (heated and cooled) were used in this experiment to simulate HTGR multi-channel core. Wall temperature, flow rate and temperature distribution in fluid were measured for natural and mixed convections. The results show that natural convection flow rate is proportional to Grashof number and that hysteresis between wall temperature and flow rate is observed for mixed convection. (author)

  18. 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.

  19. 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.

  20. 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 ...

  1. 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...

  2. Power control of water reactors using nitrogen 16 activity measurements

    International Nuclear Information System (INIS)

    At the Grenoble Nuclear Research Centre, the open-core swimming pool reactors Melusine (2 MW) and Siloe (15 MW) are controlled at a constant overall power using nitrogen-16 channels. The conventional linear control channels react instantaneously to the rapid power fluctuations, this being necessary for the safety of the reactors, but their power indications are erroneous since they are affected by local deformations of the thermal flux caused by the compensation movements of the control rods. The nitrogen-16 channels on the other hand give an indication of the overall power proportional to the mean fission flux and independent of the rod movements, but their response time is 15 seconds, A constant overall power control is thus possible by a slow correction of the reference signal given by the automatic control governed by thu linear channels by means of a correction term given by the 'N-16' channels: This is done automatically in Melusine and manually in Siloe. (authors)

  3. 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...

  4. Quantum Channels With Memory

    International Nuclear Information System (INIS)

    Quantum memory channels represent a very general, yet simple and comprehensible model for causal processes. As such they have attracted considerable research interest, mostly aimed on their transfer capabilities and structure properties. Most notably it was shown that memory channels can be implemented via physically naturally motivated collision models. We also define the concept of repeatable channels and show that only unital channels can be implemented repeat ably with pure memory channels. In the special case of qubit channels we also show that every unital qubit channel has a repeatable implementation. We also briefly explore the possibilities of stroboscopical simulation of channels and show that all random unitary channels can be stroboscopically simulated. Particularly in qubit case, all indivisible qubit channels are also random unitary, hence for qubit all indivisible channels can be stroboscopically simulated. Memory channels also naturally capture the framework of correlated experiments. We develop methods to gather and interpret data obtained in such setting and in detail examine the two qubit case. We also show that for control unitary interactions the measured data will never contradict a simple unitary evolution. Thus no memory effects can be spotted then. (author)

  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. 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.

  7. 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.)

  8. Experimental study of free flow and heat transfer in narrow vertical rectangular channels

    International Nuclear Information System (INIS)

    Free convection cooling of electronic boards and vertical channel type structures of systems in the facilities of some installations, in the abnormal working situations, is becoming important recent years. Plate type fueled pool type research reactors, such TR-2 reactor of CHNAEM have very narrow (2.1 mm width) vertical cooling channels. In case of an accident or a loss of cooling event, the heat transfer in this channels should be accomplished by natural circulation. The adequacy of this cooling should be proved for safety consideration. A simple experimental setup was constructed to simulate a TR-2 cooling channel. Plate powers and channel widths were the parameters changed. Dummy fuel plates were heated by DC current and temperature measurements were done by Cu-Constantan thermocouples located inside the channel and around it. Average Nu and Ra numbers were calculated for the channel and they are compared with the numerical results. Working fluid was air

  9. Design study on small CANDLE reactor

    International Nuclear Information System (INIS)

    A new reactor burnup strategy CANDLE was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. Here important points are that the solid fuel is fixed at each position and that any movable burnup reactivity control mechanisms such as control rods are not required. This burnup strategy can derive many merits. The change of excess reactivity along burnup is theoretically zero, and shim rods will not be required for this reactor. The reactor becomes free from accidents induced by unexpected control rods withdrawal. The core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Therefore, the operation of the reactor becomes much easier than the conventional reactors especially for high burnup reactors. The transportation and storage of replacing fuels become easy and safe, since they are free from criticality accidents. In our previous works it is appeared that application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. The average burnup of the spent fuel is about 40% that is equivalent to 40% utilization of the natural uranium without the reprocessing and enrichment. This reactor can be realized for large reactor, since the neutron leakage becomes small and its neutron economy becomes improved. In the present paper we try to design small CANDLE reactor whose performance is similar to the large reactor by increasing its fuel volume ration of the core, since its performance is strongly required for local area usage. Small long life reactor is required for some local areas. Such a characteristic that only natural uranium is required after second core is also strong merit for this case. The core with 1.0 m radius, 2.0 m length can realize CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead-Bismuth is

  10. 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)

  11. Capacities of Grassmann channels

    CERN Document Server

    Bradler, Kamil; Jauregui, Rocio

    2010-01-01

    A new class of quantum channels called Grassmann channels is introduced and their classical and quantum capacity is calculated. The channel class appears in a study of the two-mode squeezing operator constructed from operators satisfying the fermionic algebra. We compare Grassmann channels with the channels induced by the bosonic two-mode squeezing operator. Among other results, we challenge the relevance of calculating entanglement measures to assess or compare the ability of bosonic and fermionic states to send quantum information to uniformly accelerated frames.

  12. Reactor material behaviour problems when supporting the safety of RBMK reactor operation

    International Nuclear Information System (INIS)

    A number of materials science problems is defined which solution is aimed to improve the level of safety operation for RBMK type reactors. The problems are as follows: the substantiation of applicability of a leak-before-break criterion to pipelines of forced circulation coolant circuits; intergranular stress corrosion cracking in steel 08Kh18N10T pipeline welded joints; fuel channel (alloy Eh125) life extension in view of their step-by-step replacing because of channel-graphite gap narrowing; the prediction of graphite blocks integrity with accounting for the relationship between neutron and gamma radiations in a RBMK type reactor; the use of erbium-containing fuels. The measures undertaken at NPPs with RBMK type reactors to solve above-mentioned problems are described

  13. Nuclear Reactor RA Safety Report, Vol. 12, Accidents during reactor operation

    International Nuclear Information System (INIS)

    This volume includes description and analysis of typical accidents occurred during operation of RA reactor in chronological order, as follows: contamination of primary coolant circuit; leakage of heavy water from the primary coolant loop; contamination of vertical experimental channel; air contamination in the reactor building and loss of circulation of the primary coolant; failures of the vacuum pump and spent fuel packaging device; rupture of the spent fuel element cladding; dethronement's of capsule for irradiation of fuel element; rupture of the vertical experimental channel and contamination of the surroundings; swelling of a fuel element; appearance of deposits on the surface of the fuel elements cladding. The last chapter describes similar accidents occurred on nuclear reactors in the world

  14. 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...

  15. Measurements at the RA Reactor related to the VISA-2 project - Part 1, Start-up of the RA reactor and measurement of new RA reactor core parameters

    International Nuclear Information System (INIS)

    The objective of the measurements was determining the neutron flux in the RA reactor core. Since the number of fuel channels is increased from 56 to 68 within the VISA-2 project, it was necessary to attain criticality of the RA reactor and measure the neutron flux properties. The 'program of RA reactor start-up' has been prepared separately and it is included in this report. Measurements were divided in two phases. First phase was measuring of the neutron flux after the criticality was achieved but at zero power. During phase two measurements were repeated at several power levels, at equilibrium xenon poisoning. This report includes experimental data of flux distributions and absolute values of the thermal and fast neutron flux in the RA reactor experimental channels and values of cadmium ratio for determining the neutron epithermal flux. Data related to calibration of regulatory rods for cold un poisoned core are included

  16. Operation and maintenance of the RA reactor in 1964, I-II, Part II

    International Nuclear Information System (INIS)

    This volume of the report contains the following 15 Annexes: Improvement of the fuel cycle economy (record No. 37009803 in INIS DB); Analysis of neutron flux increase in horizontal experimental channels of the RA reactor record No. 37005698 in INIS DB); Application of the critical system for determining the thermal neutron flux in a research reactor with central horizontal reflector ( record No. 37055005 in INIS DB); Determining the capacity of the RA reactor heat exchanger dependent on the coolant water temperature and flow; Operation of the RA reactor in forced regime; Analysis of the CEN-132 heavy water pumps failures at the RA reactor from decontamination till present; Modifications in the vacuum loop of the distillation system; Report on decontamination of the evaporator and cleaning of the condenser of the distillation system; Operation of reactor at nominal power with reduced D2O circulation; Cooling of the RA reactor with reduced flow rate in the heavy water loop; Measurement of the heavy water level in the fuel channels of the RA reactor; Conclusions of the experts group of the RA reactor at the meeting held on November 2 and 3 1964; Conclusions of the experts group at the meeting held on November 23 1964; After heat and the cooling problem after RA reactor shut-down; Measurement of noise and vibrations on the Ra reactor heavy water system; Calculation and measurement of the uranium temperature during irradiation in the experimental channel in the reflector of the RA reactor; Temperature measurement of the reactor materials samples irradiated in the fuel channels of the RA reactor; Study of the modifications in the synchronous generators, heavy water pumps and condenser batteries of the RA reactor

  17. Graphite reactor physics

    International Nuclear Information System (INIS)

    The study of graphite-natural uranium power reactor physics, undertaken ten years ago when the Marcoule piles were built, has continued to keep in step with the development of this type of pile. From 1960 onwards the critical facility Marius has been available for a systematic study of the properties of lattices as a function of their pitch, of fuel geometry and of the diameter of cooling channels. This study has covered a very wide field: lattice pitch varying from 19 to 38 cm. uranium rods and tubes of cross-sections from 6 to 35 cm2, channels with diameters between 70 and 140 mm. The lattice calculation methods could thus be checked and where necessary adapted. The running of the Marcoule piles and the experiments carried out on them during the last few years have supplied valuable information on the overall evolution of the neutronic properties of the fuel as a function of irradiation. More detailed experiments have also been performed in Marius with plutonium-containing fuels (irradiated or synthetic fuels), and will be undertaken at the beginning of 1965 at high temperature in the critical facility Cesar, which is just being completed at Cadarache. Spent fuel analyses complement these results and help in their interpretation. The thermalization and spectra theories developed in France can thus be verified over the whole valid temperature range. The efficiency of control rods as a function of their dimensions, the materials of which they are made and the lattices surrounding them has been measured in Marius, and the results compared with calculation on the one hand and with the measurements carried out in EDF 1 on the other. Studies on the control proper of graphite piles were concerned essentially with the risks of spatial instability and the means of detecting and controlling them, and with flux distortions caused by the control rods. (authors)

  18. 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.

