Statistic techniques of process control for MTR type

International Nuclear Information System (INIS)

Oliveira, F.S.; Ferrufino, F.B.J.; Santos, G.R.T.; Lima, R.M.

2002-01-01

This work aims at introducing some improvements on the fabrication of MTR type fuel plates, applying statistic techniques of process control. The work was divided into four single steps and their data were analyzed for: fabrication of U 3 O 8 fuel plates; fabrication of U 3 Si 2 fuel plates; rolling of small lots of fuel plates; applying statistic tools and standard specifications to perform a comparative study of these processes. (author)

Final disposition of MTR fuel

International Nuclear Information System (INIS)

Jonnson, Erik B.

1996-01-01

The final disposition of power reactor fuel has been investigated for a long time and some promising solutions to the problem have been shown. The research reactor fuels are normally not compatible with the zirkonium clad power reactor fuel and can thus not rely on the same disposal methods. The MTR fuels are typically Al-clad UAl x or U 3 Si 2 , HEU resp. LEU with essentially higher remaining enrichment than the corresponding power reactor fuel after full utilization of the uranium. The problems arising when evaluating the conditions at the final repository are the high corrosion rate of aluminum and uranium metal and the risk for secondary criticality due to the high content on fissionable material in the fully burnt MTR fuel. The newly adopted US policy to take back Foreign Research Reactor Spent Fuel of US origin for a period of ten years have given the research reactor society a reasonable time to evaluate different possibilities to solve the back end of the fuel cycle. The problem is, however, complicated and requires a solid engagement from the research reactor community. The task would be a suitable continuation of the RERTR program as it involves both the development of new fuel types and collecting data for the safe long-term disposal of the spent MTR fuel. (author)

Re-qualification of MTR-type fuel plates fabrication process

International Nuclear Information System (INIS)

Elseaidy, I.M.; Ghoneim, M.M.

2010-01-01

The fabricability issues with increased uranium loading due to use low enrichment of uranium (LEU), i.e. less than 20 % of U 235 , increase the problems which occur during compact manufacturing, roll bonding of the fuel plates, potential difficulty in forming during rolling process, mechanical integrity of the core during fabrication, potential difficulty in meat homogeneity, and the ability to fabricate plates with thicker core as a means of increasing total uranium loading. To produce MTR- type fuel plates with high uranium loading (HUL) and keep the required quality of these plates, many of qualification process must be done in the commissioning step of fuel fabrication plant. After that any changing of the fabrication parameters, for example changing of any of the raw materials, devises, operators, and etc., a re- qualification process should be done in order to keep the quality of produced plates. Objective of the present work is the general description of the activities to be accomplished for re-qualification of manufacturing MTR- type nuclear fuel plates. For each process to be re-qualified, a detailed of re-qualification process were established. (author)

The use of experimental data in an MTR-type nuclear reactor safety analysis

Science.gov (United States)

Day, Simon E.

Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

The use of experimental data in an MTR-type nuclear reactor safety analysis

International Nuclear Information System (INIS)

Day, S.E.

2006-01-01

Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type (i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core. (author)

The use of experimental data in an MTR-type nuclear reactor safety analysis

Energy Technology Data Exchange (ETDEWEB)

Day, S.E

2006-07-01

Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type (i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core. (author)

Application of reactivity method to MTR fuel burn-up measurement

International Nuclear Information System (INIS)

Zuniga, A.; Ravnik, M.; Cuya, R.

2001-01-01

Fuel element burn-up has been measured for the first time by reactivity method in a MTR reactor. The measurement was performed in RP-10 reactor of Peruvian Institute for Nuclear Energy (IPEN) in Lima. It is a pool type 10MW material testing reactor using standard 20% enriched uranium plate type fuel elements. A fresh element and an element with well defined burn-up were selected as reference elements. Several elements in the core were selected for burn-up measurement. Each of them was replaced in its original position by both reference elements. Change in excess reactivity was measured using control rod calibration curve. The burn-up reactivity worth of fuel elements was plotted as a function of their calculated burnup. Corrected burn-up values of the measured fuel elements were calculated using the fitting function at experimental reactivity for all elements. Good agreement between measured and calculated burn-up values was observed indicating that the reactivity method can be successfully applied also to MTR fuel element burn-up determination.(author)

Implementation of a quality assurance system for the design and manufacturing of fuel assembly MTR-plate type

International Nuclear Information System (INIS)

Koll, J.H.

1987-01-01

Since more than 30 years ago, fuel assemblies (FA) of the MTR-Plate type, for research reactors, have been developed and produced using well known technologies, with different methods for the design, manufacturing, quality control and subsequent verification of FA behaviour, as well as of the design data. The FA and its reliability has been improved through the recycling of the obtained information. No nuclear accidents or major incidents have taken place that can be blamed to FA due to design, manufacturing or its use. Since the 70's, the use of Quality Assurance methodology has been increased, especially for Nuclear Power Plants, in order to ensure safety for these reactors. The use of QA for reactors for research, testing or other uses, has also been steadily increased, not only due to safety reasons, but also because of its convenience for a good operation, being presently a common requirement of the operator of the installation. Herewith is described the way the QA system that has been developed for the design, manufacturing, quality control and supply of MTR-plate type FA, at the Development Section of the Argentine Atomic Energy Commission (CNEA). (Author)

Concept study for interim storage of research reactor fuel elements in transport and storage casks. Transport and storage licensing procedure for the CASTOR MTR 2 cask. Final report

International Nuclear Information System (INIS)

Weiss, M.

2001-01-01

As a result of the project, a concept was to be developed for managing spent fuel elements from research reactors on the basis of the interim storage technology existing in Germany, in order to make the transition to direct disposal possible in the long term. This final report describes the studies for the spent fuel management concept as well as the development of a transport and storage cask for spent fuel elements from research reactors. The concept analyses were based on data of the fuel to be disposed of, as well as the handling conditions for casks at the German research reactors. Due to the quite different conditions for handling of casks at the individual reactors, it was necessary to examine different cask concepts as well as special solutions for loading the casks outside of the spent fuel pools. As a result of these analyses, a concept was elaborated on the basis of a newly developed transport and storage cask as well as a mobile fuel transfer system for the reactor stations, at which a direct loading of the cask is not possible, as the optimal variant. The cask necessary for this concept with the designation CASTOR trademark MTR 2 follows in ist design the tried and tested principles of the CASTOR trademark casks for transport and interim storage of spent LWR fuel. With the CASTOR trademark MTR 2, it is possible to transport and to place into long term interim storage various fuel element types, which have been and are currently used in German research reactors. The technical development of the cask has been completed, the documents for the transport license as type B(U)F package design and for obtaining the storage license at the interim storage facility of Ahaus have been prepared, submitted to the licensing authorities and to a large degree already evaluated positively. The transport license of the CASTOR trademark MTR 2 has been issued for the shipment of VKTA-contents and FRM II compact fuel elements. (orig.)

Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors

Energy Technology Data Exchange (ETDEWEB)

Khedr, A.; Abdel-Latif, Salwa H. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt); Abdel-Hadi, Eed A. [Benha Univ., Cairo (Egypt). Shobra Faculty of Engineering; D' Auria, F. [Pisa Univ. (Italy)

2016-03-15

In an attempt to understand the built-up of natural circulation in MTR pool type upward flow research reactors after loss of power, an experimental test rig was built to simulate the loop of natural circulation in MTR reactors. The test rig consisting of two vertically oriented branches, in one of them the core is simulated by two rectangular, electrically heated, parallel channels. The other branch simulates the part of the return pipe that participates in the development of core natural circulation. In the first phase of the work, many experimental runs at different conditions of channel's power and branch's initial temperatures are performed. The channel's coolant and surface temperatures were measured. The measurements and their interpretation were published by the first three authors. In the present work the thermal hydraulic behavior of the test rig is complemented by theoretical analysis using RELAP5 Mod 3.3 system code. The analysis consisting of two parts; in the first part RELAP5 model is validated against the measured values and in the second part some of the other not measured hydraulic parameters are predicted and analyzed. The test rig is typically nodalized and an input dick is prepared. In spite of the low pressure of the test rig, the results show that RELAP5 qualitatively predicts the thermal hydraulic behaviour and the accompanied phenomenon of flow inversion of such facilities. Quantitatively, there is a difference between the predicted and measured values especially the channel's surface temperature. This difference may be return to the uncertainties in initial conditions of experimental runs, the position of the thermocouples which buried inside the heat structure, and the heat transfer package in RELAP5.

Preparation of U3O8 powder for MTR type fuel from ammonium uranyl carbonate

International Nuclear Information System (INIS)

Marcondes, G.H.; Riella, H.G.

1990-08-01

In this paper it is described the research done at IPEN-CNEN/SP on the preparation of U 3 O 8 powder from calcination of the AUC, with appropriate characteristics to be used as dispersoid for MTR type fuel. The calcination in air of the AUC leads a U 3 O 8 powder that is further processed to obtain a powder with density and particle size as especifications. The important process parameters are here discussed with the variation AUC calcination temperature and sintering time of the U 3 O 8 powder. (author) [pt

The effect of code user and boundary conditions on RELAP calculations of MTR research reactor transient scenarios

Directory of Open Access Journals (Sweden)

Khedr Ahmed

2005-01-01

Full Text Available The safety evaluation of nuclear power and re search reactors is a very important step before their construction and during their operation. This evaluation based on the best estimate calculations requires qualified codes qualified users, and qualified nodalizations. The effect of code users on the RELAP5 results during the analysis of loss of flow transient in MTR research reactors is presented in this pa per. To clarify this effect, two nodalizations for research reactor different in the simulation of the open water surface boundary conditions of the reactor pool have been used. Very different results are obtained with few choices for code users. The core natural circulation flow with the be ginning of core boiling doesn't stop but in creases. The in creasing in the natural circulation flow shifts out the boiling from the core and the clad temperature decreases be low the local saturation temperature.

Simulation and transient analyses of a complete passive heat removal system in a downward cooling pool-type material testing reactor against a complete station blackout and long-term natural convection mode using the RELAP5/3.2 code

Energy Technology Data Exchange (ETDEWEB)

Hedayat, Afshin [Reactor and Nuclear Safety School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

2017-08-15

In this paper, a complete station blackout (SBO) or complete loss of electrical power supplies is simulated and analyzed in a downward cooling 5-MW pool-type Material Testing Reactor (MTR). The scenario is traced in the absence of active cooling systems and operators. The code nodalization is successfully benchmarked against experimental data of the reactor's operating parameters. The passive heat removal system includes downward water cooling after pump breakdown by the force of gravity (where the coolant streams down to the unfilled portion of the holdup tank), safety flapper opening, flow reversal from a downward to an upward cooling direction, and then the upward free convection heat removal throughout the flapper safety valve, lower plenum, and fuel assemblies. Both short-term and long-term natural core cooling conditions are simulated and investigated using the RELAP5 code. Short-term analyses focus on the safety flapper valve operation and flow reversal mode. Long-term analyses include simulation of both complete SBO and long-term operation of the free convection mode. Results are promising for pool-type MTRs because this allows operators to investigate RELAP code abilities for MTR thermal–hydraulic simulations without any oscillation; moreover, the Tehran Research Reactor is conservatively safe against the complete SBO and long-term free convection operation.

Simulation and transient analyses of a complete passive heat removal system in a downward cooling pool-type material testing reactor against a complete station blackout and long-term natural convection mode using the RELAP5/3.2 code

Directory of Open Access Journals (Sweden)

Afshin Hedayat

2017-08-01

Full Text Available In this paper, a complete station blackout (SBO or complete loss of electrical power supplies is simulated and analyzed in a downward cooling 5-MW pool-type Material Testing Reactor (MTR. The scenario is traced in the absence of active cooling systems and operators. The code nodalization is successfully benchmarked against experimental data of the reactor's operating parameters. The passive heat removal system includes downward water cooling after pump breakdown by the force of gravity (where the coolant streams down to the unfilled portion of the holdup tank, safety flapper opening, flow reversal from a downward to an upward cooling direction, and then the upward free convection heat removal throughout the flapper safety valve, lower plenum, and fuel assemblies. Both short-term and long-term natural core cooling conditions are simulated and investigated using the RELAP5 code. Short-term analyses focus on the safety flapper valve operation and flow reversal mode. Long-term analyses include simulation of both complete SBO and long-term operation of the free convection mode. Results are promising for pool-type MTRs because this allows operators to investigate RELAP code abilities for MTR thermal–hydraulic simulations without any oscillation; moreover, the Tehran Research Reactor is conservatively safe against the complete SBO and long-term free convection operation.

Non-electric applications of pool-type nuclear reactors

International Nuclear Information System (INIS)

Adamov, E.O.; Cherkashov, Yu.M.; Romenkov, A.A.

1997-01-01

This paper recommends the use of pool-type light water reactors for thermal energy production. Safety and reliability of these reactors were already demonstrated to the public by the long-term operation of swimming pool research reactors. The paper presents the design experience of two projects: Apatity Underground Nuclear Heating Plant and Nuclear Sea-Water Desalination Plant. The simplicity of pool-type reactors, the ease of their manufacturing and maintenance make this type of a heat source attractive to the countries without a developed nuclear industry. (author). 6 figs, 1 tab

Replacement of thermal column elastomeric gasket in pool type research reactors based on ageing and radiation degradation

International Nuclear Information System (INIS)

Garai, S.K.

2006-01-01

Pool type research reactors are designed with Thermal column facilities to irradiate samples at different flux levels of thermal neutrons. The sealing of demineralised pool water between stainless steel lined pool wall and the Aluminium Thermal column plate is achieved by an elastomeric gasket. The gasket joint is subjected to pool water temperature ranging from 25degC to 45degC and radiation field of the order of 104 -106 R/hr. The gasket loses its sealing properties due to ageing and radiation degradation after a few years, leading to the leakage and loss of the pool water. Though degradation of the gasket is, generally, predictable, some amount of uncertainty always remains in the leakage rate. The paper describes the study of a few elastomers in radiation environment and replacement of the Thermal column gasket of a swimming pool type research reactor. It includes the details of features like planning and scheduling, the actual sequential execution of the job, various problems encountered and corrective measures applied, engineering and radiological safety measures adopted, development of remote tools, disassembly and reassembly procedure and finally satisfactory completion of the site job in high radiation environment with minimum time and man rem consumption. (author)

MTR fuel plate qualification in OSIRIS reactor

International Nuclear Information System (INIS)

Sacristan, P.; Boulcourt, P.; Naury, S.; Marchard, L.; Carcreff, H.; Noirot, J.

2005-01-01

Qualification of new MTR fuel needs the irradiation in research reactors under representative neutronic, heat flux and thermohydraulic conditions. The experiments are performed in France in the OSIRIS reactor by irradiating MTR full size fuel plates in the IRIS device located in the reactor core. The fuel plates are easily removed from the device during the shutdown of the reactor for performing thickness measurements along the plates by means of a swelling measurement device. Beside the calculation capabilities, the experimental platform includes: the ISIS neutron mock-up for the measurement of neutron flux distribution along the plates; the γ spectrometry for the purpose of measuring the activities of the radionuclides representative of the power and the burnup and to compare with the neutronic calculation. Owing to the experience feedback, a good agreement is observed between calculation and measurement; destructive post irradiation examinations in the LECA facility (Cadarache). New irradiations with the IRIS device and at higher heat flux are under preparation for qualification of MTR fuels. (author)

Simulation of a pool type research reactor

International Nuclear Information System (INIS)

Oliveira, Andre Felipe da Silva de; Moreira, Maria de Lourdes

2011-01-01

Computational fluid dynamic is used to simulate natural circulation condition after a research reactor shutdown. A benchmark problem was used to test the viability of usage such code to simulate the reactor model. A model which contains the core, the pool, the reflector tank, the circulation pipes and chimney was simulated. The reactor core contained in the full scale model was represented by a porous media. The parameters of porous media were obtained from a separate CFD analysis of the full core model. Results demonstrate that such studies can be carried out for research and test of reactors design. (author)

Flow velocity calculation to avoid instability in a typical research reactor core

International Nuclear Information System (INIS)

Oliveira, Carlos Alberto de; Mattar Neto, Miguel

2011-01-01

Flow velocity through a research reactor core composed by MTR-type fuel elements is investigated. Core cooling capacity must be available at the same time that fuel-plate collapse must be avoided. Fuel plates do not rupture during plate collapse, but their lateral deflections can close flow channels and lead to plate over-heating. The critical flow velocity is a speed at which the plates collapse by static instability type failure. In this paper, critical velocity and coolant velocity are evaluated for a typical MTR-type flat plate fuel element. Miller's method is used for prediction of critical velocity. The coolant velocity is limited to 2/3 of the critical velocity, that is a currently used criterion. Fuel plate characteristics are based on the open pool Australian light water reactor. (author)

Convective cooling in a pool-type research reactor

Science.gov (United States)

Sipaun, Susan; Usman, Shoaib

2016-01-01

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

Chemometrics application in fuel's MTR type chemical characterization by X-ray fluorescence

International Nuclear Information System (INIS)

Silva, Clayton Pereira da

2012-01-01

In Brazil and worldwide the nuclear power has occupied a prominent position with many applications in industry, power generation, environment and medicine, improving the quality of tests and treatments, therefore people's lives. Uranium is the main element used in nuclear facilities and it s employed as base material to generation of electricity in the manufacture of radiopharmaceuticals. In the '50s, during the Cold War, the then newly created International Atomic Energy Agency proposed to oversee nuclear facilities and encourage the manufacture of nuclear fuels with low-enriched uranium (LEU) fuel came then type Material Test Reactor (MTR), manufactured initially in U 3 O 8 and U 3 Si 2 later, both dispersed in aluminum. The use of this technology requires a constant improvement of all processes involving the manufacture of MTR subject to several international protocols, which seek to ensure the reliability of the fuel from the standpoint of practical and environmental. In this context, the control of impurities, from the point of view of neutron economy, directly affects the quality of any nuclear fuel, so strict control is necessary. The literature has reported procedures which, beyond generating residues, are lengthy and costly, they need calibration curve and consequently reference materials. The aim of this work is to establish and validate a methodology for nondestructive quantitative chemical analysis, low cost and analysis time, as well as minimize the generation of waste, for multielement determination of major constituents (Utotal and Si) and impurities (B, Mg, Al, Cr, Mn, Fe, Co, Ni, Cu, Zn, Mo, Cd and others) present in U3O8 and U 3 Si 2 , meeting the needs of nuclear reactors in the nuclear fuel qualification type MTR. For that purposes, will be applied the X-ray fluorescence technique which allows fast chemical and nondestructive analysis, aside from sample preparation procedures that do not require previous chemical treatments (dissolving

Shewanella putrefaciens mtrB encodes an outer membrane protein required for Fe(III) and Mn(IV) reduction.

Science.gov (United States)

Beliaev, A S; Saffarini, D A

1998-12-01

Iron and manganese oxides or oxyhydroxides are abundant transition metals, and in aquatic environments they serve as terminal electron acceptors for a large number of bacterial species. The molecular mechanisms of anaerobic metal reduction, however, are not understood. Shewanella putrefaciens is a facultative anaerobe that uses Fe(III) and Mn(IV) as terminal electron acceptors during anaerobic respiration. Transposon mutagenesis was used to generate mutants of S. putrefaciens, and one such mutant, SR-21, was analyzed in detail. Growth and enzyme assays indicated that the mutation in SR-21 resulted in loss of Fe(III) and Mn(IV) reduction but did not affect its ability to reduce other electron acceptors used by the wild type. This deficiency was due to Tn5 inactivation of an open reading frame (ORF) designated mtrB. mtrB encodes a protein of 679 amino acids and contains a signal sequence characteristic of secreted proteins. Analysis of membrane fractions of the mutant, SR-21, and wild-type cells indicated that MtrB is located on the outer membrane of S. putrefaciens. A 5.2-kb DNA fragment that contains mtrB was isolated and completely sequenced. A second ORF, designated mtrA, was found directly upstream of mtrB. The two ORFs appear to be arranged in an operon. mtrA encodes a putative 10-heme c-type cytochrome of 333 amino acids. The N-terminal sequence of MtrA contains a potential signal sequence for secretion across the cell membrane. The amino acid sequence of MtrA exhibited 34% identity to NrfB from Escherichia coli, which is involved in formate-dependent nitrite reduction. To our knowledge, this is the first report of genes encoding proteins involved in metal reduction.

Experience with the transport and storage casks CASTOR (registered) MTR 2 for spent nuclear fuel assemblies from research reactors

Energy Technology Data Exchange (ETDEWEB)

Jack, Allen; Rettenbacher, Katharina; Skrzyppek, Juergen [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

2011-07-01

The CASTOR (registered) MTR 2 cask was designed and manufactured by the company GNS during the 1990's for the transport and interim storage of spent nuclear fuel assemblies from various types of research reactors. Casks of this type have been used at the VKTA Research Centre in Rossendorf near Dresden, Germany as well as at the European Commission's Joint Research Centre at Petten and at the HOR reactor at Delft in the Netherlands. A total of 24 units have been used for the functions of transport and storage with various spent fuel types (VVER, HFR-HEU, and HOR-HEU) for more than ten years now. This type of packaging for radioactive material is a member of the CASTOR (registered) family of spent nuclear fuel casks used worldwide. Over 1000 units are loaded and in storage in Europe, Asia, Africa and North America. This paper presents the experience from the use of the casks for transport and storage in the past, as well as the prospects for the future. (author)

Convective cooling in a pool-type research reactor

Energy Technology Data Exchange (ETDEWEB)

Sipaun, Susan, E-mail: susan@nm.gov.my [Malaysian Nuclear Agency, Industrial Technology Division, Blok 29T, Bangi 43200, Selangor (Malaysia); Usman, Shoaib, E-mail: usmans@mst.edu [Missouri University of Science and Technology, Nuclear Engineering, 222 Fulton Hall 301 W.14th St., Rolla 64509 MO (United States)

2016-01-22

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

MTR fuel testing in BR2

International Nuclear Information System (INIS)

Jacquet, P.; Verwimp, A.; Wirix, S.

2000-01-01

New fuel design for MTR 's requires to be qualified under representative conditions, that is geometry, neutron spectrum, heat flux and thermo hydraulic conditions. An irradiation device for fuel plates has been designed to derive the maximum benefit from the BR2 irradiation capacities. The fuel plates can be easily extracted from their support during a shutdown to undergo additional tests. One of these tests is the measurement of the thickness changes along the fuel plate. To that purpose, a facility in the reactor water pool has been designed to measure the fuel swelling with an accuracy of 5 μm using inductive probes. At SCK-CEN, the full range of destructive and non-destructive PIE can be performed, including γ-scanning, wet sipping, surface examination and other methods. (author)

Livermore pool-type reactor

International Nuclear Information System (INIS)

Mann, L.G.

1977-01-01

The Livermore Pool-Type Reactor (LPTR) has served a dual purpose since 1958--as an instrument for fundamental research and as a tool for measurement and calibration. Our early efforts centered on neutron-diffraction, fission, and capture gamma-ray studies. During the 1960's it was used for extensive calibration work associated with radiochemical and physical measurements on nuclear-explosive tests. Since 1970 the principal applications have been for trace-element measurements and radiation-damage studies. Today's research program is dominated by radiochemical studies of the shorter-lived fission products and by research on the mechanisms of radiation damage. Trace-element measurement for the National Uranium Resource Evaluation (NURE) program is the major measurement application today

Transportation of 33 irradiated MTR fuel assemblies from FRM/Garching to Savannah River Site, USA, using a GNS transport cask and using a new loading device

International Nuclear Information System (INIS)

Dreesen, K.; Goetze, H.G.; Holst, L.; Gerstenberg, H.; Schreckenbach, K.

2000-01-01

According to the Department of Energy program of the return spent fuel from the foreign research reactors operators, 33 irradiated MTR box shaped fuel assemblies from the Technical University Munich were shipped to SRS/USA. The fuel assemblies were irradiated for typically 800 full days and, after a sufficient cooling time, loaded into a GNS 16 cask. The GNS 16 cask is a new transport cask for box shaped MTR fuel assemblies and TRIGA fuel assemblies and was used for the first time at the FRM Garching. The capacity of the cask is 33 box shaped MTR fuel assemblies. During the loading of the fuel assemblies, a newly developed loading device was used. The main components of the loading device are the transfer flask, the shielded loading lock, adapter plate and a mobile water tank. The loading device works mechanically with manpower. For the handling of the transfer flask, a crane with a capacity of 5 metric tons is necessary. During installation of the lid the mobile water pool is filled with demineralized water and the shielded loading passage is taken away. After that the lid is put on the cask. After drainage, the mobile water pool is disassembled, and the cask is dewatered. Finally leak tests of all seals are made. The achieved leakage rate was -5 Pa x I/s. The work in FRM was done between 03.02.99 and 12.02.99 including a dry run and leak test. (author)

Studies of mixed HEU-LEU-MTR cores using 3D models

Energy Technology Data Exchange (ETDEWEB)

Haenggi, P.; Lehmann, E.; Hammer, J.; Christen, R. [Paul Scherrer Institute, Villigen (Switzerland)

1997-08-01

Several different core loadings were assembled at the SAPHIR research reactor in Switzerland combining the available types of MTR-type fuel elements, consisting mainly of both HEU and LEU fuel. Bearing in mind the well known problems which can occur in such configurations (especially power peaking), investigations have been carried out for each new loading with a 2D neutron transport code (BOXER). The axial effects were approximated by a global buckling value and therefore the radial effects could be studied in considerably detail. Some of the results were reported at earlier RERTR meetings and were compared to those obtained by other methods and with experimental values. For the explicit study of the third dimension of the core, another code (SILWER), which has been developed in PSI for LWR power plant cores, has been selected. With the help of an adapted model for the MTR-core of SAPHIR, several important questions have been addressed. Among other aspects, the estimation of the axial contribution to the hot channel factors, the influence of the control rod position and of the Xe-poisoning on the power distribution were studied. Special attention was given to a core position where a new element was assumed placed near a empty, water filled position. The comparison of elements of low and high enrichments at this position was made in terms of the induced power peaks, with explicit consideration of axial effects. The program SILWER has proven to be applicable to MTR-cores for the investigation of axial effects. For routine use as for the support of reactor operation, this 3D code is a good supplement to the standard 2D model.

Computational simulation of the natural circulation occurring in an experimental test section of a pool type research reactor

Energy Technology Data Exchange (ETDEWEB)

Nascimento, Francisco R.T. do; Lima Junior, Carlos A.S.; Oliveira, Andre F.S. de; Affonso, Renato R.W.; Faccini, Jose L.H.; Moreira, Maria L., E-mail: rogerio.tdn@gmail.com, E-mail: souzalima_ca@ien.gov.br, E-mail: oliveira.afelipe@gmail.com, E-mail: raoniwa@yahoo.com.br, E-mail: faccini@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2015-07-01

The present work presents a computational simulation of the natural circulation phenomenon developing in an experimental test section of a pool type research reactor. The test section has been designed using a reduced scale in height 1:4.7 in relation to a pool type 30 MW research reactor prototype. It comprises a cylindrical vessel, which is opened to atmosphere, and representing the reactor pool; a natural circulation pipe, a lower plenum, and a heater containing electrical resistors in rectangular plate format, which represents the fuel elements, with a chimney positioned on the top of the resistor assembly. In the computational simulation, it was used a commercial CFD software, without any turbulence model. Besides, in the presence of the natural circulation, a laminar flow has been assumed and the equations of the mass conservation, momentum and energy were solved by the finite element method. In addition, the results of the simulation are presented in terms of velocities and temperatures differences, respectively: at inlet and outlet of the heater and of the natural circulation pipe. (author)

Computational simulation of the natural circulation occurring in an experimental test section of a pool type research reactor

International Nuclear Information System (INIS)

Nascimento, Francisco R.T. do; Lima Junior, Carlos A.S.; Oliveira, Andre F.S. de; Affonso, Renato R.W.; Faccini, Jose L.H.; Moreira, Maria L.

2015-01-01

The present work presents a computational simulation of the natural circulation phenomenon developing in an experimental test section of a pool type research reactor. The test section has been designed using a reduced scale in height 1:4.7 in relation to a pool type 30 MW research reactor prototype. It comprises a cylindrical vessel, which is opened to atmosphere, and representing the reactor pool; a natural circulation pipe, a lower plenum, and a heater containing electrical resistors in rectangular plate format, which represents the fuel elements, with a chimney positioned on the top of the resistor assembly. In the computational simulation, it was used a commercial CFD software, without any turbulence model. Besides, in the presence of the natural circulation, a laminar flow has been assumed and the equations of the mass conservation, momentum and energy were solved by the finite element method. In addition, the results of the simulation are presented in terms of velocities and temperatures differences, respectively: at inlet and outlet of the heater and of the natural circulation pipe. (author)

In-pile modelling of nuclear fuel element for the MTR type reactors. Pt. 2

Energy Technology Data Exchange (ETDEWEB)

Farhadi, Kazem [AEOI, Tehran (Iran, Islamic Republic of). Radiations Application Research School

2014-06-15

In part two of the present paper, neutronic properties of the pool-type research reactor core are used to assess the similitude laws derived for out-of-pile modelling of the fuel element. The benchmark reactor used for this purpose is an IAEA 5 MW thermal pool-type research reactor currently in operation. The neutronic properties analysis are based on typical 2 200 m/sec and neutrons having 0.025 eV energy. The non-leakage capability of the system is estimated in terms of diffusion length. Also the slowing down power and the moderating ratio of the modelled methanol coolant are calculated in terms of lethargy of the diffusing medium. It is shown that the Iron which is substituted for Aluminium cladding is a relatively low absorber of neutrons but has a high neutron leakage. Methanol which replaced ordinary water as coolant is not a suitable coolant due to high neutrons absorbing substance. It is concluded that although Iron as a cladding material and methanol as a coolant meet the modelling out-of-pile criteria but are not satisfying neutronic properties. Therefore, use of them as a model clad and coolant are not suggested for research reactors. (orig.)

In-service inspection of pool type research reactors

International Nuclear Information System (INIS)

Rajamani, K.

2002-01-01

In the case of Apsara Reactor, it has been proposed to carry out major modifications in the near future. It is planned to modify the core suitably with a heavy water reflector tank to demonstrate the Multiple Purpose Research Reactor concept. The core structure will be a stationary one and will be located at the 'B' position of the pool. The modified reactor will be operated at 1 MW power level. Suitable methodologies are evolved for carrying out a planned ISI for this modified reactor

MTR loop at the MPR-GA. Siwabessy reactor of Serpong Indonesia for testing of LEU fuel

International Nuclear Information System (INIS)

Arbie, B.; Sunaryadi, D.; Supadi, S.

1991-01-01

The main objective of the MTR-Loop is for testing the specimens of MTR fuel element uprated conditions with respect to the normal conditions of the reactor fuel elements. It is intended to verify the suitability of the fuel elements for operation in a research reactor under preset temperature and pressure conditions. The most important part of the MTR loop is the test section. The fuel elements to be tested are positioned in the test section. For heat removal there is a cooling water flowing through the test section. On this paper the description of the MTR-Loop is described. Installation of the MTR-Loop will be performed in the middle of 1990. In order to facilitate the investigation of fuel behaviour and performance of the new fuel elements the supporting facilities are also already available in the RSG-GAS. (orig.)

Establishing a LEU MTR fuel manufacturing facility in South Africa

International Nuclear Information System (INIS)

Jamie, R.W.; Kocher, A.

2010-01-01

The South African MTR Fuel Manufacturing Facility was established in the 1970's to supply SAFARI-1 with Fuel Elements and Control Rods. South African capability was developed in parallel with the uranium enrichment program to meet the needs of the Reactor. Further to the July 2005 decision by the South African Governmnent to convert both SAFARI-1 and the Fuel Plant to LEU, the SAFARI-1 phase has been successfully completed and Necsa has commenced with the conversion of the MTR Fuel Manufacturing Facility. In order to establish, validate and qualify the facility, Necsa has entered into a co-operation and technology transfer agreement with AREVA CERCA, the French manufacturer of Research Reactor fuel elements. Past experiences, conversion challenges and the status of the MTR Fuel Facility Project are discussed. On-going co-operation with AREVA CERCA to implement the local manufacture of LEU fuel is explained and elaborated on. (author)

The Transcriptional Repressor, MtrR, of the mtrCDE Efflux Pump Operon of Neisseria gonorrhoeae Can Also Serve as an Activator of “off Target” Gene (glnE Expression

Directory of Open Access Journals (Sweden)

Paul J. T. Johnson

2015-06-01

Full Text Available MtrR is a well-characterized repressor of the Neisseria gonorrhoeae mtrCDE efflux pump operon. However, results from a previous transcriptional profiling study suggested that MtrR also represses or activates expression of at least sixty genes outside of the mtr locus. Evidence that MtrR can directly repress so-called “off target” genes has previously been reported; in particular, MtrR was shown to directly repress glnA, which encodes glutamine synthetase. In contrast, evidence for the ability of MtrR to directly activate expression of gonococcal genes has been lacking; herein, we provide such evidence. We now report that MtrR has the ability to directly activate expression of glnE, which encodes the dual functional adenyltransferase/deadenylase enzyme GlnE that modifies GlnA resulting in regulation of its role in glutamine biosynthesis. With its capacity to repress expression of glnA, the results presented herein emphasize the diverse and often opposing regulatory properties of MtrR that likely contributes to the overall physiology and metabolism of N. gonorrhoeae.

Study on the Safety Classification Criteria of Mechanical Systems and Components for Open Pool-Type Research Reactors

International Nuclear Information System (INIS)

Belal, Al Momani; Jo, Jong Chull

2013-01-01

This paper describes a new compromised safety classification approach based on the comparative study of the different practices in safety classification of mechanical systems and components of open pool-type RRs, which have been adopted by several developed countries in the nuclear power area. It is hoped that the proposed safety classification criteria will be used to develop a harmonized consensus international standard. Different safety classification criteria for systems, structures, and components (SSCs) of nuclear reactors are used among the countries that export or import nuclear reactor technology, which may make the nuclear technology trade and exchange difficult. Thus, such various different approaches of safety classification need to be compromised to establish a global standard. This article proposes practicable optimized criteria for safety classification of SSCs for open pool-type research reactors (RRs)

Study on the Safety Classification Criteria of Mechanical Systems and Components for Open Pool-Type Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Belal, Al Momani [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Jo, Jong Chull [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

This paper describes a new compromised safety classification approach based on the comparative study of the different practices in safety classification of mechanical systems and components of open pool-type RRs, which have been adopted by several developed countries in the nuclear power area. It is hoped that the proposed safety classification criteria will be used to develop a harmonized consensus international standard. Different safety classification criteria for systems, structures, and components (SSCs) of nuclear reactors are used among the countries that export or import nuclear reactor technology, which may make the nuclear technology trade and exchange difficult. Thus, such various different approaches of safety classification need to be compromised to establish a global standard. This article proposes practicable optimized criteria for safety classification of SSCs for open pool-type research reactors (RRs)

Chemometrics application in fuel's MTR type chemical characterization by X-ray fluorescence; Aplicacao da quimiometria para caracterizacao quimica de combustiveis tipo MTR por fluorescencia de raios-X

Energy Technology Data Exchange (ETDEWEB)

Silva, Clayton Pereira da

2012-07-01

In Brazil and worldwide the nuclear power has occupied a prominent position with many applications in industry, power generation, environment and medicine, improving the quality of tests and treatments, therefore people's lives. Uranium is the main element used in nuclear facilities and it s employed as base material to generation of electricity in the manufacture of radiopharmaceuticals. In the '50s, during the Cold War, the then newly created International Atomic Energy Agency proposed to oversee nuclear facilities and encourage the manufacture of nuclear fuels with low-enriched uranium (LEU) fuel came then type Material Test Reactor (MTR), manufactured initially in U{sub 3}O{sub 8} and U{sub 3}Si{sub 2} later, both dispersed in aluminum. The use of this technology requires a constant improvement of all processes involving the manufacture of MTR subject to several international protocols, which seek to ensure the reliability of the fuel from the standpoint of practical and environmental. In this context, the control of impurities, from the point of view of neutron economy, directly affects the quality of any nuclear fuel, so strict control is necessary. The literature has reported procedures which, beyond generating residues, are lengthy and costly, they need calibration curve and consequently reference materials. The aim of this work is to establish and validate a methodology for nondestructive quantitative chemical analysis, low cost and analysis time, as well as minimize the generation of waste, for multielement determination of major constituents (Utotal and Si) and impurities (B, Mg, Al, Cr, Mn, Fe, Co, Ni, Cu, Zn, Mo, Cd and others) present in U3O8 and U{sub 3}Si{sub 2}, meeting the needs of nuclear reactors in the nuclear fuel qualification type MTR. For that purposes, will be applied the X-ray fluorescence technique which allows fast chemical and nondestructive analysis, aside from sample preparation procedures that do not require previous chemical

Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

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

Development of Fusion Nuclear Technologies and the role of MTR's

International Nuclear Information System (INIS)

Laan, J.G. van der; Schaaf, B. van der

2006-01-01

design for the EU ITER Test Blanket Module. The duration of the irradiation is relevant for the total TBM operation during the ITER lifetime. The major result is that the basic design soundness has been demonstrated under ITER relevant conditions. Besides the ceramic breeder concept experiments with lithium lead breeder sub components are continued to measure the effects of transmutation product helium on the liquid metal properties. Similarly, activities are ongoing to perform in-pile testing of primary wall components, allowing to address fatigue type loading conditions. In the next decade 14 MeV sources such as ITER, IFMIF and maybe a volumetric source will support the crucial demonstration of components under near fusion plasma nuclear conditions. These sources have limitations in accumulated total damage (ITER) irradiation volume (IFMIF) and control. MTR's will thus continue to supply essential facts on component behaviour and materials in parallel to 14 MeV sources. The present generation of MTR's will be closed in this and next decade because they reach their end of life. The new generation will be utilised for 4 major areas of nuclear interest: energy, science, health and environmental issues. Fusion and the next generation fission (Generation 4) power plant development will share the areas energy and science in the next decades. The design and concept of the new MTR's will centre on faster development cycles, thus higher fluxes up to 5 x 10 18 nm -2 . Several MTR replacements in the EU are in different design stages such as the Reacteur Jules Horowitz in France and PALLAS in the Netherlands. The conceptual design of the replacement for the HFR, Petten, named PALLAS envisages a fruitful co-operation of the experimenters for advanced fission power reactor and fusion plant components. Materials science will also be able to use modern MTR facilities for the modelling of radiation damage in both fission and fusion environments. The development of primary fusion

Application of MTR soft-decision decoding in multiple-head ...

Indian Academy of Sciences (India)

basic MTR logic circuits, and to develop, a new one, the soft-decision MTR decoder, based on such ... of integrated circuits provides their quite simple realization. ..... recording channels, PSU-UNS International Conference on Engineering and ...

The Conserved Actinobacterial Two-Component System MtrAB Coordinates Chloramphenicol Production with Sporulation in Streptomyces venezuelae NRRL B-65442

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Nicolle F. Som

2017-06-01

Full Text Available Streptomyces bacteria make numerous secondary metabolites, including half of all known antibiotics. Production of antibiotics is usually coordinated with the onset of sporulation but the cross regulation of these processes is not fully understood. This is important because most Streptomyces antibiotics are produced at low levels or not at all under laboratory conditions and this makes large scale production of these compounds very challenging. Here, we characterize the highly conserved actinobacterial two-component system MtrAB in the model organism Streptomyces venezuelae and provide evidence that it coordinates production of the antibiotic chloramphenicol with sporulation. MtrAB are known to coordinate DNA replication and cell division in Mycobacterium tuberculosis where TB-MtrA is essential for viability but MtrB is dispensable. We deleted mtrB in S. venezuelae and this resulted in a global shift in the metabolome, including constitutive, higher-level production of chloramphenicol. We found that chloramphenicol is detectable in the wild-type strain, but only at very low levels and only after it has sporulated. ChIP-seq showed that MtrA binds upstream of DNA replication and cell division genes and genes required for chloramphenicol production. dnaA, dnaN, oriC, and wblE (whiB1 are DNA binding targets for MtrA in both M. tuberculosis and S. venezuelae. Intriguingly, over-expression of TB-MtrA and gain of function TB- and Sv-MtrA proteins in S. venezuelae also switched on higher-level production of chloramphenicol. Given the conservation of MtrAB, these constructs might be useful tools for manipulating antibiotic production in other filamentous actinomycetes.

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

International Nuclear Information System (INIS)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

2015-01-01

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)

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)

Use of heterogeneous finite elements generated by collision probability solutions to calculate a pool reactor core

International Nuclear Information System (INIS)

Calabrese, C.R.; Grant, C.R.

1990-01-01

This work presents comparisons between measured fluxes obtained by activation of Manganese foils in the light water, enriched uranium research pool reactor RA-2 MTR (Materials Testing Reactors) fuel element) and fluxes calculated by the finite element method FEM using DELFIN code, and describes the heterogeneus finite elements by a set of solutions of the transport equations for several different configurations obtained using the collision probability code HUEMUL. The agreement between calculated and measured fluxes is good, and the advantage of using FEM is showed because to obtain the flux distribution with same detail using an usual diffusion calculation it would be necessary 12000 mesh points against the 2000 points that FEM uses, hence the processing time is reduced in a factor ten. An interesting alternative to use in MTR fuel management is presented. (Author) [es

Conditioning of spent fuel assemblies from the Rossendorf RFR research reactor in transport and storage containers of the type CASTOR MTR 2

International Nuclear Information System (INIS)

Schneider, B.; Hofmann, G.

1994-09-01

Most of the spent fuel assemblies are temporarily stored in the flooded fuel ponds AB 1 and AB 2 of the RFR, and some are still in the reactor core. The conditioning task described here is part of the RFR spent fuel management concept and covers the safe emplacement of the spent fuel elements in the CASTOR MTR 2 shipping containers and the sealing of the containers in compliance with the nuclear licence issued for the conditioning task. The transfer of the spent fuel assemblies from the present wet storage conditions to the dry storage conditions in the CASTOR MTR 2 containers is done by a mobile manipulation equipment consisting essentially of the transfer sluice gate and a transfer container. Subsequent to conditioning, the shipping containers are to be transported to a licensed intermediate storage facility to await their transport to a national radwaste repository. The technical handling tools for the transfer and manipulation are briefly described, as well as the process steps involved, putting emphasis on the detailed description of processes and the accompanying time frame, so that the conditioning task can be incorporated into the work plan of the entire project. The report further presents the EDP concept established for the task, including the required data archivation and documentation. (orig.) [de

Effect of reactivity insertion rate on peak power and temperatures in swimming pool type research reactor

International Nuclear Information System (INIS)

Khan, L.A.; Jabbar, A.; Anwar, A.R.; Ahmad, N.

1998-01-01

It is essential to study the reactor behavior under different accidental conditions and take proper measures for its safe operation. We have studied the effect of reactivity insertion, with and without scram conditions, on peak power and temperatures of fuel, cladding and coolant in typical swimming pool type research reactor. The reactivity ranging from 1 $ to 2 $ and insertion times from 0.25 to 1 second have been considered. The computer code PARET has been used and results are presented in this article. (author)

Chemometrics application in fuel's MTR type chemical characterization by X-ray fluorescence; Aplicacao da quimiometria para caracterizacao quimica de combustiveis tipo MTR por fluorescencia de raios-X

Energy Technology Data Exchange (ETDEWEB)

Silva, Clayton Pereira da

2012-07-01

In Brazil and worldwide the nuclear power has occupied a prominent position with many applications in industry, power generation, environment and medicine, improving the quality of tests and treatments, therefore people's lives. Uranium is the main element used in nuclear facilities and it s employed as base material to generation of electricity in the manufacture of radiopharmaceuticals. In the '50s, during the Cold War, the then newly created International Atomic Energy Agency proposed to oversee nuclear facilities and encourage the manufacture of nuclear fuels with low-enriched uranium (LEU) fuel came then type Material Test Reactor (MTR), manufactured initially in U{sub 3}O{sub 8} and U{sub 3}Si{sub 2} later, both dispersed in aluminum. The use of this technology requires a constant improvement of all processes involving the manufacture of MTR subject to several international protocols, which seek to ensure the reliability of the fuel from the standpoint of practical and environmental. In this context, the control of impurities, from the point of view of neutron economy, directly affects the quality of any nuclear fuel, so strict control is necessary. The literature has reported procedures which, beyond generating residues, are lengthy and costly, they need calibration curve and consequently reference materials. The aim of this work is to establish and validate a methodology for nondestructive quantitative chemical analysis, low cost and analysis time, as well as minimize the generation of waste, for multielement determination of major constituents (Utotal and Si) and impurities (B, Mg, Al, Cr, Mn, Fe, Co, Ni, Cu, Zn, Mo, Cd and others) present in U3O8 and U{sub 3}Si{sub 2}, meeting the needs of nuclear reactors in the nuclear fuel qualification type MTR. For that purposes, will be applied the X-ray fluorescence technique which allows fast chemical and nondestructive analysis, aside from sample preparation procedures that do not require previous

Primary system thermal-hydraulic simulation of a experimental pool type research fast reactor

International Nuclear Information System (INIS)

Borges, E.M.; Braz Filho, F.A.

1993-01-01

The first step of the Fast Reactor Program (REARA) is the design of an experimental reactor. To this end a 5 MW t pool type reactor was adapted. The objective of this work is to evaluate the reactor behaviour at the on set protected accidents. The program NALAP was used in this study and the results showed the outstanding safety margins that this reactor type presents inherently. (author)

Reprocessing of MTR fuel at Dounreay

International Nuclear Information System (INIS)

Hough, N.

1997-01-01

UKAEA at Dounreay has been reprocessing MTR fuel for over 30 years. During that time considerable experience has been gained in the reprocessing of traditional HEU alloy fuel and more recently with dispersed fuel. Latterly a reprocessing route for silicide fuel has been demonstrated. Reprocessing of the fuel results in a recycled uranium product of either high or low enrichment and a liquid waste stream which is suitable for conditioning in a stable form for disposal. A plant to provide this conditioning, the Dounreay Cementation Plant is currently undergoing active commissioning. This paper details the plant at Dounreay involved in the reprocessing of MTR fuel and the treatment and conditioning of the liquid stream. (author)

Does Magnetization Transfer Ratio (MTR) contribute to the diagnosis and differential diagnosis of the dementias?

International Nuclear Information System (INIS)

Hentschel, F.; Kreis, M.; Damian, M.; Krumm, B.

2004-01-01

Purpose: The magnetization transfer ratio (MTR) is a MR-based neuroimaging procedure aiming at the quantification of the structural integrity of brain tissue. Its contribution to the differential diagnosis of dementias was examined and discussed in relation to the pathogenesis of age-related dementias. Materials and Methods: Sixty-one patients from a memory clinic were diagnosed by general physical and neuropsychiatric examination, and underwent neuropsychologic testing and neuroimaging using MRI. Their clinical diagnoses were based on standard operational research criteria. Additionally, the MTR in 10 defined regions of interest (ROI) was determined. This investigation was performed using a T1-weighted SE sequence. Average MTR values were determined in the individual ROI and their combinations and correlated with the age gender, cognitive impairment and clinical diagnosis. Sensitivity, specificity, positive and negative predictive value were determined, as well as the rate of correct classifications. Results: For cognitive healthy subjects, the MRT values correlate only mildly, though significantly, with age in the hippocampus and with gender in the dorsal corpus callosum. In contrast, the MTR in the frontal white matter correlates strongly and highly significantly with cognitive impairment in patients with dementia. The differential diagnostic assignment of Alzheimer's disease versus vascular dementia by MTR provides a correct classification of approximately 50% to 70%. PPV for no dementia vs. vascular dementia or the NPV for vascular vs. Alzheimer's disease are considerably higher exceeding 80%. For no dementia vs. Alzheimer's disease, the NPV was over 90%. (orig.)

MTR fuel inspection at CERCA

International Nuclear Information System (INIS)

Fanjas, Y.

1992-01-01

The stringent specifications for MTR fuel plates and fuel elements require various sophisticated inspection techniques. In particular, the development of low enriched silicide fuels made it necessary to adapt these techniques to high density plates. This paper presents the status of inspection technology at CERCA. (author)

Effect of coolant flow rate on the power at onset of nucleate boiling in a swimming pool type research reactor

International Nuclear Information System (INIS)

Khan, L.A.; Ahmad, N.; Ahmad, S.

1998-01-01

The effect of flow rate of coolant on power of Onset Nucleate Boiling (ONB) in a reference core of a swimming pool type research reactor has been studied using a as standard computer code PARET. It has been found that the decrease in the coolant flow rate results in a corresponding decrease in power at ONB. (author)

Thermal-hydraulic safety aspects related to irradiation capabilities in MTR reactors

International Nuclear Information System (INIS)

Khedr, A.

2009-01-01

MTR research reactor such as ETRR-2 is an open pool type reactor that has a capability for irradiation into a number of irradiation boxes (IBs) installed at different positions on a separate grid called irradiation grid (I G). The I B has a lower removable plug to open or close its lower nozzle according to the I B is used or not.Increasing the used No. of I Bs in irradiation means that a valuable change in the flow distribution on the I G will occur. This paper is focused on the optimum number of I Bs that could be used without deterioration the cooling of I G components and avoiding the formation of hot spots. RELAP5 system code is used for thermal hydraulic analysis of the I G cooling system. Mathematical models and fortran program is developed to calculate the heat distribution in the I G components and the equivalent nozzle diameter that compensate the I B pressure drop due to the irradiated material (I M). This equivalent diameter simulates the used I B nozzle in the RELAP5 input deck. The results show that, the internal flow into the I Bs has significant effect on the coolability of the I G components. The number of I Bs that can be used is inversely proportional with the reactor power, the IM's void fraction and directly proportional with the PCS flow rate. Different cases of operating power and void fraction at two values for PCS flow are studied. In all of the cases considered limited number of the I Bs is permissible to use in order to avoid the excessive heating of the I G components

ANALISIS POLA MANAJEMEN BAHAN BAKAR DESAIN TERAS REAKTOR RISET TIPE MTR

Directory of Open Access Journals (Sweden)

Lily Suparlina

2015-03-01

codes. Research reactor MTR type is very interested because can be usd as research and also a radioisotope production. The research reactor in Indonesia right now is already 25 years old. Therefore, it is needed to design a new research reactor as a compact core. Recent research reactor core is not enough to meet criteria acceptance in the UCD which already determined namely thernmal neutron flux in the core is 1.0x1015 n/cm2s. so that it is necessary to be redesign the alternative core design. The new research reactor design is a MTR type with 5x5 configuration core, uses U9Mo-Al fuel, 70 cm of hight and uses two certainly fuel management pattern. The aim of this research is to achieve neutron flux in the core to meet the criteria acceptance in the UCD. Calculation is done by using WIMSD-B, Batan-FUEL and Batan-3DIFF codes. The neutronic parameters to be achieved by this calculation are the power level of 50 MW thermal and core cycle of 20 days. The neutronis parameter calculation is done for new U-9Mo-Al fuel with variation of densities. The result of calculation showed that the fresh core with 5x5 configuration, 360 gram, 390 gram and 450 gram of fuel loadings have meet safety margin and acceptance criteria in the UCD at the thermal neutron flux is more then 1.0x1015 n/cm2s. But for equilibrium core is only the 450 gram of loading meet the acceptance criteria. Keywords: reactor core design design, UMo, fuel management pattern, WIMS, BATAN-FUEL

Multi-target retrieval (MTR): the simultaneous retrieval of pressure, temperature and volume mixing ratio profiles from limb-scanning atmospheric measurements

International Nuclear Information System (INIS)

Dinelli, B.M.; Alpaslan, D.; Carlotti, M.; Magnani, L.; Ridolfi, M.

2004-01-01

In this paper we describe a retrieval approach for the simultaneous determination of the altitude distributions of p, T and VMR of atmospheric constituents from limb-scanning measurements of the atmosphere. This analysis method, named multi-target retrieval (MTR), has been designed and implemented in a computer code aimed at the analysis of MIPAS-ENVISAT observations; however, the concepts implemented in MTR have a general validity and can be extended to the analysis of all type of limb-scanning observations. In order to assess performance and advantages of the proposed approach, MTR has been compared with the sequential analysis system implemented by ESA as the level-2 processor for MIPAS measurements. The comparison has been performed on a common set of target species and spectral intervals. The performed tests have shown that MTR produces results of better quality than a sequential retrieval. However, the simultaneous retrieval of p, T and water VMR has not lead to satisfactory results below the tropopause, because of the high correlation occurring between p and water VMR in the troposphere. We have shown that this problem can be fixed extending the MTR analysis to at least one further target whose spectral features decouple the retrieval of pressure and water VMR. Ozone was found to be a suitable target for this purpose. The advantages of the MTR analysis system in terms of systematic errors have also been discussed

Prediction, analysis and solution of flow inversion phenomenon in a typical MTR reactor with upward core cooling

International Nuclear Information System (INIS)

El-Morshedy, Salah El-Din

2010-01-01

Research reactors of power greater than 20 MW are usually designed to be cooled with upward coolant flow direction inside the reactor core. This is mainly to prevent flow inversion problems following a pump coast down. However, in some designs and under certain operating conditions, flow inversion phenomenon is predicted. In the present work, the best-estimate Material Testing Reactors Thermal-Hydraulic Analysis program (MTRTHA) is used to simulate a typical MTR reactor behavior with upward cooling under a hypothetical case of loss of off-site power. The flow inversion phenomenon is predicted under certain decay heat and/or pool temperature values below the design values. The reactor simulation under loss of off-site power is performed for two cases namely; two-flap valves open and one flap-valve fails to open. The model results for the flow inversion phenomenon prediction is analyzed and a solution of the problem is suggested. (orig.)

Status of development and irradiation performance of advanced proliferation resistant MTR fuel at NUKEM

International Nuclear Information System (INIS)

Hrovat, M.; Hassel, H.-W.; Wehner, E.

1985-01-01

This paper describes the current status of development and irradiation performance of fuel elements for Material Test and Research (MTR) Reactors with Medium Enriched Uranium (MEU, ≤ 45 % 235-U) and Low Enriched Uranium (LEU, ≤ 20 % 235-U). (author)

Cost of the external MTR-fuel cycle. (Uranium , reprocessing and related services)

International Nuclear Information System (INIS)

Mueller, H.; Gruber, G.

1991-01-01

This paper points out how the RERTR program has affected NUKEM's fuel supplies for MTRs and how the prices in the External MTR Fuel Cycle have developed during this period. In addition other potential fuel sources and services on the External MTR Fuel Cycle are given. (orig.)

In-vivo identification of direct electron transfer from Shewanella oneidensis MR-1 to electrodes via outer-membrane OmcA-MtrCAB protein complexes

Energy Technology Data Exchange (ETDEWEB)

Okamoto, Akihiro [Department of Applied Chemistry, School of Engineering, University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Nakamura, Ryuhei, E-mail: nakamura@light.t.u-tokyo.ac.jp [Department of Applied Chemistry, School of Engineering, University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Hashimoto, Kazuhito, E-mail: hashimoto@light.t.u-tokyo.ac.jp [Department of Applied Chemistry, School of Engineering, University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); ERATO/JST, HASHIMOTO Light Energy Conversion Project (Japan)

2011-06-30

Graphical abstract: . Display Omitted Highlights: > Monolayer biofilm of Shewanella cells was prepared on an ITO electrode. > Extracellular electron transfer (EET) process was examined with series of mutants. > Direct ET was confirmed with outer-membrane-bound OmcA-MtrCAB complex. > The EET process was not prominently influenced by capsular polysaccharide. - Abstract: The direct electron-transfer (DET) property of Shewanella bacteria has not been resolved in detail due to the complexity of in vivo electrochemistry in whole-cell systems. Here, we report the in vivo assignment of the redox signal indicative of the DET property in biofilms of Shewanella oneidensis MR-1 by cyclic voltammetry (CV) with a series of mutants and a chemical marking technique. The CV measurements of monolayer biofilms formed by deletion mutants of c-type cytochromes ({Delta}mtrA, {Delta}mtrB, {Delta}mtrC/{Delta}omcA, and {Delta}cymA), and pilin ({Delta}pilD), capsular polysaccharide ({Delta}SO3177) and menaquinone ({Delta}menD) biosynthetic proteins demonstrated that the electrochemical redox signal with a midpoint potential at 50 mV (vs. SHE) was due to an outer-membrane-bound OmcA-MtrCAB protein complex of decaheme cytochromes, and did not involve either inner-membrane-bound CymA protein or secreted menaquinone. Using the specific binding affinity of nitric monoxide for the heme groups of c-type cytochromes, we further confirmed this conclusion. The heterogeneous standard rate constant for the DET process was estimated to be 300 {+-} 10 s{sup -1}, which was two orders of magnitude higher than that previously reported for the electron shuttling process via riboflavin. Experiments using a mutant unable to produce capsular polysaccharide ({Delta}SO3177) revealed that the DET property of the OmcA-MtrCAB complex was not influenced by insulating and hydrophilic extracellular polysaccharide. Accordingly, under physiological conditions, S. oneidensis MR-1 utilizes a high density of outer

Combining different views of mammographic texture resemblance (MTR) marker of breast cancer risk

DEFF Research Database (Denmark)

Sun, S.; Karemore, Gopal; Chernoff, Konstantin

the subsequent 4 years whereas 245 cases had a diagnosis 2-4 years post mammography. We employed the MTR supervised texture learning framework to perform risk evaluation from a single mammography view. In the framework 20,000 pixels were sampled and classified by a kNN pixel classifier. A feature selection step......PURPOSE Mammographic density is a well established breast cancer risk factor. Texture analysis in terms of the Mammographoc Texture Resemblance (MTR) marker has recently shown to add to risk segregation. Hitherto only single view MTR analysis has been performed. Standard mammography examinations...

MTR core loading pattern optimization using burnup dependent group constants

Directory of Open Access Journals (Sweden)

Iqbal Masood

2008-01-01

Full Text Available A diffusion theory based MTR fuel management methodology has been developed for finding superior core loading patterns at any stage for MTR systems, keeping track of burnup of individual fuel assemblies throughout their history. It is based on using burnup dependent group constants obtained by the WIMS-D/4 computer code for standard fuel elements and control fuel elements. This methodology has been implemented in a computer program named BFMTR, which carries out detailed five group diffusion theory calculations using the CITATION code as a subroutine. The core-wide spatial flux and power profiles thus obtained are used for calculating the peak-to-average power and flux-ratios along with the available excess reactivity of the system. The fuel manager can use the BFMTR code for loading pattern optimization for maximizing the excess reactivity, keeping the peak-to-average power as well as flux-ratio within constraints. The results obtained by the BFMTR code have been found to be in good agreement with the corresponding experimental values for the equilibrium core of the Pakistan Research Reactor-1.

PcMtr, an aromatic and neutral aliphatic amino acid permease of Penicillium chrysogenum

NARCIS (Netherlands)

Trip, H; Evers, ME; Driessen, AJM

2004-01-01

The gene encoding an aromatic and neutral aliphatic amino acid permease of Penicillium chrysogenum was cloned, functionally expressed and characterized in Saccharomyces cerevisiae M4276. The permease, designated PcMtr, is structurally and functionally homologous to Mtr of Neurospora crassa, and

Performance of the MTR core with MOX fuel using the MCNP4C2 code

International Nuclear Information System (INIS)

Shaaban, Ismail; Albarhoum, Mohamad

2016-01-01

The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U 3 O 8 &PuO 2 ) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U 3 O 8 -Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U 3 O 8 -Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with 235 U and the amount of loaded 235 U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. - Highlights: • Re-cycling of the ETRR-2 reactor by MOX fuel. • Increase the number of the neutronic traps from one neutronic trap to three neutronic trap. • Calculation of the criticality safety and neutronic parameters of the ETRR-2 reactor for the U 3 O 8 -Al original fuel and the MOX fuel.

MTR fuel element burn-up measurements by the reactivity method

International Nuclear Information System (INIS)

Zuniga, A.; Cuya, T.R.; Ravnik, M.

2003-01-01

Fuel element burn-up was measured by the reactivity method in the 10 MW Peruvian MTR reactor RP-10. The main purpose of the experiment was testing the reactivity method for an MTR reactor as the reactivity method was originally developed for TRIGA reactors. The reactivity worth of each measured fuel element was measured in its original core position in order to measure the burn-up of the fuel elements that were part of the experimental core. The burn-up of each measured fuel element was derived by interpolating its reactivity worth from the reactivity worth of two reference fuel elements of known burn-up, whose reactivity worth was measured in the position of the measured fuel element. The accuracy of the method was improved by separating the reactivity effect of burn-up from the effect of the position in the core. The results of the experiment showed that the modified reactivity method for fuel element burn-up determination could be applied also to MTR reactors. (orig.)

Sipping test on a failed MTR fuel element

International Nuclear Information System (INIS)

Terremoto, Luis Antonio Albiac; Zeituni, Carlos Alberto; Silva, Antonio Teixeira e; Perrotta, Jose Augusto; Silva, Jose Eduardo Rosa da

2002-01-01

This work describes sipping tests performed on MTR fuel elements of the IEA-R1 research reactor, in order to determinate which one failed in the core during a routine operation of the reactor. radioactive iodine isotopes 131 I and 133 I, employed as failure indicators, were detected in samples corresponding to the fuel element IEA-156. The specific activity of each sample, as well as the average leaking rate, were measured for 137 Cs. The nuclear fuels U 3 O 8 - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of 137 Cs. (author)

Long term immersion test of aluminum alloy AA 6061 used for fuel cladding in MTR type reactors

International Nuclear Information System (INIS)

Linardi, Evelina M.; Rodriguez, Sebastian; Haddad, Roberto; Lanzani, Liliana

2009-01-01

In this work we present the results of long term immersion tests performed in the aluminum alloy AA 6061, used for fuel cladding in MTR type reactors. The tests were performed at open circuit potential in high purity water (ρ = 18.2 MΩ.cm) and in 10 -3 M NaCl solution. Two kinds of assemblies were studied: simple sheets and artificial crevices, immersed during 6, 12 and 18 months at room temperature. In both media and both assemblies, the aluminum hydroxide phases crystalline bayerite and bohemite were identified. It was found that a kind of localized attack named alkaline attack occurs around the iron-rich intermetallics. These particles were confirmed to control the corrosion of the AA 6061 alloy in an aerated medium. Immersion times for up to 18 months did not increase the oxide growth or the alkaline attack on the AA 6061 alloy. (author)

Conceptual Nuclear Design Of Two Models Of Research Reactor Proposed For Vietnam

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Huynh Ton Nghiem; Le Vinh Vinh; Vo Doan Hai Dang

2007-01-01

The joint study on the development of a new research reactor model for Vietnam was done. The KAERI (Korea Atomic Energy Research Institute) experts and DNRI (Dalat Nuclear Research Institute) researchers developed an advanced HANARO reactor (AHR), a 20-MW open-tank-in-pool type reactor, upward cooled and moderated by light water, reflected by heavy water and rod type fuel assemblies used. Based on the AHR model, a MTR reactor with plate fuel assemblies was developed. Computer codes named MCNP and MVP/BURN were used. Major analyses have been done for the relevant nuclear design parameters such as the neutron flux and power distributions, reactivity coefficients, control rod worth, etc. in both with clean, unperturbed core and equilibrium core condition. In case of AHR model, calculation results using MVP/BURN and MCNP codes were compared with the results using HELIOS and MCNP codes by KAERI experts and they are in a good agreement. (author)

Operation experience of the Indonesian multipurpose research reactor RSG-GAS

Energy Technology Data Exchange (ETDEWEB)

Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

1999-08-01

RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

Operation experience of the Indonesian multipurpose research reactor RSG-GAS

International Nuclear Information System (INIS)

Hastowo, Hudi; Tarigan, Alim

1999-01-01

RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U 3 O 8 -Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

Reliability assessment of emergency exhaust system in a pool-type research reactor

International Nuclear Information System (INIS)

Khan, S.A.

1991-01-01

The reliability of an extract system in a swimming-pool-type research reactor has been assessed. A global fault-tree analysis technique has been utilized. The basic event reliability data is based on both generic and reactor specific informations. The unavailability of the extract system is quantified in terms of the unavailability of the various functional requirements of the system. The unavailability is expressed as the probability of failure on demand. The computer system unavailability is determined from the minimal cutsets of the system. It is found that only three events have a major contribution to the top event, i.e., failures of compressed air supply, electric power supply and solenoid valve. A sensitivity analysis is performed to show the effects of variations in the data values of the dominant cutsets. An uncertainty analysis was also performend on the fault tree. The evaluations show that the reactor extract system lacks diversity and redundance in most of its components. It is tolerant of most minor degradations, as these are taken care of by the operating policies and procedures. However, it can not tolerate common cause failures, e.g. simultaneous compressed air and electric power supply failure. Based upon the results obtained, some recommendations are made. (orig.)

Swimming-pool piles

International Nuclear Information System (INIS)

Trioulaire, M.

1959-01-01

In France two swimming-pool piles, Melusine and Triton, have just been set in operation. The swimming-pool pile is the ideal research tool for neutron fluxes of the order of 10 13 . This type of pile can be of immediate interest to many research centres, but its cost must be reduced and a break with tradition should be observed in its design. It would be an advantage: - to bury the swimming-pool; - to reject the experimental channel; - to concentrate the cooling circuit in the swimming-pool; - to carry out all manipulations in the water; - to double the core. (author) [fr

Flow inversion and natural convection in a MTR (Materials Testing Reactor)

International Nuclear Information System (INIS)

Gimenez, M.O.; Clausse, A.

1990-01-01

The thermohydraulic evolution of a refrigerating channel of the MTR (Materials Testing Reactors) RA-6 reactor's core, at the Bariloche Atomic Center, has been studied during the transient caused by the primary system's pump decommissioning. This transient constitutes one of the reactor's operating power boundaries due to the maximum temperature permissible in fuel plates. The problem regarding the thermohydraulic code altered for the rectangular geometry calculation characteristic of the MTR design is analyzed. (Author) [es

Establishment of an authenticated physical standard for gamma spectrometric determination of the U-235 content of MTR fuel and evaluation of measurement procedures

International Nuclear Information System (INIS)

Fleck, C.M.

1979-12-01

Measurements of U-235 content in a standard MTR fuel element were carried out, using scintillation and semi-conductor spectrometers. Three different types of measurement were carried out: a) Comparison of different primary standards among one another and with single fuel plates. b) Calibration of the MTR fuel element as an authenticated physical standard. c) Evaluation of over all errors in assay measurements on MTR fuel elements. The error of the whole assay measurement will be approximately 0.9%. The Uranium distribution in the single fuel plates is the original source of error. In the case of equal Uranium contents in all fuel plates of one fuel assembly, the error of assay measurements would be about 0.3% relative to the primary standards

Neutronic calculations in core conversion of the IAN-R1 research reactor from MTR HEU to TRIGA LEU fuel

International Nuclear Information System (INIS)

Sarta Fuentes, Jose A.; Castiblanco, L.A.

2003-01-01

With cooperation of the International Atomic Energy Agency (IAEA), neutronic calculations were carried out for conversion of the Ian-R1 Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to establish a staff for neutronic calculation at the Instituto de Cancan's Nucleares y Energia s Alternatives (INEA) a program was established. This program included training, acquisition of hardware, software and calculation for the core with MTR-HEU fuel , enriched nominally to 93% and calculation for several arrangements with the TRIGA-LEU fuel, enriched to 19.7%. The results were verified and compared with several groups of calculation at the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, and General Atomics (GA) in United States. As a result of this program, several technical reports have been wrote. (author)

L-Area STS MTR/NRU/NRX Grapple Assembly Closure Mechanics Review

International Nuclear Information System (INIS)

Huizenga, D. J.

2016-01-01

A review of the closure mechanics associated with the Shielded Transfer System (STS) MTR/NRU/NRX grapple assembly utilized at the Savannah River Site (SRS) was performed. This review was prompted by an operational event which occurred at the Canadian Nuclear Laboratories (CNL) utilizing a DTS-XL grapple assembly which is essentially identical to the STS MTR/NRU/NRX grapple assembly used at the SRS. The CNL operational event occurred when a NRU/NRX fuel basket containing spent nuclear fuel assemblies was inadvertently released by the DTS-XL grapple assembly during a transfer. The SM review of the STS MTR/NRU/NRX grapple assembly will examine the operational aspects of the STS and the engineered features of the STS which prevent such an event at the SRS. The design requirements for the STS NRU/NRX modifications and the overall layout of the STS are provided in other documents.

Immobilisation of MTR waste in cement (product evaluation). Annual report March 1985

International Nuclear Information System (INIS)

Howard, C.G.; Smith, D.L.G.; Williams, J.R.A.

1986-01-01

This report describes work performed at Winfrith under the UKAEA's research and development programme on radioactive waste management. The work carried out during April 1984 to March 1985 on the evaluation of laboratory and 200 dm 3 scale products of cemented MTR waste was sponsored by the Department of the Environment as part of radioactive waste management research programme. The results will be used in the formulation of Government policy but at this stage they do not necessarily represent Government policy. (author)

Transportation of spent MTR fuels

Energy Technology Data Exchange (ETDEWEB)

Raisonnier, D.

1997-08-01

This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs.

Transportation of spent MTR fuels

International Nuclear Information System (INIS)

Raisonnier, D.

1997-01-01

This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs

Design and Construction of Pool Door for Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Jung, Kwangsub; Lee, Sangjin; Choi, Jinbok; Oh, Jinho; Lee, Jongmin [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The pool door is a structure to isolate the reactor pool from the service pool for maintenance. The pool door is installed before the reactor pool is drained. The pool door consists of structural component and sealing component. The main structures of the pool door are stainless steel plates and side frames. The plates and frames are assembled by welded joints. Lug is welded at the top of the plate. The pool door is submerged in the pool water when it is used. Materials of the pool door should be resistive to corrosion and radiation. Stainless steel is used in structural components and air nozzle assemblies. Features of design and construction of the pool door for the research reactor are introduced. The pool door is designed to isolate the reactor pool for maintenance. Structural analysis is performed to evaluate the structural integrity during earthquake. Tests and inspections are also carried out during construction to identify the safety and function of the pool door.

Design and Construction of Pool Door for Research Reactor

International Nuclear Information System (INIS)

Jung, Kwangsub; Lee, Sangjin; Choi, Jinbok; Oh, Jinho; Lee, Jongmin

2016-01-01

The pool door is a structure to isolate the reactor pool from the service pool for maintenance. The pool door is installed before the reactor pool is drained. The pool door consists of structural component and sealing component. The main structures of the pool door are stainless steel plates and side frames. The plates and frames are assembled by welded joints. Lug is welded at the top of the plate. The pool door is submerged in the pool water when it is used. Materials of the pool door should be resistive to corrosion and radiation. Stainless steel is used in structural components and air nozzle assemblies. Features of design and construction of the pool door for the research reactor are introduced. The pool door is designed to isolate the reactor pool for maintenance. Structural analysis is performed to evaluate the structural integrity during earthquake. Tests and inspections are also carried out during construction to identify the safety and function of the pool door

The obtainment of highly concentrated uranium pellets for plate type (MTR) fuel by dispersion of uranium aluminides in aluminium

International Nuclear Information System (INIS)

Morando, R.A.; Raffaeli, H.A.; Balzaretti, D.E.

1980-01-01

The use of the intermetallic UAl 3 for manufacturing plate type MTR fuel with 20% U 235 enriched uranium and a density of about 20 kg/m 3 is analyzed. The technique used is the dispersion of UAl 3 particles in aluminium powder. The obtainment of the UAl 3 intermetallic was performed by fusion in an induction furnace in an atmosphere of argon at a pressure of 0.7 BAR (400 mm) using an alumina melting pot. To make the aluminide powder and attain the wished granulometry a cutting and a rotating crusher were used. Aluminide powders of different granulometries and different pressures of compactation were analyzed. In each case the densities were measured. The compacts were colaminated with the 'Picture Frame' technique at temperatures of 490 and 0 deg C with excellent results from the manufacturing view point. (M.E.L.) [es

Determination of neutron energy spectrum at a pneumatic rabbit station of a typical swimming pool type material test research reactor

International Nuclear Information System (INIS)

Malkawi, S.R.; Ahmad, N.

2002-01-01

The method of multiple foil activation was used to measure the neutron energy spectrum, experimentally, at a rabbit station of Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type material test research reactor. The computer codes MSITER and SANDBP were used to adjust the spectrum. The pre-information required by the adjustment codes was obtained by modelling the core and its surroundings in three-dimensions by using the one dimensional transport theory code WIMS-D/4 and the multidimensional finite difference diffusion theory code CITATION. The input spectrum covariance information required by MSITER code was also calculated from the CITATION output. A comparison between calculated and adjusted spectra shows a good agreement

Review of events at large pool-type irradiators

International Nuclear Information System (INIS)

Trager, E.A. Jr.