  19. Survey of research reactors

    International Nuclear Information System (INIS)

    A survey of reasearch reactors based on the IAEA Nuclear Research Reactor Data Base (RRDB) was done. This database includes information on 273 operating research reactors ranging in power from zero to several hundred MW. From these 273 operating research reactors 205 reactors have a power level below 5 MW, the remaining 68 reactors range from 5 MW up to several 100 MW thermal power. The major reactor types with common design are: Siemens Unterrichtsreaktors, 1.2 Argonaut reactors, Slowpoke reactors, the miniature neutron source reactors, TRIGA reactors, material testing reactors and high flux reactors. Technical data such as: power, fuel material, fuel type, enrichment, maximum neutron flux density and experimental facilities for each reactor type as well as a description of their utilization in physics and chemistry, medicine and biology, academic research and teaching, training purposes (students and physicists, operating personnel), industrial application (neutron radiography, silicon neutron transmutation doping facilities) are provided. The geographically distribution of these reactors is also shown. As conclusions the author discussed the advantages (low capital cost, low operating cost, low burn up, simple to operate, safe, less restrictive containment and sitting requirements, versatility) and disadvantages (lower sensitivity for NAA, limited radioisotope production, limited use of neutron beams, limited access to the core, licensing) of low power research reactors. 24 figs., refs. 15, Tab. 1 (nevyjel)

  20. Department of reactor technology

    International Nuclear Information System (INIS)

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  1. RB reactor noise analysis

    International Nuclear Information System (INIS)

    Statistical fluctuations of reactivity represent reactor noise. Analysis of reactor noise enables determining a series of reactor kinetic parameters. Fluctuations of power was measured by ionization chamber placed next to the tank of the RB reactor. The signal was digitized by an analog-digital converter. After calculation of the mean power, 3000 data obtained by sampling were analysed

  2. Comparison between new thermohydraulic one-channel models and experiments

    Science.gov (United States)

    Blender, H.; Elzmann, J.

    1981-11-01

    Five different thermohydraulic one-channel models, COCHA, FRANCESCA, MARMITA, STASWR and THS, were tested bu experimentally checking two-phase flows along a boiling water reactor fuel element. As regards the evolution of the vapor content along the cooling channel, the agreement between all the programs and the measurements is satisfactory for small to middle entrance undercooling in the domain of undercooled boiling. For high undercooling, only the COCHA program gives satisfactory results. For the middle part of the cooling channel, all programs are satisfactory, while in the upper part, especially for increasing outlet vapor contents, the calculated values are generally too low for all programs, and especially for FRANCESCA.

  3. Method of testing fuel assemblies for nuclear reactors

    International Nuclear Information System (INIS)

    The stresses occurring in the fuel assemblies are simulated by power excursions. For this purpose the fuel assembly is placed in the neutron field of a test reactor and for a short time can be exposed to the much higher neutron field of a pulsed reactor. One possibility of design provides for the test and the pulsed reactor lying one above the other, separated by a neutron absorber and penetrated by a common irradiation channel. The fuel assembly then is to be moved from the position in the test reactor to the position in the pulsed reactor. The other possibility is to make the irradiation duct pass along the gap between both reactors and, by means of a tube-shaped absorber, open one or the other irradiation field. (DG)

  4. 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)

  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. 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.

  7. Investigation of serviceability of rod carbide fuel pins on energy mode of high power of NEMF reactor

    International Nuclear Information System (INIS)

    There were conducted post reactor material and science investigations of NRM (Nuclear Rocket Motor) standard fuel pins. There were used in complex test in ETC (Experimental Technological Channel) flowing technological channel of IV reactor on moving and energy modes of NEMF (Nuclear Energy Moving Facility) reactor. By comparing of condition of fuel pins which were tested in the course of different number of start-ups of IV-1 reactor on energy mode of high power (MHP) of NEMF reactor (number of start-ups on moving mode and on energy mode of low power is equal) there was determined the main factors of fuel pins serviceability on MHP. (author)

  8. Measurements at the RA Reactor related to the VISA-2 project - Part 4, Calorimetry and chemical dosimetry of the new partly filled RA reactor core

    International Nuclear Information System (INIS)

    This report contains the results of values of chemical and calorimetry measurements of absorbed doses in the experimental channels VK-5, VK-9, GF-34 and fuel channels (0706 and 0607) of the Ra reactor. Calorimetry measurements were during reactor operation at 900 kW or 1 MW power, dependent on the type of samples in the calorimeters. For the chemical measurements the power was kept at 500 kW

  9. Small sized reactor for laser radiation

    International Nuclear Information System (INIS)

    Substantiation of the possibility of the creation of an autonomous nuclear power plant generating laser radiation is given in this paper. The work of the power plant is based on the use of the small sized reactor, generating electric energy, with non-self maintained discharge lasers built-in reactor core or reflector, having high flux density of a thermal neutron ΦTh≥1,0.1013 cm-2.s-1 in the channels, where the lasers are located. A thermionic fast reactor-converter with beryllium reflector of the space nuclear-energetic installation was chosen for consideration by us, because the neutron radiation and electric energy necessary for operating of nonself maintained discharge lasers should be produced in one nuclear reactor. To prove the possibility of operation of the non-self maintained discharge lasers in the reactor and estimate parameters of the laser systems in the reactor, we used the results of experimental and computing researches of the neutron characteristics of non-self maintained discharge lasers built-in the beryllium reflector. These experiments were carried out by us on the critical assembly PhS-1, simulating that thermionic fast reactor-converter. Moreover, we used the research results of the in-reactor diagnostics of the nuclear-excited plasma of the laser gas mixtures, data of experimental characteristics of non-self maintained discharge, threshold and output data of lasers of different waves lengths, carried out by us earlier on the Kazakhstan Research Nuclear Reactor WWR-K. Thus the possibility of achieving an autonomous, compact nuclear power plant generating not only electrical energy, but also laser radiation concerning the large capacities of infra-red and ultra-violet range of waves lengths in stationary and pulse modes is shown. (author)

  10. Operational reactor physics analysis codes (ORPAC)

    International Nuclear Information System (INIS)

    For efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are regularly required. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation of samples requires a prior estimation of the reactivity load due to the sample, the heating rate and the activity developed in it during irradiation. For the safety of personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr.Therefore, a proper shielding and radioactive cooling of the irradiated sample are required to meet the said requirement. Knowledge of xenon load variation with time (Startup-curve) helps in estimating Xenon override time. Monitoring of power in individual fuel channels during reactor operation is essential to know any abnormal power distribution to avoid unsafe situations. Complexities in the estimation of above mentioned reactor parameters and their frequent requirement compel one to use computer codes to avoid possible human errors. For efficient and quick evaluation of parameters related to reactor operations such as xenon load, critical moderator height and nuclear heating and reactivity load of isotope samples/experimental assembly, a computer code ORPAC (Operational Reactor Physics Analysis Codes) has been developed. This code is being used for regular assessment of reactor physics parameters in Dhruva and Cirus. The code ORPAC written in Visual Basic 6.0 environment incorporates several important operational reactor physics aspects on a single platform with graphical user interfaces (GUI) to make it more user-friendly and presentable. (author)

  11. 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)

  12. Remotely operated inspection equipment for the Candu fuel channels

    International Nuclear Information System (INIS)

    Equipment is described which has been successfully used for the nondestructive inspection of fuel channel components within Ontario Hydro's CANDU nuclear reactors. By the use of automated systems, significant savings in personnel radiation exposure and unit outage duration have been realized, with improved quality and quantity of nondestructive examination information. (author)

  13. 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

  14. Upgrade of Dhruva fuel channel flow instrumentation

    International Nuclear Information System (INIS)

    Dhruva, a 100 MW Heavy Water moderated and cooled, vertical tank-type Research Reactor, using metallic natural Uranium fuel has flow instrumentation for all the 144 fuel channels, consisting of venturi and triplicate DP gauges for each fuel channel. These gauges provide contacts for generation of reactor trip on low flow through fuel channel. These DP gauges were facing numerous generic and ageing related failures over the years and was also difficult to maintain owing to obsolescence. While considering an upgrade for these DP gauges, it was also planned to replace the existing Coolant Low Flow Trip (CLFT) system with a computer based Reactor Trip Logic System (RTLS). Being a retrofit job, the existing panels for mounting the gauges, cable layout, impulse tubing layout, etc. were retained, thereby simplifying the site execution, reducing reactor down time and also reducing person-milli-Sievert consumption. A customized Electronic DP Indicating Switch (EDPIS) was conceptualized for achieving these objectives. Such a design, utilizing a standard DP transmitter with customized electronic circuitry, was developed, evaluated and finalized after a series of factory trials, field trials and prototyping. The instrument design included contact input for existing CLFT system and also provision for 4-20 mA current output for the proposed computer based RTLS. The display and form factor of the instrument remained identical to older one and ensures familiarity of O and M personnel. Since EDPIS is classified as Safety Class IA, stringent type tests, hardware FMEA and V and V of the micro-controller software were carried out as per the requirements laid down by relevant standards for qualification of these instruments. Being a customized instrument, the manufacturing process was closely monitored and was followed by stringent QA plan and acceptance tests. A total of 396 gauges were replaced in a phased manner during scheduled fuelling outages and thereby did not affect reactor

  15. 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)

  16. 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...

  17. Computer code for nuclear reactor core thermal reliability calculation

    International Nuclear Information System (INIS)

    RASTENAR program was described for computing heat-engineering reliability of cores in nuclear reactors operating under stationary conditions. The following factors of heat-engineering reliability were found to be computable: rated critical margin; limiting critical margin; probability of initiation of critical heat removal in channel (inferior conditions of heat transfer); probability that no channel would be subject to critical heat removal; and reactor power reserve coefficient. The probability that no channel in the core would experience critical heat removal when boiling during operation of the reactor at fixed power level was taken for the principal quantitative criterion. The structure and limitations of the program were described together with the computation algorithm. The program was written for an M-220 computer

  18. Advanced heavy water reactor pressure tube-easy replaceability

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor (AHWR) is a 300 MWe vertical pressure tube type reactor. A coolant channel consists of pressure tube, made of Zr-2.5 % Nb, which is separated from cold calandria tube using garter spring spacers. The principal function of pressure tube is to support and locate the fuel assembly and allows light water coolant through fuel assembly by natural circulation. Since AHWR is designed for life of 100 years, it necessitates the replacement of pressure tubes during service life. Easy replaceability of pressure tube, along with surveillance requirements, has major bearing on the design of coolant channel assembly. The several systems and tools have been conceptualised to cater the needs for easy and quick replacement of a pressure tube during reactor shut down. This paper gives the highlights of the innovative design features of coolant channel, preliminary design and pre-requisites for replacement, and experimental programme for demonstration of easy replaceability. (author)

  19. 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...

  20. Investigation on Nodalization for Analysis of SFR Channel Blockage Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Won Pyo; Kwon, Young Min; Ha, Ki Suk; Lee, Kwi Lim; Jeong, Hae Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    The present paper demonstrates nodalization analysis results obtained in application of the MATRA-LMR/FB to channel blockage accidents for a SFR (Sodium cooled Fast Reactor), KALIMER-150. In the earlier study, a uniform node size over the total sub-channel length in a subassembly was used. The study was carried out not only for the radially different positions, i.e. central, medium between the center and the duct wall, and edge sub-channels in the assembly, but also for larger blockage sizes larger than 6 sub-channels, the blockage size of which was classified into a DBE(Design Basis Event) in the KALIMER-150 design. The present investigation focuses mainly on the identification of conservatism as well as consistency in the analyses of the maximum coolant temperature for the 6 sub-channels blockage accidents

  1. Open-channel hydraulics

    International Nuclear Information System (INIS)

    This text discusses the following: concepts of fluid flow; the momentum principle; computation of uniform flow; design of channels; and turbulent diffusion and dispersion in steady open-channel flow. Emphasis is concerned only with the flow of water in channels where the water is not transporting significant quantities of air or sediment. The text contains quite a few examples demonstrating the application of the presented principles

  2. Bank Liabilities Channel

    OpenAIRE

    Quadrini, Vincenzo

    2014-01-01

    The financial intermediation sector is important not only for channeling resources from agents in excess of funds to agents in need of funds (lending channel). By issuing liabilities it also creates financial assets held by other sectors of the economy for insurance purpose. When the intermediation sector creates less liabilities or their value falls, agents are less willing to engage in activities that are individually risky but desirable in aggregate (bank liabilities channel). The paper st...

  3. Bank Liabilities Channel

    OpenAIRE

    Vincenzo Quadrini

    2015-01-01

    The financial intermediation sector is important not only for channeling resources from agents in excess of funds to agents in need of funds (lending channel). By issuing liabilities it also creates financial assets held by other sectors of the economy for insurance purpose. When the intermediation sector creates less liabilities or their value falls, agents are less willing to engage in activities that are individually risky but desirable in aggregate (bank liabilities channel). The paper st...

  4. 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

  5. N Reactor Lessons Learned workshop

    International Nuclear Information System (INIS)

    This report describes a workshop designed to introduce participants to a process, or model, for adapting LWR Safety Standards and Analysis Methods for use on rector designs significantly different than LWR. The focus of the workshop is on the ''Lessons Learned'' from the multi-year experience in the operation of N Reactor and the efforts to adapt the safety standards developed for commercial light water reactors to a graphite moderated, water cooled, channel type reactor. It must be recognized that the objective of the workshop is to introduce the participants to the operation of a non-LWR in a LWR regulatory world. The total scope of this topic would take weeks to provide a through overview. The objective of this workshop is to provide an introduction and hopefully establish a means to develop a longer term dialogue for technical exchange. This report provides outline of the workshop, a proposed schedule of the workshop, and a description of the tasks will be required to achieve successful completion of the project

  6. Performance of Thorium ORGEL Reactor

    International Nuclear Information System (INIS)

    The aim of this paper is to show the general characteristics of a thorium ORGEL reactor. Reference is deliberately made to existing technology or techniques that are easily developable, avoiding any advanced solutions; in particular, the general design of the reactor is entirely similar to one operating on uranium, and includes an organic coolant, a gas-insulation and SAP pressure tube channel, and a continuous fuel cycle with one zone and one type of fuel. The criteria governing the choice of fuel element are reviewed, and an element consisting of a cluster of 37 SAP-clad thorium oxide rods is chosen. The technical and economic performance of the fuel cycle - investment in fissionable material, specific consumption, cost of the cycle - is analysed as a function of various parameters such as lattice pitch, power density and burn-up. For every case considered the uranium is entirely recycled; the supplementary fissionable material is uranium-235. Finally, the performance is compared with that obtained in a uranium reactor, and an analysis is made of the economic conditions under which the thorium cycle would become more advantageous than the uranium. (author)

  7. Russian RBMK reactor design information

    International Nuclear Information System (INIS)

    This document concerns the systems, design, and operations of the graphite-moderated, boiling, water-cooled, channel-type (RBMK) reactors located in the former Soviet Union (FSU). The Russian Academy of Sciences Nuclear Safety Institute (NSI) in Moscow, Russia, researched specific technical questions that were formulated by the Pacific Northwest Laboratory (PNL) and provided detailed technical answers to those questions. The Russian response was prepared in English by NSI in a question-and-answer format. This report presents the results of that technical exchange in the context they were received from the NSI organization. Pacific Northwest Laboratory is generating this document to support the US Department of Energy (DOE) community in responding to requests from FSU states, which are seeking Western technological and financial assistance to improve the safety systems of the Russian-designed reactors. This report expands upon information that was previously available to the United States through bilateral information exchanges, international nuclear society meetings, International Atomic Energy Agency (IAEA) reactor safety programs, and Research and Development Institute of Power Engineering (RDIPE) reports. The response to the PNL questions have not been edited or reviewed for technical consistency or accuracy by PNL staff or other US organizations, but are provided for use by the DOE community in the form they were received

  8. 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

  9. Introduction of Nuclear Reactor Engineering

    International Nuclear Information System (INIS)

    This book introduces development, status, supply and demand and resource of nuclear reactor. It deals with basic knowledge of nuclear reactor, which are reactor system, heat recovery in reactor core, structural feature in reactor, materials of structure in reactor, shielding of gamma ray, shielding of reactor, safety and environmental problem of nuclear power plant, nuclear fuel and economical efficiency of nuclear energy.

  10. Symmetrization for redundant channels

    Science.gov (United States)

    Tulplue, Bhalchandra R. (Inventor); Collins, Robert E. (Inventor)

    1988-01-01

    A plurality of redundant channels in a system each contain a global image of all the configuration data bases in each of the channels in the system. Each global image is updated periodically from each of the other channels via cross channel data links. The global images of the local configuration data bases in each channel are separately symmetrized using a voting process to generate a system signal configuration data base which is not written into by any other routine and is available for indicating the status of the system within each channel. Equalization may be imposed on a suspect signal and a number of chances for that signal to heal itself are provided before excluding it from future votes. Reconfiguration is accomplished upon detecting a channel which is deemed invalid. A reset function is provided which permits an externally generated reset signal to permit a previously excluded channel to be reincluded within the system. The updating of global images and/or the symmetrization process may be accomplished at substantially the same time within a synchronized time frame common to all channels.

  11. Initiation of long electrical discharge in ICF reactor dense atmospheres

    International Nuclear Information System (INIS)

    There is significant interest in using light-ion beams (LIB) as drivers for inertial confinement fusion (ICF). Typical commercial target yields are on the order of 0.1 to 1 GJ, which requires the driver to be shielded adequately from the implosion-associated energy. The LIB can be guided through a gaseous shield via current carrying reduced density plasma channels surrounded with ambient gas envelop. The plasma serves as a conductor for the confinement current and for beam neutralization while the envelope counteracts inertially the core pressure. The optimum channel density for reactor-grade LIB propagation (a few torr) has triggered researchers to choose low-pressure gas doped with high-Z material to fill the reactor. Channeling simplicity and minimum pellet heating from the channels were among the motivations for this choice. Although such a shield absorbs the target soft x ray, it is incompatible with power reactor requirements. Factors like channel susceptibility to various hydrodynamic instabilities, maximum attainable LIB power density, and the reactor dimensions seem to require much higher gas pressure. An additional shielding and breeding blanket has to be built around the activated reactor vessel, adding cost and maintenance complexity to the plant design. Recently, a new concept for a commercial LIB reactor has been suggested. The concept is based on a rotating body of boiling heavy water, which serves simultaneously as the first wall, working fluid, shock wave absorber, biological shielding, and tritium breeding blanket. The LIB propagate in 0.5- to 1-m-long reduced density channels formed in the high-pressure superheated steam, which fills a vortex cavity formed at the water surface center

  12. 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.)