1989-03-01

Large pool-type gamma irradiators are used in applications such as the ''cold'' sterilization of medical and pharmaceutical supplies, and recent changes in federal regulations make it possible they will be used extensively in the preservation of foodstuffs. Because of this possible large increase in the use of irradiators, the Office of Nuclear Materials Safety and Safeguards was interested in knowing what events had occurred at irradiators. The event data would be used as background in developing new regulations on irradiators. Therefore, AEOD began a study of the operating experience at large, wet source storage gamma irradiators. The scope of the study was to assess all available operating information on large (≥ 250,000 curie), pool-type irradiators licensed by both the NRC and the Agreement States, and events at foreign facilities. The study found that about 0.12 events have been reported per irradiator-year. Most of these events were precursor events, in that there was no evidence of damage to the radioactive sources or degradation in the level of safety of the facility. Events with more significant impacts had a reported frequency of about 0.01 event per irradiator-year. However, the actual rate of occurrence of events of concern to the staff may be higher because there are few specific reporting requirements for events at irradiators. It is suggested that during development of a regulation for large pool-type irradiators consideration be given to specifying requirements for: reporting breakdowns in access control systems; periodic inspection of the source movement and suspension system; systems to detect source leakage and product contamination; allowable pool leakage; and feedback of information on operational events involving safety-important systems

Feasibility study on large pool-type LMFBR

International Nuclear Information System (INIS)

1984-01-01

A feasibility study has been conducted from 1981 FY to 1983 FY, in order to evaluate the feasibility of a large pool-type LMFBR under the Japanese seismic design condition and safety design condition, etc. This study was aimed to establish an original reactor structure concept which meets those design conditions especially required in Japan. In the first year, preceding design concepts had been reviewed and several concepts were originated to be suitable to Japan. For typical two of them being selected by preliminary analysis, test programs were planned. In the second year, more than twenty tests with basic models had been conducted under severe conditions, concurrently analytical approaches were promoted. In the last year, larger model tests were conducted and analytical methods have been verified concerning hydrodynamic effects on structure vibration, thermo-hydraulic behaviours in reactor plena and so on. Finally the reactor structure concepts for a large pool-type LMFBR have been acknowledged to be feasible in Japan. (author)

Structure and Function of Neisseria gonorrhoeae MtrF Illuminates a Class of Antimetabolite Efflux Pumps

Directory of Open Access Journals (Sweden)

Chih-Chia Su

2015-04-01

Full Text Available Neisseria gonorrhoeae is an obligate human pathogen and the causative agent of the sexually transmitted disease gonorrhea. The control of this disease has been compromised by the increasing proportion of infections due to antibiotic-resistant strains, which are growing at an alarming rate. N. gonorrhoeae MtrF is an integral membrane protein that belongs to the AbgT family of transporters for which no structural information is available. Here, we describe the crystal structure of MtrF, revealing a dimeric molecule with architecture distinct from all other families of transporters. MtrF is a bowl-shaped dimer with a solvent-filled basin extending from the cytoplasm to halfway across the membrane bilayer. Each subunit of the transporter contains nine transmembrane helices and two hairpins, posing a plausible pathway for substrate transport. A combination of the crystal structure and biochemical functional assays suggests that MtrF is an antibiotic efflux pump mediating bacterial resistance to sulfonamide antimetabolite drugs.

JHR. A high performance MTR under construction for a sustainable nuclear energy

International Nuclear Information System (INIS)

Iracane, Daniel; Cordier, Pierre-Yves

2009-01-01

The Access to an up-to-date Material Testing Reactor (MTR) is essential to support a sustainable nuclear energy, meeting industry and public needs, and keeping a high level of scientific expertise. This includes services to existing and coming reactor technologies for major stakes such as safety and competitiveness, lifetime management, operation optimization, development of innovative structural material and fuel required for future systems (innovative Gen III, Gen IV, fusion...), etc. The JHR copes with this context. Design phase has been completed by the end of 2005 and JHR is now under construction. Start of operation is scheduled in 2014. As a new MTR taking benefit of a large available worldwide experience, JHR offers new major experimental capability that will be presented. JHR will be operated within an international users' consortium that will guarantee effective and cost-effective operation. This innovative way to operate a MTR, as a user-facility for the benefit of industry and public bodies, will be presented. (author)

Loading 076 assemblies in two IV-04 transport casks for transport to the U.S. Savannah River Site (SC); Trasferimento di 72 elementi irraggiati MTR dalla piscina dell`impianto EUREX a due contenitori IU-04 per il trasporto al Savannah River Site-Department of Energy (USA)

Energy Technology Data Exchange (ETDEWEB)

Gili, Michele [ENEA, Centro Ricerche Saluggia, Vercelli (Italy). Dipt. Energia

1997-09-01

The National Agency for New Technologies and the Environments has signed with the US Department of Energy a contract for the transfer of 150 irradiated MTR fuel assemblies stored in the EUREX plant pool at The National Agency for New Technologies and the Environments Research Centre of Saluggia. The first scheduled transport has been made in february 1997 and has involved the successful loading of 76 assemblies in two IU-04 (Pegase) transport casks. The loaded casks have been shipped to the U.S. Savannah River Site (SC).

A model development for a thermohydraulic calculation material convection of MTR (Materials Testing Reactors)

International Nuclear Information System (INIS)

Abbate, P.

1990-01-01

The CONVEC program developed for the thermohydraulic calculation under a natural convection regime for MTR type reactors is presented. The program is based on a stationary, one dimensional model of finite differences that allow to calculate the temperatures of cooler, cladding and fuel as well as the flow for a power level specified by the user. This model has been satisfactorily validated by a water cooling (liquid phase) and air system. (Author) [es

E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

Energy Technology Data Exchange (ETDEWEB)

Van Tichelen, Katrien, E-mail: kvtichel@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Mirelli, Fabio, E-mail: fmirelli@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Greco, Matteo, E-mail: mgreco@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Viviani, Giorgia, E-mail: giorgiaviviani@gmail.com [University of Pisa, Lungarno Pacinotti 43, 56126 Pisa (Italy)

2015-08-15

Highlights: • The E-SCAPE facility is a thermal hydraulic scale model of the MYRRHA fast reactor. • The focus is on mixing and stratification in liquid-metal pool-type reactors. • Forced convection, natural convection and the transition are investigated. • Extensive instrumentation allows validation of computational models. • System thermal hydraulic and CFD models have been used for facility design. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant. The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing. The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE. The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

Criticality Studies in a Pilot Plant for Processing MTR-Type Irradiated Fuels; Estudios de Criticidad de una Planta Piloto para el Tratamiento de Combustibles Irradiados Tipo ' MTR '

Energy Technology Data Exchange (ETDEWEB)

Pereira Sanchez, G.; Uriarte Hueda, A. [Junta de Energia Nuclear, Division de Materiales Madrid (Spain)

1966-05-15

A number of theoretical studies on nuclear safety have been carried out in a pilot plant being constructed at the Junta de Energia Nuclear in Madrid for processing irradiated fuels from the MTR-type experimental reactor JEN-1. The study was carried out working with aqueous and organic solutions at two levels of {sup 235}U enrichment - 20% and 93%. The paper is divided into two main parts: the first deals with the individual items of equipment, and the interactions between these are studied in the second part. The calculations in this second part have been made using three different methods to make it more certain that the system as a whole can never be critical. The first method employed is based on the solid angle concept and makes it possible to fix the maximum {sup 235}U concentrations within the system. The second method, based on the albedo, supplies the value of the multiplication factor K of the whole assembly as a function of the concentration of {sup 235}U. In the last part, the distribution of the equipment is compared with other similar systems and experimental tests from other sources. Finally, the paper specifies the conditions for working the installation which ensure that a nuclear accident can never occur. (author) [Spanish] Se ha efectuado una serie de estudios teoricos sobre la seguridad nuclear de una planta piloto, que se encuentra en construccion en la Junta de Energfa Nuclear situada en Madrid, para el tratamiento de combustibles irradiados procedentes del reactor experimental JEN-1 del tipo MTR. El estudio se ha realizado utilizando disoluciones, tanto acuosas como organicas, con dos grados de enriquecimiento, 20% y 93% en {sup 235}U. Este trabajo comprende dos partes principales: en la primera se han considerado las distintas unidades del equipo individualmente y en la segunda se han estudiado las interacciones entre ellas. El calculo de esta segunda parte se ha hecho por tres metodos diferentes para tener una mayor seguridad de que el

Detection of mutations in mtrR gene in quinolone resistant strains of N.gonorrhoeae isolated from India

Directory of Open Access Journals (Sweden)

S V Kulkarni

2015-01-01

Full Text Available Background and Objectives: Emergence of multi-drug resistant Neisseria gonorrhoeae resulting from new genetic mutation is a serious threat in controlling gonorrhea. This study was undertaken to identify and characterise mutations in the mtrR genes in N.gonorrhoeae isolates resistant to six different antibiotics in the quinolone group. Materials and Methods: The Minimum inhibitory concentrations (MIC of five quinolones for 64 N.gonorrhoeae isolates isolated during Jan 2007-Jun 2009 were determined by E-test method. Mutations in MtrR loci were examined by deoxyribonucleic acid (DNA sequencing. Results: The proportion of N.gonorrhoeae strains resistant to anti-microbials was 98.4% for norfloxacin and ofloxacin, 96.8% for enoxacin and ciprofloxacin, 95.3% for lomefloxacin. Thirty-one (48.4% strains showed mutation (single/multiple in mtrR gene. Ten different mutations were observed and Gly-45 → Asp, Tyr-105 → His being the most common observed mutation. Conclusion: This is the first report from India on quinolone resistance mutations in MtrRCDE efflux system in N.gonorrhoeae. In conclusion, the high level of resistance to quinolone and single or multiple mutations in mtrR gene could limit the drug choices for gonorrhoea.

MTR and PWR/PHWR in-pile loop safety in integration with the operation of multipurpose reactor - GAS

International Nuclear Information System (INIS)

Suharno; Aji, Bintoro; Sugiyanto; Rohman, Budi; Zarkasi, Amin S.; Giarno

1998-01-01

MTR and PWR/PHWR In-Pile Loop safety analysis in integration with the operation of Multipurpose Reactor - Gas has been carried out and completed. The assessment is emphasized on the function of the interface systems from the dependence of the operation and the evaluation to the possibility of leakage or failure of the in-pile part inside the reactor pool and reactor core. The analysis is refers to the logic function of the interface system and the possibility of leakage or failure of the in-pile part inside reactor pool and reactor core to consider the integrity of the core qualitatively. The results show that in normal and in transient conditions , the interface system meet the function requirement in safe integrated operation of in-pile loop and reactor. And the results of the possibility analysis of the leakage shows that the possibility based on mechanically assessment is very low and the impact to core integrity is nothing or can be eliminated. The possible position for leakage is on the flen on which one meter above the top level of the core, therefore no influence of leakage to the core

Mutation of a Rice Gene Encoding a Phenylalanine Biosynthetic Enzyme Results in Accumulation of Phenylalanine and Tryptophan[W

Science.gov (United States)

Yamada, Tetsuya; Matsuda, Fumio; Kasai, Koji; Fukuoka, Shuichi; Kitamura, Keisuke; Tozawa, Yuzuru; Miyagawa, Hisashi; Wakasa, Kyo

2008-01-01

Two distinct biosynthetic pathways for Phe in plants have been proposed: conversion of prephenate to Phe via phenylpyruvate or arogenate. The reactions catalyzed by prephenate dehydratase (PDT) and arogenate dehydratase (ADT) contribute to these respective pathways. The Mtr1 mutant of rice (Oryza sativa) manifests accumulation of Phe, Trp, and several phenylpropanoids, suggesting a link between the synthesis of Phe and Trp. Here, we show that the Mtr1 mutant gene (mtr1-D) encodes a form of rice PDT with a point mutation in the putative allosteric regulatory region of the protein. Transformed callus lines expressing mtr1-D exhibited all the characteristics of Mtr1 callus tissue. Biochemical analysis revealed that rice PDT possesses both PDT and ADT activities, with a preference for arogenate as substrate, suggesting that it functions primarily as an ADT. The wild-type enzyme is feedback regulated by Phe, whereas the mutant enzyme showed a reduced feedback sensitivity, resulting in Phe accumulation. In addition, these observations indicate that rice PDT is critical for regulating the size of the Phe pool in plant cells. Feeding external Phe to wild-type callus tissue and seedlings resulted in Trp accumulation, demonstrating a connection between Phe accumulation and Trp pool size. PMID:18487352

Irradiation of novel MTR fuel plates in BR2

International Nuclear Information System (INIS)

Verboomen, B.; Aoust, Th.; Beeckmans De Westmeerbeeck, A.; De Raedt, Ch.

2000-01-01

Since the end of 1999, novel MTR fuel plates with very high-density meat are being irradiated in BR2. The purpose of the irradiation is to investigate the behaviour of these fuel plates under very severe reactor operation conditions. The novel fuel plates are inserted in two standard six-tube BR2 fuel elements in the locations normally occupied by the standard outer fuel plates. The irradiation in BR2 was prepared by carrying out detailed neutron Monte Carlo calculations of the whole BR2 core containing the two experimental fuel elements for various positions in the reactor and for various azimuthal orientations of the fuel elements. Comparing the thus determined fission density levels and azimuthal profiles in the new MTR fuel plates irradiated in the various channels allowed the experimenters to choose the most appropriate BR2 channel and the most appropriate fuel element orientation. (author)

RRFC hardware operation manual

International Nuclear Information System (INIS)

Abhold, M.E.; Hsue, S.T.; Menlove, H.O.; Walton, G.

1996-05-01

The Research Reactor Fuel Counter (RRFC) system was developed to assay the 235 U content in spent Material Test Reactor (MTR) type fuel elements underwater in a spent fuel pool. RRFC assays the 235 U content using active neutron coincidence counting and also incorporates an ion chamber for gross gamma-ray measurements. This manual describes RRFC hardware, including detectors, electronics, and performance characteristics

Simplified analysis of trasients in pool type liquid metal reactors

International Nuclear Information System (INIS)

Botelho, D.A.

1987-01-01

The conceptual design of a liquid metal fast breeder reactor will require a great effort of development in several technical disciplines. One of them is the thermal-hydraulic design of the reactor and of the heat and fluid transport components inside the reactor vessel. A simplified model to calculate the maximum sodium temperatures is presented in this paper. This model can be used to optimize the layout of components inside the reactor vessel and was easily programmed in a small computer. Illustrative calculations of two transients of a typical hot pool type fast reactor are presented and compared with the results of other researchers. (author) [pt

FRG compact core - one year experience

International Nuclear Information System (INIS)

Knop, W.; Schreiner, P.

2001-01-01

The GKSS research centre Geesthacht GmbH operates the MTR-type swimming pool reactor FRG-1 (5 MW) for more than 40 years. The FRG-1 has been upgraded and refurbished many times to follow the changing demands of safe operation and today's needs of high neutron flux for scientific research. High neutron fluxes with highest availability is the permanent demand of the science on the operation of a neutron source. (orig.)

Some equipment for graphite research in swimming pool reactors

International Nuclear Information System (INIS)

Seguin, M.; Arragon, Ph.; Dupont, G.; Gentil, J.; Tanis, G.

1964-01-01

The irradiation devices described are used for research concerning reactors of the natural uranium type, moderated by graphite and cooled by carbon dioxide. The devices are generally designed for use in swimming pool reactors. The following points have been particularly studied: - maximum use of the irradiation volume, - use of the simplest technological solutions, - standardization of certain constituent parts. This standardization calls for precision machining and careful assembling; these requirements are also true when a relatively low irradiation temperature is required and the nuclear heating is pronounced. Finally, the design of these devices is suitable for the irradiation of other fissile or non-fissile materials. (authors) [fr

Nodalization effects on RELAP5 results related to MTR research reactor transient scenarios

Directory of Open Access Journals (Sweden)

Khedr Ahmed

2005-01-01

Full Text Available The present work deals with the anal y sis of RELAP5 results obtained from the evaluation study of the total loss of flow transient with the deficiency of the heat removal system in a research reactor using two different nodalizations. It focuses on the effect of nodalization on the thermal-hydraulic evaluation of the re search reactor. The analysis of RELAP5 results has shown that nodalization has a big effect on the predicted scenario of the postulated transient. There fore, great care should be taken during the nodalization of the reactor, especially when the avail able experimental or measured data are insufficient for making a complete qualification of the nodalization. Our analysis also shows that the research reactor pool simulation has a great effect on the evaluation of natural circulation flow and on other thermal-hydraulic parameters during the loss of flow transient. For example, the on set time of core boiling changes from less than 2000 s to 15000 s, starting from the beginning of the transient. This occurs if the pool is simulated by two vertical volumes in stead of one vertical volume.

Control of gdhR Expression in Neisseria gonorrhoeae via Autoregulation and a Master Repressor (MtrR of a Drug Efflux Pump Operon

Directory of Open Access Journals (Sweden)

Corinne E. Rouquette-Loughlin

2017-04-01

Full Text Available The MtrCDE efflux pump of Neisseria gonorrhoeae contributes to gonococcal resistance to a number of antibiotics used previously or currently in treatment of gonorrhea, as well as to host-derived antimicrobials that participate in innate defense. Overexpression of the MtrCDE efflux pump increases gonococcal survival and fitness during experimental lower genital tract infection of female mice. Transcription of mtrCDE can be repressed by the DNA-binding protein MtrR, which also acts as a global regulator of genes involved in important metabolic, physiologic, or regulatory processes. Here, we investigated whether a gene downstream of mtrCDE, previously annotated gdhR in Neisseria meningitidis, is a target for regulation by MtrR. In meningococci, GdhR serves as a regulator of genes involved in glucose catabolism, amino acid transport, and biosynthesis, including gdhA, which encodes an l-glutamate dehydrogenase and is located next to gdhR but is transcriptionally divergent. We report here that in N. gonorrhoeae, expression of gdhR is subject to autoregulation by GdhR and direct repression by MtrR. Importantly, loss of GdhR significantly increased gonococcal fitness compared to a complemented mutant strain during experimental murine infection. Interestingly, loss of GdhR did not influence expression of gdhA, as reported for meningococci. This variance is most likely due to differences in promoter localization and utilization between gonococci and meningococci. We propose that transcriptional control of gonococcal genes through the action of MtrR and GdhR contributes to fitness of N. gonorrhoeae during infection.

Pool type liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Guthrie, B.M.

1978-08-01

Various technical aspects of the liquid metal fast breeder reactor (LMFBR), specifically pool type LMFBR's, are summarized. The information presented, for the most part, draws upon existing data. Special sections are devoted to design, technical feasibility (normal operating conditions), and safety (accident conditions). A survey of world fast reactors is presented in tabular form, as are two sets of reference reactor parameters based on available data from present and conceptual LMFBR's. (auth)

New options to fuel plate for MTR reactor

International Nuclear Information System (INIS)

Macedo, C.R.

1988-01-01

The main datas of fuel elements and the new materials for good performance of the MTR reactor are described. A study to verify the possibility of introduction a new element on the alloy is presented. After verification the stages of nucleus fabrication with dispersion cermets of uranium oxide is gave a special emphasis to cermet fabrication of uranium-aluminium alloys. (C.G.C.) [pt

New high density MTR fuel. The CEA-CERCA-COGEMA development program

International Nuclear Information System (INIS)

Languille, A.; Durand, J.P.; Gay, A.

1999-01-01

The development of a new generation of LEU, high in density and with reprocessing capacities MTR fuel, is a key issue to provide reactor operators with a smooth operation which is necessary for a long term development of Nuclear Energy. In the RRFM'98 meeting, a joint contribution of CEA, CERCA and COGEMA presented a technical classification of the potential candidates uranium alloys. In this paper this MTR working group presents the development program of a new high density fuel. This program is composed of three main steps: Basic Data analysis and collection, Plate Tests (Irradiation and Post Irradiation Examinations) and Lead Test Assemblies (Irradiation and Post Irradiation Examinations). The goal to be reached is to make this new fuel available before the end of the present US return policy. (author)

Structure and reconstitution of yeast Mpp6-nuclear exosome complexes reveals that Mpp6 stimulates RNA decay and recruits the Mtr4 helicase

Energy Technology Data Exchange (ETDEWEB)

Wasmuth, Elizabeth V. [Structural Biology Program, Sloan Kettering Institute, Memorial Sloan Kettering Cancer Center, New York, United States; Zinder, John C. [Structural Biology Program, Sloan Kettering Institute, Memorial Sloan Kettering Cancer Center, New York, United States; Tri-Institutional Training Program in Chemical Biology, Memorial Sloan Kettering Cancer Center, New York, United States; Zattas, Dimitrios [Structural Biology Program, Sloan Kettering Institute, Memorial Sloan Kettering Cancer Center, New York, United States; Das, Mom [Structural Biology Program, Sloan Kettering Institute, Memorial Sloan Kettering Cancer Center, New York, United States; Lima, Christopher D. [Structural Biology Program, Sloan Kettering Institute, Memorial Sloan Kettering Cancer Center, New York, United States; Howard Hughes Medical Institute, Memorial Sloan Kettering Cancer Center, New York, United States

2017-07-25

Nuclear RNA exosomes catalyze a range of RNA processing and decay activities that are coordinated in part by cofactors, including Mpp6, Rrp47, and the Mtr4 RNA helicase. Mpp6 interacts with the nine-subunit exosome core, while Rrp47 stabilizes the exoribonuclease Rrp6 and recruits Mtr4, but it is less clear if these cofactors work together. Using biochemistry with Saccharomyces cerevisiae proteins, we show that Rrp47 and Mpp6 stimulate exosome-mediated RNA decay, albeit with unique dependencies on elements within the nuclear exosome. Mpp6-exosomes can recruit Mtr4, while Mpp6 and Rrp47 each contribute to Mtr4-dependent RNA decay, with maximal Mtr4-dependent decay observed with both cofactors. The 3.3 Å structure of a twelve-subunit nuclear Mpp6 exosome bound to RNA shows the central region of Mpp6 bound to the exosome core, positioning its Mtr4 recruitment domain next to Rrp6 and the exosome central channel. Genetic analysis reveals interactions that are largely consistent with our model.

Determination of the burn-up in fuels of the MTR type by means of gamma spectroscopy with crystal of INa(Tl)

International Nuclear Information System (INIS)

Kestelman, A.J.

1988-01-01

One of the responsibilities of the Laboratory of Analysis by Neutronic Activation of the RA-6 reactor is to determine the burn-up in fuels of the MTR type. In order to gain experience, up to the arrival of the hyperpure Germanium detector (HPGe) to be used in normal operation, preliminary measurements with a crystal of INa(Tl) were made. The fuel elements used are originated in the RA-3 reactor, with a decay superior to the thirteen years. For this reason, the unique visible photoelectric peak is the one of Cs-137, owing to the low resolution of the INa(Tl). After preliminary measurements, the profiles of burn-up, rectified by attenuation, were measured. Once the efficiency of the detector was determined, the calculation of the burn-up was made; for the element No. 144, a value of 21.6 ± 2.9 g was obtained to be compared with the value 21.9 g which was the evaluation made by the operators. (Author) [es

Radiological performance of hot water layer system in open pool type reactor

OpenAIRE

Amr Abdelhady

2013-01-01

The paper presents the calculated dose rate carried out by using MicroShield code to show the importance of hot water layer system (HWL) in 22 MW open pool type reactor from the radiation protection safety point of view. The paper presents the dose rate profiles over the pool surface in normal and abnormal operations of HWL system. The results show that, in case of losing the hot water layer effect, the radiation dose rate profiles over the pool surface will increase from values lower than th...

Development of system design and seismic performance evaluation for reactor pool working platform of a research reactor

International Nuclear Information System (INIS)

Kwag, Shinyoung; Lee, Jong-Min; Oh, Jinho; Ryu, Jeong-Soo

2014-01-01

Highlights: • Design of reactor pool working platform (RPWP) is newly proposed for an open-tank-in-pool type research reactor. • Main concept of RPWP is to minimize the pool top radiation level. • Framework for seismic performance evaluation of nuclear SSCs in a deterministic and a probabilistic manner is proposed. • Structural integrity, serviceability, and seismic margin of the RPWP are evaluated during and after seismic events. -- Abstract: The reactor pool working platform (RPWP) has been newly designed for an open-tank-in-pool type research reactor, and its seismic response, structural integrity, serviceability, and seismic margin have been evaluated during and after seismic events in this paper. The main important concept of the RPWP is to minimize the pool top radiation level by physically covering the reactor pool of the open-tank-in-pool type research reactor and suppressing the rise of flow induced by the primary cooling system. It is also to provide easy handling of the irradiated objects under the pool water by providing guide tubes and refueling cover to make the radioisotopes irradiated and protect the reactor structure assembly. For this concept, the new three dimensional design model of the RPWP is established for manufacturing, installation and operation, and the analytical model is developed to analyze the seismic performance. Since it is submerged under and influenced by water, the hydrodynamic effect is taken into account by using the hydrodynamic added mass method. To investigate the dynamic characteristics of the RPWP, a modal analysis of the developed analytical model is performed. To evaluate the structural integrity and serviceability of the RPWP, the response spectrum analysis and response time history analysis have been performed under the static load and the seismic load of a safe shutdown earthquake (SSE). Their stresses are analyzed for the structural integrity. The possibility of an impact between the RPWP and the most

Refurbishment of Pakistan research reactor (PARR-1) for stainless steel lining of the reactor pool

International Nuclear Information System (INIS)

Salahuddin, A.; Israr, M.; Hussain, M.

2002-01-01

Pakistan Research Reactor-1 (PARR-1) is a pool-type research reactor. Reactor aging has resulted in the increase of water seepage from the concrete walls of the reactor pool. To stop the seepage, it was decided to augment the existing pool walls with an inner lining of stainless steel. This could be achieved only if the pool walls could be accessed unhindered and without excessive radiation doses. For this purpose a partial decommissioning was done by removing all active core components including standard/control fuel elements, reflector elements, beam tubes, thermal shield, core support structure, grid plate and the pool's ceramic tiles, etc. An overall decommissioning program was devised which included procedures specific to each item. This led to the development of a fuel transport cask for transportation, and an interim fuel storage bay for temporary storage of fuel elements (until final disposal). The safety of workers and the environment was ensured by the use of specially designed remote handling tools, appropriate shielding and pre-planned exposure reduction procedures based on the ALARA principle. During the implementation of this program, liquid and solid wastes generated were legally disposed of. It is felt that the experience gained during the refurbishment of PARR-1 to install the stainless steel liner will prove useful and better planning and execution for the future decommissioning of PARR-1, in particular, and for other research reactors like PARR-2 (27 kW MNSR), in general. Furthermore, due to the worldwide activities on decommissioning, especially those communicated through the IAEA CRP on 'Decommissioning Techniques for Research Reactors', the importance of early planning has been well recognized. This has made possible the implementation of some early steps like better record keeping, rehiring of trained manpower, and creation of interim and final waste storage. (author)

Development of the uranium recovery process from rejected fuel plates in the fabrication of MTR type nuclear fuel

International Nuclear Information System (INIS)

Fleming Rubio, Peter Alex

2010-01-01

The current work was made in Conversion laboratory belonging to Chilean Nuclear Energy Commission, CCHEN. This is constituted by the development of three hydrometallurgical processes, belonging to the recovery of uranium from fuel plates based on uranium silicide (U_3Si_2) process, for nuclear research reactors MTR (Material Testing Reactor) type, those that come from the Fuel Elements Manufacture Plant, PEC. In the manufacturing process some of these plates are subjected to destructive tests by quality requirement or others are rejected for non-compliance with technical specifications, such as: lack of homogenization of the dispersion of uraniferous compound in the meat, as well as the appearance of the defects, such as blisters, so-called "dog bone", "fish tail", "remote islands", among others. Because the uranium used is enriched in 19.75% U_2_3_5 isotope, which explains the high value in the market, it must be recovered for reuse, returning to the production line of fuel elements. The uranium silicide, contained in the plates, is dispersed in an aluminum matrix and covered with plates and frames of ASTM 6061 Aluminum, as a sandwich coating, commonly referred to as 'meat' (sandwich meat). As aluminum is the main impurity, the process begins with this metal dissolution, present in meat and plates, by NaOH reaction, followed by a vacuum filtration, washing and drying, obtaining a powder of uranium silicide, with a small impurities percentage. Then, the crude uranium silicide reacts with a solution of hydrofluoric acid, dissolving the silicon and simultaneously precipitating UF_4 by reaction with HNO_3, obtaining an impure UO_2(NO_3)_2 solution. The experimental work was developed and implemented at laboratory scale for the three stages pertaining to the uranium recovery process, determining for each one the optimum operation conditions: temperature, molarity or concentration, reagent excess, among others (author)

Simulating a partial LOCA in a narrow channel using the DSNP simulating system

International Nuclear Information System (INIS)

Saphier, D.

2007-01-01

A partial LOCA accident in a pool type research reactor was investigated. A new MTR type fuel channel model for the DSNP simulation system was developed; permitting detailed axial and radial temperature distribution. New and older heat transfer correlations were incorporated in the model. Simulation for water levels of 14 and 35 cm in a 62 cm channel were performed. The resulting maximum temperatures remain significantly below the aluminium melting point, and no damage to the core will take place under these conditions

Burnup measurements on spent fuel elements of the RP-10 research reactor

Energy Technology Data Exchange (ETDEWEB)

Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

Burnup measurements on spent fuel elements of the RP-10 research reactor

International Nuclear Information System (INIS)

Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro

2011-01-01

This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using 137 Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

Crystal structure of the Neisseria gonorrhoeae MtrD inner membrane multidrug efflux pump.

Directory of Open Access Journals (Sweden)

Jani Reddy Bolla

Full Text Available Neisseria gonorrhoeae is an obligate human pathogen and the causative agent of the sexually-transmitted disease gonorrhea. The control of this disease has been compromised by the increasing proportion of infections due to antibiotic-resistant strains, which are growing at an alarming rate. The MtrCDE tripartite multidrug efflux pump, belonging to the hydrophobic and amphiphilic efflux resistance-nodulation-cell division (HAE-RND family, spans both the inner and outer membranes of N. gonorrhoeae and confers resistance to a variety of antibiotics and toxic compounds. We here report the crystal structure of the inner membrane MtrD multidrug efflux pump, which reveals a novel structural feature that is not found in other RND efflux pumps.

CFD investigation of flow inversion in typical MTR research reactor undergoing thermal-hydraulic transients

International Nuclear Information System (INIS)

Salama, Amgad

2011-01-01

Highlights: → The 3D, CFD simulation of FLOFA accident in the generic IAEA 10 MW research reactor is carried out. → The different flow and heat transfer mechanisms involved in this process were elucidated. → The transition between these mechanisms during the course of FLOFA is discussed and investigated. → The interesting inversion process upon the transition from downward flow to upward flow is shown. → The temperature field and the friction coefficient during the whole transient process were shown. - Abstract: Three dimensional CFD full simulations of the fast loss of flow accident (FLOFA) of the IAEA 10 MW generic MTR research reactor are conducted. In this system the flow is initially downward. The transient scenario starts when the pump coasts down exponentially with a time constant of 1 s. As a result the temperatures of the heating element, the clad, and the coolant rise. When the flow reaches 85% of its nominal value the control rod system scrams and the power drops sharply resulting in the temperatures of the different components to drop. As the coolant flow continues to drop, the decay heat causes the temperatures to increase at a slower rate in the beginning. When the flow becomes laminar, the rate of temperature increase becomes larger and when the pumps completely stop a flow inversion occurs because of natural convection. The temperature will continue to rise at even higher rates until natural convection is established, that is when the temperatures settle off. The interesting 3D patterns of the flow during the inversion process are shown and investigated. The temperature history is also reported and is compared with those estimated by one-dimensional codes. Generally, very good agreement is achieved which provides confidence in the modeling approach.

Experience and prospects for developing research reactors of different types

International Nuclear Information System (INIS)

Kuatbekov, R.P.; Tretyakov, I.T.; Romanov, N.V.; Lukasevich, I.B.

2015-01-01

NIKIET has a 60-year experience in the development of research reactors. Altogether, there have been more than 25 NIKIET-designed plants of different types built in Russia and 20 more in other countries, including pool-type water-cooled and water moderated research reactors, tank-type and pressure-tube research reactors, pressurized high-flux, heavy-water, pulsed and other research reactors. Most of the research reactors were designed as multipurpose plants for operation at research centers in a broad range of applications. Besides, unique research reactors were developed for specific application fields. Apart from the experience in the development of research reactor designs and the participation in the reactor construction, a unique amount of knowledge has been gained on the operation of research reactors. This makes it possible to use highly reliable technical solutions in the designs of new research reactors to ensure increased safety, greater economic efficiency and maintainability of the reactor systems. A multipurpose pool-type research reactor of a new generation is planned to be built at the Center for Nuclear Energy Science & Technology (CNEST) in the Socialist Republic of Vietnam to be used to support a spectrum of research activities, training of skilled personnel for Vietnam nuclear industry and efficient production of isotopes. It is exactly the applications a research reactor is designed for that defines the reactor type, design and capacity, and the selection of fuel and components subject to all requirements of industry regulations. The design of the new research reactor has a great potential in terms of upgrading and installation of extra experimental devices. (author)

Design of inventory pools in spare part support operation systems

Science.gov (United States)

Mo, Daniel Y.; Tseng, Mitchell M.; Cheung, Raymond K.

2014-06-01

The objective of a spare part support operation is to fulfill the part request order with different service contracts in the agreed response time. With this objective to achieve different service targets for multiple service contracts and the considerations of inventory investment, it is not only important to determine the inventory policy but also to design the structure of inventory pools and the order fulfilment strategies. In this research, we focused on two types of inventory pools: multiple inventory pool (MIP) and consolidated inventory pool (CIP). The idea of MIP is to maintain separated inventory pools based on the types of service contract, while CIP solely maintains a single inventory pool regardless of service contract. Our research aims to design the inventory pool analytically and propose reserve strategies to manage the order fulfilment risks in CIP. Mathematical models and simulation experiments would be applied for analysis and evaluation.

Core neutronics of a swimming pool research reactor

International Nuclear Information System (INIS)

Mannan, M.A.; Mondal, M.A.W.; Pervini, M.E.

1981-01-01

The initial cores of the 5 MW swimming pool research reactor of the Nuclear Research Centre, Tehran have been analyzed using the computer codes METHUSELAH and EQUIPOISE. The effective multiplication factor, critical mass, moderator temperature and void coefficients of the core have been calculated and compared with vendor's values. Calculated values agree reasonably well with the vendor's results. (author)

Seismic responses of a pool-type fast reactor with different core support designs

International Nuclear Information System (INIS)

Wu, Ting-shu; Seidensticker, R.W.

1989-01-01

In designing the core support system for a pool-type fast reactor, there are many issues which must be considered in order to achieve an optimum and balanced design. These issues include safety, reliability, as well as costs. Several design options are possible to support the reactor core. Different core support options yield different frequency ranges and responses. Seismic responses of a large pool-type fast reactor incorporated with different core support designs have been investigated. 4 refs., 3 figs

Methodological study for management of the generated effluents during MTR-type fuel elements fabrication at IPEN/CNEN-SP plant

International Nuclear Information System (INIS)

Tanzillo Santos, Glaucia Regina

2008-01-01

Full text: The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the main programs of the Institute of Energetic and Nuclear Research of the National Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel -CCN- is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt % 235 U), to supply its IEA-R1 research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts. This study aims to implant the sustainability concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to

Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel

International Nuclear Information System (INIS)

Deen, J.R.; Snelgrove, J.L.; Papastergiou, C.

1992-01-01

The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed

MTR radiological database for SRS spent nuclear fuel facilities

International Nuclear Information System (INIS)

Blanchard, A.

2000-01-01

A database for radiological characterization of incoming Material Test Reactor (MTR) fuel has been developed for application to the Receiving Basin for Offsite Fuels (RBOF) and L-Basin spent fuel storage facilities at the Savannah River Site (SRS). This database provides a quick quantitative check to determine if SRS bound spent fuel is radiologically bounded by the Reference Fuel Assembly used in the L-Basin and RBOF authorization bases. The developed database considers pertinent characteristics of domestic and foreign research reactor fuel including exposure, fuel enrichment, irradiation time, cooling time, and fuel-to-moderator ratio. The supplied tables replace the time-consuming studies associated with authorization of SRS bound spent fuel with simple hand calculations. Additionally, the comprehensive database provides the means to overcome resource limitations, since a series of simple, yet conservative, hand calculations can now be performed in a timely manner and replace computational and technical staff requirements

Cash pooling

OpenAIRE

Lozovaya, Karina

2009-01-01

This work makes a mention of cash management. At next chapter describes two most known theoretical models of cash management -- Baumol Model and Miller-Orr Model. Principal part of work is about cash pooling, types of cash pooling, cash pooling at Czech Republic and influence of cash pooling over accounting and taxes.

Radiological performance of hot water layer system in open pool type reactor

Directory of Open Access Journals (Sweden)

Amr Abdelhady

2013-06-01

Full Text Available The paper presents the calculated dose rate carried out by using MicroShield code to show the importance of hot water layer system (HWL in 22 MW open pool type reactor from the radiation protection safety point of view. The paper presents the dose rate profiles over the pool surface in normal and abnormal operations of HWL system. The results show that, in case of losing the hot water layer effect, the radiation dose rate profiles over the pool surface will increase from values lower than the worker permissible dose limits to values very higher than the permissible dose limits.