  13. Luncheon address: Development of the CANDU reactor

    International Nuclear Information System (INIS)

    The paper is a highlight of the some of the achievements in the development of the CANDU Reactor, taken from the book Canada Enters the Nuclear Age. The CANDU reactor is one of Canada's greatest scientific/engineering achievements, that started in the 1940's and bore fruit with the reactors of the 60's, 70's, and 80's. The Government decided in the 1950's to proceed with a demonstration nuclear power reactor (NPD), AECL invited 7 Canadian corporations to bid on a contract to design and construct the NPD plant. General Electric was selected. A utility was also essential for participation and Ontario Hydro was chosen. In May 1957 it was concluded that the minimum commercial size would be about 200MWe and it should use horizontal pressure tubes to contain the fuel and pressurized heavy water coolant. The book also talks of standard out-reactor components such as pumps, valves, steam generators and piping. A major in-reactor component of interest was the fuel, fuel channels and pressure tubes. A very high level of cooperation was required for the success of the CANDU program

  14. Thermal hydraulic analysis of nuclear research reactors

    International Nuclear Information System (INIS)

    A loss of coolant accident (LOCA) can cause total or partial core uncovery which is followed by substantial fuel element temperature increase due to fuel residual heat. It is essential to demonstrate that such a temperature increase does not lead to excessive core melting and to significant radioactive material release into the reactor building and consequently to the environment. The THEAP computer codes able to perform reliable analysis of such accidents have been developed. THEAP-I is a computer code developed with the aim to contribute to the safety analysis of the MTR open pool research reactors. THEAP-I is designed for three dimensional, transient thermal/hydraulic analysis of a thermally interacting channel bundle totally immersed into water or air, such as the reactor core. The mathematical and physical models and methods of the solution are given as well as the code description and the input data. A sample problem is included, referring to the Greek Research Reactor analysis, under a hypothetical severe loss of coolant accident. The micro computer version of the code is also described. More emphasis is given in the new features of the code (i.e. input data structure). A set of instructions for running in an IBM-AT2 computer with the microsoft FORTRAN V4.0 is included together with a sample problem referring to the Greek Research Reactor. THEAP-I can be used also for other MTR open pool research reactors. Refs and figs

  15. Shakedown analysis of fusion reactor first wall

    International Nuclear Information System (INIS)

    Shakedown analyses of a typical fusion reactor first wall including coolant channels subjected to cyclic thermal/steady primary and cyclic primary/steady thermal stresses are carried out. The stresses are assumed to be predominantly of the bending type. The first cycle of loading/unloading is analyzed using elastic-plastic beam bending theory. The general problem of shakedown is solved using the shakedown theorem of perfect plasticity

  16. 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

  17. Study on Methanol Conversion Efficiency and Mass Transfer of Steam-Methanol Reforming on Flow Rate Variation in Curved Channel

    International Nuclear Information System (INIS)

    In this study, numerical analysis of curved channel steam-methanol reformer was conducted using the computational fluid dynamics (CFD) commercial code STAR-CCM. A pre-numerical analysis of reference model with a cylindrical channel reactor was performed to validate the combustion model of the CFD commercial code. The result of advance validation was in agreement with reference model over 95%. After completing the validation, a curved channel reactor was designed to determine the effects of shape and length of flow path on methanol conversion efficiency and generation of hydrogen. Numerical analysis of the curved-channel reformer was conducted under various flow rate (10/15/20 μl/min). As a result, the characteristics of flow and mass transfer were confirmed in the cylindrical channel and curved channel reactor, and useful information about methanol conversion efficiency and hydrogen generation was obtained for various flow rate.

  18. 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.

  19. 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.

  20. Verification of SOPHT for parallel channel flow stability

    International Nuclear Information System (INIS)

    As part of a continuing program to verify the thermalhydraulic computer code SOPHT, a number of experimental tests designed to study parallel channel flow instability were simulated. This phenomenon of flow instability could cause dryout of reactor fuel resulting in poor heat transfer to the coolant and increased fuel temperatures. Selected tests from two out reactor experimental programs are predicted with the SOPHT code. Both of these test programs were carried out at Westinghouse Canada Incorporated, Hamilton, Ontario, under contract to Atomic Energy of Canada. The first test facility consisted of a vertical heated channel in parallel with an unheated bypass connected between a common inlet and outlet header. The second test facility was similar to the first, however there were three parallel heated channels as well as an unheated bypass

  1. 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)

  2. Fast neutron benchmark proposal at TRIGA-ACPR Reactor

    International Nuclear Information System (INIS)

    The development of fast neutron benchmarks is a historical aim of reactor physics. The dry experimental tube situated in the central region of the core in TRIGA Annular-Core Pulsing Reactor (ACPR) offers a suitable neutron source for fast neutron benchmark development. Our proposal consists in mounting a high-enriched uranium annular converter into the dry channel of the core. Preliminary computations and measurements are presented in this paper. Neutron flux computations in the dry channel and the uranium converter were performed using MCNP and WIMS codes. Also neutron flux spectrum measurements and fast and thermal neutron flux distribution measurements were performed using foil activation techniques. (authors)

  3. Development of activation analysis on the IBR-2 reactor

    International Nuclear Information System (INIS)

    Different examples of activation analysis (AA) application and probabilities of its further development using IBR-2 reactor (Dubna) with two facilities: REGATA pneumotransport facility desigued for instrumental AA and biophysical channel designed for element analysis using capture prompt quanta and radiography-are considered. Characteristics of irradiation channels, values of flux densities for thermal, resonance and fast neutrons are given. Application advantages concerning instrumental AA of resonance neutrons are considered. New application trend of AA for composition optimization of concretes used in shielding structures of nuclear reactors, for reduction of long-lived directed activity is pointed out. 20 refs.; 6 figs.; 7 tabs

  4. 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)

  5. Nuclear reactor building

    International Nuclear Information System (INIS)

    Purpose: To prevent seismic vibrations of external buildings from transmitting to the side walls of a reactor container in a tank type FBR reactor building. Constitution: The reactor building is structured such that the base mat for a reactor container chamber and a reactor container is separated from the base mat for the walls of building, and gas-tight material such as silicon rubber is filled in the gap therebetween. With such a constitution, even if the crane-supporting wall vibrates violently upon occurrence of earthqualkes, the seismic vibrations do not transmit toward the reactor container chamber. (Horiuchi, T.)

  6. 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

  7. 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

  8. Mechanosensitive ion channels

    Directory of Open Access Journals (Sweden)

    Ken Takahashi

    2016-01-01

    Full Text Available Cell surface receptors are involved in numerous important biological processes including embryogenesis, tissue differentiation, and cellular homeostasis. Among them, mechanosensitive ion channels play an essential role in cellular functions of every cell including neurons, cardiomyocytes, and osteocytes. Here, we discuss types, roles, structures, and biophysical factors that affect the functions of mechanosensitive ion channels.

  9. 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.

  10. Thermoemission reactor-converters for nuclear power units in outer space

    International Nuclear Information System (INIS)

    In a thermoemission reactor-converter, the direct conversion of thermal into electrical energy is based on the thermoelectric emission of electrons in electricity-generating channels. The paper describes the following: the electricity-generating channel, characteristics of the electricity-generating channels and the reactor-converter, a thermal neutron reactor-converter, a fast neutron reactor-converter, and estimating the mass-limit characteristics of the thermoemission units. Thermoemission nuclear power units with built-in generators in the nuclear reactor core can be regarded as a promising source of electric power for supplying the needs of space equipment for various purposes with a wide range of electric power demands over a long service life and with acceptable mass-limit characteristics

  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. Incompatibility breaking quantum channels

    International Nuclear Information System (INIS)

    A typical bipartite quantum protocol, such as EPR-steering, relies on two quantum features, entanglement of states and incompatibility of measurements. Noise can delete both of these quantum features. In this work we study the behavior of incompatibility under noisy quantum channels. The starting point for our investigation is the observation that compatible measurements cannot become incompatible by the action of any channel. We focus our attention to channels which completely destroy the incompatibility of various relevant sets of measurements. We call such channels incompatibility breaking, in analogy to the concept of entanglement breaking channels. This notion is relevant especially for the understanding of noise-robustness of the local measurement resources for steering. (paper)

  13. 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...... drive the late repolarization of the ventricle with some redundancy, and in atria this repolarization reserve is supplemented by the fairly atrial-specific KV1.5, Kir3, KCa, and K2P channels. The role of the latter two subtypes in atria is currently being clarified, and several findings indicate that...... they could constitute targets for new pharmacological treatment of atrial fibrillation. The interplay between the different K(+) channel subtypes in both atria and ventricle is dynamic, and a significant up- and downregulation occurs in disease states such as atrial fibrillation or heart failure. The...

  14. 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.

  15. Reactor Physics Programme

    International Nuclear Information System (INIS)

    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

  16. 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)

  17. 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.)

  18. 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.

  19. 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.

  20. Frictional drop in pressure micro-channel

    International Nuclear Information System (INIS)

    Recently, the flow in a sub-millimeter scale channel has been pointed out as an issue for safety of nuclear reactors. In this work the friction characteristics of water in a sub-millimeter scale channel were investigated experimentally. The friction factors and the critical Reynolds number were measured using water flow through circular tubes with diameters of 0.5, 0.25 and 0.17 mm. The experimental results show that the measured friction factor for water agreed well with the conventional Poiseuille (λ = 64/Re) and Blasius (λ =0.316*Re-0.25) equations in laminar and turbulent flow regime; the laminar-turbulent transition Reynolds number was approximately 2300 for diameter 0.5 mm. For diameter 0.25 mm, the friction factor evaluated by the form pressure drop also agreed well with the Poiseuille equation. For diameter 0.17 mm, the measured total friction factor was close to the Poiseuille prediction. (authors)

  1. Parallel channel effects under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Due to parallel channel effects, different flow patterns such as liquid down-flow and gas up-flow appear simultaneously in fuel bundles of a BWR core during postulated LOCAs. Applying the parallel channel effects to the fuel bundle, water drain tubes with a restricted bottom end have been developed in order to mitigate counter-current flow limiting and to increase the falling water flow rate at the upper tie plate. The upper tie plate with water drain tubes is an especially effective means of increasing the safety margin of a reactor with narrow gaps between fuel rods and high steam velocity at the upper tie plate. The characteristics of the water drain tubes have been experimentally investigated using a small-scaled steam-water system simulating a BWR core. Then, their effect on the fuel cladding temperature was evaluated using the LOCA analysis program SAFER. (orig.)

  2. 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.)

  3. 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

  4. Process heat reactors

    International Nuclear Information System (INIS)

    The consumption of heat, for industrial and domestic needs, takes up half of the national energy supply; direct utilization of the heat produced by nuclear reactors could therefore contribute to reduce the deficit in the energetic results. The restraints proper to heat consumption (dispersal and variety of consumers, irregular demand) involve the development of the heat transport system structures and adequate nuclear reactors. With this in view, the Commissariat a l'Energie Atomique and Technicatome are developing the CAS reactor series, pressurized water reactors (PWR), (CAS 3G reactor with a power of 420 MW.th.), and the Thermos reactor (100 MW.th.), directly conceived to produce heat at 1200C and whose technology derives from the experimental pool reactors type. In order to prove the value of the Thermos design, an experimental reactor should soon be constructed in the Saclay nuclear research centre

  5. 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

  6. 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.

  7. 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)

  8. Nuclear Reactor RA Safety Report, Vol. 11, Reactor operation

    International Nuclear Information System (INIS)

    This volume includes the following chapters describing: Organisation of reactor operation (including operational safety, fuel management, and regulatory rules for RA reactor operation); Control and maintenance of reactor components (reactor core, nuclear fuel, heavy water and cover gas systems, mechanical structures, electric power supply system, reactor instrumentation); Quality assurance and Training of the reactor personnel

  9. Testing the reactor charging machine

    International Nuclear Information System (INIS)

    One of the main objective of the R - D technological engineering program devoted to the Fuel Handling System is domestic production of equipment and technology for testing the ends of the reactor charging machine (MID) destined to Cernavoda NPP, beginning with Unit 2. To achieve the objective based on an own design, a bench-scale testing stand of MIDs which can simulate the pressure, flow-rate, and temperature conditions proper to fuel channels in operating CANDU 600 reactors. The main components of this testing facility are: - fuel channels, cold also test sections, allowing the coupling of MID end upwardly and downwardly, corresponding to the direction of the water flow through the channel; - technological installation feeding with light water the testing sections of the facility in thermohydraulic conditions, similar to those in the reactor, allowing the cold and hot testings, respectively, of the MID end; - cold testing installation, water supply and oil control panel, feeding the hydraulic drives of the MID's end during the testings; - fixed bridge and mobile carrier for MID's end positioning against testing sections; - installation for functional testing of MID thrusters, before pre-admission and reception tests; - dedicated tools and devices; - raising and transport mechanical devices for handling and positioning the MID's end upon the carrier; - automation panel for controlling the stand equipment and MID's end; - process computer for conducting on-line tests. MID's end testing implies mainly the following operations: - regulation, calibration and functional testing of the MID thrusters carried out independently on a specialised stand; - regulation and calibration of MID's end sub-assemblages; - carrying out the cold and hot pre-admission tests consisting in automatic performing, without operator intervention, of 12 fuel changes, two of which being successive; - performing the cold and hot reception tests, consisting in automatic accomplishment of 4

  10. The Chernobylsk reactor accident

    International Nuclear Information System (INIS)

    The construction, the safety philosophy, the major reactor physical parameters of RBMK-1000 type reactor units and the detailed description of the Chernobylsk-4 reactor accident, its causes and conclusions, the efforts to reduce the consequences on the reactor site and in the surroundings are discussed based on different types of Soviet documents including the report presented to the IAEA by the Soviet Atomic Energy Agency in August 1986. (V.N.)

  11. Zero energy reactor 'RB'

    International Nuclear Information System (INIS)

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  12. Ion channel screening.

    Science.gov (United States)

    Dunlop, John; Bowlby, Mark; Peri, Ravikumar; Tawa, Gregory; LaRocque, James; Soloveva, Veronica; Morin, John

    2008-08-01

    Ion channels are attractive targets for drug discovery with recent estimates indicating that voltage and ligand-gated channels account for the third and fourth largest gene families represented in company portfolios after the G protein coupled and nuclear hormone receptor families. A historical limitation on ion channel targeted drug discovery in the form of the extremely low throughput nature of the gold standard assay for assessing functional activity, patch clamp electrophysiology in mammalian cells, has been overcome by the implementation of multi-well plate format cell-based screening strategies for ion channels. These have taken advantage of various approaches to monitor ion flux or membrane potential using radioactive, non-radioactive, spectroscopic and fluorescence measurements and have significantly impacted both high-throughput screening and lead optimization efforts. In addition, major advances have been made in the development of automated electrophysiological platforms to increase capacity for cell-based screening using formats aimed at recapitulating the gold standard assay. This review addresses the options available for cell-based screening of ion channels with examples of their utility and presents case studies on the successful implementation of high-throughput screening campaigns for a ligand-gated ion channel using a fluorescent calcium indicator, and a voltage-gated ion channel using a fluorescent membrane potential sensitive dye. PMID:18694388

  13. The Failure Effect of Primary Coolant Pump to Thermo-Hydraulic Characteristic of TRIGA 2000 Reactor

    International Nuclear Information System (INIS)

    Has been done analysis of transient, when TRIGA 2000 reactor loss of primary coolant flow because primary pump loss of electric power, so fail in function.The calculation using RELAP5/MOD32 computer code with reactor core is modeled in the form of different seven channels as representation of different seven areas in core with 116 fuels. This reactor model also considers position of tip of primary pipe of input tank which is below of core, form of lower part core geometry influencing direction and coolant flow rate into core, and existence of diffuser system. The result of calculation in condition of steady state is obtained initiation condition of steady state is reached after 2500 seconds from reactor starts operation on 2000 kW power. On steady state, the channel-3 cladding temperature (hottest) is 149.63℃, the coolant temperature outlet from the channel-3 (hottest) is 105.66℃ , reactor inlet temperature is 32.2℃, and reactor outlet temperature is 46.79℃. The primary coolant entering reactor with flow rate 59.64 kg/s, distributed to core 31.44 kg/s and to by-pass of core or by-pass of chimney 28.20 kg/s. The result of calculation transient is obtained, before scram occur the channel-3 cladding temperature (hottest) is 161.03℃ and the coolant temperature outlet from the channel-3 (hottest) is 117.66℃. In the reactor core is a natural circulation as well (from reactor core, to chimney, to by-pass of chimney, to by-pass of core and back to platform) which is cooling reactor core. Scram occur on 250 seconds after failure of the primary pump. Based on result of this study is known that, when transient condition is happened because primary pump failure, reactor is predicted to stays in safety margin. (author)

  14. High solids fermentation reactor

    Science.gov (United States)

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  15. 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)

  16. Fusion reactor studies

    International Nuclear Information System (INIS)

    A review is given of fusion reactor systems studies, the objectives of these studies are outlined and some recent conceptual reactor designs are described. The need for further studies in greater depth is indicated so that progress towards a commercial fusion reactor may be consolidated. (U.K.)

  17. Reactor power measuring device

    International Nuclear Information System (INIS)

    The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

  18. Uncertainties treatment in the water-cooled nuclear research reactor-thermal design and analysis

    International Nuclear Information System (INIS)

    This paper describes methods of uncertainties and its calculationprocedures for the water-cooled nuclear research reactor (i. e. WWR-M2) with a 10 MWth, in the thermal design and analysis, where the uncertainties are due to the reactor fuel coolant channel design fabrication defects (fuel meat and clad thickness uncertainties). The results show: (1) The effect of fuel meat and cladding thickness may have great influence on the distribution of the axial temperatures (cladding surface, and fuel centerline) and other parameters in the reactor core and more intense in the reactor thermal design, (2) The selection of new reactor core operating conditions and parameters due to the fuel coolant channel fabrication defects, and (3) Calculation values of the hot spot and temperatures of the WWR-M2 reactor by using different methods.(author)

  19. Nondestructive testing of ampoules with lithium ceramics designed for blanket of thermonuclear reactor

    International Nuclear Information System (INIS)

    Full text: There are carried out prolonged radiation tests on research reactor WWR-K of ceramic materials made of lithium titanate Li2Ti03 with enrichment 36Li to 90 % manufactured in the form of sintered small balls and cylinder tablets put in experimental assembles (ampoules). At the present time tritium titanate is considered as one of the possible candidates of tritium production zone for demonstration international thermonuclear reactor blanket. Before feeding into the reactor experimental assemblages with Li2TiO3 were exposed to nondestructive control on horizontal channel of reactor with 'Agava' plant use by the neutron radiography method. The purpose of this work is on the one hand feeding quality control of tablets and small balls of lithium ceramics into experimental assembles, on the other hand the efficiency test of neutron radiography plant work after long stoppage of WWR-K reactor and the geometry change of irradiative channels and active zone of reactor

  20. 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)

  1. Critical channel power calculation for nominal operation in the CNE (Embalse nuclear power plant): sensitivity study

    International Nuclear Information System (INIS)

    In the Embalse nuclear power plant (CNE), the Regional Overpower Protection System acting on the Shutdown Systems number 1 and number 2 protects the reactor against overpowers in the reactor field for a localized peaking or a power increase in the reactor as a whole. This report summarizes the results of the critical channel power calculation for the time average powers configuration for the 380 reactor field channels. The final purpose of this work is to analyze and eventually modify the detector set points. Other reactor configurations are being analyzed. The report also presents a sensitivity analysis in order to evaluate potential sources of error and uncertainties which could affect the ROP performance. (author)

  2. Counter-current flow limitation in thin rectangular channels

    International Nuclear Information System (INIS)

    The phenomenon of counter-current flow limitation (CCFL) in thin rectangular channels is important in determining the heat removal capability of research reactors which use plate-type fuel elements similar to the MTR design. An analytical expression for predicting CCFL in narrow rectangular channels was derived from the momentum equations for the liquid and gas phase. The model assumes that the liquid downflow is in the form of a film along the narrower side walls of the channel, while the gas flow occupies the wide span of the rectangular channel. The average thickness of liquid film is related to the rate of gas flow through a stability criterion for the liquid film. The CCFL correlation agrees with air/water data taken at relatively high gas velocities. Depending on the magnitude of the dimensionless channel width W*, the new CCFL correlation approaches zero liquid penetration either in the form of a Wallis correlation or in terms of a Kutateladze number. The new correlation indicates that for a thin rectangular channel, the constant C in the Wallis flooding correlation depends on the aspect ratio of the channel. The approach to the appropriate asymptotic solutions also justifies the use of twice the wide span as the correct length scale for thin rectangular channels. 14 refs., 6 figs

  3. 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

  4. Calculations of two-phase flows in the liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Mathematical models used for the safety analysis of liquid metal cooled fast breeder reactors are considered. Models, taking into account sodium boiling in reactor channels (one-dimensional and many-dimensional approaches), fuel cladding melting, and movement of molten materials during loss of coolant, accidents are described

  5. 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)

  6. 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

  7. 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...... 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....

  8. 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

  9. 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.)