Assessment of fuel damage of pool type research reactor in the case of fuel plates blockage

Energy Technology Data Exchange (ETDEWEB)

Jalil, Jafari; Samad, Khakshournia [AEOI, Karegar Ave. School of R and D of Nuclear Reactors and Accelerators, Teheran (Iran, Islamic Republic of); D' Auria, F. [Pisa Univ., DIMNP (Italy)

2007-07-01

Tehran Research Reactor (TRR) is a pool type 5 MW research reactor. It is assumed that external objects or debris that may fall down to reactor core cause obstruction of coolant flow through one of the fuel assemblies. Thermal hydraulic analysis of this event, using the RELAP5 system code has been studied. The reported transient is related to the partial and total obstruction of a single Fuel Element (FE) cooling channel of 27 FE equilibrium core of TRR. Such event constitutes a severe accident for this type of reactor since it may lead to local dryout and eventually to loss of the FE integrity. Two scenarios are analysed to emphasize the severity of the accident. The first one is a partial blockage of an average FE considering four different obstruction levels: 25%, 50%, 75% and 97% of nominal flow area. The second one is an extreme scenario consisting of total blockage of the same FE. This study constitutes the first step of a larger work which consists of performing a 3-dimensional simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic calculation included in the RELAP5 code. Main results obtained from the RELAP5 calculations are as following. First, in the case of flow blockage under 97% of the nominal flow area of an average FE, only an increase of the coolant and clad temperatures is observed without any consequences for the integrity of the FE. The mass flow rate remains sufficient to cool the clad safely. Secondly, in the case of total obstruction of the nominal flow area, it is seen that transient turns out to be a severe accident due to the dryout conditions are reached shortly and melting of the cladding occurs. Thirdly, the use of the point kinetic approach leads to conservative results. A best estimate simulation of such kind of transients requires the use of 3-dimensional kinetic calculations, which could be done using the current Coupled Codes

Reactivity accident analysis in MTR cores

International Nuclear Information System (INIS)

Waldman, R.M.; Vertullo, A.C.

1987-01-01

The purpose of the present work is the analysis of reactivity transients in MTR cores with LEU and HEU fuels. The analysis includes the following aspects: the phenomenology of the principal events of the accident that takes place, when a reactivity of more than 1$ is inserted in a critical core in less than 1 second. The description of the accident that happened in the RA-2 critical facility in September 1983. The evaluation of the accident from different points of view: a) Theoretical and qualitative analysis; b) Paret Code calculations; c) Comparison with Spert I and Cabri experiments and with post-accident inspections. Differences between LEU and HEU RA-2 cores. (Author)

Determination of power density distribution of fuel assemblies for research reactor by directly measuring the strontium-91 activities

International Nuclear Information System (INIS)

Yuan, Liq-Ji

1987-01-01

This work described the investigations of reactor core power peaking and three dimensional power density distribution of present core configuration of Tsing Hua Open-pool reactor (THOR). An experimental program, based on non-destructive fuel gamma scanning of 91 Sr activities, provides the data of fission density distribution for individual fuel pin of four-rod TRIGA-LEU cluster or for MTR-type fuel assembly. The informations are essentially important for the safety of reactor operation and for fuel management especially for the mixed loading with three different types of fuel at present. The relative power peaking values and the power density distribution for present core are discussed. (author)

Swimming-pool piles; Piles piscines

Energy Technology Data Exchange (ETDEWEB)

Trioulaire, M [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

In France two swimming-pool piles, Melusine and Triton, have just been set in operation. The swimming-pool pile is the ideal research tool for neutron fluxes of the order of 10{sup 13}. This type of pile can be of immediate interest to many research centres, but its cost must be reduced and a break with tradition should be observed in its design. It would be an advantage: - to bury the swimming-pool; - to reject the experimental channel; - to concentrate the cooling circuit in the swimming-pool; - to carry out all manipulations in the water; - to double the core. (author) [French] En France, deux piles piscines, Melusine et Triton, viennent d'entrer en service. La pile piscine est l'outil de recherche ideal pour des flux de neutrons de l'ordre de 10{sup 13}. Ce type de pile peut interesser des maintenant de nombreux centres de recherches mais il faut reduire son prix de revient et rompre avec le conformisme de sa conception. Il y a avantage: - a enterrer la piscine; - a supprimer les canaux experimentaux; - a concentrer le circuit de refrigeration dans la piscine; - a effectuer toutes les manipulations dans l'eau; - a doubler le coeur. (auteur)

Experience and research with the IEA-R1 Brazilian reactor

International Nuclear Information System (INIS)

Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W.

1982-06-01

The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lightwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. The research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology, are presented. (Author) [pt

Experience and research with the IEA-R1 Brazilian reactor

International Nuclear Information System (INIS)

Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W. de.

1982-06-01

The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lighwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. It is also presented the research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology. (Author) [pt

Simulation of the Gamma Dose Rate in Loss of Pool Water Accident of the Second Egyptian Research Reactor ETRR-2

International Nuclear Information System (INIS)

Amin, E.; Saleh, H.; Ashoub, N.

2000-01-01

The Second Egyptian Research Reactor ETRR-2, is a pool type reactor, a sudden loss of pool water resulting of leaving the core region un-covered. The reactor core is surrounded by chimney chambers whose water is isolated from pool water. This accident would lead to significant external dose. A model is developed and is used to calculate the dose rates for key access and traffic plans from indirect line of sight of the core have a maximum dose rate. The model developed uses the discrete ordinate method as implemented in the code DOT 3.5

Research progresses and future directions on pool boiling heat transfer

Directory of Open Access Journals (Sweden)

M. Kumar

2015-12-01

Full Text Available This paper reviews the previous work carried on pool boiling heat transfer during heating of various liquids and commodities categorized as refrigerants and dielectric fluids, pure liquids, nanofluids, hydrocarbons and additive mixtures, as well as natural and synthetic colloidal solutions. Nucleate pool boiling is an efficient and effective method of boiling because high heat fluxes are possible with moderate temperature differences. It is characterized by the growth of bubbles on a heated surface. It occurs during boiling of liquids for excess temperature ranging from 5 to 30 °C in various processes related to high vaporization of liquid for specific purposes like sugarcane juice heating for jaggery making, milk heating for khoa making, steam generation, cooling of electronic equipments, refrigeration and etcetera. In this review paper, pool boiling method during heating of liquids for specific purpose is depicted. It is inferred that enhancement in pool boiling heat transfer is a challenging and complex task. Also, recent research and use of various correlations for natural convection pool boiling is reviewed.

Validation and verification of the MTR{sub P}C thermohydraulic package

Energy Technology Data Exchange (ETDEWEB)

Doval, Alicia [INVAP S.E., Bariloche, Rio Negro (Argentina). Nuclear Engineering Dept.]. E-mail: doval@invap.com.ar

1998-07-01

The MTR{sub P}C v2.6 is a computational package developed for research reactor design and calculation. It covers three of the main aspects of a research reactor: neutronic, shielding and thermohydraulic. In this work only the thermohydraulic package will be covered, dealing with verification and validation aspects. The package consists of the following steady state programs: CAUDVAP 2.60 for the hydraulic calculus, estimates the velocity distribution through different parallel channels connected to a common inlet and outlet common plenum. TERMIC 1H v3.0, used for the thermal design of research reactors, provides information about heat flux for a given maximum wall temperature, onset of nucleate boiling, redistribution phenomena and departure from nucleate boiling. CONVEC V3.0 allows natural convection calculations, giving information on heat fluxes for onset of nucleate boiling, pulsed and burn-out phenomena as well as total coolant flow. Results have been validated against experimental values and verified against theoretical and computational programmes results, showing a good agreement. (author)

Seismic sloshing experiments of large pool-type fast breeder reactors

International Nuclear Information System (INIS)

Sakurai, A.; Masuko, Y.; Kurihara, C.; Ishihama, K.; Yashiro, T.; Rodwell, E.

1989-01-01

This paper presents the results of seismic sloshing experiments performed on large pool-type LMFBR vessels. Two types of tests were performed. The first type of test was designed to understand the basis phenomena of sloshing (limited to linear sloshing only) and evaluate the effects of the deck-mounted components (i.e., IHXs, pumps, and UIS) on sloshing wave heights using a 1/10-scale model (diameter 2.23 m x H 1.03 m) of the LSPB 1340 MWe pool plant. The second type of test was designed to evaluate the structural integrity of the thermal baffles of the roof-deck to withstand sloshing impulsive pressures (focused on nonlinear sloshing), using a two-dimensional 1/3-scale model (L 8 m x W 3 m x H 2.6 m) of a typical 1000 MWe pool plant. The results of the linear sloshing tests have shown that: 1. the vessel wall stiffness has no effect on the sloshing natural frequency; 2. sloshing wave heights are lowered by 30% to 50% in the presence of the deck-mounted components; and 3. damping factors of sloshing are not influenced by the wall stiffness while they are increased by the presence of the deck-mounted components. The results of the nonlinear sloshing tests are that: 1. the maximum impulsive pressure occurs when the first effective wave strikes at the roof-deck, and thereafter the impulsive pressure decreases irrespective of the impact velocity of the fluid; 2. the first effective wave refers to the case in which the height of the fluid free surface becomes nearly twice the height of the cover gas space; and 3. the structural integrity of the thermal baffles for the roof-deck against the sloshing load was confirmed. In addition to these results, two sloshing-caused problems were identified. The first one is the spillover of hot sodium into the gas-dam type thermal insulator. The second one is cover-gas entrainment into sodium which might lead to a transient overpower (TOP) incident because of the presence of gas bubbles in the reactor core. (orig./HP)

Diatomite Type Filters for Swimming Pools. Standard No. 9, Revised October, 1966.

Science.gov (United States)

National Sanitation Foundation, Ann Arbor, MI.

Pressure and vacuum diatomite type filters are covered in this standard. The filters herein described are intended to be designed and used specifically for swimming pool water filtration, both public and residential. Included are the basic components which are a necessary part of the diatomite type filter such as filter housing, element supports,…

Immobilisation of MTR waste in cement (product evaluation). Final report. December 1987

International Nuclear Information System (INIS)

Howard, C.G.; Lee, D.J.

1988-01-01

The enriched uranium/aluminium fuel used in Material Testing Reactors is reprocessed at Dounreay Nuclear Power Development Establishment (DNE). The main chemical component of the liquid waste produced by this process is acid deficient aluminium nitrate. This is stored in stainless steel tanks at DNE. As a result of work carried out under the UKAEA radioactive waste management programme a decision was taken to immobilise the waste in cement. The programme had two main components, plant design and development of the cementation process. The plant for the cementation of MTR waste is under construction and will be commissioned in 1988/9. The primary objective of this project is to find a suitable process for changing the highly mobile radioactive waste into an inert stable solid. Work carried out on the development of the immobilisation process showed that a conditioning stage (neutralisation) is required to make the acid waste compatible with cement. Small scale experiments showed that adding Ordinary Portland Cement blended with ground granulated Blast Furnace Slag to Simulant MTR Liquor produces an acceptable product. The process has been demonstrated at full scale (200 litres) and the products have been subjected to an extensive programme of destructive and non-destructive testing. Specimens have been tested up to 1200 days after manufacture and show no significant signs of deterioration even when stored underwater or when subjected to freeze thaw cycling. Development work has also shown that the process can successfully immobilise simulant MTR liquor over a wide range of liquor concentrations. The programme therefore successfully produced a formulation that met all the requirements of both the process and product specification. (author)

Evaluation of analysis method standardless by WDXRF and EDXRF of aluminum powder used in MTR type fuel

International Nuclear Information System (INIS)

Scapin, Valdirene O.; Salvador, Vera L.R.; Cotrim, Marycel E.B.; Pires, Maria A.F.; Scapin, Marcos A.

2011-01-01

The nuclear fuel used in IEA-R1m reactor at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP) is the MTR type. This fuel is compound of a core (U 3 Si 2 -Al dispersion briquette) wrapped in an aluminum plate with two cladding (superior and inferior) both in aluminum. The fuel element efficiency depends on the quality control of U 3 Si 2 and aluminum. For aluminum should be checked the impurities levels such as Si, Mn, Fe, Co, Cu, Zn and others and Al total . Aiming to provide a quick method, multielemental and non-destructive, the performance of the wavelength dispersive (WDXRF) and energy dispersive (EDXRF) X-ray fluorescence techniques, using the curve instrument sensitivity curve method, also known like standard less analysis, was evaluated. This method allows the determination from the element boron (Z=5) to uranium (Z=92) with concentrations ranging from 0.001 to 99.99% without the need for individual calibration curve and chemical pretreatments in the sample preparation. The results were compared with calibration curve method data, using statistical tests tools. By multivariate analysis of all the experimental data, especially by the discriminant analysis (DA) and cluster analysis (CA), respectively, it was possible to evaluate a correlation between variables of the applied analytical methods could be interpreted in context to qualify the fuels by XRF technique and method standard less. The results showed that the proposed method is satisfactory for both spectrometers; however it was found that the WDXRF presents the greatest conformity degree. (author)

AAA-ATPase NVL2 acts on MTR4-exosome complex to dissociate the nucleolar protein WDR74

Energy Technology Data Exchange (ETDEWEB)

Hiraishi, Nobuhiro; Ishida, Yo-ichi; Nagahama, Masami, E-mail: nagahama@my-pharm.ac.jp

2015-11-20

Nuclear VCP-like 2 (NVL2) is a chaperone-like nucleolar ATPase of the AAA (ATPase associated with diverse cellular activities) family, which exhibits a high level of amino acid sequence similarity with the cytosolic AAA-ATPase VCP/p97. These proteins generally act on macromolecular complexes to stimulate energy-dependent release of their constituents. We previously showed that NVL2 interacts with RNA processing/degradation machinery containing an RNA helicase MTR4/DOB1 and an exonuclease complex, nuclear exosome, and involved in the biogenesis of 60S ribosomal subunits. These observations implicate NVL2 as a remodeling factor for the MTR4-exosome complex during the maturation of pre-ribosomal particles. Here, we used a proteomic screen and identified a WD repeat-containing protein 74 (WDR74) as a factor that specifically dissociates from this complex depending on the ATPase activity of NVL2. WDR74 shows weak amino acid sequence similarity with the yeast ribosome biogenesis protein Nsa1 and is co-localized with NVL2 in the nucleolus. Knockdown of WDR74 decreases 60S ribosome levels. Taken together, our results suggest that WDR74 is a novel regulatory protein of the MTR4-exsosome complex whose interaction is regulated by NVL2 and is involved in ribosome biogenesis. - Highlights: • WDR74 accumulates in MTR4-exosome complex upon expression of dominant-negative NVL2. • WDR74 is co-localized with NVL2 in the nucleolus. • WDR74, along with NVL2, is involved in the synthesis of 60S ribosomal subunits.

AAA-ATPase NVL2 acts on MTR4-exosome complex to dissociate the nucleolar protein WDR74

International Nuclear Information System (INIS)

Hiraishi, Nobuhiro; Ishida, Yo-ichi; Nagahama, Masami

2015-01-01

Nuclear VCP-like 2 (NVL2) is a chaperone-like nucleolar ATPase of the AAA (ATPase associated with diverse cellular activities) family, which exhibits a high level of amino acid sequence similarity with the cytosolic AAA-ATPase VCP/p97. These proteins generally act on macromolecular complexes to stimulate energy-dependent release of their constituents. We previously showed that NVL2 interacts with RNA processing/degradation machinery containing an RNA helicase MTR4/DOB1 and an exonuclease complex, nuclear exosome, and involved in the biogenesis of 60S ribosomal subunits. These observations implicate NVL2 as a remodeling factor for the MTR4-exosome complex during the maturation of pre-ribosomal particles. Here, we used a proteomic screen and identified a WD repeat-containing protein 74 (WDR74) as a factor that specifically dissociates from this complex depending on the ATPase activity of NVL2. WDR74 shows weak amino acid sequence similarity with the yeast ribosome biogenesis protein Nsa1 and is co-localized with NVL2 in the nucleolus. Knockdown of WDR74 decreases 60S ribosome levels. Taken together, our results suggest that WDR74 is a novel regulatory protein of the MTR4-exsosome complex whose interaction is regulated by NVL2 and is involved in ribosome biogenesis. - Highlights: • WDR74 accumulates in MTR4-exosome complex upon expression of dominant-negative NVL2. • WDR74 is co-localized with NVL2 in the nucleolus. • WDR74, along with NVL2, is involved in the synthesis of 60S ribosomal subunits.

Sharing the load: Mex67-Mtr2 cofunctions with Los1 in primary tRNA nuclear export.

Science.gov (United States)

Chatterjee, Kunal; Majumder, Shubhra; Wan, Yao; Shah, Vijay; Wu, Jingyan; Huang, Hsiao-Yun; Hopper, Anita K

2017-11-01

Eukaryotic transfer RNAs (tRNAs) are exported from the nucleus, their site of synthesis, to the cytoplasm, their site of function for protein synthesis. The evolutionarily conserved β-importin family member Los1 (Exportin-t) has been the only exporter known to execute nuclear export of newly transcribed intron-containing pre-tRNAs. Interestingly, LOS1 is unessential in all tested organisms. As tRNA nuclear export is essential, we previously interrogated the budding yeast proteome to identify candidates that function in tRNA nuclear export. Here, we provide molecular, genetic, cytological, and biochemical evidence that the Mex67-Mtr2 (TAP-p15) heterodimer, best characterized for its essential role in mRNA nuclear export, cofunctions with Los1 in tRNA nuclear export. Inactivation of Mex67 or Mtr2 leads to rapid accumulation of end-matured unspliced tRNAs in the nucleus. Remarkably, merely fivefold overexpression of Mex67-Mtr2 can substitute for Los1 in los1 Δ cells. Moreover, in vivo coimmunoprecipitation assays with tagged Mex67 document that the Mex67 binds tRNAs. Our data also show that tRNA exporters surprisingly exhibit differential tRNA substrate preferences. The existence of multiple tRNA exporters, each with different tRNA preferences, may indicate that the proteome can be regulated by tRNA nuclear export. Thus, our data show that Mex67-Mtr2 functions in primary nuclear export for a subset of yeast tRNAs. © 2017 Chatterjee et al.; Published by Cold Spring Harbor Laboratory Press.

Jordan Research and Training Reactor (JRTR) Utilization Facilities

International Nuclear Information System (INIS)

Xoubi, N.

2013-01-01

Jordan Research and Training Reactor (JRTR) is a 5 MW light water open pool multipurpose reactor that serves as the focal point for Jordan National Nuclear Centre, and is designed to be utilized in three main areas: Education and training, nuclear research, and radioisotopes production and other commercial and industrial services. The reactor core is composed of 18 fuel assemblies, MTR plate type 19.75% enriched uranium silicide (U 3 Si 2 ) in aluminium matrix, and is reflected on all sides by beryllium and graphite. The reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45×10 14 cm -2 s -1 , and is controlled by a Hafnium control absorber rod and B 4 C shutdown rod. The reactor is designed to include laboratories and classrooms that will support the establishment of a nuclear reactor school for educating and training students in disciplines like nuclear engineering, reactor physics, radiochemistry, nuclear technology, radiation protection, and other related scientific fields where classroom instruction and laboratory experiments will be related in a very practical and realistic manner to the actual operation of the reactor. JRTR is designed to support advanced nuclear research as well as commercial and industrial services, which can be preformed utilizing any of its 35 experimental facilities. (author)

Protection system for minimizing the consequences of a flow blockage incident at a pool-type research reactor

International Nuclear Information System (INIS)

de Vries, J.W.; van Dam, H.; Gysler, G.

1990-01-01

Safety analysis activities were performed for the HOR, a pool-type research reactor with plate-type fuel elements and a maximum licensed power of 3 MW. Following internationally accepted guidelines, a wide variety of possible process disturbances has been considered. For the HOR the most aggravating accident conditions could result from a sudden flow blockage of cooling channels. If this event occurs in the high power density region of the core, a decrease of the hot channel flow either causes flow reversal or prompts burnout. Unless the reactor is scrammed in time, the fuel plates will heat up rapidly and local melting will occur with possible propagation of voiding and burnout to adjacent channels. In the analysis, melting of the cladding has been considered by using a simplified model approach. The number of voided coolant channels, as well as the propagation rate of fuel plates reaching locally the melting temperature, were calculated for different conditions of operation. In order to reduce the risk of a fuel melt accident occurring at the HOR, the protection system features a special design option. The system recognizes cooling channel voiding by detection of a sudden decrease of neutron flux. In the present work, it has been shown that a flow blockage incident can be detected in the early stages of development. Also, in accordance with the results of experimental tests, it can be concluded that in many cases melting of fuel plates will be effectively prevented. If such an accident occurs on a very fast time scale, at least the radiological consequences are significantly mitigated by preventing propagation, thus limiting the number of molten fuel plates

Application of nonlinear nodal diffusion method for a small research reactor

International Nuclear Information System (INIS)

Jaradat, Mustafa K.; Alawneh, Luay M.; Park, Chang Je; Lee, Byungchul

2014-01-01

Highlights: • We applied nonlinear unified nodal method for 10 MW IAEA MTR benchmark problem. • TRITION–NEWT system was used to obtain two-group burnup dependent cross sections. • The criticality and power distribution compared with reference (IAEA-TECDOC-233). • Comparison between different fuel materials was conducted. • Satisfactory results were provided using UNM for MTR core calculations. - Abstract: Nodal diffusion methods are usually used for LWR calculations and rarely used for research reactor calculations. A unified nodal method with an implementation of the coarse mesh finite difference acceleration was developed for use in plate type research reactor calculations. It was validated for two PWR benchmark problems and then applied for IAEA MTR benchmark problem for static calculations to check the validity and accuracy of the method. This work was conducted to investigate the unified nodal method capability to treat material testing reactor cores. A 10 MW research reactor core is considered with three calculation cases for low enriched uranium fuel depending on the core burnup status of fresh, beginning-of-life, and end-of-life cores. The validation work included criticality calculations, flux distribution, and power distribution; in addition, a comparison between different fuel materials with the same uranium content was conducted. The homogenized two-group cross sections were generated using the TRITON–NEWT system. The results were compared with a reference, which was taken from IAEA-TECDOC-233. The unified nodal method provides satisfactory results for an all-rod out case, and the three-dimensional, two-group diffusion model can be considered accurate enough for MTR core calculations

Preliminary developments of MTR plates with uranium nitride

Energy Technology Data Exchange (ETDEWEB)

Durand, J.P.; Laudamy, P. [CERCA, Romans (France); Richter, K. [Institut fuer Transurane, Karlsruhe (Germany)

1997-08-01

In the opinion of CERCA, the total weight of Uranium per MTR plate (without changing the external dimensions) cannot be further increased using U{sub 3}Si{sub 2}. Limits have been reached on plates with a thicker meat or loaded to 6g Ut/cm{sup 3}. The use of a denser fuel like Uranium mononitride could permit an increase in these limits. A collaboration between the Institute for Transuranium Elements (ITU), Joint Research Centre of the European Commission, and CERCA has been set ut. The preliminary studies at the ITU to check compatibility between aluminium and UN proved that there are no metallurgical interactions below 500{degrees}C. Feasibility of the manufacturing, on a laboratory scale at CERCA, of depleted Uranium mononitride plates loaded to 7 g Ut/cm{sup 3} has been demonstrated. The manufacturing process, however, is only one aspect of the development of a new fuel. The experience gained in the case of U{sub 3}Si{sub 2} has shown that the development of a new fuel requires considerable time and financial investment. Such a development certainly represents an effort of about 10 years.

Preliminary developments of MTR plates with uranium nitride

International Nuclear Information System (INIS)

Durand, J.P.; Laudamy, P.; Richter, K.

1997-01-01

In the opinion of CERCA, the total weight of Uranium per MTR plate (without changing the external dimensions) cannot be further increased using U 3 Si 2 . Limits have been reached on plates with a thicker meat or loaded to 6g Ut/cm 3 . The use of a denser fuel like Uranium mononitride could permit an increase in these limits. A collaboration between the Institute for Transuranium Elements (ITU), Joint Research Centre of the European Commission, and CERCA has been set ut. The preliminary studies at the ITU to check compatibility between aluminium and UN proved that there are no metallurgical interactions below 500 degrees C. Feasibility of the manufacturing, on a laboratory scale at CERCA, of depleted Uranium mononitride plates loaded to 7 g Ut/cm 3 has been demonstrated. The manufacturing process, however, is only one aspect of the development of a new fuel. The experience gained in the case of U 3 Si 2 has shown that the development of a new fuel requires considerable time and financial investment. Such a development certainly represents an effort of about 10 years

Simulation of the gamma dose rate in a loss of pool water accident of the second Egyptian research reactor ET-RR-2

International Nuclear Information System (INIS)

Amin, E.; Saleh, H.G.; Ashoub, N.

2002-01-01

The second Egyptian research reactor ET-RR-2, is a pool type reactor. A sudden loss of pool water would leave the core region uncovered. The reactor core is surrounded by chimney chambers with water isolated from the pool water. This accident would lead to significant external doses. A model is developed and used to calculate the dose rates for key access-areas and traffic plans from indirect line of sight of the core which have a maximum dose rate. The model developed uses the discrete ordinate method as implemented in the code DOT3.5. (orig.) [de

Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

International Nuclear Information System (INIS)

Pond, R.B.; Matos, J.E.

1996-05-01

As part of the Department of Energy's spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, are not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report

Pool-type reactor

International Nuclear Information System (INIS)

Hopkins, S.R.

1977-01-01

This invention relates to a pool nuclear reactor fitted with a perfected system to raise the buckets into a vertical position at the bottom of a channel. This reactor has an inclined channel to guide a bucket containing a fuel assembly to introduce it into the reactor jacket or extract it therefrom and a damper at the bottom of the channel to stop the drop of the bucket. An upright vertically movable rod has a horizontally articulated arm with a hook. This can pivot to touch a radial lug on the bucket and pivot the bucket around its base in a vertical position, when the rod moves up [fr

Structural characterization of the principal mRNA-export factor Mex67–Mtr2 from Chaetomium thermophilum

Energy Technology Data Exchange (ETDEWEB)

Aibara, Shintaro; Valkov, Eugene; Lamers, Meindert H. [MRC Laboratory of Molecular Biology, Francis Crick Avenue, Cambridge Biomedical Campus, Cambridge CB2 0QH (United Kingdom); Dimitrova, Lyudmila; Hurt, Ed [Biochemie-Zentrum der Universität Heidelberg, Im Neuenheimer Feld 328, 69120 Heidelberg (Germany); Stewart, Murray, E-mail: ms@mrc-lmb.cam.ac.uk [MRC Laboratory of Molecular Biology, Francis Crick Avenue, Cambridge Biomedical Campus, Cambridge CB2 0QH (United Kingdom)

2015-06-27

The crystal structures of the individual domains of the Mex67–Mtr2 complex from C. thermophilum have been determined and their arrangement in solution has been studied by SAXS. Members of the Mex67–Mtr2/NXF–NXT1 family are the principal mediators of the nuclear export of mRNA. Mex67/NXF1 has a modular structure based on four domains (RRM, LRR, NTF2-like and UBA) that are thought to be present across species, although the level of sequence conservation between organisms, especially in lower eukaryotes, is low. Here, the crystal structures of these domains from the thermophilic fungus Chaetomium thermophilum are presented together with small-angle X-ray scattering (SAXS) and in vitro RNA-binding data that indicate that, not withstanding the limited sequence conservation between different NXF family members, the molecules retain similar structural and RNA-binding properties. Moreover, the resolution of crystal structures obtained with the C. thermophilum domains was often higher than that obtained previously and, when combined with solution and biochemical studies, provided insight into the structural organization, self-association and RNA-binding properties of Mex67–Mtr2 that facilitate mRNA nuclear export.

Structural characterization of the principal mRNA-export factor Mex67–Mtr2 from Chaetomium thermophilum

International Nuclear Information System (INIS)

Aibara, Shintaro; Valkov, Eugene; Lamers, Meindert H.; Dimitrova, Lyudmila; Hurt, Ed; Stewart, Murray

2015-01-01

The crystal structures of the individual domains of the Mex67–Mtr2 complex from C. thermophilum have been determined and their arrangement in solution has been studied by SAXS. Members of the Mex67–Mtr2/NXF–NXT1 family are the principal mediators of the nuclear export of mRNA. Mex67/NXF1 has a modular structure based on four domains (RRM, LRR, NTF2-like and UBA) that are thought to be present across species, although the level of sequence conservation between organisms, especially in lower eukaryotes, is low. Here, the crystal structures of these domains from the thermophilic fungus Chaetomium thermophilum are presented together with small-angle X-ray scattering (SAXS) and in vitro RNA-binding data that indicate that, not withstanding the limited sequence conservation between different NXF family members, the molecules retain similar structural and RNA-binding properties. Moreover, the resolution of crystal structures obtained with the C. thermophilum domains was often higher than that obtained previously and, when combined with solution and biochemical studies, provided insight into the structural organization, self-association and RNA-binding properties of Mex67–Mtr2 that facilitate mRNA nuclear export

Design of the Demineralized Water Make-up Line to Maintain the Normal Pool Water Level of the Reactor Pool in the Research Reactor

International Nuclear Information System (INIS)

Yoon, Hyun Gi; Choi, Jung Woon; Yoon, Ju Hyeon; Chi, Dae Young

2012-01-01

In many research reactors, hot water layer system (HWLS) is used to minimize the pool top radiation level. Reactor pool divided into the hot water layer at the upper part of pool and the cold part below the hot water layer with lower temperature during normal operation. Water mixing between these layers is minimized because the hot water layer is formed above cold water. Therefore the hot water layer suppresses floatation of cold water and reduces the pool top radiation level. Pool water is evaporated form the surface to the building hall because of high temperature of the hot water layer; consequently the pool level is continuously fallen. Therefore, make-up water is necessary to maintain the normal pool level. There are two way to supply demineralized water to the pool, continuous and intermittent methods. In this system design, the continuous water make-up method is adopted to minimize the disturbance of the reactor pool flow. Also, demineralized water make-up is connected to the suction line of the hot water layer system to raise the temperature of make-up water. In conclusion, make-up demineralized water with high temperature is continuously supplied to the hot water layer in the pool

Pool Structures: A New Type of Interaction Zones of Lithospheric Plate Flows

Science.gov (United States)

Garetskyi, R. G.; Leonov, M. G.

2018-02-01

Study of tectono-geodynamic clusters of the continental lithosphere (the Sloboda cluster of the East European Platform and the Pamir cluster of Central Asia) permitted identification of pool structures, which are a specific type of zone of intraplate interaction of rock masses.

Corrosion behavior of spent MTR fuel elements in a drowned salt mine repository

International Nuclear Information System (INIS)

Brodda, B.G.; Fachinger, J.

1995-01-01

Spent MTR fuel from German Material Test Reactors will not be reprocessed, but stored in a final salt repository in the deep geologic underground. Fuel elements will be placed in POLLUX containers, which are assumed to resist the corrosive attack of an accidentally formed concentrated salt brine for about 500 years. After a container failure the brine would contact the fuel element, corrode the aluminum plating and possibly leach radionuclides from the fuel. A source term for the calculation of radionuclide mobilization results from the investigation of the behavior of MTR fuel in this scenario, which has to be considered for the long-term safety analysis of a deep mined rock salt repository. Experiments with the different plating materials show that the considered aluminum alloys will not resist the corrosive attack of a brine solution, especially in the presence of iron, under the conditions in a drowned salt mine repository. Although differences in the corrosion rates of about two orders of magnitude were observed when applying different parameter sets, the deterioration must be considered to be almost instantaneous in geological terms. Radionuclides are mobilized from irradiated MTR fuel, when the meat of the fuel element becomes accessible to the brine solution. It seems, however, that the radionuclides are effectively trapped by the aluminum hydroxide formed, as the activity concentrations in the brine solution soon reach a constant level with the progressing corrosion of the cladding aluminum. In the presence of iron a more significant initial release was observed, but also in this case an equilibrium activity seems to be reached as a consequence of radionuclide trapping

Comparison Of ²⁵²Cf Time Correlated Induced Fisssion With AmLi Induced Fission On Fresh MTR Research Reactor Fuel

Energy Technology Data Exchange (ETDEWEB)

Joshi, Jay Prakash [Los Alamos National Laboratory

2017-03-30

The effective application of international safeguards to research reactors requires verification of spent fuel as well as fresh fuel. To accomplish this goal various nondestructive and destructive assay techniques have been developed in the US and around the world. The Advanced Experimental Fuel Counter (AEFC) is a nondestructive assay (NDA) system developed at Los Alamos National Laboratory (LANL) combining both neutron and gamma measurement capabilities. Since spent fuel assemblies are stored in water, the system was designed to be watertight to facilitate underwater measurements by inspectors. The AEFC is comprised of six ³He detectors as well as a shielded and collimated ion chamber. The ³He detectors are used for active and passive neutron coincidence counting while the ion chamber is used for gross gamma counting. Active coincidence measurement data is used to measure residual fissile mass, whereas the passive coincidence measurement data along with passive gamma measurement can provide information about burnup, cooling time, and initial enrichment. In the past, most of the active interrogation systems along with the AEFC used an AmLi neutron interrogation source. Owing to the difficulty in obtaining an AmLi source, a ²⁵²Cf spontaneous fission (SF) source was used during a 2014 field trail in Uzbekistan as an alternative. In this study, experiments were performed to calibrate the AEFC instrument and compare use of the ²⁵²Cf spontaneous fission source and the AmLi (α,n) neutron emission source. The ²⁵²Cf source spontaneously emits bursts of time-correlated prompt fission neutrons that thermalize in the water and induce fission in the fuel assembly. The induced fission (IF) neutrons are also time correlated resulting in more correlated neutron detections inside the ³He detector, which helps reduce the statistical errors in doubles when using the ²⁵²Cf interrogation source instead of

Neutronic analysis of HEU to LEU conversion calculation for AEOI 5 MW pool-type MTR fuel research reactor core

International Nuclear Information System (INIS)

Pazirandeh, A.; Lutz, D.; Bartsch, G.

1987-07-01

The possibility of converting HEU(93%) fuel to LEU(20%) fuel without or with slight alteration to the fuel element geometry is discussed. The fuel density varies between 1.7 to 4.1 g U-235/cm. In cross section generation a unit cell with an extra zone to account for extra Al and water was considered. In burnup calculations a sequential shuffling pattern was assumed with fixed position control fuel elements. A cross section data set in 45 energy groups were generated using RSYST/CGM system using the cross section library JFET. Then for 2D-diffusion calculations homogenized and condensed 5 energy group cross sections were prepared. (orig./HP)

Back-end of the research reactor fuel cycle

International Nuclear Information System (INIS)

Gruber, Gehard J.

1996-01-01

This paper outlines the status of topics and issues related to: (1) Research Reactor Spent Nuclear Fuel Return to the U.S., including policy, shipments and ports of entry, management sites, fees, storage technologies, contracts, actual shipment, and legal process, (2) UKAEA: MTR Spent Nuclear Fuel Reprocessing, (3) COGEMA: MTR Spent Nuclear Fuel Reprocessing, and (4) Intermediate Storage + Direct Disposal for Research Reactors. (author)

Activity of corrosion products in pool type reactors with ascending flow in the core

International Nuclear Information System (INIS)

Andrade e Silva, Graciete S. de; Queiroz Bogado Leite, Sergio de

1995-01-01

A model for the activity of corrosion products in the water of a pool type reactor with ascending flow is presented. The problem is described by a set of coupled differential equations relating the radioisotope concentrations in the core and pool circuits and taking into account two types of radioactive sources: i) those from radioactive species formed in the fuel cladding, control elements, reflector, etc, and afterwards released to the primary stream by corrosion (named reactor sources) and ii) those formed from non radioactive isotopes entering the primary stream by corrosion of the circuit components and being activated when passing through the core (named circuit sources). (author). 6 refs, 3 figs, 4 tabs

MTR2: a discriminator and dead-time module used in counting systems

International Nuclear Information System (INIS)

Bouchard, J.

2000-01-01

In the field of radioactivity measurement, there is a constant need for highly specialized electronic modules such as ADCs, amplifiers, discriminators, dead-time modules, etc. But sometimes it is almost impossible to find on the market the modules having the performances corresponding to our needs. The purpose of the module presented here, called MTR2 (Module de Temps-mort Reconductible), is to process, in terms of pulse height discrimination and dead-time corrections, the pulses delivered by the detectors used in counting systems. This dead-time, of the extendible type, is triggered by both the positive and negative parts of the incoming pulse and the dead-time corrections are made according to the live-time method. This module, which has been developed and tested at LPRI, can be used alone in simple counting channels or in more complex systems such as coincidence systems. The philosophy governing the choice and the implementation of this type of dead-time as well as the system used for the dead-time corrections is presented. The electronic scheme and the performances are also presented. This module is available in the NIM standard

Experience on wet storage spent fuel sipping at IEA-R1 Brazilian research reactor

International Nuclear Information System (INIS)

Perrotta, J.A.; Terremoto, L.A.A.; Zeituni, C.A.