  10. Fundamentals of reactor chemistry

    International Nuclear Information System (INIS)

    In the Nuclear Engineering School of JAERI, many courses are presented for the people working in and around the nuclear reactors. The curricula of the courses contain also the subject material of chemistry. With reference to the foreign curricula, a plan of educational subject material of chemistry in the Nuclear Engineering School of JAERI was considered, and the fundamental part of reactor chemistry was reviewed in this report. Since the students of the Nuclear Engineering School are not chemists, the knowledge necessary in and around the nuclear reactors was emphasized in order to familiarize the students with the reactor chemistry. The teaching experience of the fundamentals of reactor chemistry is also given. (author)

  11. 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

  12. 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

  13. 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...

  14. TRP channels: an overview

    DEFF Research Database (Denmark)

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

    2005-01-01

    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.......The TRP ("transient receptor potential") family of ion channels now comprises more than 30 cation channels, most of which are permeable for Ca2+, and some also for Mg2+. On the basis of sequence homology, the TRP family can be divided in seven main subfamilies: the TRPC ('Canonical') family, the...... TRPV ('Vanilloid') family, the TRPM ('Melastatin') family, the TRPP ('Polycystin') family, the TRPML ('Mucolipin') family, the TRPA ('Ankyrin') family, and the TRPN ('NOMPC') family. The cloning and characterization of members of this cation channel family has exploded during recent years, leading to a...

  15. 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.

  16. Side-Channel Oscilloscope

    CERN Document Server

    Chaudhuri, Sumanta

    2011-01-01

    Side-Channel Analysis used for codebreaking could be used constructively as a probing tool for internal gates in integrated circuits. This paper outlines basic methods and mathematics for that purpose

  17. 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...

  18. 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

  19. Volume Regulated Channels

    DEFF Research Database (Denmark)

    Klausen, Thomas Kjær

    - serves a multitude of functions in the mammalian cell, regulating the membrane potential (Em), cell volume, protein activity and the driving force for facilitated transporters giving Cl- and Cl- channels a major potential of regulating cellular function. These functions include control of the cell cycle...... of volume perturbations evolution have developed system of channels and transporters to tightly control volume homeostasis. In the past decades evidence has been mounting, that the importance of these volume regulated channels and transporters are not restricted to the defense of cellular volume......, controlled cell death and cellular migration. Volume regulatory mechanisms has long been in focus for regulating cellular proliferation and my thesis work have been focusing on the role of Cl- channels in proliferation with specific emphasis on ICl, swell. Pharmacological blockage of the ubiquitously...

  20. 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.

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

    International Nuclear Information System (INIS)

    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

  2. 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.

  3. The disposition of can thermocouples in a nuclear reactor

    International Nuclear Information System (INIS)

    A philosophy is presented for deciding the distribution of can thermocouples within channels and of instrumented channels throughout the core of a reactor with cluster-type fuel elements when only a few thermocouples can be located in any one channel. The arrangement is made according to a 'factorial' design in which all fuel element positions of interest are covered in a group of channels. Two types of factorial design can be applied: the unconfounded design by which the thermocouples in each channel are chosen at random from the possible positions available, with the results that the temperatures have attached to them an uncertainty determined by the differences among channels; and the confounded design by which the positions are chosen so as to give temperatures whose uncertainty is determined only by the random variations within channels. It is also necessary to estimate standard deviations in order to predict the number of cans likely to reach a given temperature. The standard deviation can be expected to vary with channel position, and since there will also be systematic variations in temperature with channel position it is necessary to arrange channels into groups having similar mean fluxes and flux distributions. Each group is instrumented according to the pattern of a confounded design. The information that such an arrangement provides is an estimate of the systematic temperature variations within channels, estimates of within-channel variation of can temperature, of between-channel variation of can temperature, and of the variation of these quantities among groups of channels grouped according to similarity of mean flux and flux profile. (author)

  4. Device for reactor control system

    International Nuclear Information System (INIS)

    A device for nuclear reactor control system is described. The device comprises a channel with control column of neutron absorbing liquid displacer and drain throttle. To increase the reliability and stabilization the control in the flow of liquid, the displacer is fixed to the bar with the help of a rod which length is not less than the half of the core height. The displacer occupies the lower section of the core and divides the column of liquid in two parts consisting of the control column above the displacer and protective column below the drain throttle. The proposed device provides the control of energy distribution along the core height and can be used for leveling energy distribution field or its shaping. A reliable operation of the device is insured, in particular, the stability of such important characteristics as the position and height of the column of liquid, the magnitude of introduced reactivity, the range of controlled parameters

  5. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPRTM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENATM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENATM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  6. 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.

  7. 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 ...

  8. 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...

  9. 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.

  10. 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...

  11. Channeling in quasicrystals

    Energy Technology Data Exchange (ETDEWEB)

    Du Marchie van Voorthuysen, E.H.; Smulders, P.J.M. (Vakgroep Nucleaire Vaste Stof Fysica, Univ. of Groningen (Netherlands)); Werkman, R.D. (Vakgroep Vaste Stof Fysica, Univ. of Groningen (Netherlands)); Boer, J.L. de; Smaalen, S. van (Lab. of Inorganic Chemistry, Univ. of Groningen (Netherlands))

    1992-02-01

    Ion-beam channeling has been observed in quasicrystals. For 1 MeV {sup 4}He{sup +} ions in icosahedral AlCuFe the maximum effect found is 36%. The full width at half maximum of the observed dips is 1.3deg. The effect persists up to great depths (>200 nm), thus showing a high degree of ordering in this phase. The channeling effect is sensitive to radiation damage. (orig.).

  12. Channeling in quasicrystals

    International Nuclear Information System (INIS)

    Ion-beam channeling has been observed in quasicrystals. For 1 MeV 4He+ ions in icosahedral AlCuFe the maximum effect found is 36%. The full width at half maximum of the observed dips is 1.3deg. The effect persists up to great depths (>200 nm), thus showing a high degree of ordering in this phase. The channeling effect is sensitive to radiation damage. (orig.)

  13. The Maple reactor project

    International Nuclear Information System (INIS)

    MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

  14. High temperature reactors

    International Nuclear Information System (INIS)

    With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements

  15. 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.

  16. 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

  17. 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

  18. 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...

  19. Reactor Safety: Introduction

    International Nuclear Information System (INIS)

    The programme of the Reactor Safety Division focuses on the development of expertise on materials behaviour under irradiation for fission and fusion oriented applications. Furthermore, as nuclear energy needs international public acceptance with respect to safety and efficient management of natural resources and wants to reduce the burden of nuclear waste, the Reactor Safety Division enhanced its efforts to develop the MYRRHA project. MYRRHA, an accelerator driven sub-critical system, might have the potential to cope in Europe with the above mentioned constraints on acceptability and might serve as a technological platform for GEN IV reactor development, in particular the Liquid Metal Fast Reactor.The Reactor Safety Division gathers three research entities that are internationally recognised: the Reactor Materials Research department, the Reactor Physics and MYRRHA department and the Instrumentation department.The objectives of Reactor Materials Research are: to evaluate the integrity and behaviour of structural materials and nuclear fuels used in present and future nuclear power industry; to perform research to unravel and understand the parameters that determine the material and fuel behaviour under or after irradiation; to contribute to the interpretation and modelling of the materials and fuels behaviour in order to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the Reactor Materials Research department concentrate on four distinct disciplines: Reactor Pressure Vessel Steel embrittlement Stress corrosion cracking in reactor coolant environment, including Irradiation Assisted Stress Corrosion Cracking; Nuclear Fuel characterisation and development of new fuel types for commercial and test reactors. Development of materials for Fusion and advanced nuclear fission reactors. The safe operation of present nuclear power plants relies primarily on the integrity of the reactor pressure vessel

  20. 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

  1. 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.

  2. Measurement of spectrum at the experimental 6.5 MW reactor in Vinca

    International Nuclear Information System (INIS)

    Since RA reactor is supplied with horizontal experimental channels which lead directly to the core fast neutron spectrum in the channel does not differ much from the neutron spectrum in the core. Spectrum was measured by 'telescope' for detecting scattered protons. Measuring procedure together with the measured spectrum are presented in this paper

  3. 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.

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

    International Nuclear Information System (INIS)

    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.

  5. The HOR Nuclear Instrument Channel Refit

    Energy Technology Data Exchange (ETDEWEB)

    Kaaijk, C.N.J. [Reactor Institute Delft, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands)

    2011-07-01

    The research reactor in Delft, the HOR was built around 1960. Because of ageing effects, the nuclear instrumentation was completely replaced in 1980, together with the construction of a new control room outside the containment. In 2010, after 30 years of successful operation, it became a real challenge to repair the nuclear channels because of obsolete components. Of course, this problem was identified earlier, and a project was started in 2008 to select and replace the electronics of the nuclear channels. For this purpose a European tender was started to select a manufacturer for the new electronics in accordance with the requirements. The boundary conditions to be fulfilled by the manufacturer were: a) The functionality of the instrumentation and the interface to the plant should remain the same, and b) The proposed type of equipment should have been installed and commissioned successfully at other research reactors of comparable type earlier. Only the electronics should be replaced, detectors and cabling are reused. Parallel to this tender we started discussions with the authorities to clarify which standards the instrumentation should fulfil. We selected a digital system based on two microcontrollers, each one checking the other one. It turns out to be a flexible system. It was easily adapted to our needs, showing adequate provisions for guaranteeing data integrity. In the summer maintenance period of 2010 the instrumentation was successfully installed and commissioned. This paper will describe the steps taken and the tests performed. (author)

  6. Research performed at the ETRR-1 reactor using TOF technique

    International Nuclear Information System (INIS)

    This paper represents the results of studies of neutron transmission at several single crystals, performed at ETRR-1 reactor horizontal channels. The results of these studies starting in 1984 and continuing to date are discussed; the use of large single crystals as a band pass filter is also assessed

  7. G2 and G3 reactors design

    International Nuclear Information System (INIS)

    The 'FRANCE ATOME' Manufacturers Party has been entrusted with the G2 and G3 reactors engineering by the french A.E.C., for the first-five-year french project. Although these reactors are essentially plutonium generators, everyone has been linked with a power station which is supposed to supply with 40 MW, 'Electricite de France' has taken the liability upon itself. The reactor core includes most of G1 reactor parts (central gap excluded): horizontal channels, graphite parallelepipedic bricks stacking, steel thermal shield. The cooling is provided with CO2 under a 15 atmospheres pressure. This pressure is kept steady in a press-stressed concrete packing-case which is a cylinder horizontally shaped. Steel strips tightened encircle the concrete cylinder; itself protected by sole-plates. The cylinder bottom has brought about unusual problems which have been solved by the choice of an hemispheric shape. Packing-case tightness is provided by a 30 mm iron-plate connected with the inner wall of concrete. One of the reactor's special characteristics is the possibility of loading and unloading while operating. On loading side, barrel locks, each weighting 50 tons, allow new cans, at a pressure of 15 atmospheres, to pass. The cans process almost in a steady way through the channel, and finally drop down through bent spouts, then through spiral toboggans into a new lock. The cooling CO2 flow is provided with 3 turbo-bellows, these are actuated by average pressure-steam, obtained from exchangers. Every reactor supplies 4 exchangers which have been very difficult to build and to set up. The secondary cycle is standard and contains 3 stages (pressure 10,3: 2 and 0,5 kg/cm2). Steam can be condensed in the event of a group turbo-generator stopping, with no modification for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO2, its storage and drain. 49 boron carbide rods are used to control the operating power

  8. 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

  9. 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

  10. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

  11. Numerical Investigation of Startup Instabilities in Parallel-Channel Natural Circulation Boiling Systems

    Directory of Open Access Journals (Sweden)

    S. P. Lakshmanan

    2010-01-01

    Full Text Available The behaviour of a parallel-channel natural circulation boiling water reactor under a low-pressure low-power startup condition has been studied numerically (using RELAP5 and compared with its scaled model. The parallel-channel RELAP5 model is an extension of a single-channel model developed and validated with experimental results. Existence of in-phase and out-of-phase flashing instabilities in the parallel-channel systems is investigated through simulations under equal and unequal power boundary conditions in the channels. The effect of flow resistance on Type-I oscillations is explored. For nonidentical condition in the channels, the flow fluctuations in the parallel-channel systems are found to be out-of-phase.

  12. New assembly of ionization chambers for the RBMK reactor control systems

    International Nuclear Information System (INIS)

    A new assembly consisting of ionization chambers (IC) and cables is developed for the RBMK reactor control systems (RCS). The assembly together with the channel of RCS ensures reliable power control in the range from 10-10 to 1.5 nominal power and measuring the neutron flux density in the reactor core height. The design of the assembly with IC is presented. Tests showed that the new IC assembly provides high sensitivity of measurements and RCS channel noise immunity in all range of reactor power variation

  13. RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 1013 cm-2 s-1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

  14. Thai research reactor

    International Nuclear Information System (INIS)

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  15. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  16. Fuel channel replacement experience at Wolsong Unit 1

    International Nuclear Information System (INIS)

    The Wolsong Nuclear Power Station is located on the shore of the Eastern Sea near the ancient capital city of Kyongju in the Republic of Korea. Wolsong Unit 1 is the only operating CANDU 6 Pressurized Heavy Water Reactor at the four unit station owned and operated by the Korea Electric Power Corporation (KEPCO). The other three CANDU 6 units are under various stages of construction and all three are scheduled to be in-service during the 1997-1999 period. The Wolsong 1 fuel channels were subjected to periodic inspections in accordance with the regulatory requirements and the inspection findings revealed acceptable indications until 1990. The debris induced wear marks identified during pressure tube inspections in 1990, 1992, and 1994 led to the decision to replace fuel channels E11, M11, and O08 during the scheduled plant outage in 1994. The three fuel channels were replaced in 26.5 days (8.8 days per channel) and a total of 75.8 Rem (25.3 Rem per channel) dose was expended. Unshielded radiation fields on reactor face ranged from 450 mRem/hour on contact tithe the end fitting face to 2000 mRem/hour on contact with the feeder ports. The work was performed on two twelve hour shifts involving 20 technicians and engineers from AECL; and 4-4 engineers, technicians, and tradesmen from KEPCO and KPS. (author)

  17. Computer code for thermal hydraulic analysis of light water reactors

    International Nuclear Information System (INIS)

    A computer programme (THAL) has been developed to perform thermal hydraulic analysis of a single channel in a light water moderated core. In this code the hydrodynamic and thermodynamic equations describing one-dimensional axial flow have been discretized and solved explicitly stepwise up the coolant channel for an arbitrary power profile. THAL has been developed for use on small computers and it is capable of predicting the coolant, clad and fuel temperature profiles, steam quality, void fraction, pressure drop, critical heat flux and DNB ratio throughout the core. A boiling water reactor and a pressurized water reactor have been analyzed as test cases. The results obtained through the use of THAL compare favourably with those given in the design reports of these reactor systems. (author)

  18. Investigating a Partial LOCA in the IAEA Generic Research Reactor

    International Nuclear Information System (INIS)

    The behavior of the IAEA research reactor under partial LOCA (Loss of Coolant Accident conditions scats investigated. The reactor is a pool-type light water 10 MW research reactor employing MTR-type fuel elements. The extremely rare LOCA scenario is assumed to take place when a guillotine break in one of the Malta coolant loops occurs. Under these conditions the water level in the pool decreases, reaching a level that covers only part of the core. With no flow through the cooling channels, decay-heat will raise the temperature of the partly covered fuel elements. This study demonstrates that if water level remains at or above 1/4 of the fuel channel length, no damage to the core will occur. This has also been shown by an experiment performed at Livermore

  19. 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

  20. 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

  1. Soreq Nuclear Reactor Fuel Element Flow Distribution

    International Nuclear Information System (INIS)

    Flow of cold water through the Soreq Nuclear Reactor fuel element was simulated numerically. The main objective of the present study was to obtain the flow distribution among the rectangular channels of the element. The results of the simulations were compared to the overall pressure drop on the element measured in Soreq Nuclear Reactor. The numerical model chosen has succeeded in predicting the pressure drop on the fuel element of up to 5% from the measured values. Flow through the IPEN IEA-R1 MTR fuel element was also simulated as a part of a model validation procedure. The numerical results were compared to the measurements available in the literature [1]. It was found that the water pool above the fuel element has a significant influence on the flow distribution among the channels of the element. The flow distribution reported in [1] was closely predicted numerically when the water pool was included into the simulated geometry. It can be concluded that flow distribution in the Soreq Nuclear Reactor fuel element is flatter than that in the IPEN IEA-R1 MTR fuel element

  2. Nuclear graphite for high temperature reactors

    International Nuclear Information System (INIS)

    The cores and reflectors in modern High Temperature Gas Cooled Reactors (HTRs) are constructed from graphite components. There are two main designs; the Pebble Bed design and the Prism design. In both of these designs the graphite not only acts as a moderator, but is also a major structural component that may provide channels for the fuel and coolant gas, channels for control and safety shut off devices and provide thermal and neutron shielding. In addition, graphite components may act as a heat sink or conduction path during reactor trips and transients. During reactor operation, many of the graphite component physical properties are significantly changed by irradiation. These changes lead to the generation of significant internal shrinkage stresses and thermal shut down stresses that could lead to component failure. In addition, if the graphite is irradiated to a very high irradiation dose, irradiation swelling can lead to a rapid reduction in modulus and strength, making the component friable.The irradiation behaviour of graphite is strongly dependent on its virgin microstructure, which is determined by the manufacturing route. Nevertheless, there are available, irradiation data on many obsolete graphites of known microstructures. There is also a well-developed physical understanding of the process of irradiation damage in graphite. This paper proposes a specification for graphite suitable for modern HTRs. (author)

  3. Review of Ontario Hydro program to reduce fuel channel gap

    International Nuclear Information System (INIS)

    The consultant was contracted to review the 'Pressure Tube Fretting and Power Pulse Program' undertaken by Ontario Hydro. They were requested to report primarily on the adequacy of quality assurance activities in the project management, engineering, and manufacturing parts of the project, but in addition were asked to review and comment on the technical adequacy of the methods. The program involved three potential solutions or approaches. Each was designed to prevent fuel bundles shifting into the reactor core if there is fuel channel flow reversal. Each approach was also intended to reduce or eliminate fuel vibration and the consequential fretting observed in fuel channels with significant creep. (author). 4 figs

  4. Repair cycle optimization for multi-channel engineering systems

    International Nuclear Information System (INIS)

    Conceptual definition of mathematical problem related to the evaluation of residual service-life of multi-channel engineering systems (systems comprising N similar process channels for the RBMK-type reactors are included, in particular) and optimization of the cycles of overhaul and scheduled repairs is considered. Integral economic index of aging permitting not only to predict residual service-life of the system but also to evaluate from technico-economical viewpoint the optimal strategy of overhaul schedulled repairs is obtained. 2 refs

  5. 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.

  6. 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

  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. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.)