1998-01-01

The IEA-R1 research reactor of the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) is a pool type reactor of B and W design, that has been operating since 1957 at a power of 2 MW. Irradiated (spent) fuels have been stored at the facility during the various years of operation. At present there are 40 spent fuel assemblies at dry storage, 79 spent fuel assemblies at wet storage and 30 fuel assemblies in the core. The oldest fuels are of United States origin, made with U-Al alloy, both of LEU and HEU MTR fuel type. Many of these fuel assemblies have corrosion pits along their lateral fuel plates. These pits originate by galvanic corrosion between the fuel plate and the stainless steel storage racks. As a consequence of the possibility of sending the irradiated old fuels back the U.S.A., sipping tests were performed with the spent fuel assemblies. The reason for this was to evaluate their 137 Cs leaking rate, if any. This work describes the procedure and methodology used to perform the sipping tests with the fuel assemblies at the storage pool, and presents the results obtained for the 137 Cs sipping water activity for each fuel assembly. A correlation is made between the corrosion pits and the activity values measured. A 137 Cs leaking rate is determined and compared to the criteria established for canning spent fuel assemblies before shipment

Characterisation of the corrosion products of non-irradiated material test reactors fuel elements (MTR-FE)

Energy Technology Data Exchange (ETDEWEB)

Mazeina, L.; Curtius, H.; Fachinger, J. [Inst. for Safety Research and Reactor Technology, Research Centre Juelich (Germany)

2003-07-01

In a high concentrated Mg-rich brine a non-irradiated MTR-FE corroded. The formed corrosion products consists of an amorphous part and of hydrotalcites, which were identified as Mg-Al-hydrotalcites with chloride anions in the interlayer. (orig.)

Depletion Calculations for MTR Core Using MCNPX and Multi-Group Nodal Diffusion Methods

International Nuclear Information System (INIS)

Jaradata, Mustafa K.; Park, Chang Je; Lee, Byungchul

2013-01-01

In order to maintain a self-sustaining steady-state chain reaction, more fuel than is necessary in order to maintain a steady state chain reaction must be loaded. The introduction of this excess fuel increases the net multiplication capability of the system. In this paper MCNPX and multi-group nodal diffusion theory will be used for depletion calculations for MTR core. The eigenvalue and power distribution in the core will be compared for different burnup. Multi-group nodal diffusion theory with combination of NEWT-TRITON system was used to perform depletion calculations for 3Χ3 MTR core. 2G and 6G approximations were used and compared with MCNPX results for 2G approximation the maximum difference from MCNPX was 40 mk and for 6G approximation was 6 mk which is comparable to the MCNPX results. The calculated power using nodal code was almost the same MCNPX results. Finally the results of the multi-group nodal theory were acceptable and comparable to the calculated using MCNPX

Plenum separator system for pool-type nuclear reactors

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1983-01-01

This invention provides a plenum separator system for pool-type nuclear reactors which substantially lessens undesirable thermal effects on major components. A primary feature of the invention is the addition of one or more intermediate plena, containing substantially stagnant and stratified coolant, which separate the hot and cold plena and particularly the hot plena from critical reactor components. This plenum separator system also includes a plurality of components which together form a dual pass flow path annular region spaced from the reactor vessel wall by an annular gas space. The bypass flow through the flow path is relatively small and is drawn from the main coolant pumps and discharged to an intermediate plenum

Core conversion from rod to plate type fuel elements in research reactors

International Nuclear Information System (INIS)

Khattab, M.S.; Mina, A.R.

1997-01-01

Core thermalhydraulic analysis have been performed for rod and plate types fuel elements without altering the core bundles square grid spacer (68 mm, side) and coolant mass flow rate. The U O 2 -Mg, 10% enrichment rod type fuel elements are replaced by the MTR plate type, U-Al alloy of 20% enrichment. Coolant mass flux increased from 2000 kg/m 2 S to 5000 kg/m 2 S. Reactor power could be upgraded from 2 to 10 MW without significantly altering the steady state, thermal-hydraulic safety margins. Fuel, clad and coolant transient temperatures are determined inside the core hot channel during flow coast down using paret code. Residual heat removal system of 20% coolant capacity is necessary for upgrading reactor power to encounter the case of pumps off at 10 MW nominal operation. 6 figs., 2 tabs

Maternal age at birth and childhood type 1 diabetes: a pooled analysis of 30 observational studies

DEFF Research Database (Denmark)

Cardwell, Chris R; Stene, Lars C; Joner, Geir

2009-01-01

for potential confounders. Meta-analysis techniques were used to derive combined odds ratios and to investigate heterogeneity among studies. RESULTS: Data were available for 5 cohort and 25 case-control studies, including 14,724 cases of type 1 diabetes. Overall, there was, on average, a 5% (95% CI 2......OBJECTIVE: The aim if the study was to investigate whether children born to older mothers have an increased risk of type 1 diabetes by performing a pooled analysis of previous studies using individual patient data to adjust for recognized confounders. RESEARCH DESIGN AND METHODS: Relevant studies...... published before June 2009 were identified from MEDLINE, Web of Science, and EMBASE. Authors of studies were contacted and asked to provide individual patient data or conduct prespecified analyses. Risk estimates of type 1 diabetes by maternal age were calculated for each study, before and after adjustment...

MTR fuel element supply by CERCA through CECCN after the production transfer from NUKEM

International Nuclear Information System (INIS)

Hassel, H.W.

1991-01-01

The transfer of fuel element supply contracts, the corresponding Al-materials, structure parts, documents, uranium metal, customers related know-how, tools and equipment from NUKEM to CERCA has been completed, thus now giving a high flexibility for CERCA's workshop to fabricate and inspect large quantities of several types of fuel elements simultaneously. Based on this fact, on strategic planning for the next couple of years and on the fact that after 10 years of RERTR program the necessary high density fuel has been successfully developed and implemented, 'business as usual' in the field of fabrication has well become possible. The RERTR community should now use the great chance to concentrate all its efforts on problems which still strongly influence the fabrication and the use of MTR fuel elements: supply of enriched uranium,reprocessing capabilities and politics, transports of nuclear materials. (author)

Backfitting swimming pool reactors

International Nuclear Information System (INIS)

Roebert, G.A.

1978-01-01

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

Neutronics design of the next tokamak. (Swimming pool type)

International Nuclear Information System (INIS)

Seki, Y.; Iida, H.; Kitamura, K.; Minato, A.; Sako, K.; Mori, S.; Nishida, H.

1983-01-01

A swimming pool type tokamak reactor (SPTR) has been proposed in the Japan Atomic Energy Research Institute as a candidate for the next generation tokamak reactor after the JT-60. The concept of the SPTR evolved from an incentive to relieve the difficulties of repair and maintenance procedures of a tokamak reactor. After about two years of the reactor design studies, several advantages of the SPTR over the conventional tokamak reactors such as the ease of penetration shielding, reduction in solid radwaste have been shown. On the other hand, some drawbacks and uncertainties of the SPTR have also been pointed out but so far no serious defect negating the concept has been found. This paper describes the neutronics aspect of the SPTR based mostly on the result of one dimensional calculations. At first, the radiation shielding capability of water is compared with those of other candidate materials used in the blanket and shield of fusion reactors. Based on the result of the comparison and other requirements such as tritium breeding, thermal mechanical design, repair and maintenance procedures, the material arrangements of the blanket and shield are determined. The result of the blanket neutronics calculations, the radiation shielding calculations for the superconducting magnets, shutdown dose calculations are given together with major penetration shielding considerations. (author)

Evaporation rate measurement in the pool of IEAR-1 reactor

International Nuclear Information System (INIS)

Torres, Walmir Maximo; Cegalla, Miriam A.; Baptista Filho, Benedito Dias

2000-01-01

The surface water evaporation in pool type reactors affects the ventilation system operation and the ambient conditions and dose rates in the operation room. This paper shows the results of evaporation rate experiment in the pool of IEA-R1 research reactor. The experiment is based on the demineralized water mass variation inside cylindrical metallic recipients during a time interval. Other parameters were measured, such as: barometric pressure, relative humidity, environmental temperature, water temperature inside the recipients and water temperature in the reactor pool. The pool level variation due to water contraction/expansion was calculated. (author)

Irradiation experience of IPEN fuel at IEA-R1 research reactor

International Nuclear Information System (INIS)

Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

1998-01-01

IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

MTR (Materials Testing Reactors) cores fuel management. Application of a low enrichment reactor for the equilibrium and transitory core calculation

International Nuclear Information System (INIS)

Relloso, J.M.

1990-01-01

This work describes a methodology to define the equilibrium core and a MTR (Materials Testing Reactors) type reactor's fuel management upon multiple boundary conditions, such as: end cycle and permitted maximum reactivities, burn-up extraction and maximun number of movements by rechange. The methodology proposed allows to determine the best options through conceptual relations, prior to a detailed calculation with the core code, reducing the test number with these codes and minimizing in this way CPU cost. The way to better systematized search of transient cores from the first one to the equilibrium one is presented. (Author) [es

Consideration of BORAX-type reactivity accidents applied to research reactors

International Nuclear Information System (INIS)

Couturier, Jean; Meignen, Renaud; Bourgois, Thierry; Biaut, Guillaume; Mireau, Jean-Pierre; Natta, Marc

2011-01-01

Most of the research reactors discussed in this document are pool-type reactors in which the reactor vessel and some of the reactor coolant systems are located in a pool of water. These reactors generally use fuel in plate assemblies formed by a compact layer of uranium (or U 3 Si 2 ) and aluminium particles, sandwiched between two thin layers of aluminium serving as cladding. The fuel melting process begins at 660 deg. C when the aluminium melts, while the uranium (or U 3 Si 2 ) particles may remain solid. The accident that occurred in the American SL-1 reactor in 1961, together with tests carried out in the United States as of 1954 in the BORAX-1 reactor and then, in 1962, in the SPERT-1 reactor, showed that a sudden substantial addition of reactivity in this type of reactor could lead to explosive mechanisms caused by degradation, or even fast meltdown, of part of the reactor core. This is what is known as a 'BORAX-type' accident. The aim of this document is first to briefly recall the circumstances of the SL-1 reactor accident, the lessons learned, how this operational feedback has been factored into the design of various research reactors around the world and, second, to describe the approach taken by France with regard to this type of accident and how, led by IRSN, this approach has evolved in the last decade. (authors)

Study of structural attachments of a pool type LMFBR vessel through seismic analysis of a simplified three dimensional finite element model

International Nuclear Information System (INIS)

Ahmed, H.; Ma, D.

1979-01-01

A simplified three dimensional finite element model of a pool type LMFBR in conjunction with the computer program ANSYS is developed and scoping results of seismic analysis are produced. Through this study various structural attachments of a pool type LMFBR like the reactor vessel skirt support, the pump support and reactor shell-support structure interfaces are studied. This study also provides some useful results on equivalent viscous damping approach and some improvements to the treatment of equivalent viscous damping are recommended. This study also sets forth pertinent guidelines for detailed three dimensional finite element seismic analysis of pool type LMFBR

Roof loading and response following a HCDA in a pool-type reactor

International Nuclear Information System (INIS)

Lancefield, M.J.; Leigh, K.M.; Potter, R.; Staniforth, R.

1979-01-01

In a pool-type reactor the loading and response of the roof structure to a HCDA is important to safety analysis and design. The U.K. programme of experimental and theoretical work on this topic is described. Good progress in understanding and evaluating the complex processes has been made and this is illustrated by results from experimental and theoretical work. 5 refs

Adoption of ASME Code Section XI for ISI to Research Reactors

International Nuclear Information System (INIS)

Tawfik, Y.E.; El-sesy, I.A.; Shaban, H.I.; Ibrahim, M.M.

2002-01-01

ETRR-2 (Second Egyptian thermal research reactor) is a multi-purpose, pool- type reactor with an open water surface and variable core arrangement. The core power is 22 MWth, cooled and moderated by light water and with beryllium reflectors. It contains plate- type fuel elements (MTR type, 19.7% enriched uranium) with aluminum clad. The ETRR-2 reactor consist of 57 systems and around 200 subsystems. These systems contain many mechanical components such as tanks, pipes, valves, pumps, heat exchangers, cooling tower, air compressors, and supports. In this present work, a trial was made to adopt the general requirements of ASME code, section XI to ETRR-2 research reactor. ASME (American Society of Mechanical Engineers) boiler and pressure vessel Code, section XI, provides requirements for in-service inspection (ISI) and in-service testing (IST) of components and systems, and repair/replacement activities in a nuclear power plant. Also, IAEA (International Atomic Energy Authority) has published some recommendations for ISI for research reactors similar to that rules and requirements specified in ASME. The complete ISI program requires several steps that have to be performed in sequence. These steps are described in many logic flow charts (LFC's). These logic flow charts include; the general LFC's for all steps required to complete ISI program, the LFC's for examination requirements, the LFC's for flaw evaluation modules, and the LFC's for acceptability of welds for class 1 components. This program includes, also, the inspection program for welded parts of the reactor components during its lifetime. This inspection program is applied for each system and subsystem of ETRR-2 reactor. It includes the examination area type, the component type, the part to be examined, the weld type, the examination method, the inspection program schedule, and the detailed figures of the welded components. (authors)

High temperature ultrasonic sensor for fission gas characterization in MTR harsh environment

Science.gov (United States)

Gatsa, O.; Combette, P.; Rozenkrantz, E.; Fourmentel, D.; Destouches, C.; Ferrandis, J. Y. AD(; )

2018-01-01

In the contemporary world, the measurements in hostile environment is one of the predominant necessity for automotive, aerospace, metallurgy and nuclear plant. The measurement of different parameters in experimental reactors is an important point in nuclear power strategy. In the near past, IES (Institut d'Électronique et des Systèmes) on collaboration with CEA (Commissariat à l'Energie Atomique et aux Energies Alternatives) have developed the first ultrasonic sensor for the application of gas quantity determination that has been tested in a Materials Testing Reactor (MTR). Modern requirements state to labor with the materials that possess stability on its parameters around 350°C in operation temperature. Previous work on PZT components elaboration by screen printing method established the new basis in thick film fabrication and characterization in our laboratory. Our trials on Bismuth Titanate ceramics showed the difficulties related to high electrical conductivity of fabricated samples that postponed further research on this material. Among piezoceramics, the requirements on finding an alternative solution on ceramics that might be easily polarized and fabricated by screen printing approach were resolved by the fabrication of thick film from Sodium Bismuth Titanate (NBT) piezoelectric powder. This material exhibits high Curie temperature, relatively good piezoelectric and coupling coefficients, and it stands to be a good solution for the anticipated application. In this paper, we present NBT thick film fabrication by screen printing, characterization of piezoelectric, dielectric properties and material parameters studies in dependence of temperature. Relatively high resistivity in the range of 1.1013 Ohm.cm for fabricated thick film is explained by Aurivillius structure in which a-and b-layers form perovskite structure between oxides of c-layer. Main results of this study are presented and discussed in terms of feasibility for an application to a new sensor

High enrichment to low enrichment core's conversion. Technical securities

International Nuclear Information System (INIS)

Abbate, P.; Madariaga, M.R.

1990-01-01

This work presents the fulfillment of the technical securities subscribed by INVAP S.E. for the conversion of a high enriched uranium core. The reactor (of 5 thermal Mw), built in the 50's and 60's, is of the 'swimming pool' type, with light water and fuel elements of the curve plates MTR type, enriched at 93.15 %. These are neutronic and thermohydraulic securities. (Author) [es

Detection of fission products release in the research reactor 'RA' spent fuel storage pool

International Nuclear Information System (INIS)

Matausek, M.V.; Vukadin, Z.; Pavlovic, S.; Maksin, T.; Idakovic, Z.; Marinkovic, N.

1997-05-01

Spent fuel resulting from 25 years of operating the 6.5/10 MW thermal heavy water moderated and cooled research reactor RA at the VINCA Institute is presently all stored in the temporary spent fuel storage pool in the basement of the reactor building. In 1984, the reactor was shut down for refurbishment, which for a number of reasons has not yet been completed. Recent investigations show that independent of the future status of the research reactor, safe disposal of the so far irradiated fuel must be the subject of primary concern. The present status of the research reactor RA spent fuel storage pool at the VINCA Institute presents a serious safety problem. Action is therefore initiated in two directions. First, safety of the existing spent fuel storage should be improved. Second, transferring spent fuel into another, presumably dry storage space should be considered. By storing the previously irradiated fuel of the research reactor RA in a newly built storage space, sufficient free space will be provided in the existing spent fuel storage pool for the newly irradiated fuel when the reactor starts operation again. In the case that it would be decided to decommission the research reactor RA, the newly built storage space would provide safe disposal for the fuel irradiated so far

Swimming pool cleaner poisoning

Science.gov (United States)

Swimming pool cleaner poisoning occurs when someone swallows this type of cleaner, touches it, or breathes in ... The harmful substances in swimming pool cleaner are: Bromine ... copper Chlorine Soda ash Sodium bicarbonate Various mild acids

Blanket and vacuum vessel design of the next tokamak. (Swimming pool type)

International Nuclear Information System (INIS)

Iida, H.; Minato, A.; Kitamura, K.

1983-01-01

The structural design study of a reactor module for a swimming pool type reactor (SPTR) was conducted. Since pool water plays the role of radiation shielding in the SPTR, the module does not have a solid shield. It consists of tritium breeding blankets, divertor collector plates and a vacuum vessel. The object of this study is to show the reactor module design which has a simple structure and a sufficient tritium breeding ratio. A large coverage of the plasma chamber surface with tritium breeding blanket is essential in order to obtain a high tritium breeding ratio. A breeding blanket is also placed behind the divertor collector plate, i.e. in the upper and lower region, as well as in the outboard and inboard regions of the module. A concept in which the first wall is an integral part of the blanket is employed to minimize the thickness of structural and cooling material brazed in front of the breeding material (Li 2 O) and to enhance the tritium breeding capability. In order to simplify the module structure the vacuum vessel and breeding blanket is also integrated in the inboard region. One of the features inherent in the swimming pool type reactor is an additional external force on the vacuum vessel, namely hydraulic pressure. A detailed structural analysis of the vacuum vessel is performed. Divertor collector plates are assemblies of co-axial tubes. They minimize the electromagnetic force on the plate induced by the plasma disruption. A thermal and structural analysis and life time estimation of the first wall and divertor collector plates are performed. (author)

Atomistic simulation on charge mobility of amorphous tris(8-hydroxyquinoline) aluminum (Alq3): origin of Poole-Frenkel-type behavior.

Science.gov (United States)

Nagata, Yuki; Lennartz, Christian

2008-07-21

The atomistic simulation of charge transfer process for an amorphous Alq(3) system is reported. By employing electrostatic potential charges, we calculate site energies and find that the standard deviation of site energy distribution is about twice as large as predicted in previous research. The charge mobility is calculated via the Miller-Abrahams formalism and the master equation approach. We find that the wide site energy distribution governs Poole-Frenkel-type behavior of charge mobility against electric field, while the spatially correlated site energy is not a dominant mechanism of Poole-Frenkel behavior in the range from 2x10(5) to 1.4x10(6) V/cm. Also we reveal that randomly meshed connectivities are, in principle, required to account for the Poole-Frenkel mechanism. Charge carriers find a zigzag pathway at low electric field, while they find a straight pathway along electric field when a high electric field is applied. In the space-charge-limited current scheme, the charge-carrier density increases with electric field strength so that the nonlinear behavior of charge mobility is enhanced through the strong charge-carrier density dependence of charge mobility.

Decommissioning of the MTR-605 process water building at the Idaho National Engineering Laboratory. Final report

International Nuclear Information System (INIS)

Browder, J.H.; Wills, E.L.

1985-01-01

Decontamination and decommissioning (D and D) of the unused radioactively contaminated portions of the MTR-605 building at the Test Reactor Area of the Idaho National Engineering Laboratory has been completed; this final report describes the D and D project. The building is a two-story concrete structure that was used to house piping systems to channel and control coolant water flow for the Materials Testing Reactor (MTR), a 40 MW (thermal) light water test reactor that was operated from 1952 until 1970 and then deactivated. D and D project objectives were to reduce potential environmental and radioactive contamination hazards to levels as low a reasonably achievable. Primary tasks of the D and D project were: to remove contaminated piping (about 400 linear ft of 36- and 30-in.-dia stainless steel pipe) and valves from the primary coolant pipe tunnels, to remove a primary coolant pump and piping, and to remove the three 8-ft-dia by 25-ft-long evaporators from the building second floor

Immobilisation of MTR waste in cement (product evaluation)

International Nuclear Information System (INIS)

Howard, C.G.; Lee, D.J.

1988-01-01

The enriched uranium/aluminium fuel used in Material Testing Reactors is reprocessed at Dounreay Nuclear Power Development Establishment (DNE). The main chemical component of the liquid waste produced by this process is acid deficient aluminium nitrate. The primary objective of this project is to find a suitable process for changing the highly mobile radioactive waste into an inert stable solid. Work carried out on the development of the immobilisation process showed that a conditioning stage (neutralisation) is required to make the acid waste compatible with cement. Small scale experiments showed that adding Ordinary Portland Cement blended with ground granulated Blast Furnace Slag to Simulant MTR Liquor produces an acceptable product. The process has been demonstrated at full scale (200 litres) and the products have been subjected to an extensive programme of destructive and non-destructive testing. (author)

Analysis of a total flow blockage of a Fuel Assembly in a typical MTR Research Reactor by RELAP5/MOD3.3

International Nuclear Information System (INIS)

Adorni, M.; Salah, A.B.; Di Maro, B.; Pierro, F.; D'Auria, F.; Hamidouche, T.

2004-01-01

The lack of full understanding of complex mechanisms connected with the interaction between thermal-hydraulics and neutronics still challenge the design and the operation of nuclear reactors by the adoption of conservative safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience imposes the revisiting of those areas and eventually the identification of design/safety requirements that can be relaxed [1]. Currently, the enlarged commercial exploitation of nuclear Research Reactors (RR) has increased the consideration to their corresponding safety issues. Almost all of the safety analyses have so far been performed using conservative computational tools [2]. Nowadays, the application of Best-Estimate (BE) methods constitutes a real necessity in order to increase their commercial productivity. In this framework, an attempt is made to apply the BE technique to perform a safety evaluation under research reactors operational conditions. In fact, this technique has been largely verified and validated for power reactors using coupled system thermal-hydraulic and three-dimensional neutron kinetics [1]. For this purpose, as typical representative of research reactors, the IAEA 10 MW MTR Research Reactors problem [3] is considered. The system thermal-hydraulic RELAP5 [4] code was developed to simulate transient scenarios in Power reactors such PWR, BWR, VVER, etc. However, only limited work was performed to access the applicability of the code to Research Reactors operating conditions (low pressure, mass flow rates, power, etc) [5]. Previous works performed in this field are reported in [5], [6] and [7]. In this framework, total and partial blockage of a single Fuel Assembly cooling channel are investigated. As a first attempt the calculations are performed by applying the BE thermal-hydraulic system code RELAP5 alone using its point kinetic model to derive the instantaneous core

Pump/heat exchanger assembly for pool-type reactor

International Nuclear Information System (INIS)

Nathenson, R.D.; Slepian, R.M.

1987-01-01

A heat exchanger and pump assembly comprising a heat exchanger including a housing for defining an annularly shaped cavity and supporting therein a plurality of heat transfer tubes. A pump is disposed beneath the heat exchanger and is comprised of a plurality of flow couplers disposed in a circular array. Each flow coupler is comprised of a pump duct for receiving a first electrically conductive fluid, i.e. the primary liquid metal, from a pool thereof, and a generator duct for receiving a second electrically conductive fluid, i.e. the intermediate liquid metal. The primary liquid metal is introduced from the reactor pool into the top, inlet ends of the tubes, flowing downward therethrough to be discharged from the tubes' bottom ends directly into the reactor pool. The primary liquid metal is variously introduced into the pump ducts directly from the reactor pool, either from the bottom or top end of the flow coupler. The intermediate fluid introduced into the generator ducts via the inlet duct and inlet plenum and after leaving the generator ducts passes through the annular cavity of the exchanger to cool the primary liquid in the tubes. The annular magnetic field of the pump is produced by a circular array of electromagnets having hollow windings cooled by a flow of the intermediate metal. (author)

The THMIS-MTR observation of a active region filament

Science.gov (United States)

Zong, W. G.; Tang, Y. H.; Fang, C.

We present some THMIS-MTR observations of a active region filament on September 4, 2002. The full stokes parameters of the filament were obtained in Hα, CaII 8542 and FeI 6302. By use of the data with high spatial resolution(0.44" per pixel), we probed the fine structure of the filament and gave out the parameters at the barbs' endpoints, including intensity, velocity and longitudinal magnetic field. Comparing the quiescent filament which we have discussed before, we find that: 1)The velocities of the barbs' endpoints are much bigger in the active region filament, the values are more than one thousand meters per second. 2)The barbs' endpoints terminate at the low logitudinal magnetic field in the active region filament, too.

Gamma spectrum measurement in a swimming-pool-type reactor

International Nuclear Information System (INIS)

Pla, E.

1969-01-01

After recalling the various modes of interaction of gamma rays with matter, the authors describe the design of a spectrometer for gamma energies of between 0.3 and 10 MeV. This spectrometer makes use of the Compton and pair-production effects without eliminating them. The collimator, the crystals and the electronics have been studied in detail and are described in their final form. The problem of calibrating the apparatus is then considered ; numerous graphs are given. The sensitivity of the spectrometer for different energies is determined mainly for the 'Compton effect' group. Finally, in the last part of the report, are given results of an experimental measurement of the gamma spectrum of a swimming-pool type reactor with new elements. (author) [fr

Resumption of transport of KUR spent fuel from Japan to USA - Very long-term storage and public acceptance for transport

International Nuclear Information System (INIS)

Nakagome, Yoshihiro; Nishimaki, Kenzo; Kanda, Keiji

1999-01-01

The Research Reactor Institute, Kyoto University (KURRI) has more than 250 MTR-type HEU spent fuel elements. They have been stored in water pools after irradiation in the Kyoto University Research Reactor (KUR) core. The longest pool residence time is 25 years. In accordance with the Foreign Research Reactor Spent Nuclear Fuel Receipt Program of the United States, sixty KUR spent fuel elements were shipped from KURRI to the Savannah River Site of the USDOE in August, 1999. This shipment was done successfully through a public port in Osaka Prefecture, Japan. This is the first shipment in the past twenty-six years after the last shipment through the Yokohama Port. Concerning the use of a public port, we had to solve many issues for public acceptance. In this paper, we describe how we have stored the spent fuels for a long time with high integrity and how we have obtained public acceptance for the transport. (author)

21 CFR 1250.89 - Swimming pools.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Swimming pools. 1250.89 Section 1250.89 Food and... SANITATION Sanitation Facilities and Conditions on Vessels § 1250.89 Swimming pools. (a) Fill and draw swimming pools shall not be installed or used. (b) Swimming pools of the recirculation type shall be...

OSIRIS, a MTR adapted and well fitted to LEU utilization qualification and development

International Nuclear Information System (INIS)

Barnier, M.; Beylot, J.P.

1984-01-01

The MTR OSIRIS has been successfully operated for 4 years using the ''Caramel'' low enriched uranium dioxyde fuel for the whole core loading. In the first part we examine the performance and operating experience obtained up to the present time with ''Caramel''. In a second part the paper discusses the results of the calculations for a complete OSIRIS core loaded with 20 % silicide fuel and makes a comparison with UAl 93 % and ''Caramel'' 7 % fuels. (author)

Conceptual analyses of neutronic and equilibrium refueling parameters to develop a cost-effective multi-purpose pool-type research reactor using WIMSD and CITVAP codes

Energy Technology Data Exchange (ETDEWEB)

Hedayat, Afshin, E-mail: ahedayat@aeoi.org.ir

2016-12-01

Highlights: • Introducing a high-beneficent and low-cost multipurpose research reactor. • High technical documents and standard safety issues are introduced coherently. • High effective conceptual neutronic analyses and fuel management strategy. • Gaining high score design criteria and safety margins via 3-D core modeling. • Capacity and capability to produce all medical and industrial radioisotopes. - Abstract: In this paper, neutronic and equilibrium refueling parameters of a multi-purpose cost-effective research reactor have been studied and analyzed. It has been tried to provide periodic and long-term requirements of the irradiating applications coherently. The WIMSD5B and CITVAP codes are used to calculate neutronic parameters and simulate fuel management strategy. The used nuclear data, codes, and calculating methods have been severally benchmarked and verified, successfully. Fundamental concepts, design criteria, and safety issues are introduced and discussed, coherently. Design criteria are selected to gain the most economic benefits per capital costs via minimum required reactor power. Accurate, fast and simplified models have been tried for an integrated decision making and analyses using deterministic codes. Core management, power effects, fuel consumption and burn up effects, and also a complete simulation of the fuel management strategy are presented and analyzed. Results show that the supposed reactor core design can be promisingly suitable in accordance with the commercial multi-purpose irradiating applications. It also retains Operating Limits and Conditions (OLCs) due to standard safety issues, conservatively where safety parameters are calculated using best estimate tools. Such reactor core configuration and integrated refueling task can effectively enhance the Quality Assurance (QA) of the general irradiating applications of the current MTR within their power limits and corresponding OLCs.

Immobilization of radioactive waste sludge from spent fuel storage pool

International Nuclear Information System (INIS)

Pavlovic, R.; Plecas, I.

1998-01-01

In the last forty years, in FR Yugoslavia, as result of the research reactors' operation and radionuclides application in medicine, industry and agriculture, radioactive waste materials of the different categories and various levels of specific activities were generated. As a temporary solution, these radioactive waste materials are stored in the two hanger type interim storages for solid waste and some type of liquid waste packed in plastic barrels, and one of three stainless steal underground containers for other types of liquid waste. Spent fuel elements from nuclear reactors in the Vinca Institute have been temporary stored in water filled storage pool. Due to the fact that the water in the spent fuel elements storage pool have not been purified for a long time, all metallic components submerged in the water have been hardly corroded and significant amount of the sludge has been settled on the bottom of the pool. As a first step in improving spent fuel elements storage conditions and slowing down corrosion in the storage spent fuel elements pool we have decided to remove the sludge from the bottom of the pool. Although not high, but slightly radioactive, this sludge had to be treated as radioactive waste material. Some aspects of immobilisation, conditioning and storage of this sludge are presented in this paper. (author

Biodiversity of 52 chicken populations assessed by microsatellite typing of DNA pools

Directory of Open Access Journals (Sweden)

Thomson Pippa

2003-09-01

Full Text Available Abstract In a project on the biodiversity of chickens funded by the European Commission (EC, eight laboratories collaborated to assess the genetic variation within and between 52 populations from a wide range of chicken types. Twenty-two di-nucleotide microsatellite markers were used to genotype DNA pools of 50 birds from each population. The polymorphism measures for the average, the least polymorphic population (inbred C line and the most polymorphic population (Gallus gallus spadiceus were, respectively, as follows: number of alleles per locus, per population: 3.5, 1.3 and 5.2; average gene diversity across markers: 0.47, 0.05 and 0.64; and proportion of polymorphic markers: 0.91, 0.25 and 1.0. These were in good agreement with the breeding history of the populations. For instance, unselected populations were found to be more polymorphic than selected breeds such as layers. Thus DNA pools are effective in the preliminary assessment of genetic variation of populations and markers. Mean genetic distance indicates the extent to which a given population shares its genetic diversity with that of the whole tested gene pool and is a useful criterion for conservation of diversity. The distribution of population-specific (private alleles and the amount of genetic variation shared among populations supports the hypothesis that the red jungle fowl is the main progenitor of the domesticated chicken.

Heat transfer performance of multilayer insulation system under roof slab of pool-type LMFBR

International Nuclear Information System (INIS)

Kinoshita, Izumi; Naohara, Nobuyuki; Uotani, Masaki

1986-01-01

To cope with thermal expansion of stainless steel plate, about 90 insulation structures are installed under the roof-slab of pool-type LMFBR. The objective of this study is to evaluate from heat transfer experiment and visualized experiment, the effect of distance between each thermal insulation structure on heat transfer characteristics of insulation system under roof-slab. Two types of insulation structures are selected, one is open type and the other is closed type. Distance between each thermal insulation structure and hot surface temperatures are varied as a parameter. Furthermore, heat flux of the roof-slab insulation system of reactor are estimated from the results of heat transfer experiment. (author)

MTR fuel plate qualification capabilities at SCK-CEN

International Nuclear Information System (INIS)

Koonen, E.; Jacquet, P.

2002-01-01

In order to enhance the capabilities of BR2 in the field of MTR fuel plate testing, a dedicated irradiation device has been designed. In its basic version this device allows the irradiation of 3 fuel plates. The central fuel plate may be replaced by a dummy plate or a plate carrying dosimeters. A first FUTURE device has been built. A benchmark irradiation has been executed with standard BR2 fuel plates in order to qualify this device. Detailed neutronic calculations were performed and the results compared to the results of the post-irradiation examinations of the plates. These comparisons demonstrate the capability to conduct a fuel plate irradiation program under requested and well-known irradiation conditions. Further improvements are presently being designed in order to extend the ranges of heat flux and surface temperature of the fuel plates that can be handled with the FUTURE device. (author)

Monte-Carlo validation of secondary sodium activation in a pool-type LMFBR

International Nuclear Information System (INIS)

Plamiotti, G.; Rado, V.; Salvatores, M.

1980-09-01

The secondary sodium activation in a pool-type LMFBR is the main parameter to be accurately evaluated in the shield design. In the present work a complete two dimensional description of the system, including core, shielding and sodium up to Heat Exchangers, is coupled to local Heat Exchanger Monte-Carlo calculations. This refined calculation is used to deduce a simplified method to take into account the coupling of radial propagation in the Heat Exchanger and its finite cylindrical structure

Trace element analysis at the Livermore pool-type reactor using neutron activation techniques

International Nuclear Information System (INIS)

Ragaini, R.C.; Ralston, R.; Garvis, D.

1975-01-01

The capabilities of trace element analysis at the Livermore Pool-Type Reactor (LPTR) using instrumental neutron activation analysis (INAA) are discussed. A description is given of the technology and the methods employed, including sample preparation, irradiation, and analysis. Applications of the INAA technique in past and current projects are described. A computer program, GAMANAL, has been used for nuclide identification and quantification. (U.S.)

Neutronic modelling of the Harwell MTR's: some recent problems

International Nuclear Information System (INIS)

Taylor, N.P.

1984-01-01

Use of the Harwell Materials Testing Reactors for the irradiation of experimental rigs gives rise to a number of requirements for calculations of neutron fluxes. In addition photon fluxes are required for estimates of nuclear heating rates. A range of calculational methods are employed, from simple cell to whole reactor models, and the latter have been extended for preliminary design studies for the next generation of MTR to replace DIDO and PLUTO. The technique used for these various models are described in this note, with emphasis on the areas in which modelling problems are encountered. The applications divide into three distinct areas: calculations concerning rigs irradiated within the reactor core, those for rigs positioned in the D 2 O reflector surrounding the core, and design studies for a replacement reactor. (Auth.)

Qualification of high-density fuel manufacturing for research reactors at CNEA

Energy Technology Data Exchange (ETDEWEB)

Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; De La Fuente, M.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H. [CNEA, Buenos Aires (Argentina)

2001-07-01

CNEA, the National Atomic Energy Commission of Argentina, is at the present a qualified supplier of uranium oxide fuel for research reactors. A new objective in this field is to develop and qualify the manufacturing of LEU high-density fuel for this type of reactors. According with the international trend Silicide fuel and U-xMo fuel are included in our program as the most suitable options. The facilities to complete the qualification of high-density MTR fuels, like the manufacturing plant installations, the reactor, the pool side fuel examination station and the hot cells are fully operational and equipped to perform all the activities required within the program. The programs for both type of fuels include similar activities: development and set up of the fuel material manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of miniplates, fabrication and irradiation of full scale fuel elements, post-irradiation examination and feedback for manufacturing improvements. For silicide fuels most of these steps have already been completed. For U-xMo fuel the activities also include the development of alternative ways to obtain U-xMo powder, feasibility studies for large-scale manufacturing and the economical assessment. Set up of U-xMo fuel plate manufacturing is also well advanced and the fabrication of the first full scale prototype is foreseen during this year. (author)

Experience on Maintenance of Thai Research Reactor's 'Small-Section' Pool

International Nuclear Information System (INIS)

Tippayakul, Chanatip

2013-01-01

The reactor pool of TRR-1/M1 has been used since 1962 when the reactor building was constructed. Periodic maintenance of the reactor pool has been conducted by cleaning the pool surface and re-painting with epoxy coating. The TRR-1/M1 pool basically consists of two sections referred as 'large-section' and 'small-section'. The latest re-painting activity of the 'large-section' pool was performed in 2006 but the 'small-section' pool had not been re-painted for more than 10 years. Therefore, to assure that the 'small-section' pool can maintain leak-proof condition, the re-painting of the 'small-section' pool was performed in the early 2012. A project team was organized specially for this project and a detailed execution plan was developed. The project activities include removing foreign objects and highly activated materials from the pool section, cleaning, inspecting, re-painting the pool surface and testing for water leaks. Preparation of the repainting activities had begun 2 years in advance. During the time, the reactor core had been relocated to operate in the large-section pool away from the working area in order to minimize radioactivity. The challenge of this project was to handle 4 sets of highly radioactive bolts and nuts which support the weight of the 'void tank' irradiation facility. These bolts and nuts were made from stainless steel and had been in the flux region since the installation of the 'void tank' irradiation facility approximately 30 years ago. Dose rate measurement at the contacts of these bolts and nuts were found to be in the range of 10 . 20 R/hr. The strategy to minimize the dose rate of the workers to conduct the pool repainting in the area was to remove the bolts and nuts and replace with new ones before entering the area. Special tools were improvised in order to remove the bolts and nuts under water. During the execution of the project, close radiation monitoring was performed by the radiation protection team. The project was conducted

Birth order and childhood type 1 diabetes risk: a pooled analysis of 31 observational studies

DEFF Research Database (Denmark)

Cardwell, Chris R; Stene, Lars C; Joner, Geir

2011-01-01

The incidence rates of childhood onset type 1 diabetes are almost universally increasing across the globe but the aetiology of the disease remains largely unknown. We investigated whether birth order is associated with the risk of childhood diabetes by performing a pooled analysis of previous...

Loading procedures for shipment of irradiated fuel

Energy Technology Data Exchange (ETDEWEB)

Bates, E F; Feltz, D E; Sandel, P S; Schoenbucher, B [Texas A and M University (United States)

1974-07-01

The Nuclear Science Center at Texas A and M does not have proper equipment and facilities for transferring irradiated fuel from the reactor pool to the transport vehicle. To accomplish the transfer of 23 MTR type fuel elements procedures were developed using a modified fork lift and flex-lift obtained locally. The transfer was accomplished without incident and with negligible personnel exposure. (author)

Loading procedures for shipment of irradiated fuel

International Nuclear Information System (INIS)

Bates, E.F.; Feltz, D.E.; Sandel, P.S.; Schoenbucher, B.

1974-01-01

The Nuclear Science Center at Texas A and M does not have proper equipment and facilities for transferring irradiated fuel from the reactor pool to the transport vehicle. To accomplish the transfer of 23 MTR type fuel elements procedures were developed using a modified fork lift and flex-lift obtained locally. The transfer was accomplished without incident and with negligible personnel exposure. (author)

Effectiveness of trigger point dry needling for plantar heel pain: a meta-analysis of seven randomized controlled trials

Directory of Open Access Journals (Sweden)

He C

2017-08-01

Full Text Available Chunhui He,1,* Hua Ma2,* 1Internal Medicine of Traditional Chinese Medicine, 2Medical Image Center, The First Affiliated Hospital of Xinjiang Medical University, Wulumuqi, People’s Republic of China *These authors contributed equally to this work Background: Plantar heel pain can be managed with dry needling of myofascial trigger points (MTrPs; however, whether MTrP needling is effective remains controversial. Thus, we conducted this meta-analysis to evaluate the effect of MTrP needling in patients with plantar heel pain. Materials and methods: PubMed, Embase, Web of Science, SinoMed (Chinese BioMedical Literature Service System, People’s Republic of China, and CNKI (National Knowledge Infrastructure, People’s Republic of China databases were systematically reviewed for randomized controlled trials (RCTs that assessed the effects of MTrP needling. Pooled weighted mean difference (WMD with 95% CIs were calculated for change in visual analog scale (VAS score, and pooled risk ratio (RR with 95% CIs were calculated for success rate for pain and incidence of adverse events. A fixed-effects model or random-effects model was used to pool the estimates, depending on the heterogeneity among the included studies. Results: Extensive literature search yielded 1,941 articles, of which only seven RCTs met the inclusion criteria and were included in this meta-analysis. The pooled results showed that MTrP needling significantly reduced the VAS score (WMD =–15.50, 95% CI: –19.48, –11.53; P<0.001 compared with control, but it had a similar success rate for pain with control (risk ratio [RR] =1.15, 95% CI: 0.87, 1.51; P=0.320. Moreover, MTrP needling was associated with a similar incidence of adverse events with control (RR =1.89, 95% CI: 0.38, 9.39; P=0.438. Conclusion: MTrP needling effectively reduced the heel pain due to plantar fasciitis. However, considering the potential limitations in this study, more large-scale, adequately powered, good

Assessment of structural materials inside the reactor pool of the Dalat research reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Luong Ba Vien; Nguyen Minh Tuan; Trang Cao Su

2010-01-01

Originally the Dalat Nuclear Research Reactor (DNRR) was a 250-kW TRIGA MARK II reactor, started building from early 1960s and achieved the first criticality on February 26, 1963. During the 1982-1984 period, the reactor was reconstructed and upgraded to 500kW, and restarted operation on March 20, 1984. From the original TRIGA reactor, only the pool liner, beam ports, thermal columns, and graphite reflector have been remained. The structural materials of pool liner and other components of TRIGA were made of aluminum alloy 6061 and aluminum cladding fuel assemblies. Some other parts, such as reactor core, irradiation rotary rack around the core, vertical irradiation facilities, etc. were replaced by the former Soviet Union's design with structural materials of aluminum alloy CAV-1. WWR-M2 fuel assemblies of U-Al alloy 36% and 19.75% 235 U enrichment and aluminum cladding have been used. In its original version, the reactor was setting upon an all-welded aluminum frame supported by four legs attached to the bottom of the pool. After the modification made, the new core is now suspended from the top of the pool liner by means of three aluminum concentric cylindrical shells. The upper one has a diameter of 1.9m, a length of 3.5m and a thickness of 10mm. This shell prevents from any visual access to the upper part of the pool liner, but is provided with some holes to facilitate water circulation in the 4cm gap between itself and the reactor pool liner. The lower cylindrical shells act as an extracting well for water circulation. As reactor has been operated at low power of 500 kW, it was no any problem with degradation of core structural materials due to neutron irradiation and thermal heat, but there are some ageing issues with aluminum liner and other structures (for example, corrosion of tightening-up steel bolt in the fourth beam port and flood of neutron detector housing) inside the reactor pool. In this report, the authors give an overview and assessment of

1968 Listing of Swimming Pool Equipment.

Science.gov (United States)

National Sanitation Foundation, Ann Arbor, MI. Testing Lab.

An up-to-date listing of swimming pool equipment including--(1) companies authorized to display the National Sanitation Foundation seal of approval, (2) equipment listed as meeting NSF swimming pool equipment standards relating to diatomite type filters, (3) equipment listed as meeting NSF swimming pool equipment standard relating to sand type…

Preliminary estimation of bryophyte biomass and carbon pool from three contrasting different vegetation types

Czech Academy of Sciences Publication Activity Database

Singh, M.K.; Juhász, A.; Csintalan, Z.; Kaligaric, M.; Marek, Michal V.; Urban, Otmar; Tuba, Z.

2005-01-01

Roč. 33, č. 1 (2005), s. 267-270 ISSN 0133-3720 Grant - others:EU(CZ) HPRI-CT-2002-00197 Institutional research plan: CEZ:AV0Z60870520 Keywords : bryophyte * carbon pool * rain forest Subject RIV: EH - Ecology, Behaviour Impact factor: 0.320, year: 2005

Decontamination and decommissioning of the MTR-603 HB-2 cubicle

International Nuclear Information System (INIS)

Smith, D.L.

1987-01-01

The decontamination and decommissioning (D and D) of the MTR-603 HB-2 cubicle located at the Idaho National Engineering Laboratory (INEL) are described. The HP-2 cubicle became radioactively contaminated during out-of-pile circulating water loop experiments conducted in the Materials Testing Reactor in the 1950s and 1960s. The work performed to accomplish the D and D objectives of reducing the high radiation fields caused by contamination inside the cubicle, preventing future contamination spread, and making about 1400 ft 2 of floor space available for reuse are discussed. Decommissioning of the HB-2 cubicle consisted of total dismantlement of the cubicle and its contents and was performed without disrupting ongoing laboratory work being conducted in areas surrounding the HB-2 cubicle

Progress towards a new Canadian irradiation-research facility

International Nuclear Information System (INIS)

Lee, A.G.; Lidstone, R.F.

1993-01-01

As reported at the second meeting of the International Group on Research Reactors, Atomic Energy of Canada Limited (AECL) is evaluating its options for future irradiation facilities. During the past year significant progress has been made towards achieving consensus on the irradiation requirements for AECL's major research programs and interpreting those requirements in terms of desirable characteristics for experimental facilities in a research reactor. The next stage of the study involves identifying near-term and long-term options for irradiation-research facilities to meet the requirements. The near-term options include assessing the availability of the NRU reactor and the capabilities of existing research reactors. The long-term options include developing a new irradiation-research facility by adapting the technology base for the MAPLE-X10 reactor design. Because materials testing in support of CANDU power reactors dominates AECL's irradiation requirements, the new reactor concept is called the MAPLE Materials Testing Reactor (MAPLE-MTR). Parametric physics and engineering studies are in progress on alternative MAPLE-MTR configurations to assess the capabilities for the following types of test facilities: - fast-neutron sites, that accommodate materials-irradiation assemblies, - small-diameter vertical fuel test loops that accommodate multielement assemblies, - large-diameter vertical fuel test loops, each able to hold one or more CANDU fuel bundles, - horizontal test loops, each able to hold full-size CANDU fuel bundles or small-diameter multi-element assemblies, and - horizontal beam tubes

Pump/heat exchanger assembly for pool-type reactor

International Nuclear Information System (INIS)

Nathenson, R.D.; Slepian, R.M.

1989-01-01

This patent describes a heat exchanger and pump assembly for transferring thermal energy from a heated, first electrically conductive fluid to a pumped, second electrically conductive fluid and for transferring internal energy from the pumped, second electrically conductive fluid to the first electrically conductive fluid, the assembly adapted to be disposed within a pool of the first electrically conductive fluid and comprising: a heat exchanger comprising means for defining a first annularly shaped cavity for receiving a flow of the second electrically conductive fluid and a plurality of tubes disposed within the cavity, whereby the second electrically conductive fluid in the cavity is heated, each of the tubes having an input and an output end. The input ends being disposed at the top of the heat exchanger for receiving from the pool a flow of the first electrically conductive fluid therein. The output ends being disposed at the bottom of and free of the cavity defining means for discharging the first electrically conductive fluid directly into the pool; a pump disposed beneath the heat exchanger and comprised of a plurality of flow couplers disposed in a circular array, each flow coupler comprised of a pump duct for receiving the first electrically conductive fluid and a generator duct for receiving the second electrically conductive fluid

The Jules Horowitz Reactor (JHR), a European Material Testing Reactor (MTR), with extended experimental capabilities

International Nuclear Information System (INIS)

Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P.

2003-01-01

The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation. To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.10 14 ncm -2 s -1 and a fast flux of 6,4.10 14 ncm -2 s -1 , it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = deplacement per atom). The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (author)

Feynman-α technique for measurement of detector dead time using a 30 kW tank-in-pool research reactor

International Nuclear Information System (INIS)

Akaho, E.H.K.; Intsiful, J.D.K.; Maakuu, B.T.; Anim-Sampong, S.; Nyarko, B.J.B.

2002-01-01

Reactor noise analysis was carried out for Ghana Research Reactor-1 GHARR-1, a tank-in-pool type reactor using the Feynman-α technique (variance-to-mean method). Measurements made at different detector positions and under subcritical conditions showed that the technique could not be used to determine the prompt decay constant for the reactor which is Be reflected with photo-neutron background. However, for very low dwell times the technique was used to measure the dead time of the detector which compares favourably with the value obtained using the α-conventional method

Feynman-alpha technique for measurement of detector dead time using a 30 kW tank-in-pool research reactor

CERN Document Server

Akaho, E H K; Intsiful, J D K; Maakuu, B T; Nyarko, B J B

2002-01-01

Reactor noise analysis was carried out for Ghana Research Reactor-1 GHARR-1, a tank-in-pool type reactor using the Feynman-alpha technique (variance-to-mean method). Measurements made at different detector positions and under subcritical conditions showed that the technique could not be used to determine the prompt decay constant for the reactor which is Be reflected with photo-neutron background. However, for very low dwell times the technique was used to measure the dead time of the detector which compares favourably with the value obtained using the alpha-conventional method.

Integrity assessment of research reactor fuel cladding and material testing using eddy current inspection

International Nuclear Information System (INIS)

Alencar, Donizete Anderson de

2004-01-01

A methodology to perform the integrity assessment of research reactors nuclear fuels cladding, such as those installed in IPR-Rl (TRIGA) and IEA-R1 (MTR), using nondestructive electromagnetic inspection (eddy current) is presented. This methodology is constituted by: the development of calibration reference standards, specific for each type of fuel; the development of special test probes; the recommendations for the inspection equipment calibration; the construction of voltage based evaluation curves and the inspection procedures developed for the characterization of detected flaws. The test probes development, specially those designed for the inspection of MTR fuels cladding, which present access difficulties due to the narrow gap between fuel plates (2,89 mm for IEAR-R1), constituted a challenge that demanded the introduction of unusual materials and constructive techniques. The operational performance of the developed resources, as well as the special operative characteristics of the test probes, such as their immunity to adjacent fuel plates interference and electrical resistivity changes of the fuels meat are experimentally demonstrated. The practical applicability of the developed methodology is verified in non radioactive environment, using a dummy MTR fuel element model, similar to an IEA-R1 reactor fuel element, produced and installed in IPEN, Sao Paulo. The efficacy of the proposed methodology was verified by the achieved results. (author)

Design of hydrotherapy exercise pools.

Science.gov (United States)

Edlich, R F; Abidin, M R; Becker, D G; Pavlovich, L J; Dang, M T

1988-01-01

Several hydrotherapy pools have been designed specifically for a variety of aquatic exercise. Aqua-Ark positions the exerciser in the center of the pool for deep-water exercise. Aqua-Trex is a shallow underwater treadmill system for water walking or jogging. Swim-Ex generates an adjustable laminar flow that permits swimming without turning. Musculoskeletal conditioning can be accomplished in the above-ground Arjo shallow-water exercise pool. A hydrotherapy pool also can be custom designed for musculoskeletal conditioning in its shallow part and cardiovascular conditioning in a deeper portion of the pool. Regardless of the type of exercise, there is general agreement that the specific exercise conducted in water requires significantly more energy expenditure than when the same exercise is performed on land.

Decontamination and decommissioning of the MTR-603 HB-2 cubicle. Final report

International Nuclear Information System (INIS)

Smith, D.L.

1985-12-01

This report describes the decontamination and decommissioning (D and D) of the MTR-603 HB-2 cubicle located at the Idaho National Engineering Laboratory (INEL). The HB-2 cubicle became radioactively contaminated during out-of-pile circulating water loop experiments conducted in the Materials Testing Reactor in the 1950s and 1960s. This report describes work performed to accomplish the D and D objectives of reducing the high radiation fields caused by contamination inside the cubicle, preventing future contamination spread, and making about 1400 ft 2 of floor space available for reuse. D and D of the HB-2 cubicle consisted of total dismantlement of the cubicle and its contents

Technical assessment study on pool-type LMFBR

International Nuclear Information System (INIS)

1986-01-01

Technical assessment study on pool-type LMFBR was started in 1984 FY, inheriting the products from the Feasibility study, in order to accomplish cost reduction of reactor structure and enhanced structural reliability. This study consists of four major subjects; aseismic design development, component design optimization, high temperature structural design optimization and thermal hydraulics design optimization. In 1985 FY numbers of large model tests and analytical evaluations have been performed based on the prospects obtained in the first year's study. These tests and analyses have produced a lot of findings in each subject. They are concerning; (1) the effect of various building structures and analysis methods on floor response reduction, and data for evaluation of aseismic design concepts and structural integrity to seismic loading in the aseismic design development study. (2) data for evaluation of size reduction of main components in the reactor vessel, and heat transfer data required for structural integrity evaluation in the component design optimization study. (3) data for verification of inelastic analysis method, and assurance of technical applicability of disimilar weld in the high temperature structural design optimization study. (4) the effect of component size and location on thermal hydraulic characteristics, and data of thermal hydraulic similarity in thermal hydraulic design optimization study. This report summarizes the results obtained in 1985 FY. (author)

In-tank examination and experience with MTR fuel integrity at the Imperial College reactor

Energy Technology Data Exchange (ETDEWEB)

Franklin, S.; Chapman, N.; Robertson, B.; Shields, A.; Velez-Moss, S. [Imperial College of Science Technology and Medicine, Silwood Park, Ascot (United Kingdom); Boeck, H.; Schachner, H.; Klapfer, E. [Atominstitut of the Austrian Universities, Vienna (Austria)

2000-07-01

Many changes have occurred in the UK nuclear industry over the past 10 years: nuclear power/radiation research groups have closed, the fast reactor program ceased, and the United Kingdom Atomic Energy Authority (UKAEA) changed emphasis to decommissioning. Many UK research reactors and associated facilities have closed. In 1997, the 100 kW CONSORT pool-type reactor became the last civil nuclear research reactor surviving in the UK. Although VIPER, NEPTUNE and VULCAN remain in the defense field, they have lower steady state neutron fluxes. With so many reactors closing, CONSORT has a strong future. In fact, it underpins many research projects, monitoring schemes and power plants - but each provides a relatively small amount of business. The future strategy of the reactor is being reviewed this year. First criticality took place April 1965, and so in parallel, it is important to understand what the residual technical life of the reactor might be. This paper presents the results of an in-service inspection, which took place in August 1999. (author)

In-tank examination and experience with MTR fuel integrity at the Imperial College reactor

International Nuclear Information System (INIS)

Franklin, S.; Chapman, N.; Robertson, B.; Shields, A.; Velez-Moss, S.; Boeck, H.; Schachner, H.; Klapfer, E.

2000-01-01

Many changes have occurred in the UK nuclear industry over the past 10 years: nuclear power/radiation research groups have closed, the fast reactor program ceased, and the United Kingdom Atomic Energy Authority (UKAEA) changed emphasis to decommissioning. Many UK research reactors and associated facilities have closed. In 1997, the 100 kW CONSORT pool-type reactor became the last civil nuclear research reactor surviving in the UK. Although VIPER, NEPTUNE and VULCAN remain in the defense field, they have lower steady state neutron fluxes. With so many reactors closing, CONSORT has a strong future. In fact, it underpins many research projects, monitoring schemes and power plants - but each provides a relatively small amount of business. The future strategy of the reactor is being reviewed this year. First criticality took place April 1965, and so in parallel, it is important to understand what the residual technical life of the reactor might be. This paper presents the results of an in-service inspection, which took place in August 1999. (author)

The rehabilitation/upgrading of Philippine Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Renato, T Banaga [Philippines Nuclear Research Inst., Quezon (Philippines)

1998-10-01

The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E{sub 1}-U-Z{sub 1}-H{sub 1.6} TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

The rehabilitation/upgrading of Philippine Research Reactor

International Nuclear Information System (INIS)

Renato T, Banaga

1998-01-01

The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E 1 -U-Z 1 -H 1.6 TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

High enrichment to low enrichment core's conversion. Accidents analysis

International Nuclear Information System (INIS)

Abbate, P.; Rubio, R.; Doval, A.; Lovotti, O.

1990-01-01

This work analyzes the different accidents that may occur in the reactor's facility after the 20% high-enriched uranium core's conversion. The reactor (of 5 thermal Mw), built in the 50's and 60's, is of the 'swimming pool' type, with light water and fuel elements of the curve plates MTR type, enriched at 93.15 %. This analysis includes: a) accidents by reactivity insertion; b) accidents by coolant loss; c) analysis by flow loss and d) fission products release. (Author) [es

Estimating flow characteristics of different weir types and optimum dimensions of downstream receiving pool

OpenAIRE

Emiroglu, M. Emin

2010-01-01

This paper presents the results of a laboratory study on the flow characteristics of sharp-crested weirs, broad-crested weirs, and labyrinth weirs. The variation of the maximum bubble penetration depth for different weir types is investigated depending on overfall jet expansion, discharge, and drop height. Moreover, most efficient depth, length and width of the downstream receiving pool in an open channel system are studied by considering the penetration depth, overfall jet expansion, jet tra...

Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility; Estudo termohidraulico de um elemento combustivel tipo MTR visando a construcao de um dispositivo de irradiacao

Energy Technology Data Exchange (ETDEWEB)

Coragem, Helio Boemer de Oliveira

1980-07-01

A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility; Estudo termohidraulico de um elemento combustivel tipo MTR visando a construcao de um dispositivo de irradiacao

Energy Technology Data Exchange (ETDEWEB)

Coragem, Helio Boemer de Oliveira

1980-07-01

A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

Gene pool conservation of teak in Myanmar

International Nuclear Information System (INIS)

Tin-Tun

1995-01-01

Myanmar with an area of 261, 228 Sq. miles is endowed with various types of forests which occupied nearly 50% of the country. Teak (Tectona grandis Linn. f.) is one of the most valuable timber species for its excellent wood quality and properties which are not observed with other timbers. Gene pool can be defined as a group of individual trees growing over a wide range of environmental conditions, and constituting different genetic complexes which can be transmitted to the offsprings. Topics such as: objectives of gene pool conservation, genetically improved seeds for large scale forest plantations, methodology of conservation, are discussed in the article. Myanmar teak dominates the world's teak market, and thus it is crucial to maintain the superiority in the conservation of gene complexes of teak. To some extent, the conservation of gene pools of teak and tree improvements are being undertaken by the Forest Research Institute of Myanmar. It is felt that the dissemination of the philosophy and concept of gene conservation to the personal involved in the forestry activities of the country are still inadequate

The thermal performance of a loop-type heat pipe for passively removing residual heat from spent fuel pool

Energy Technology Data Exchange (ETDEWEB)

Xiong, Zhenqin [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Gu, Hanyang, E-mail: guhanyang@stu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Wang, Minglu [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Cheng, Ye [Shanghai Nuclear Engineering Research and Design Institute, Shanghai 200233 (China)

2014-12-15

Highlights: • Feasibility of applying loop-type heat pipes for SFP is studied. • The heat transfer rate of the heat pipes was tested. • The heat transfer coefficient was between 200 and 490 W/m{sup 2}/s. • The effect of the water temperature is dominant. • Three kinds of the filling ratio 27%, 21% and 14% are compared. - Abstract: Heat pipe is an efficient heat transfer device without electrically driven parts. Therefore large-scale loop type heat pipe systems have potential uses for passively removing heat from spent fuel pools and reactor cores under the accidental conditions to improve the safety of the nuclear power station. However, temperature difference between the hot water in the spent fuel pool and the ambient air which is the heat sink is small, in the range of 20–60 °C. To understand and predict the heat removal capacity of such a large scale loop type heat pipe in the situation similar to the accidental condition of the spent fuel pool (SFP) for the design purpose, a loop-type heat pipe with a very high and large evaporator has been fabricated and was tested using ammonia as the working fluid. The evaporator with inner diameter of 65 mm and length of 7.6 m is immersed in a hot water tube which simulate the spent fuel pool. The condenser of the loop-type heat pipe is cooled by the air. The tests were performed with the velocity of the hot water in the tube in the range of 0.7–2.1 × 10{sup −2} m/s, the hot water inlet temperature between 50 and 90 °C and the air velocity ranging from 0.5 m/s to 2.5 m/s. Three kinds of the ammonia volumetric filling ratio in the heat pipe were tested, i.e. 27%, 21% and 14%. It is found that the heat transfer rate was in the range of 1.5–14.9 kW, and the heat transfer coefficient of evaporator was between 200 and 490 W/m{sup 2}/s. It is feasible to use the large scale loop type heat pipe to passively remove the residual heat from SFP. Furthermore, the effect of air velocity, air temperature, water flow

The thermal performance of a loop-type heat pipe for passively removing residual heat from spent fuel pool

International Nuclear Information System (INIS)

Xiong, Zhenqin; Gu, Hanyang; Wang, Minglu; Cheng, Ye

2014-01-01

Highlights: • Feasibility of applying loop-type heat pipes for SFP is studied. • The heat transfer rate of the heat pipes was tested. • The heat transfer coefficient was between 200 and 490 W/m 2 /s. • The effect of the water temperature is dominant. • Three kinds of the filling ratio 27%, 21% and 14% are compared. - Abstract: Heat pipe is an efficient heat transfer device without electrically driven parts. Therefore large-scale loop type heat pipe systems have potential uses for passively removing heat from spent fuel pools and reactor cores under the accidental conditions to improve the safety of the nuclear power station. However, temperature difference between the hot water in the spent fuel pool and the ambient air which is the heat sink is small, in the range of 20–60 °C. To understand and predict the heat removal capacity of such a large scale loop type heat pipe in the situation similar to the accidental condition of the spent fuel pool (SFP) for the design purpose, a loop-type heat pipe with a very high and large evaporator has been fabricated and was tested using ammonia as the working fluid. The evaporator with inner diameter of 65 mm and length of 7.6 m is immersed in a hot water tube which simulate the spent fuel pool. The condenser of the loop-type heat pipe is cooled by the air. The tests were performed with the velocity of the hot water in the tube in the range of 0.7–2.1 × 10 −2 m/s, the hot water inlet temperature between 50 and 90 °C and the air velocity ranging from 0.5 m/s to 2.5 m/s. Three kinds of the ammonia volumetric filling ratio in the heat pipe were tested, i.e. 27%, 21% and 14%. It is found that the heat transfer rate was in the range of 1.5–14.9 kW, and the heat transfer coefficient of evaporator was between 200 and 490 W/m 2 /s. It is feasible to use the large scale loop type heat pipe to passively remove the residual heat from SFP. Furthermore, the effect of air velocity, air temperature, water flow rate and

System for uranium superficial density measurement in U3Si2 MTR fuel plates using radiography

International Nuclear Information System (INIS)

Hey, Martin A.; Gomez Marlasca, Fernando

2003-01-01

The paper describes a method for measuring uranium superficial density in high density uranium silicide (U 3 Si 2 ) MTR fuel plates, through the use of industrial radiography, a set of patterns built for this purpose, a transmission optical densitometer, and a quantitative model of analysis and measurement. Our choice for this particular method responds to its high accuracy, low cost and easy implementation according to the standing quality control systems. (author)

Study on the seismic response of reactor vessel of pool type LMFBR including fluid-structure interaction

International Nuclear Information System (INIS)

Tanimoto, K.; Ito, T.; Fujita, K.; Kurihara, C.; Sawada, Y.; Sakurai, A.

1988-01-01

The paper presents the seismic response of reactor vessel of pool type LMFBR with fluid-structure interaction. The reactor vessel has bottom support arrangement, the same core support system as Super-Phenix in France. Due to the bottom support arrangement, the level of core support is lower than that of the side support arrangement. So, in this reactor vessel, the displacement of the core top tends to increase because of the core's rocking. In this study, we investigated the vibration and seismic response characteristics of the reactor vessel. Therefore, the seismic experiments were carried out using one-eighth scale model and the seismic response including FSI and sloshing were investigated. From this study, the effect of liquid on the vibration characteristics and the seismic response characteristics of reactor vessel were clarified and sloshing characteristics were also clarified. It was confirmed that FEM analysis with FSI can reproduce the seismic behavior of the reactor vessel and is applicable to seismic design of the pool type LMFBR with bottom support arrangement. (author). 5 refs, 14 figs, 2 tabs

High temperature ultrasonic sensor for fission gas characterization in MTR harsh environment

Directory of Open Access Journals (Sweden)

Gatsa O.

2018-01-01

In this paper, we present NBT thick film fabrication by screen printing, characterization of piezoelectric, dielectric properties and material parameters studies in dependence of temperature. Relatively high resistivity in the range of 1.1013 Ohm.cm for fabricated thick film is explained by Aurivillius structure in which a-and b-layers form perovskite structure between oxides of c-layer. Main results of this study are presented and discussed in terms of feasibility for an application to a new sensor device operating at high temperature level (400°. Piezoelectric parameters enhancement and loss reduction at elevated temperatures are envisaged to be optimized. Further sensor development and test in MTR are expected to be realized in the near future.

Design of make-up water system for Tehran research reactor spent nuclear fuels storage pool

Energy Technology Data Exchange (ETDEWEB)

Aghoyeh, Reza Gholizadeh [Reactor Research Group, Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran (AEOI), North Amirabad, P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of); Khalafi, Hosein, E-mail: hkhalafi@aeoi.org.i [Reactor Research Group, Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran (AEOI), North Amirabad, P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of)

2010-10-15

Spent nuclear fuels storage (SNFS) is an essential auxiliary system in nuclear facility. Following discharge from a nuclear reactor, spent nuclear fuels have to be stored in water pool of SNFS away from reactor to allow for radioactive to decay and removal of generated heat. To prevent corrosion damage of fuels and other equipments, the storage pool is filled with de-ionized water which serves as moderator, coolant and shielding. The de-ionized water will be provided from make-up water system. In this paper, design of a make-up water system for optimal water supply and its chemical properties in SNFS pool is presented. The main concern of design is to provide proper make-up water throughout the storage time. For design of make-up water system, characteristics of activated carbon purifier, anionic, cationic and mixed-bed ion-exchangers have been determined. Inlet water to make-up system provide from Tehran municipal water system. Regulatory Guide 1.13 of the and graver company manual that manufactured the Tehran research reactor (TRR) make-up water system have been used for make-up water system of TRR spent nuclear fuels storage pool design.

Design of make-up water system for Tehran research reactor spent nuclear fuels storage pool

International Nuclear Information System (INIS)

Aghoyeh, Reza Gholizadeh; Khalafi, Hosein

2010-01-01

Spent nuclear fuels storage (SNFS) is an essential auxiliary system in nuclear facility. Following discharge from a nuclear reactor, spent nuclear fuels have to be stored in water pool of SNFS away from reactor to allow for radioactive to decay and removal of generated heat. To prevent corrosion damage of fuels and other equipments, the storage pool is filled with de-ionized water which serves as moderator, coolant and shielding. The de-ionized water will be provided from make-up water system. In this paper, design of a make-up water system for optimal water supply and its chemical properties in SNFS pool is presented. The main concern of design is to provide proper make-up water throughout the storage time. For design of make-up water system, characteristics of activated carbon purifier, anionic, cationic and mixed-bed ion-exchangers have been determined. Inlet water to make-up system provide from Tehran municipal water system. Regulatory Guide 1.13 of the and graver company manual that manufactured the Tehran research reactor (TRR) make-up water system have been used for make-up water system of TRR spent nuclear fuels storage pool design.

Pakistan upgrades PARR-1 and converts to LEU

International Nuclear Information System (INIS)

Anon.

1990-01-01

The Pakistan Research Reactor, PARR-1, is a 5MW swimming pool type reactor originally designed to use MTR type fuel elements fabricated from uranium enriched to more than 90%. After about 24 years of satisfactory operation it is now planned to convert the reactor to use low enriched (20%) uranium fuel. The opportunity will also be taken to upgrade the reactor power to about 9MW. This power upgrading will meet the demand for higher neutron fluxes for experimental and radioisotope production as well as compensating for the neutron flux penalty arising from conversion from high enriched to low enriched fuel. During the process of conversion and upgrading it is also proposed to renovate existing services and associated systems and to add certain new safety related engineering. (author)

Stability of inner baffle-shell of pool type LMFBR - experimental and theoretical studies

International Nuclear Information System (INIS)

Lebey, J.; Combescure, A.

1987-01-01

I pool type LMFBR, the primary coolant circuit, inside the main vessel, comprises a hot plenum separated from a cold plenum by an inner baffle. For Superphenix 1 reactor, it was judged advisable to built a double-shell baffle, each shell withstanding only one type of loading (primary loading for one shell, secondary loading for the other). Due to the size and intricacy of the structure, this design involves unnegligible supplementary costs and manufacturing difficulties. Thus, an alternative solution has been studied for future plants projects. It consists of a single shell baffle having a shape especially studied to sustain the two types of applied loadings (thermal plus primary loadings). Such a shape was calculated by NOVATOME, and it was decided to check the ability of methods of analysis to predict the ruin of this structure under primary loading. For this purpose, a mock-up has been tested, and the experimental results compared with the calculated ones. (orig./GL)

Coupled 3D neutronics/thermal hydraulics modeling of the SAFARI-1 MTR

International Nuclear Information System (INIS)

Rosenkrantz, Adam; Avramova, Maria; Ivanov, Kostadin; Prinsloo, Rian; Botes, Danniëll; Elsakhawy, Khalid

2014-01-01

Highlights: • Development of 3D coupled neutronics/thermal–hydraulic model of SAFARI-1. • Verification of 3D steady-state NEM based neutronics model for SAFARI-1. • Verification of 3D COBRA-TF based thermal–hydraulic model of SAFARI-1. • Quantification of the effect of correct modeling of thermal–hydraulic feedback. - Abstract: The purpose of this study was to develop a coupled accurate multi-physics model of the SAFARI-1 Material Testing Reactor (MTR), a facility that is used for both research and the production of medical isotopes. The model was developed as part of the SAFARI-1 benchmarking project as a cooperative effort between the Pennsylvania State University (PSU) and the South African Nuclear Energy Corporation (Necsa). It was created using a multi-physics coupling of state of the art nuclear reactor simulation tools, consisting of a neutronics code and a thermal hydraulics code. The neutronics tool used was the PSU code NEM, and the results from this component were verified using the Necsa neutronics code OSCAR-4, which is utilized for SAFARI-1 core design and fuel management. On average, the multiplication factors of the neutronics models agreed to within 5 pcm and the radial assembly-averaged powers agreed to within 0.2%. The thermal hydraulics tool used was the PSU version of COBRA-TF (CTF) sub-channel code, and the results of this component were verified against another thermal hydraulics code, the RELAP5-3D system code, used at Necsa for thermal–hydraulics analysis of SAFARI-1. Although only assembly-averaged results from RELAP5-3D were available, they fell within the range of values for the corresponding assemblies in the comprehensive CTF solution. This comparison allows for the first time to perform a quantification of steady-state errors for a low-powered MTR with an advanced thermal–hydraulic code such as CTF on a per-channel basis as compared to simpler and coarser-mesh RELAP5-3D modeling. Additionally, a new cross section

Pool scrubbing

International Nuclear Information System (INIS)

Lopez-Jimenez, J.; Herranz, J.; Escudero, M.J.; Espigares, M.M.; Peyres, V.; Polo, J.; Kortz, Ch.; Koch, M.K.; Brockmeier, U.; Unger, H.; Dutton, L.M.C.; Smedley, Ch.; Trow, W.; Jones, A.V.; Bonanni, E.; Calvo, M.; Alonso, A.

1996-12-01

The Source Term Project in the Third Frame Work Programme of the European Union Was conducted under and important joined effort on pool scrubbing research. CIEMAT was the Task Manager of the project and several other organizations participated in it: JRC-Ispra, NNC Limited, RUB-NES and UPM. The project was divided into several tasks. A peer review of the models in the pool scrubbing codes SPARC90 and BUSCA-AUG92 was made, considering the different aspects in the hydrodynamic phenomenology, particle retention and fission product vapor abortions. Several dominant risk accident sequences were analyzed with MAAP, SPARC90 and BUSCA-AUG92 codes, and the predictions were compared. A churn-turbulent model was developed for the hydrodynamic behaviour of the pool. Finally, an experimental programme in the PECA facility of CIEMAT was conducted in order to study the decontamination factor under jet injection regime, and the experimental observations were compared with the SPARC and BUSCA codes. (Author)

Rules for the licensing of new experiments in BR2: application to the test irradiation of new MTR-fuels

International Nuclear Information System (INIS)

Joppen, F.

2000-01-01

New types of MTR fuel elements are being developed and require a qualification before routine operation could be authorized. During the test irradiation the new fuel elements .are considered as experimental devices and their irradiation is allowed according to the procedures for experiments. Authorization is based on the advice .of a consultative committee on experiments. This procedure is valid as long as the irradiation is covered by the actual reactor license. An additional license or an amendment is only required if due to the experiment the risk for the workers or the environment is increased in a significant way. A few experimental fuel plates loaded in the primary loop of the reactor will not increase this risk. The source term for potential radioactive releases remains more or less the same. The probability for an accident can be limited by restricting the heat flux and surface temperature. (author)

Validation concerns for dry storage of foreign research reactor spent nuclear fuel

International Nuclear Information System (INIS)

Trumble, E.F.

1994-01-01

Recent decisions by the Department of Energy have accelerated the need for storage options to support the return of foreign research reactor (FRR) fuel to the United States. Many of these returns consist of fuel types which contain highly enriched uranium and are aluminum clad. These attributes present many challenges not experienced in the fuel storage designs for commercial nuclear fuels where the fuels have lower enrichment and the cladding is more robust. Historically, returned FRR fuel has been stored for short periods in basins where it is cooled and then sent to be reprocessed. However, a severe lack of basin space and questionable availability of reprocessing facilities necessitates the development of other proposals. One proposed option is to store the FRR fuel in a dry state, thus reducing the corrosion problems associated with aluminum cladding. A drawback to this type of storage, however, is the lack of experimental data for this type of fuel under dry storage conditions. This lack of data has led to recent discussions over the accuracy of some of the current multigroup cross section libraries when applied to dry, fast systems of uranium and aluminum. This concern is evaluated for the specific case of Material Test Reactor (MTR) fuel (MTR is >60% of FRR fuel), a review of applicable experiments is presented and a new experiment is proposed

Final qualification of an industrial wide range neutron instrumentation in the Osiris MTR reactor

Energy Technology Data Exchange (ETDEWEB)

Barbot, L.; Normand, S. [CEA, LIST, Laboratoire Capteur et Architectures Electroniques, F-91191 Gif Sur Yvette (France); Pasdeloup, P. [AREVA TA, Controle Commande and Mesures, F-13762 Les Milles (France); Lescop, B. [CEA, INSTN, UEIN, F-91191 Gif Sur Yvette (France)

2009-07-01

This work deals with the final qualification of the IRINA in-core neutron flux measurement system in the MTR Osiris reactor. A specific irradiation device has been set up to validate the last changes in the complete system (electronic, transmitting cable and monitor). Experimental results show the IRINA measurement system meet entirely the in-core reactor conditions requirements: a thermal neutron flux from 10{sup 7} n.cm{sup -2}.s{sup -1} up to 10{sup 14} n.cm{sup -2}.s{sup -1} and a temperature of 300 C degrees during a minimum operating time of 1000 hours. (authors)

Some safety considerations in the selection of redans for pool-type LMR plants

International Nuclear Information System (INIS)

Pan, Y.C.; Pedersen, D.R.; Wang, C.Y.