  9. Fuel channel life limiting factors that dictate fuel channel maintenance requirements

    International Nuclear Information System (INIS)

    CANDU reactors have been operating for 33 years. The Nuclear Power Demonstration (NPD) Unit started up in 1962 and the prototype of CANDU, Douglas Point, started in 1967. The first commercial reactors, Pickering Units 1 and 2 both went into service in 1971 closely followed by Units 3 and 4 in 1972 and 1973 respectively. Operating commercial reactor experience represents over 10,000 pressure tubes, not including the replaced channels in all the Pickering A Units, and nearly 130,000 pressure tube operating years. No pressure tube has yet operated for its 30 year design lifetime of 210 KEFPH at 80% capacity factor. The longest operating time for pressure tubes to-date is about 120 KEFPH in Pickering Unit 4. Many lessons have been learned regarding pressure tube life limiting factors from the early CANDU units and these, together with the information obtained from an extensive pressure tube R and D program, have resulted in many design changes and improvements in material properties, mainly from manufacturing route changes. Reactors built recently are expected to achieve their 30 year design life. The development of Periodic and In-service Inspection programs and equipment, assessment methodologies and acceptance criteria, and the development of maintenance tooling and procedures are enabling the life limiting factors to be addressed in the currently operating units. The life limiting factors in currently operating Units are reviewed in relation to the experience gained from the early units, the R and D programs and the inspection and maintenance performed to date. (author)

  10. Development of a digital card to simulate period transients in research reactors

    International Nuclear Information System (INIS)

    This work presents the development of a card to be used in a 'slot' of a micro-computer for evaluation of a nuclear channel used to monitor the start up of nuclear reactors. The results of the bench tests showed good linearity and 2% error deviation in the entire range of operation. Fields tests, performed with the start up channel of IEA-R1 research reactor showed that the card is an excellent device to verify the performance of the channel during steady state, and transient conditions. (author)

  11. Innovative Pressure Tube Light Water Reactor with Variable Moderator Control

    International Nuclear Information System (INIS)

    The features of a reactor based on multiple pressure tubes, rather than a single pressure vessel, provide the reactor with considerable flexibility for continuous design improvements and developments. This paper presents the development of innovative pressure tube light water reactor, which has the ability to advance the current pressure tubes reactors. The proposed design is aimed to simplify the pressure tubes reactors by: - replacing heavy water by a light water as a coolant and moderator, - adopting batch refueling instead of on-line refueling. Furthermore, the design is based on proven technologies, existing fuel and structure materials. Therefore, it is reasonable to expect significant capital cost savings, short licensing and introduction period of the proposed concept into the power production grid. The basic novelty of the proposed design is based on an idea of variable moderator content in the core and 'breed and burn' mode of operation. Both concepts were extensively investigated and reported in the past (2) (3) (4). In order to evaluate a practical reactor design build on proven technology, several features of the advanced CANDU reactor (ACR-1000) were adopted. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The proposed design is basically pressure tube light water reactor with variable moderator Control (PTVM LWR). This paper presents a detailed description of the PTVM core design and demonstrates the reactivity control and the 'breed and burn' mode of operation, which are implemented by the variation of the moderator in the core, from a

  12. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  13. Evaluation of research reactors

    International Nuclear Information System (INIS)

    The present status of research reactors with highly enriched (93%) uranium fuel at JAERI, JRR-2 and JMTR is described. JRR-2 is a heterogeneous type of reactor, using heavy water as moderator and coolant. It uses both MTR type and cylindrical type of fuel elements. The maximum thermal power and the thermal neutron flux are 10 MW and 2x1014 n/cm2 see respectively. The reactor has been used for various experiments such as solid state physics, material irradiation, reactor fuel irradiation and radioisotope production. The JMTR is a multi-purpose tank type material testing reactor, and light water moderator and coolant, operated at 50 MW. The evaluation of lower enriched fuel and its consequences for both reactors is considered more especially

  14. 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

  15. The nuclear soliton reactor

    International Nuclear Information System (INIS)

    The basic reactor physics of a completely novel nuclear fission reactor design - the soliton-reactor - is presented on the basis of a simple model. In such a reactor, the neutrons in the critical region convert either fertile material in the adjacent layers into fissile material or reduce the poisoning of fissile material in such a manner that successively new critical regions emerge. The result is an autocatalytically driven burn-up wave which propagates throughout the reactor. Thereby, the relevant characteristic spatial distributions (neutron flux, specific power density and the associated particle densities) are solitons - wave phenomena resulting from non-linear partial differential equations which do not change their shape during propagation. A qualitativley new kind of harnessing nuclear fission energy may become possible with fuel residence times comparable with the useful lifetime of the reactor system. In the long run, fast breeder systems which exploit the natural uranium and thorium resources, without any reprocessing capacity are imaginable. (orig.)

  16. 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...

  17. Computer-aided testing and operational aids for PARR-1 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ansari, S.A. (Nuclear Engineering Div., Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (PK))

    1990-06-01

    The utilization of the plant computer of Pakistan Research Reactor (PARR-1) for automatic periodic testing of nuclear instrumentation in the reactor is described. Computer algorithms have been developed for on-line acquisition and real-time processing of nuclear channel signals. The mean value, standard deviation, and probability distributions of nuclear channel signals are obtained in real time, and the computer generates a warning message if the signal error exceeds the maximum permissible error. In this way a faulty channel is automatically identified. Other real-time algorithms are also described that assist the operator in safe reactor operation by automatically computing approach-to-criticality during reactor start-up and the control rod worth determination.

  18. Computer-aided testing and operational aids for PARR-1 nuclear reactor

    International Nuclear Information System (INIS)

    The utilization of the plant computer of Pakistan Research Reactor (PARR-1) for automatic periodic testing of nuclear instrumentation in the reactor is described. Computer algorithms have been developed for on-line acquisition and real-time processing of nuclear channel signals. The mean value, standard deviation, and probability distributions of nuclear channel signals are obtained in real time, and the computer generates a warning message if the signal error exceeds the maximum permissible error. In this way a faulty channel is automatically identified. Other real-time algorithms are also described that assist the operator in safe reactor operation by automatically computing approach-to-criticality during reactor start-up and the control rod worth determination

  19. 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)

  20. Development of a central PC-based system for reactor signal monitoring and analysis

    International Nuclear Information System (INIS)

    A personal computer based system was developed for on-line monitoring, signal processing and display of important reactor parameters of the Pakistan Research Reactor-1. The system was designed for assistance to both reactor operator and users. It performs three main functions. The first is the centralized radiation monitoring in and around the reactor building. The computer acquires signals from radiation monitoring channels and continuously displays them on distributed monitors. Trend monitoring and alarm generation is also done. In case of any abnormal condition the radiation level data is automatically stored in computer memory for detailed off-line analysis. In the second part the computer does the performance testing of nuclear instrumentation channels by signal statistical analysis, and generates alarm in case the channel standard deviation error exceeds the permissible error. Mean values of important nuclear signals are also displayed on distributed monitors as a part of reactor safety parameters display system. The third function is on-line computation of reactor physics parameters of the core which are important from operational and safety points-of-view. The signals from radiation protection system and nuclear instrumentation channels in the reactor were interfaced with the computer for this purpose. The development work was done under an IAEA research contract as a part of coordinated research programme. (author)

  1. Fusion reactor research

    International Nuclear Information System (INIS)

    This work covers four separate areas: (1) development of technology for processing liquid lithium from blankets, (2) investigation of hydrogen isotope permeation in candidate structural metals and alloys for near-term fusion reactors, (3) analytical studies encompassing fusion reactor thermal hydraulics, tritium facility design, and fusion reactor safety, and (4) studies involving dosimetry and damage analysis. Recent accomplishments in each of these areas are summarized

  2. 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

  3. The replacement research reactor

    International Nuclear Information System (INIS)

    As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described

  4. PFBR reactor protection

    International Nuclear Information System (INIS)

    Design philosophy adopted for Prototype Fast breeder Reactor (PFBR) is a classical one and has the following features: triplicated sensors for measuring important safety parameters; two independent reactor protection Logic Systems based on solid state devices; reactivity control achieved by control rods; gas equipped modules at the core blanket interface providing negative reactivity. Design verification of these features showed that safety of the reactor can be achieved by a traditional approach since the inherent features of LMFBR make this easy

  5. Reactor BR2

    International Nuclear Information System (INIS)

    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

  6. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    This module describes the general design, characteristics and parameters of TRIGA reactors and fuels. It is recommended that most of this information should be incorporated into any reactor operator training program and, in many cases, the facility Safety Analysis Report. It is oriented to teach the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and the differences between TRIGA fuels and others more traditional reactor fuels. (nevyjel)

  7. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

  8. 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.

  9. 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.

  10. Trp channels and itch.

    Science.gov (United States)

    Sun, Shuohao; Dong, Xinzhong

    2016-05-01

    Itch is a unique sensation associated with the scratch reflex. Although the scratch reflex plays a protective role in daily life by removing irritants, chronic itch remains a clinical challenge. Despite urgent clinical need, itch has received relatively little research attention and its mechanisms have remained poorly understood until recently. The goal of the present review is to summarize our current understanding of the mechanisms of acute as well as chronic itch and classifications of the primary itch populations in relationship to transient receptor potential (Trp) channels, which play pivotal roles in multiple somatosensations. The convergent involvement of Trp channels in diverse itch signaling pathways suggests that Trp channels may serve as promising targets for chronic itch treatments. PMID:26385480

  11. 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.

  12. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1984 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, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

  13. 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)

  14. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose 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 Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

  15. 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

  16. New reactor concepts

    International Nuclear Information System (INIS)

    The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

  17. Reactor construction steels

    International Nuclear Information System (INIS)

    The basic functions of light water reactor components are shown on the example of a pressurized water reactor and the requirements resulting therefrom for steel, the basic structural material, are derived. A detailed analysis of three main groups of reactor steels is presented and the applications are indicated of low-alloyed steels, high-alloyed austenitic steels, and steels with a high content of Ni and of alloying additions for steam generator pipes. An outline is given of prospective fast breeder reactor steels. (J.K.)

  18. Commercialization of fast reactors

    International Nuclear Information System (INIS)

    Comparative analysis has been performed of capital and fuel cycle costs for fast BN-type and pressurized light water VVER-type reactors. As a result of materials demand and components costs comparison of NPPs with VVER-1000 and BN-600 reactors, respectively, conclusion was made, that under equal conditions of the comparison, NPP with fast reactor had surpassed the specific capital cost of NPP with VVER by about 30 - 40 %. Ways were determined for further decrease of this difference, as well as for the fuel cycle cost reduction, because at present it is higher than that of VVER-type reactors. (author)

  19. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

  20. Natural convection type reactor

    International Nuclear Information System (INIS)

    In a natural convection type nuclear reactor, a reactor core is disposed such that the top of the reactor core is always situated in a flooded position even if pipelines connected to the pressure vessel are ruptured and the level at the inside of the reactor vessel is reduced due to flashing. Further, a lower dry well situated below the pressure vessel is disposed such that it is in communication with a through hole to a pressure suppression chamber situated therearound and the reactor core is situated at the level lower than that of the through hole. If pipelines connected to the pressure vessel are ruptured to cause loss of water, although the water level is lowered after the end of the flashing, the reactor core is always flooded till the operation of a pressure accummulation water injection system to prevent the top of the reactor core even from temporary exposure. Further, injected water is discharged to the outside of the pressure vessel, transferred to the lower dry well, and flows through the through hole to the pressure control chamber and cools the surface of the reactor pressure vessel from the outside. Accordingly, the reactor core is cooled to surely and efficiently remove the after-heat. (N.H.)