1985-01-01

Three basic safety issues in the selection of the redan design for a pool type liquid metal fast breeder reactor plant are examined. The first area examined is the effect of the redan selection on the integrity of the primary system pressure boundary in normal and offset conditions. The second area is on the consequence of the hypothetical core disruptive accident. The third area is on the consequence of the loss of heat sink accident. Some general discussion and numerical results are presented which may help in the selection of an optimum redan design. 3 refs., 7 figs

Pressure drop calculation in a fuel element of a pool type reactor

International Nuclear Information System (INIS)

Lassance, Victor; Oliveira, Andre F.; Moreira, Maria de L.

2013-01-01

Even with the advances of hardware in computer sciences, sometimes it is necessary to simplify the simulation in order to optimize the results given the same calculation runtime. The object of this study is a thermodynamic analysis of the core of a pool type research reactor, focusing on natural circulation. Due to the high geometrical complexity of the core, the scale transfer process becomes an essential step to the thermodynamic study of the reactor. This process takes place by determining the effective equivalent properties obtained from a detailed simulation of the core and transferring them to a porous medium having a coarse mesh while preserving the overall characteristics. In this way, it will be able to obtain the quadratic resistance coefficient KQ by calculating the pressure drop inside the fuel element. To observe in detail the behavior of this flow, longitudinal and transversal cross sections will be made in different points, thereby observing the velocity and pressure distributions. The analysis will provide detailed data on the fluid flow between the fuel plates enabling the observation of possible critical points or undesired behavior. The whole analysis was made by using the commercial code ANSYS CFX ver. 12.1. This is study will provide data, as a first step to enable future simulations which will consider the entire reactor. (author)

Sustainability of common pool resources

OpenAIRE

Timilsina, Raja Rajendra; Kotani, Koji; Kamijo, Yoshio

2017-01-01

Sustainability has become a key issue in managing natural resources together with growing concerns for capitalism, environmental and resource problems. We hypothesize that the ongoing modernization of competitive societies, which we refer to as "capitalism," affects human nature for utilizing common pool resources, thus compromising sustainability. To test this hypothesis, we design and implement a set of dynamic common pool resource games and experiments in the following two types of Nepales...

Justify of implementation of a hot water layer system in swimming pool research reactor IEA-R1m

International Nuclear Information System (INIS)

Toyoda, Eduardo Yoshio; Gordon, Ana Maria Pinho Leite; Sordi, Gian-Maria A.A.

2001-01-01

The IPEN/CNEN-SP has a swimming pool research reactor (IEA-R1m) in operation since 1957 at 2 MW. In 1998, after some modifications, its nominal power increased to 5 MW. Among these modifications some adaptations had to be accomplished in the radiological protection and operational procedure. The present work aim to study the need of implementation of a hot water layer in order to reduce the dose in the workers in the vicinity of the reactor swimming pool. Applying the principles of radioprotection optimization, it was concluded that the decision of the construction of one hot water layer system in the reactor swimming pool, is not necessary. (author)

Tools evaluation and development for loss of coolant accidents analysis in research reactors

International Nuclear Information System (INIS)

Maprelian, Eduardo; Cabral, Eduardo L.L.; Silva, Antonio T. e

1999-01-01

The loss of coolant accidents (LOCA) in pool type research reactors are normally considered as limiting in the licensing process. This paper verifies the viability of the computer code 3D-AIRLOCA to analyze LOCA in a pool type research reactor, and also develops two computer codes LOSS and TEMPLOCA. The computer code LOSS determines the time tom drawn the pool down to the level of the bottom of the core, and the computer code TEMPLOCA calculates the peak fuel element temperature during the transient. These two coders substitutes the 3D-AIRLOCA in the LOCA analysis for pool type research reactors. (author)

Feasibility analysis of the Primary Loop of Pool-Type Natural Circulating Nuclear Reactor Dedicated to Seawater Desalination

Energy Technology Data Exchange (ETDEWEB)

Jeong, Woonho; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

2016-05-15

In this study, the feasibility of natural circulation was evaluated for the reference plant AHR400 (Advanced Heating Reactor 400MWth). AHR400 is a pool-type desalination-dedicated nuclear reactor. As a consequence, AHR400 has low operating pressure and temperature which provides large safety margin. Removal of the reactor coolant pump from the AHR400 will enforce integrity of the reactor vessel and passive safety feature. Therefore, the study also tried to find out optimized primary loop design to achieve total natural circulation of the coolant. Natural circulation capacity of the primary loop of the desalination dedicated nuclear reactor AHR400 was evaluated. It was concluded that to remove RCP from the AHR400 and operates the reactor only by natural circulation of the coolant is impossible. Decreased core power as half make removal of RCP possible with 15m central height difference between the core and IHXs. Furthermore, validation and modification of pressure loss coefficients by small-scaled natural circulation experiment at a pool-type reactor would provide more accurate results.

Modal analysis of pool door in water tank

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Soo; Jeong, Kyeong Hoon; Park, Chan Gook; Koo, In Soo [KAERI, Daejeon (Korea, Republic of)

2012-10-15

A pool door is installed at the chase of the pool gate by means of an overhead crane in the building of a research reactor. The principal function of the pool door, which is located between the reactor pool and service pool, is to separate the reactor pool from the service pool for the maintenance and/or the removal of the equipment either in the reactor pool or service pool. The pool door consists of stainless steel plates supported by structural steel frames and sealing components. The pool door is equipped with double inflatable gaskets. The configuration of the pool door is shown in Figure 1. The FEM analysis and theoretical calculation by the formula were performed to evaluate the natural frequency for the pool door in the water. The results from the two methods were compared.

Economical analysis to utilize MTR fuel elements using silicides in research reactors

International Nuclear Information System (INIS)

Bergallo, Juan E.; Novara, Oscar E.; Adelfang, Pablo

2000-01-01

According to international programs on reducing enrichment in research reactors and the necessity to maintain their operation, new fuel elements have been developed in order to meet both objectives. Thus, U-Si alloy fuel elements for research reactors are becoming of greater interest for the international markets. It became necessary to make an economic study about the convenience of introducing this type of fuel elements in the RA-3 reactor and to know the potentiality of this fuel. The economical behavior of the reactor operation has been evaluated comparing the actual U 3 O 8 nuclear fuel cycle with U 3 Si 2 nuclear fuels. Results obtained show that the main economical factor to determine the change of fuels is the cost of fabrication, and the change is advisable up to an 80% difference. The other factors related to the cost of nuclear fuel cycle are not relevant or have real minor impacts. (author)

MTR spent fuel back-end - Cogema's long-term commitment

International Nuclear Information System (INIS)

Thomasson, J.

1998-01-01

MTR spent fuel back end has been subject to many reversal and uncertainties in the past 10 years. Until the end of 1988, US obligated materials were subject to the Off site Fuels Policy (OFP). Under this policy, spent fuels were returned to USA, and were reprocessed there. This OFP took end the 31th of December 1988, and Research Reactor's operators had to implement others solutions: On site storage or Reprocessing in Europe. Meanwhile the RERTR Program was leading to a new LEU fuel to replace HEU aluminide. This new silicide fuel has one main drawback: it cannot be reprocessed in working plants without some process main line modifications. Fortunately, a new Research Reactors spent fuels return policy has been set up by the US in the early 1996. This new policy applies to all reactors converted or that have agreed to convert to LEU, and reactors operating with HEU for which no suitable LEU is available. It covers all the spent fuels discharged until 2006/05/12. But after that period of time, each reactor will be fully responsible for its spent fuels. Since the end of 1996, COGEMA is proposing reprocessing services for Aluminides spent fuels, based on the La Hague capability. This COGEMA answer is for the long term, as the La Hague plant has a good load for the coming years, including the first decade of the next century. Further, this activity benefits from a strong R and D support, that allowed fulfilling the evolutive needs of our customers, and gives us the ability to adapt the plant to the future market. Taking advantage of this flexibility, COGEMA offers Research Reactors' operators a long-term commitment. Already two reactors' operators have chosen to contract with COGEMA for the whole life of their reactors. The contracts execution is under progress and the first transportation will take place soon. Beside today's services, COGEMA is involved in R and D activities to support new fuels development enhancing present LEU performances and having the ability to

Safety challenges encountered during the operating life of the almost 40 year old research reactor BR2

International Nuclear Information System (INIS)

Koonen, E.; Joppen, F.; Gubel, P.

2001-01-01

The BR2 reactor is one of the major MTR-type research reactors in the world. Its operation started in the early 1960's. Two major refurbishment operations have been carried out since then. Several safety reassessments were carried out over the years in order to keep the safety level in line with modern standards and to enhance operational safety. This paper gives an overview of the safety challenges encountered over the years and how those were met. (author)

Sparse feature learning for instrument identification: Effects of sampling and pooling methods.

Science.gov (United States)

Han, Yoonchang; Lee, Subin; Nam, Juhan; Lee, Kyogu

2016-05-01

Feature learning for music applications has recently received considerable attention from many researchers. This paper reports on the sparse feature learning algorithm for musical instrument identification, and in particular, focuses on the effects of the frame sampling techniques for dictionary learning and the pooling methods for feature aggregation. To this end, two frame sampling techniques are examined that are fixed and proportional random sampling. Furthermore, the effect of using onset frame was analyzed for both of proposed sampling methods. Regarding summarization of the feature activation, a standard deviation pooling method is used and compared with the commonly used max- and average-pooling techniques. Using more than 47 000 recordings of 24 instruments from various performers, playing styles, and dynamics, a number of tuning parameters are experimented including the analysis frame size, the dictionary size, and the type of frequency scaling as well as the different sampling and pooling methods. The results show that the combination of proportional sampling and standard deviation pooling achieve the best overall performance of 95.62% while the optimal parameter set varies among the instrument classes.

Experimental validation of thermal design of top shield for a pool type SFR

International Nuclear Information System (INIS)

Aithal, Sriramachandra; Babu, V. Rajan; Balasubramaniyan, V.; Velusamy, K.; Chellapandi, P.

2016-01-01

Highlights: • Overall thermal design of top shield in a SFR is experimentally verified. • Air jet cooling is effective in ensuring the temperatures limits for top shield. • Convection patterns in narrow annulus are in line with published CFD results. • Wire mesh insulation ensures gradual thermal gradient at top portion of main vessel. • Under loss of cooling scenario, sufficient time is available for corrective action. - Abstract: An Integrated Top Shield Test Facility towards validation of thermal design of top shield for a pool type SFR has been conceived, constructed & commissioned. Detailed experiments were performed in this experimental facility having full-scale features. Steady state temperature distribution within the facility is measured for various heater plate temperatures in addition to simulating different operating states of the reactor. Following are the important observations (i) jet cooling system is effective in regulating the roof slab bottom plate temperature and thermal gradient across roof slab simulating normal operation of reactor, (ii) wire mesh insulation provided in roof slab-main vessel annulus is effective in obtaining gradual thermal gradient along main vessel top portion and inhibiting the setting up of cellular convection within annulus and (iii) cellular convection with four distinct convective cells sets in the annular gap between roof slab and small rotatable plug measuring ∼ϕ4 m in diameter & gap width varying from 16 mm to 30 mm. Repeatability of results is also ensured during all the above tests. The results presented in this paper is expected to provide reference data for validation of thermal hydraulic models in addition to serving as design validation of jet cooling system for pool type SFR.

Reconstruction of Extracellular Respiratory Pathways for Iron(III Reduction in Shewanella oneidensis strain MR-1

Directory of Open Access Journals (Sweden)

Dan eCoursolle

2012-02-01

Full Text Available Shewanella oneidensis strain MR-1 is a facultative anaerobic bacterium capable of respiring a multitude of electron acceptors, many of which require the Mtr respiratory pathway. The core Mtr respiratory pathway includes a periplasmic c-type cytochrome (MtrA, an integral outer membrane β-barrel protein (MtrB and an outer membrane-anchored c-type cytochrome (MtrC. Together, these components facilitate transfer of electrons from the c-type cytochrome CymA in the cytoplasmic membrane to electron acceptors at and beyond the outer membrane. The genes encoding these core proteins have paralogs in the S. oneidensis genome (mtrB and mtrA each have four while mtrC has three and some of the paralogs of mtrC and mtrA are able to form functional Mtr complexes. We demonstrate that of the additional three mtrB paralogs found in the S. oneidensis genome, only MtrE can replace MtrB to form a functional respiratory pathway to soluble iron(III citrate. We also evaluate which mtrC / mtrA paralog pairs (a total of 12 combinations are able to form functional complexes with endogenous levels of mtrB paralog expression. Finally, we reconstruct all possible functional Mtr complexes and test them in a S. oneidensis mutant strain where all paralogs have been eliminated from the genome. We find that each combination tested with the exception of MtrA / MtrE / OmcA is able to reduce iron(III citrate at a level significantly above background. The results presented here have implications towards the evolution of anaerobic extracellular respiration in Shewanella and for future studies looking to increase the rates of substrate reduction for water treatment, bioremediation, or electricity production.

13 CFR 120.611 - Pools backing Pool Certificates.

Science.gov (United States)

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Pools backing Pool Certificates. 120.611 Section 120.611 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION BUSINESS LOANS Secondary Market Certificates § 120.611 Pools backing Pool Certificates. (a) Pool characteristics. As set...

Design study of an IHX support structure for a POOL-TYPE Sodium-cooled fast reactor

International Nuclear Information System (INIS)

Park, Chang Gyu; Kim, Jong Bum; Lee, Jae Han

2009-01-01

The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity

Mathematical model development of heat and mass exchange processes in the outdoor swimming pool

Directory of Open Access Journals (Sweden)

M. V. Shaptala

2014-12-01

Full Text Available Purpose. Currently exploitation of outdoor swimming pools is often not cost-effective and, despite of their relevance, such pools are closed in large quantities. At this time there is no the whole mathematical model which would allow assessing qualitatively the effect of energy-saving measures. The aim of this work is to develop a mathematical model of heat and mass exchange processes for calculating basic heat and mass losses that occur during its exploitation. Methodology. The method for determination of heat and mass loses based on the theory of similarity criteria equations is used. Findings. The main types of heat and mass losses of outdoor pool were analyzed. The most significant types were allocated and mathematically described. Namely: by evaporation of water from the surface of the pool, by natural and forced convection, by radiation to the environment, heat consumption for water heating. Originality. The mathematical model of heat and mass exchange process of the outdoor swimming pool was developed, which allows calculating the basic heat and mass loses that occur during its exploitation. Practical value. The method of determining heat and mass loses of outdoor swimming pool as a software system was developed and implemented. It is based on the mathematical model proposed by the authors. This method can be used for the conceptual design of energy-efficient structures of outdoor pools, to assess their use of energy-intensive and selecting the optimum energy-saving measures. A further step in research in this area is the experimental validation of the method of calculation of heat losses in outdoor swimming pools with its use as an example the pool of Dnipropetrovsk National University of Railway Transport named after Academician V. Lazaryan. The outdoor pool, with water heating- up from the boiler room of the university, is operated year-round.

Research progresses and future directions on pool boiling heat transfer

OpenAIRE

M. Kumar; V. Bhutani; P. Khatak

2015-01-01

This paper reviews the previous work carried on pool boiling heat transfer during heating of various liquids and commodities categorized as refrigerants and dielectric fluids, pure liquids, nanofluids, hydrocarbons and additive mixtures, as well as natural and synthetic colloidal solutions. Nucleate pool boiling is an efficient and effective method of boiling because high heat fluxes are possible with moderate temperature differences. It is characterized by the growth of bubbles on a heated s...

Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility

International Nuclear Information System (INIS)

Coragem, Helio Boemer de Oliveira

1980-01-01

A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

dynamic performance of research reactors

International Nuclear Information System (INIS)

Abo elnor, A.G.M.

2007-01-01

this work studies the dynamic performance of material testing reactor (MTR), where the dynamic performance of any reactor reflects its safety behavior and it should enhance its intrinsic characteristics s ystem corrects itself internally without introducing external corrective action . the present work analyzes and studies the dynamic performance of mtr through the transfer function. the servo system parameters can be changed to fit the system demand. the servo system is an excellent approximation to some of the practical servo system currently use in reactor control system, and a quadratic form of this sort should closely approximate the behavior of almost any type of physical equipment which might be chosen to drive a control rod. proposed changes in servo system parameters could enhance the dynamic performance of the system , but the suitable parameters can be evaluated by using the automatic reactor power control system model

Dominant seismic sloshing mode in a pool-type reactor tank

International Nuclear Information System (INIS)

Ma, D.C.; Gvildys, J.; Chang, Y.W.

1987-01-01

Large-diameter LMR (Liquid Metal Reactor) tanks contain a large volume of sodium coolant and many in-tank components. A reactor tank of 70 ft. in diameter contains 5,000,000 of sodium coolant. Under seismic events, the sloshing wave may easily reach several feet. If sufficient free board is not provided to accommodate the wave height, several safety problems may occur such as damage to tank cover due to sloshing impact and thermal shocks due to hot sodium, etc. Therefore, the sloshing response should be properly considered in the reactor design. This paper presents the results of the sloshing analysis of a pool-type reactor tank with a diameter of 39 ft. The results of the fluid-structure interaction analysis are presented in a companion paper. Five sections are contained in this paper. The reactor system and mathematical model are described. The dominant sloshing mode and the calculated maximum wave heights are presented. The sloshing pressures and sloshing forces acting on the submerged components are described. The conclusions are given

Influence of reactor design on the establishment of natural circulation in pool-type LMFBR

International Nuclear Information System (INIS)

Durham, M.E.

1976-01-01

The general principles involved in establishing natural circulation in a pool-type liquid metal cooled fast breeder reactor following loss of a.c. supplies are elucidated and the effects of design features by use of the computer code MELANI are quantified. It is shown that natural circulation can provide a feasible means of emergency core cooling in addition to that provided by pony motors. The choice of primary pump rundown time has a significant effect in controlling peak core outlet temperatures in the hypothetical case of natural circulation alone being the core heat removal process. (author)

A CFD numerical model for the flow distribution in a MTR fuel element

International Nuclear Information System (INIS)

Andrade, Delvonei Alves de; Santos, Pedro Henrique Di Giovanni; Oliveira, Fabio Branco Vaz de; Torres, Walmir Maximo; Umbehaun, Pedro Ernesto; Souza, Jose Antonio Batista de; Belchior Junior, Antonio; Sabundjian, Gaiane; Prado, Adelk de Carvalho; Angelo, Gabriel

2015-01-01

Previously, an instrumented dummy fuel element (DMPV-01), with the same geometric characteristics of a MTR fuel element, was designed and constructed for pressure drop and flow distribution measurement experiments at the IEA-R1 reactor core. This dummy element was also used to measure the flow distribution among the rectangular flow channels formed by element fuel plates. A CFD numerical model was developed to complement the studies. This work presents the proposed CFD model as well as a comparison between numerical and experimental results of flow rate distribution among the internal flow channels. Numerical results show that the model reproduces the experiments very well and can be used for the studies as a more convenient and complementary tool. (author)

A CFD numerical model for the flow distribution in a MTR fuel element

Energy Technology Data Exchange (ETDEWEB)

Andrade, Delvonei Alves de; Santos, Pedro Henrique Di Giovanni; Oliveira, Fabio Branco Vaz de; Torres, Walmir Maximo; Umbehaun, Pedro Ernesto; Souza, Jose Antonio Batista de; Belchior Junior, Antonio; Sabundjian, Gaiane; Prado, Adelk de Carvalho, E-mail: acprado@ipen.br, E-mail: delvonei@ipen.br, E-mail: dpedro_digiovanni_s@hotmail.com, E-mail: fabio@ipen.br, E-mail: wmtorres@ipen.br, E-mail: umbehaun@ipen.br, E-mail: jasouza@ipen.br, E-mail: abelchior@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear; Angelo, Edvaldo, E-mail: eangelo@mackenzie.br [Universidade Presbiteriana Mackenzie, Sao Paulo, SP (Brazil); Angelo, Gabriel, E-mail: gangelo@fei.edu.br [Fundacao Educacional Inaciana (FEI), Sao Bernardo do Campo, SP (Brazil)

2015-07-01

Previously, an instrumented dummy fuel element (DMPV-01), with the same geometric characteristics of a MTR fuel element, was designed and constructed for pressure drop and flow distribution measurement experiments at the IEA-R1 reactor core. This dummy element was also used to measure the flow distribution among the rectangular flow channels formed by element fuel plates. A CFD numerical model was developed to complement the studies. This work presents the proposed CFD model as well as a comparison between numerical and experimental results of flow rate distribution among the internal flow channels. Numerical results show that the model reproduces the experiments very well and can be used for the studies as a more convenient and complementary tool. (author)

c-Type cytochrome-dependent formation of U(IV nanoparticles by Shewanella oneidensis.

Directory of Open Access Journals (Sweden)

Matthew J Marshall

2006-09-01

Full Text Available Modern approaches for bioremediation of radionuclide contaminated environments are based on the ability of microorganisms to effectively catalyze changes in the oxidation states of metals that in turn influence their solubility. Although microbial metal reduction has been identified as an effective means for immobilizing highly-soluble uranium(VI complexes in situ, the biomolecular mechanisms of U(VI reduction are not well understood. Here, we show that c-type cytochromes of a dissimilatory metal-reducing bacterium, Shewanella oneidensis MR-1, are essential for the reduction of U(VI and formation of extracellular UO(2 nanoparticles. In particular, the outer membrane (OM decaheme cytochrome MtrC (metal reduction, previously implicated in Mn(IV and Fe(III reduction, directly transferred electrons to U(VI. Additionally, deletions of mtrC and/or omcA significantly affected the in vivo U(VI reduction rate relative to wild-type MR-1. Similar to the wild-type, the mutants accumulated UO(2 nanoparticles extracellularly to high densities in association with an extracellular polymeric substance (EPS. In wild-type cells, this UO(2-EPS matrix exhibited glycocalyx-like properties and contained multiple elements of the OM, polysaccharide, and heme-containing proteins. Using a novel combination of methods including synchrotron-based X-ray fluorescence microscopy and high-resolution immune-electron microscopy, we demonstrate a close association of the extracellular UO(2 nanoparticles with MtrC and OmcA (outer membrane cytochrome. This is the first study to our knowledge to directly localize the OM-associated cytochromes with EPS, which contains biogenic UO(2 nanoparticles. In the environment, such association of UO(2 nanoparticles with biopolymers may exert a strong influence on subsequent behavior including susceptibility to oxidation by O(2 or transport in soils and sediments.

FRG-1: new millenium - new compact core

International Nuclear Information System (INIS)

Schreiner, P.; Knop, W.

2001-01-01

The GKSS research center Geesthacht GmbH operates the MTR-type swimming pool research reactor FRG-1 (5 MW) for more than 40 years. The FRG-1 has been converted in February 1991 from HEU (93 %) to LEU (20 %) in one step and at that time the core size was reduced from 49 to 26 fuel elements. Consequently the thermal neutron flux in beam tube positions could be increased by more than a factor of two. It is the strong intention of GKSS to continue the operation of the FRG-1 research reactor for at least an additional 15 years with high availability and utilization. The reactor has been operated during the last years for approximately 250 full power days per year. To prepare the FRG-1 for an efficient future use, the core size has been reduced in a second step from 26 fuel elements to 12 fuel elements. (author)

The Nuclear Insurance Pools: Operations and Covers

International Nuclear Information System (INIS)

Tetley, M.

2008-01-01

Nuclear insurance pools have provided insurance for the nuclear industry for over fifty years and it is fair to say that the development of civil nuclear power would not have been possible without the support of the commercial insurance market. The unknown risks presented by the nascent nuclear power industry in the 1950s required a leap of faith by insurers who developed specialist pooled insurance capacity to ensure adequate capacity to back up the operators' compensation obligations. Since then, nuclear insurance pools have evolved to become comprehensive suppliers of most types of insurance for nuclear plant globally. This paper will outline the structure, development, products and current operations of nuclear insurance pools.(author)

Code Development of Radioactive Aerosol Scrubbing in Pool-Injection Zone

Energy Technology Data Exchange (ETDEWEB)

Jo, Hyun Joung; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jang, Dong Soon [Chungnam National University, Daejeon (Korea, Republic of)

2015-10-15

The pool scrubbing models were reviewed and an aerosol scrubbing code has been prepared to calculate decontamination factor through the injection zone. The developed code has been verified using the experimental results and evaluated parametrically on the input variables. In injection zone, the initial steam condensation was most effective mechanism for the aerosol removal, and the steam fraction and pool temperature were highly affected on the decontamination factor by initial steam condensation. The aerosol scrubbing code will be updated to evaluate the decontamination factor at rise zone and finally whole pool scrubber phenomena. If a severe accident occurs in a nuclear power plant (NPP), the aerosol and gaseous fission products might be produced in the reactor vessel, and then released to the environment after the containment failure. FCVS (Filtered Containment Venting System) is one of the severe accident mitigation systems for retaining the containment integrity by discharging the high-temperature and high-pressure fission products to the environment after passing through the filtration system. In general, the FCVS is categorized into two types, wet and dry types. The scrubbing pool could play an important role in the wet type FCVS because a large amount of aerosol is captured in the water pool. The pool scrubbing phenomena have been modelled and embedded in several computer codes, such as SPARC (Suppression Pool Aerosol Removal Code), BUSCA (BUbble Scrubbing Algorithm) and SUPRA (Suppression Pool Retention Analysis). These codes aim at simulating the pool scrubbing process and estimating the decontamination factors (DFs) of the radioactive aerosol and iodine gas in the water pool, which is defined as the ratio of initial mass of the specific radioactive material to final massy after passing through the water pool. The pool scrubbing models were reviewed and an aerosol scrubbing code has been prepared to calculate decontamination factor through the injection

Thai research reactor

International Nuclear Information System (INIS)

Aramrattana, M.

1987-01-01

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)

Backfitting of the FRG reactors

Energy Technology Data Exchange (ETDEWEB)

Krull, W [GKSS-Forschungszentrum Geesthacht GmbH, Geesthacht (Germany)

1990-05-01

The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U{sub 3}Si{sub 2} fuel. Both cooling towers were repaired. Replacement of instrumentation is planned.

Backfitting of the FRG reactors

International Nuclear Information System (INIS)

Krull, W.

1990-01-01

The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U 3 Si 2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

International Nuclear Information System (INIS)

Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

2008-01-01

The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

Decontamination and decommissioning of the MTR [Materials Testing Reactor]-603 HB-2 cubicle

International Nuclear Information System (INIS)

Smith, D.L.

1987-10-01

This paper describes the decontamination and decommissioning (D and D) of the MTR-603 HB-2 cubicle located at the Idaho National Engineering Laboratory (INEL). The HB-2 cubicle became radioactively contaminated during out-of-pile circulating water loop experiments conducted in the Materials Testing Reactor in the 1950s and 1960s. This paper describes work performed to accomplish the D and D objectives of reducing the high radiation fields caused by contamination inside the cubicle, preventing future contamination spread, and making about 1400 ft 2 of floor space available for reuse. Decommissioning of the HB-2 cubicle consisted of total dismantlement of the cubicle and its contents and was performed without disrupting ongoing laboratory work being conducted in areas surrounding the HB-2 cubicle. 3 refs., 7 figs., 4 tabs

Investigation of the condition of spent-fuel pool components

International Nuclear Information System (INIS)

Kustas, F.M.; Bates, S.O.; Opitz, B.E.; Johnson, A.B. Jr.; Perez, J.M. Jr.; Farnsworth, R.K.

1981-09-01

It is currently projected that spent nuclear fuel, which is discharged from the reactor and then stored in water pools, may remain in those pools for several decades. Other studies have addressed the expected integrity of the spent fuel during extended water storage; this study assesses the integrity of metallic spent fuel pool components. Results from metallurgical examinations of specimens taken from stainless steel and aluminum components exposed in spent fuel pools are presented. Licensee Event Reports (LERs) relating to problems with spent fuel components were assessed and are summarized to define the types of operational problems that have occurred. The major conclusions of this study are: aluminum and stainless steel spent fuel pool components have a good history of performance in both deionized and borated water pools. Although some operational problems involving pool components have occurred, these problems have had minimal impacts

Investigation of the condition of spent-fuel pool components

Energy Technology Data Exchange (ETDEWEB)

Kustas, F.M.; Bates, S.O.; Opitz, B.E.; Johnson, A.B. Jr.; Perez, J.M. Jr.; Farnsworth, R.K.

1981-09-01

It is currently projected that spent nuclear fuel, which is discharged from the reactor and then stored in water pools, may remain in those pools for several decades. Other studies have addressed the expected integrity of the spent fuel during extended water storage; this study assesses the integrity of metallic spent fuel pool components. Results from metallurgical examinations of specimens taken from stainless steel and aluminum components exposed in spent fuel pools are presented. Licensee Event Reports (LERs) relating to problems with spent fuel components were assessed and are summarized to define the types of operational problems that have occurred. The major conclusions of this study are: aluminum and stainless steel spent fuel pool components have a good history of performance in both deionized and borated water pools. Although some operational problems involving pool components have occurred, these problems have had minimal impacts.

Corrosion of aluminium alloy test coupons in water of spent fuel storage pool at RA reactor

International Nuclear Information System (INIS)

Pesic, M.; Maksin, T.; Jordanov, G.; Dobrijevic, R.

2004-12-01

Study on corrosion of aluminium cladding, of the TVR-S type of enriched uranium spent fuel elements of the research reactor RA in the storage water pool is examined in the framework nr the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) 'Corrosion of Research Reactor Clad-Clad Spent Fuel in Water' since 2002. Standard racks with aluminium coupons are exposed to water in the spent fuel pools of the research reactor RA. After predetermined exposure times along with periodic monitoring of the water parameters, the coupons are examined according to the strategy and the protocol supplied by the IAEA. Description of the standard corrosion racks, experimental protocols, test procedures, water quality monitoring and compilation of results of visual examination of corrosion effects are present in this article. (author)

Pool scrubbing models for iodine components

Energy Technology Data Exchange (ETDEWEB)

Fischer, K [Battelle Ingenieurtechnik GmbH, Eschborn (Germany)

1996-12-01

Pool scrubbing is an important mechanism to retain radioactive fission products from being carried into the containment atmosphere or into the secondary piping system. A number of models and computer codes has been developed to predict the retention of aerosols and fission product vapours that are released from the core and injected into water pools of BWR and PWR type reactors during severe accidents. Important codes in this field are BUSCA, SPARC and SUPRA. The present paper summarizes the models for scrubbing of gaseous Iodine components in these codes, discusses the experimental validation, and gives an assessment of the state of knowledge reached and the open questions which persist. The retention of gaseous Iodine components is modelled by the various codes in a very heterogeneous manner. Differences show up in the chemical species considered, the treatment of mass transfer boundary layers on the gaseous and liquid sides, the gas-liquid interface geometry, calculation of equilibrium concentrations and numerical procedures. Especially important is the determination of the pool water pH value. This value is affected by basic aerosols deposited in the water, e.g. Cesium and Rubidium compounds. A consistent model requires a mass balance of these compounds in the pool, thus effectively coupling the pool scrubbing phenomena of aerosols and gaseous Iodine species. Since the water pool conditions are also affected by drainage flow of condensate water from different regions in the containment, and desorption of dissolved gases on the pool surface is determined by the gas concentrations above the pool, some basic limitations of specialized pool scrubbing codes are given. The paper draws conclusions about the necessity of coupling between containment thermal-hydraulics and pool scrubbing models, and proposes ways of further simulation model development in order to improve source term predictions. (author) 2 tabs., refs.

Pool scrubbing models for iodine components

International Nuclear Information System (INIS)

Fischer, K.

1996-01-01

Pool scrubbing is an important mechanism to retain radioactive fission products from being carried into the containment atmosphere or into the secondary piping system. A number of models and computer codes has been developed to predict the retention of aerosols and fission product vapours that are released from the core and injected into water pools of BWR and PWR type reactors during severe accidents. Important codes in this field are BUSCA, SPARC and SUPRA. The present paper summarizes the models for scrubbing of gaseous Iodine components in these codes, discusses the experimental validation, and gives an assessment of the state of knowledge reached and the open questions which persist. The retention of gaseous Iodine components is modelled by the various codes in a very heterogeneous manner. Differences show up in the chemical species considered, the treatment of mass transfer boundary layers on the gaseous and liquid sides, the gas-liquid interface geometry, calculation of equilibrium concentrations and numerical procedures. Especially important is the determination of the pool water pH value. This value is affected by basic aerosols deposited in the water, e.g. Cesium and Rubidium compounds. A consistent model requires a mass balance of these compounds in the pool, thus effectively coupling the pool scrubbing phenomena of aerosols and gaseous Iodine species. Since the water pool conditions are also affected by drainage flow of condensate water from different regions in the containment, and desorption of dissolved gases on the pool surface is determined by the gas concentrations above the pool, some basic limitations of specialized pool scrubbing codes are given. The paper draws conclusions about the necessity of coupling between containment thermal-hydraulics and pool scrubbing models, and proposes ways of further simulation model development in order to improve source term predictions. (author) 2 tabs., refs

Dynamic design load of type 2 water-flow capsule in Nuclear Safety Research Reactor in Tokai Research Establishment of Japan Atomic Energy Research Institute, and its reuse test

International Nuclear Information System (INIS)

1981-01-01

A report by the Nuclear Safety Bureau of the Science and Technology Agency to the Nuclear Safety Commission was presented on the validity of the dynamic design load of type 2 water-flow capsule and the method of its reuse test. The safety in both aspects of the capsule was confirmed. The Nuclear Safety Research Reactor (NSRR), in which the water-flow capsule is set, is a swimming pool type reactor, fueled with enriched uranium, having heat output of 300 kW in normal operation and maximum instantaneous heat output of 23,000 MW in pulse operation. The type 2 water-flow capsule, with the initial conditions simulating a power generating LWR plant and being appropriately set, is used to acquire the data on fuel behavior and destructive power in pulse irradiation. (J.P.N.)

Dynamic design load of type 2 water-flow capsule in Nuclear Safety Research Reactor in Tokai Research Establishment of Japan Atomic Energy Research Institute, and its reuse test

Energy Technology Data Exchange (ETDEWEB)

1981-01-01

A report by the Nuclear Safety Bureau of the Science and Technology Agency to the Nuclear Safety Commission was presented on the validity of the dynamic design load of type 2 water-flow capsule and the method of its reuse test. The safety in both aspects of the capsule was confirmed. The Nuclear Safety Research Reactor (NSRR), in which the water-flow capsule is set, is a swimming pool type reactor, fueled with enriched uranium, having heat output of 300 kW in normal operation and maximum instantaneous heat output of 23,000 MW in pulse operation. The type 2 water-flow capsule, with the initial conditions simulating a power generating LWR plant and being appropriately set, is used to acquire the data on fuel behavior and destructive power in pulse irradiation.

Hawaii ESI: POOLS (Anchialine Pool Points)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains sensitive biological resource data for anchialine pools in Hawaii. Anchialine pools are small, relatively shallow coastal ponds that occur...

Technical ability of new MTR high-density fuel alloys regarding the whole fuel cycle

International Nuclear Information System (INIS)

Durand, J.P.; Maugard, B.; Gay, A.

1998-01-01

The development of new fuel alloys could provide a good opportunity to improve drastically the fuel cycle on the neutronic performances and the reprocessing point of view. Nevertheless, those parameters can only be considered if the fuel manufacture feasibility has been previously demonstrated. As a matter of fact, a MTR work group involving French partners (CEA, CERCA, COGEMA) has been set up in order to evaluate the technical ability of new fuels considering the whole fuel cycle. In this paper CERCA is presenting the preliminary results of UMo and UNbZr fuel plate manufacture, CEA is comparing to U 3 Si 2 the neutronic performances of fuels such as UMo, UN, UNbZr, while COGEMA is dealing with the reprocessing feasibility. (author)

Analysis of pressure distribution originated over the external plate window of the RA-10 nuclear fuel

International Nuclear Information System (INIS)

Gramajo, M A; Garcia, J.C

2012-01-01

The RA10 is a pool type multipurpose research reactor. The core consists of a rectangular array of MTR fuel type. The refrigeration system at full power and normal operations conditions is carried out by an ascendant flow through the core. To ensure the refrigeration in the sub-channel formed between two adjacent fuels, there is a window orifice over the outer fuel plate. Part of the coolant flow that gets into the fuel will be derived by the window orifice to the sub-channel. Due to the change in the coolant flow direction is necessary to establish the pressure distribution originated over the window In order to achieve this goal a CFD commercial code (FLUENT v6.3.26) was used to perform numerical simulations to obtain the pressure distribution over the window. A quarter of the fuel was modeled using proper symmetry and boundaries conditions (author)

Safety features of TR-2 reactor

International Nuclear Information System (INIS)

Tuerker, T.

2001-01-01

TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

Decontamination of outdoor school swimming pools in Fukushima

International Nuclear Information System (INIS)

Saegusa, Jun

2013-01-01

After the Fukushima Daiichi NPP accident following the Great East Japan Earthquake, many school swimming pools in Fukushima have suspended water discharge, due to concerns that pool water which contains radioactive fallout is discharged into a river or waterway for agricultural use. The Japan Atomic Energy Agency conducted researches and examinations on the existing absorbent method and the flocculation method as ways for decontaminating pool water. By reviewing and improving these methods through decontamination demonstrations at eight pools in Fukushima, a practical decontamination method for outdoor pools has been established. This report summarizes the methods and results of the decontamination demonstrations carried out at the schools. Also, the surface density of fallout estimated at one of the pools is also presented and discussed in connection with the overall collection ratio of radiocesium at the pool. (author)

On the lability and functional significance of the type 1 copper pool in ceruloplasmin.

Science.gov (United States)

Musci, G; Fraterrigo, T Z; Calabrese, L; McMillin, D R

1999-08-01

The possibility that ceruloplasmin (CP) functions as a copper transferase has fueled a continuing interest in studies of the copper release process. The principal goal of the current investigation has been to identify the most labile copper centers in sheep protein. In fact, subjecting the enzyme to a slow flux of cyanide at pH 5.2 under nitrogen in the presence of ascorbate and a phenanthroline ligand produces partially demetalated forms of the protein. By standard chromatographic techniques it is possible to isolate protein with a Cu/CP ratio of approximately 4 or approximately 5 as opposed to the native protein which has Cu/CP = 5.8. In contrast to other blue oxidases, analysis suggests that CP preferentially loses its type 1 coppers under these conditions. Thus, the spectroscopic signals from the type 1 centers exhibit a loss of intensity while the EPR signal of the type 2 copper becomes stronger. Furthermore, the Cu/CP approximately 4 and Cu/ CP approximately 5 components retain about 50% of the activity of the native protein, consistent with an intact type 2/type 3 cluster. All three type 1 copper sites appear to suffer copper loss. Reconstitution with a copper(I) reagent restores the spectroscopic properties of the native protein and 90% of the original activity. The results suggest a possible functional significance for the presence of three type 1 coppers in CP. By employing a pool of redox-active but relatively labile type 1 copper centers, the enzyme can serve as a copper donor, if necessary, without completely sacrificing its oxidase activity.

Myofascial trigger point-focused head and neck massage for recurrent tension-type headache: A randomized, placebo-controlled clinical trial

Science.gov (United States)

Moraska, Albert F.; Stenerson, Lea; Butryn, Nathan; Krutsch, Jason P.; Schmiege, Sarah J.; Mann, J. Douglas

2014-01-01

Objective Myofascial trigger points (MTrPs) are focal disruptions in skeletal muscle that can refer pain to the head and reproduce the pain patterns of tension-type headache (TTH). The present study applied massage focused on MTrPs of subjects with TTH in a placebo-controlled, clinical trial to assess efficacy on reducing headache pain. Methods Fifty-six subjects with TTH were randomized to receive 12 massage or placebo (detuned ultrasound) sessions over six weeks, or to wait-list. Trigger point release (TPR) massage focused on MTrPs in cervical musculature. Headache pain (frequency, intensity and duration) was recorded in a daily headache diary. Additional outcome measures included self-report of perceived clinical change in headache pain and pressure-pain threshold (PPT) at MTrPs in the upper trapezius and sub-occipital muscles. Results From diary recordings, group differences across time were detected in headache frequency (p=0.026), but not for intensity or duration. Post hoc analysis indicated headache frequency decreased from baseline for both massage (pheadache pain for massage than placebo or wait-list groups (p=0.002). PPT improved in all muscles tested for massage only (all p'streatment of TTH, and 2) TTH, like other chronic conditions, is responsive to placebo. Clinical trials on headache that do not include a placebo group are at risk for overestimating the specific contribution from the active intervention. PMID:25329141

A neutronic feasibility study for LEU conversion of the Brookhaven Medical Research Reactor (BMRR).

Energy Technology Data Exchange (ETDEWEB)

Hanan, N. A.

1998-01-14

A neutronic feasibility study for converting the Brookhaven Medical Research Reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with Brookhaven National Laboratory. Two possible LEU cores were identified that would provide nearly the same neutron flux and spectrum as the present HEU core at irradiation facilities that are used for Boron Neutron Capture Therapy and for animal research. One core has 17 and the other has 18 LEU MTR-type fuel assemblies with uranium densities of 2.5g U/cm{sup 3} or less in the fuel meat. This LEU fuel is fully-qualified for routine use. Thermal hydraulics and safety analyses need to be performed to complete the feasibility study.

Addressing data privacy in matched studies via virtual pooling.

Science.gov (United States)

Saha-Chaudhuri, P; Weinberg, C R

2017-09-07

Data confidentiality and shared use of research data are two desirable but sometimes conflicting goals in research with multi-center studies and distributed data. While ideal for straightforward analysis, confidentiality restrictions forbid creation of a single dataset that includes covariate information of all participants. Current approaches such as aggregate data sharing, distributed regression, meta-analysis and score-based methods can have important limitations. We propose a novel application of an existing epidemiologic tool, specimen pooling, to enable confidentiality-preserving analysis of data arising from a matched case-control, multi-center design. Instead of pooling specimens prior to assay, we apply the methodology to virtually pool (aggregate) covariates within nodes. Such virtual pooling retains most of the information used in an analysis with individual data and since individual participant data is not shared externally, within-node virtual pooling preserves data confidentiality. We show that aggregated covariate levels can be used in a conditional logistic regression model to estimate individual-level odds ratios of interest. The parameter estimates from the standard conditional logistic regression are compared to the estimates based on a conditional logistic regression model with aggregated data. The parameter estimates are shown to be similar to those without pooling and to have comparable standard errors and confidence interval coverage. Virtual data pooling can be used to maintain confidentiality of data from multi-center study and can be particularly useful in research with large-scale distributed data.

Using proliferation assessment methodologies for Safeguards-by-Design

International Nuclear Information System (INIS)

Van der Meer, K.; Rossa, R.; Turcanu, C.; Borella, A.

2013-01-01

MYRRHA, an accelerator driven system (ADS) is designed as a proton accelerator coupled to a liquid Pb-Bi spallation target, surrounded by a Pb-Bi cooled sub-critical neutron multiplying medium in a pool type configuration. An assessment based on three methodologies was made of the proliferation risks of the MYRRHA ADS in comparison with the BR2 MTR, an existing research reactor at the Belgian Nuclear Research Centre SCK-CEN. The used methodologies were the TOPS (Technical Opportunities to Increase the Proliferation Resistance of Nuclear Power Systems), the PR-PP and the INPRO methodologies. The various features of the methodologies are described and the results of the assessments are given and discussed. It is concluded that it would be useful to define one single methodology with two options to perform a quick and a more detailed assessment. The paper is followed by the slides of the presentation

Research Campus Types | Climate Neutral Research Campuses | NREL

Science.gov (United States)

Research Campus Types Research Campus Types Research campuses and laboratories come in all shapes and sizes, but have one thing in common; performing vital research and development. These campuses Private sector industries Federal, State, and Local Government Laboratories and research campuses operate

Cooling Performance of Natural Circulation for a Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

Birth order and childhood type 1 diabetes risk: a pooled analysis of 31 observational studies

DEFF Research Database (Denmark)

Cardwell, Chris R; Stene, Lars C; Joner, Geir

2010-01-01

BACKGROUND: The incidence rates of childhood onset type 1 diabetes are almost universally increasing across the globe but the aetiology of the disease remains largely unknown. We investigated whether birth order is associated with the risk of childhood diabetes by performing a pooled analysis...... and after adjustment for confounders, and investigate heterogeneity. RESULTS: Data were available for 6 cohort and 25 case-control studies, including 11¿955 cases of type 1 diabetes. Overall, there was no evidence of an association prior to adjustment for confounders. After adjustment for maternal age...... at birth and other confounders, a reduction in the risk of diabetes in second- or later born children became apparent [fully adjusted OR¿=¿0.90 95% confidence interval (CI) 0.83-0.98; P¿=¿0.02] but this association varied markedly between studies (I(2)¿=¿67%). An a priori subgroup analysis showed...

Seismic isolation structure for pool-type LMFBR - isolation building with vertically isolated floor for NSSS

International Nuclear Information System (INIS)

Sakurai, A.; Shiojiri, H.; Aoyagi, S.; Matsuda, T.; Fujimoto, S.; Sasaki, Y.; Hirayama, H.

1987-01-01

The NSSS isolation floor vibration characteristics were made clear. Especially, the side support bearing (rubber bearing) is effective for horizontal floor motion restraint and rocking motion control. Seismic isolation effects for responses of the reactor components can be sufficiently expected, using the vertical seismic isolation floor. From the analytical and experimental studies, the following has been concluded: (1) Seismic isolation structure, which is suitable for large pool-type LMFBR, were proposed. (2) Seismic response characteristics of the seismic isolation structure were investigated. It was made clear that the proposed seismic isolation (Combination of the isolated building and the isolated NSSS floor) was effective. (orig./HP)

Variation of biomass and carbon pools with forest type in temperate forests of Kashmir Himalaya, India.

Science.gov (United States)

Dar, Javid Ahmad; Sundarapandian, Somaiah

2015-02-01

An accurate characterization of tree, understory, deadwood, floor litter, and soil organic carbon (SOC) pools in temperate forest ecosystems is important to estimate their contribution to global carbon (C) stocks. However, this information on temperate forests of the Himalayas is lacking and fragmented. In this study, we measured C stocks of tree (aboveground and belowground biomass), understory (shrubs and herbaceous), deadwood (standing and fallen trees and stumps), floor litter, and soil from 111 plots of 50 m × 50 m each, in seven forest types: Populus deltoides (PD), Juglans regia (JR), Cedrus deodara (CD), Pinus wallichiana (PW), mixed coniferous (MC), Abies pindrow (AP), and Betula utilis (BU) in temperate forests of Kashmir Himalaya, India. The main objective of the present study is to quantify the ecosystem C pool in these seven forest types. The results showed that the tree biomass ranged from 100.8 Mg ha(-1) in BU forest to 294.8 Mg ha(-1) for the AP forest. The understory biomass ranged from 0.16 Mg ha(-1) in PD forest to 2.36 Mg ha(-1) in PW forest. Deadwood biomass ranged from 1.5 Mg ha(-1) in PD forest to 14.9 Mg ha(-1) for the AP forest, whereas forest floor litter ranged from 2.5 Mg ha(-1) in BU and JR forests to 3.1 Mg ha(-1) in MC forest. The total ecosystem carbon stocks varied from 112.5 to 205.7 Mg C ha(-1) across all the forest types. The C stocks of tree, understory, deadwood, litter, and soil ranged from 45.4 to 135.6, 0.08 to 1.18, 0.7 to 6.8, 1.1 to 1.4, and 39.1-91.4 Mg ha(-1), respectively, which accounted for 61.3, 0.2, 1.4, 0.8, and 36.3 % of the total carbon stock. BU forest accounted 65 % from soil C and 35 % from biomass, whereas PD forest contributed only 26 % from soil C and 74 % from biomass. Of the total C stock in the 0-30-cm soil, about 55 % was stored in the upper 0-10 cm. Soil C stocks in BU forest were significantly higher than those in other forests. The variability of C pools of different ecosystem components is

Solar swimming pool

Energy Technology Data Exchange (ETDEWEB)

1985-01-01

This report examines the feasibility of using solar collectors to heat the water in a previously unheated outdoor swimming pool. The solar system is used in conjunction with a pool blanket, to conserve heat when the pool is not in use. Energy losses through evaporation can be reduced by as much as 70% by a pool blanket. A total of 130 m{sup 2} of highly durable black synthetic collectors were installed on a support structure at a 30{degree} angle from the horizontal, oriented to the south. Circulation of pool water though the collectors, which is controlled by a differential thermostat, was done with the existing pool pump. Before installation the pool temperature averaged 16{degree}C; after installation it ranged from 20{degree} to 26{degree}C. It was hard to distinguish how much pool heating was due to the solar system and how much heat was retained by the pool blanket. However, the pool season was extended by five weeks and attendance tripled. 2 figs.

HEU and LEU MTR fuel elements as target materials for the production of fission molybdenum

International Nuclear Information System (INIS)

Sameh, A.A.; Bertram-Berg, A.

1993-01-01

The processing of irradiated MTR-fuels for the production of fission nuclides for nuclear medicine presents a significantly increasing task in the field of chemical separation technology of high activity levels. By far the most required product is MO-99, the mother nuclide of Tc-99m which is used in over 90% of the organ function tests in nuclear medicine. Because of the short half life of Mo-99 (66 h) the separation has to be carried out from shortly cooled neutron irradiated U-targets. The needed product purity, the extremely high radiation level, the presence of fission gases like xenon-133 and of volatile toxic isotopes such as iodine-131 and its compounds in kCi-scale require a sophisticated process technology

A fundamental study on sodium-water reaction in the double-pool-type LMFBR

International Nuclear Information System (INIS)

Yoshida, Kazuo; Akimoto, Tokuzo

1987-01-01

In order to evaluate the pressure rise by large sodium-water reaction in the Double-Pool LMFBR, basic tests on pressure wave celerity in rectangular tube are carried out. The initial spike pressure in rectangular-shelled steam generator of the Double Pool reactor, strongly depends on pressure wave celerity. In this study, celerity was measured as a function of pressure wave rising time and pulse height, and influence of water around the test section on celerity was investigated. (author)

Pressure tube type research reactor

International Nuclear Information System (INIS)

Ueda, Hiroshi.

1975-01-01

Object: To permit safe and reliable replacement of primary pipes by providing a reactor container so as to surround a pressure pipe, with upper portions of the two separably coupled together, and coupling the pressure pipe and primary piping by joint coupling above and below the reactor container, with the lower coupling joint surrounded by drain receptacle. Structure: At the time of replacement of a pressure pipe, a partition valve is opened to exhaust primary cooling water within pressure pipe and upper and lower portions of the primary piping and replace the decelerator within the reactor container with water of the same quality as that of pool water within an upper shield pool. Thereafter, the entire space above the drain receptacle is filled with pool water by closing a partition valve and opening a water supply valve. Then, upper portion seal cover, pool bottom lid, upper joint and upper portion primary piping are removed, then bolts and nuts are loosened, and the pressure pipe is taken out together with the shield block. (Kamimura, M.)

Integrity assessment of research reactor fuel cladding and material testing using eddy current inspection; Avaliacao de integridade de revestimentos de combustiveis de reatores de pesquisa e teste de materiais utilizando o ensaio de correntes parasitas

Energy Technology Data Exchange (ETDEWEB)

Alencar, Donizete Anderson de

2004-07-01

A methodology to perform the integrity assessment of research reactors nuclear fuels cladding, such as those installed in IPR-Rl (TRIGA) and IEA-R1 (MTR), using nondestructive electromagnetic inspection (eddy current) is presented. This methodology is constituted by: the development of calibration reference standards, specific for each type of fuel; the development of special test probes; the recommendations for the inspection equipment calibration; the construction of voltage based evaluation curves and the inspection procedures developed for the characterization of detected flaws. The test probes development, specially those designed for the inspection of MTR fuels cladding, which present access difficulties due to the narrow gap between fuel plates (2,89 mm for IEAR-R1), constituted a challenge that demanded the introduction of unusual materials and constructive techniques. The operational performance of the developed resources, as well as the special operative characteristics of the test probes, such as their immunity to adjacent fuel plates interference and electrical resistivity changes of the fuels meat are experimentally demonstrated. The practical applicability of the developed methodology is verified in non radioactive environment, using a dummy MTR fuel element model, similar to an IEA-R1 reactor fuel element, produced and installed in IPEN, Sao Paulo. The efficacy of the proposed methodology was verified by the achieved results. (author)

Livestock Grazing as a Driver of Vernal Pool Ecohydrology

Science.gov (United States)

Michaels, J.; McCarten, N. F.

2017-12-01

Vernal pools are seasonal wetlands that host rare plant communities of high conservation priority. Plant community composition is largely driven by pool hydroperiod. A previous study found that vernal pools grazed by livestock had longer hydroperiods compared with pools excluded from grazing for 10 years, and suggests that livestock grazing can be used to protect plant diversity. It is important to assess whether observed differences are due to the grazing or due to water balance variables including upland discharge into or out of the pools since no a priori measurements were made of the hydrology prior to grazing. To address this question, in 2016 we compared 15 pools that have been grazed continuously and 15 pools that have been fenced off for over 40 years at a site in Sacramento County. We paired pools based on abiotic characteristics (size, shape, slope, soil type) to minimize natural variation. We sampled vegetation and water depth using Solinst level loggers. We found that plant diversity and average hydroperiod was significantly higher in the grazed pools. We are currently measuring groundwater connectivity and upland inputs in order to compare the relative strength of livestock grazing as a driver of hydroperiod to these other drivers.

Application of neutron noise analysis to a swimming pool research reactor

International Nuclear Information System (INIS)

Behringer, K.; Lescano, V.H.; Meier, F.; Phildius, J.; Winkler, H.

1982-01-01

This work is part of a programme of establishing practical applications of neutron noise techniques to a swimming pool research reactor and deals with two different items: (1) The identification of local boiling caused e.g. by a partial blockage of the coolant flow in a fuel element. Local boiling can easily lead to a burn-out situation. The onset of boiling can be detected by neutron noise analysis and a boiling detection system is presently under development. (2) The measurement of the time evolution of the reactivity induced by xenon after reactor shut-down by an on-line reactivity meter based on neutron noise analysis. From the data, the prompt neutron decay constant at delayed critical, the equilibrium xenon reactivity worth, and an estimate of the average steady-state power flux in the core before reactor shut-down were obtained. (author)

Pooled biological specimens for human biomonitoring of environmental chemicals: opportunities and limitations.

Science.gov (United States)

Heffernan, Amy L; Aylward, Lesa L; Toms, Leisa-Maree L; Sly, Peter D; Macleod, Matthew; Mueller, Jochen F

2014-01-01

Biomonitoring has become the "gold standard" in assessing chemical exposures, and has an important role in risk assessment. The pooling of biological specimens-combining multiple individual specimens into a single sample-can be used in biomonitoring studies to monitor levels of exposure and identify exposure trends or to identify susceptible populations in a cost-effective manner. Pooled samples provide an estimate of central tendency and may also reveal information about variation within the population. The development of a pooling strategy requires careful consideration of the type and number of samples collected, the number of pools required and the number of specimens to combine per pool in order to maximise the type and robustness of the data. Creative pooling strategies can be used to explore exposure-outcome associations, and extrapolation from other larger studies can be useful in identifying elevated exposures in specific individuals. The use of pooled specimens is advantageous as it saves significantly on analytical costs, may reduce the time and resources required for recruitment and, in certain circumstances, allows quantification of samples approaching the limit of detection. In addition, the use of pooled samples can provide population estimates while avoiding ethical difficulties that may be associated with reporting individual results.

Reduction of the pool-top radiation level in HANARO

International Nuclear Information System (INIS)

Lee, Choong-Sung; Park, Sang-Jun; Kim, Heonil; Park, Yong-Chul; Choi, Young-San

1999-01-01

HANARO is an open-tank-in-pool type reactor. Pool water is the only shielding to minimize the pool top radiation level. During the power ascension test of HANARO, the measured pool top radiation level was higher than the design value because some of the activation products in the coolant reached the pool surface. In order to suppress this rising coolant, the hot water layer system (HWL) was designed and installed to maintain l.2 meter-deep hot water layer whose temperature is 5degC higher than that of the underneath pool surface. After the installation of the HWL system, however, the radiation level of the pool-top did not satisfy the design value. The operation modes of the hot water layer system and the other systems in the reactor pool, which had an effect on the formation of the hot water layer, were changed to reduce pool-top radiation level. After the above efforts, the temperature and the radioactivity distribution in the pool was measured to confirm whether this system blocked the rising coolant. The radiation level at the pool-top was significantly reduced below one tenth of that before installing the HWL and satisfied the design value. It was also confirmed by calculation that this hot water layer system would significantly reduce the release of fission gases to the reactor hall and the environment during the hypothetical accident as well. (author)

Declassification of radioactive water from a pool type reactor after nuclear facility dismantling

Science.gov (United States)

Arnal, J. M.; Sancho, M.; García-Fayos, B.; Verdú, G.; Serrano, C.; Ruiz-Martínez, J. T.

2017-09-01

This work is aimed to the treatment of the radioactive water from a dismantled nuclear facility with an experimental pool type reactor. The main objective of the treatment is to declassify the maximum volume of water and thus decrease the volume of radioactive liquid waste to be managed. In a preliminary stage, simulation of treatment by the combination of reverse osmosis (RO) and evaporation have been performed. Predicted results showed that the combination of membrane and evaporation technologies would result in a volume reduction factor higher than 600. The estimated time to complete the treatment was around 650 h (25-30 days). For different economical and organizational reasons which are explained in this paper, the final treatment of the real waste had to be reduced and only evaporation was applied. The volume reduction factor achieved in the real treatment was around 170, and the time spent for treatment was 194 days.

Research reactor fuel transport in the U.K

Energy Technology Data Exchange (ETDEWEB)

Panter, R [U.K. Atomic Energy Authority, Harwell (United Kingdom)

1983-09-01

This paper describes the containers currently used for transport of fresh or spent fuel elements for Research and Materials Test Reactors in the U.K., their status, operating procedures and some of the practical difficulties. In the U.K., MTR fuel cycle work is almost entirely the responsibility of the U.K. Atomic Energy Authority.

Inter-progenitor pool wiring: An evolutionarily conserved strategy that expands neural circuit diversity.

Science.gov (United States)

Suzuki, Takumi; Sato, Makoto

2017-11-15

Diversification of neuronal types is key to establishing functional variations in neural circuits. The first critical step to generate neuronal diversity is to organize the compartmental domains of developing brains into spatially distinct neural progenitor pools. Neural progenitors in each pool then generate a unique set of diverse neurons through specific spatiotemporal specification processes. In this review article, we focus on an additional mechanism, 'inter-progenitor pool wiring', that further expands the diversity of neural circuits. After diverse types of neurons are generated in one progenitor pool, a fraction of these neurons start migrating toward a remote brain region containing neurons that originate from another progenitor pool. Finally, neurons of different origins are intermingled and eventually form complex but precise neural circuits. The developing cerebral cortex of mammalian brains is one of the best examples of inter-progenitor pool wiring. However, Drosophila visual system development has revealed similar mechanisms in invertebrate brains, suggesting that inter-progenitor pool wiring is an evolutionarily conserved strategy that expands neural circuit diversity. Here, we will discuss how inter-progenitor pool wiring is accomplished in mammalian and fly brain systems. Copyright © 2017 Elsevier Inc. All rights reserved.

Development of a transport cask for spent fuel elements of research reactors

International Nuclear Information System (INIS)

Quintana, F.; Saliba, R.O.; Furnari, J.C.; Mourao, R.P; Leite da Silva, L.; Novara, O.; Alexandre Miranda, C.; Mattar Neto, M.

2012-01-01

This article presents an overview of the development of a research reactor spent fuel transport cask. Through a project funded by the IAEA, Argentina, Brazil and Chile have collaborated to enhance regional capacity in the management of spent fuel elements from research reactors operated in the region. A packaging for the transport of research reactors spent fuel was developed. It was designed by a team of researchers from the countries mentioned and a 1:2 scale model for MTR type fuel was constructed in Argentina and subsequently tested in CDTN facilities in Belo Horizonte, Brazil. There were three test sequences to test the cask for normal transport and hypothetical accident conditions. It has successfully passed the tests and the overall performance was considered satisfactory. As part of the licensing process, a test sequence with the presence of regulatory authorities is scheduled for December, 2012 (author)

Successful completion of a time sensitive MTR and TRIGA Indonesian shipment

International Nuclear Information System (INIS)

Anne, Catherine; Patterson, John; Messick, Chuck

2005-01-01

Early this year, a shipment of 109 MTR fuel assemblies was received at the Department of Energy's Savannah River Site from the BATAN reactor in Serpong, Indonesia and another of 181 TRIGA fuel assemblies was received at the Idaho National Laboratory from the two BATAN Indonesian TRIGA reactors in Bandung and Yogyakarta, Indonesia. These were the first Other-Than- High-Income Countries shipments under the FRR program since the Spring 2001. The Global Threat Reduction Initiative announced by Secretary Abraham will require expeditious scheduling and extreme sensitivity to shipment security. The subject shipments demonstrated exceptional performance in both respects. Indonesian terrorist acts and 9/11 impacted the security requirements for the spent nuclear fuel shipments. Internal Indonesian security issues and an upcoming Indonesian election led to a request to perform the shipment with a very short schedule. Preliminary site assessments were performed in November 2003. The DOE awarded a task order to NAC for shipment performance just before Christmas 2003. The casks departed the US in January and the fuel elements were delivered at the DOE sites by the end of April 2004. The paper will present how the team completed a successful shipment in a timely manner. (author)

Analysis of sodium pool fire in SFEF for assessing the limiting pool fire

International Nuclear Information System (INIS)

Mangarjuna Rao, P.; Ramesh, S.S.; Nashine, B.K.; Kasinathan, N.; Chellapandi, P.

2011-01-01

Accidental sodium leaks and resultant sodium fires in Liquid Metal Fast Breeder Reactor (LMFBR) systems can create a threat to the safe operation of the plant. To avoid this defence-in depth approach is implemented from the design stage of reactor itself. Rapid detection of sodium leak and fast dumping of the sodium into the storage tank of a defective circuit, leak collection trays, adequate lining of load bearing structural concrete and extinguishment of the sodium fire are the important defensive measures in the design, construction and operation of a LMFBR for protection against sodium leaks and their resultant fires. Evaluation of sodium leak events and their consequences by conducting large scale engineering experiments is very essential for effective implementation of the above protection measures for sodium fire safety. For this purpose a Sodium Fire Experimental Facility (SFEF) is constructed at SED, IGCAR. SFEF is having an experimental hall of size 9 m x 6 m x 10 m with 540 m 3 volume and its design pressure is 50 kPa. It is a concrete structure and provided with SS 304 liner, which is fixed to the inside surfaces of walls, ceiling and floor. A leak tight door of size (1.8 m x 2.0 m) is provided to the experimental hall and the facility is provided with a sodium equipment hall and a control room. Experimental evaluation of sodium pool fire consequences is an important activity in the LMFBR sodium fire safety related studies. An experimental program has been planned for different types of sodium fire studies in SFEF. A prior to that numerical analysis have been carried out for enclosed sodium pool fires using SOFIRE-II sodium pool fire code for SFEF experimental hall configuration to evaluate the limiting pool fire. This paper brings out results of the analysis carried out for this purpose. Limiting pool fire of SFEF depends on the exposed surface area of the pool, amount of sodium in the pool, oxygen concentration and initial sodium temperature. Limiting

French experience in design, operation and revamping of nuclear research reactors, in support of advanced reactors development

International Nuclear Information System (INIS)

Barre, B.; Bergeonneau, P.; Merchie, F.; Minguet, J.L.; Rousselle, P.

1996-01-01

The French nuclear program is strongly based on the R and D work performed in the CEA nuclear research centers and particularly on the various experimental programs carried out in its research reactors in the frame of cooperative actions between the Commissariat a l'Energie Atomique (CEA), Framatome and Electricite de France (EDF). Several types of research reactors have been built by Technicatome and CEA to carry out successfully this considerable R and D work on fuels and materials, among them the socalled Materials Testing Reactors (MTR) SILOE (35 MW) and OSIRIS (70 MW) which are indeed very well suited for technological irradiations. Their simple and flexible design and the large irradiation space available around the core, the SILOE and OSIRIS reactors can be shared by several types of applications such as fuel and material testings for nuclear power plants, radioisotopes production, silicon doping and fundamental research. It is worthwhile recalling that Technicatome and CEA have also built research reactors fully dedicated to safety experimental studies, such as the CABRI, SCARABEE and PHEBUS reactors at Cadarache, and others dedicated to fundamental research such as ORPHEE (14 MW) and the Reacteur a Haut Flux -High Flux Reactor- (RHF 57 MW). This paper will present some of the most significant conceptual and design features of all these reactors as well as the main improvements brought to most of them in the last years. Based on this wide experience, CEA and Technicatome have specially designed for export a new multipurpose research reactor named SIRIUS, with two versions depending on the utilization spectrum and the power range (5 MW to 30 MW). At last, CEA has recently launched the preliminary project study of a new MTR, the Jules Horowitz Reactor, to meet the future needs of fuels and materials irradiations in the next 4 or 5 decades, in support of the French long term nuclear power program. (J.P.N.)

Laser surveillance systems for fuel storage pools

International Nuclear Information System (INIS)

Boeck, H.

1985-06-01

A Laser Surveillance System (LASSY) as a new safeguards device has been developed under the IAEA research contract No. 3458/RB at the Atominstitut Wien using earlier results by S. Fiarman. This system is designed to act as a sheet of light covering spent fuel assemblies in spent fuel storage pools. When movement of assemblies takes place, LASSY detects and locates the position of the movement in the pool and when interrogated, presents a list of pool positions and times of movement to the safeguards inspector. A complete prototype system was developed and built. Full scale tests showed the principal working capabilities of a LASSY underwater

Laboratory investigation and simulation of breakthrough curves in karst conduits with pools

Science.gov (United States)

Zhao, Xiaoer; Chang, Yong; Wu, Jichun; Peng, Fu

2017-12-01

A series of laboratory experiments are performed under various hydrological conditions to analyze the effect of pools in pipes on breakthrough curves (BTCs). The BTCs are generated after instantaneous injections of NaCl tracer solution. In order to test the feasibility of reproducing the BTCs and obtain transport parameters, three modeling approaches have been applied: the equilibrium model, the linear graphical method and the two-region nonequilibrium model. The investigation results show that pools induce tailing of the BTCs, and the shapes of BTCs depend on pool geometries and hydrological conditions. The simulations reveal that the two-region nonequilibrium model yields the best fits to experimental BTCs because the model can describe the transient storage in pools by the partition coefficient and the mass transfer coefficient. The model parameters indicate that pools produce high dispersion. The increased tailing occurs mainly because the partition coefficient decreases, as the number of pools increases. When comparing the tracer BTCs obtained using the two types of pools with the same size, the more appreciable BTC tails that occur for symmetrical pools likely result mainly from the less intense exchange between the water in the pools and the water in the pipe, because the partition coefficients for the two types of pools are virtually identical. Dispersivity values decrease as flow rates increase; however, the trend in dispersion is not clear. The reduced tailing is attributed to a decrease in immobile water with increasing flow rate. It provides evidence for hydrodynamically controlled tailing effects.

ITHNA.SYS: An Integrated Thermal Hydraulic and Neutronic Analyzer SYStem for NUR research reactor

Energy Technology Data Exchange (ETDEWEB)

Mazidi, S., E-mail: samirmazidi@gmail.com [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Meftah, B., E-mail: b_meftah@yahoo.com [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Belgaid, M., E-mail: belgaidm@yahoo.com [Faculté de Physique, Université Houari Boumediene, USTHB, BP 31, Bab Ezzouar, Alger (Algeria); Letaim, F., E-mail: fletaim@yahoo.fr [Faculté des Sciences et Technologies, Université d’El-oued, PO Box 789, El-oued (Algeria); Halilou, A., E-mail: hal_rane@yahoo.fr [Division Réacteur NUR, Centre de Recherche Nucléaire de Draria, BP 43 Sebala, Draria, Alger (Algeria)

2015-08-15

Highlights: • We develop a neutronic and thermal hydraulic MTR reactor analyzer. • The analyzer allows a rapid determination of the reactor core parameters. • Some NUR reactor parameters have been analyzed. - Abstract: This paper introduces the Integrated Thermal Hydraulic and Neutronic Analyzer SYStem (ITHNA.SYS) that has been developed for the Algerian research reactor NUR. It is used both as an operating aid tool and as a core physics engineering analysis tool. The system embeds three modules of the MTR-PC software package developed by INVAP SE: the cell calculation code WIMSD, the core calculation code CITVAP and the program TERMIC for thermal hydraulic analysis of a material testing reactor (MTR) core in forced convection. ITHNA.SYS operates both in on-line and off-line modes. In the on-line mode, the system is linked, via the computer parallel port, to the data acquisition console of the reactor control room and allows a real time monitoring of major physical and safety parameters of the NUR core. PC-based ITHNA.SYS provides a viable and convenient way of using an accumulated and often complex reactor physics stock of knowledge and frees the user from the intricacy of adequate reactor core modeling. This guaranties an accurate, though rapid, determination of a variety of neutronic and thermal hydraulic parameters of importance for the operation and safety analysis of the NUR research reactor. Instead of the several hours usually required, the processing time for the determination of such parameters is now reduced to few seconds. Validation of the system was performed with respect to experimental measurements and to calculations using reference codes. ITHNA.SYS can be easily adapted to accommodate other kinds of MTR reactors.

A report on the transport of MTR-type spent fuel assemblies of the Philippine Research Reactor (PRR-1)

International Nuclear Information System (INIS)

Yoshisaki, Magno B.; Leopando, Leonardo S.

1999-03-01

Fifty one (51) fuel assemblies of mixed enrichment from the Philippine Research Reactor (PRR-1), consisting of 50 spent and 1 fresh, were shipped to the United States last 14 March 1999 under the U.S. Return of Foreign Research Reactor (FRR) fuel policy. The shipment was in line with the U.S. initiative to implement its Record of Decision (ROD) which took effect on 13 May 1996 to accept and manage all FRR uranium fuel of U.S. origin and enriched in the United States. The shipment program would last10 years, ending midnight of 13 May 2006. The ROD provided a 3 year extension period within which to accept FRR spent nuclear fuel (SNF) withdrawn from reactors after 2006. The U.S. policy gave priority to the NPT significance of high enriched U, as the prime target of the return of FRR policy. Classified as a developing country, the Philippines, through the PNRI, signed a contract with the U.S. Department of Energy for the cost-free shipment of PRR-1 spent fuel to the United States. Spent fuel loading and transport operations to the port area lasted seven (7) days, from 8 to 14 March 1999. (Author)

Tax credits and purchasing pools: will this marriage work?

Science.gov (United States)

Trude, S; Ginsburg, P B

2001-04-01

Bipartisan interest is growing in Congress for using federal tax credits to help low-income families buy health insurance. Regardless of the approach taken, tax credit policies must address risk selection issues to ensure coverage for the chronically ill. Proposals that link tax credits to purchasing pools would avoid risk selection by grouping risks similar to the way large employers do. Voluntary purchasing pools have had only limited success, however. This Issue Brief discusses linking tax credits to purchasing pools. It uses information from the Center for Studying Health System Change's (HSC) site visits to 12 communities as well as other research to assess the role of purchasing pools nationwide and the key issues and implications of linking tax credits and pools.

What the difference to use LEU and HEU fuel elements separately or together in a research reactor

International Nuclear Information System (INIS)

Kaya, S.; Uestuen, G.

2005-01-01

Concerning of nuclear material safety, most of the research reactors are advised to shift from HEU (high enriched-%93 U-235) to LEU (low enriched-%20 U-235) fuel elements. When LEU and HEU fuel elements are to be used together in a research reactor, some design and safety problems are encountered. According to use of the reactor, some research reactors such as MTR type may not show any considerable difference for HEU or LEU fuel elements, but the efficiency of radioisotope production generated by thermal neutron interaction may decrease about twenty-thirty percent when LEU fuel elements are used. Here, fine mesh-sized 3D neutronic analysis of TR-2 research reactor is presented to indicate the arising problem when LEU end HEU fuel elements are used together in a research reactor. Partial thermohydraulic analysis of the reactor is also given to show the betterness of the LEU fuel element design. However, there might be some points that should be noticed for safer operation of plate type fuelled research reactors. (author)

Laser spectroscopic analysis in atmospheric pollution research

CSIR Research Space (South Africa)

Forbes, PBC

2008-01-01

Full Text Available stream_source_info ForbesP_2008.pdf.txt stream_content_type text/plain stream_size 3174 Content-Encoding ISO-8859-1 stream_name ForbesP_2008.pdf.txt Content-Type text/plain; charset=ISO-8859-1 Laser spectroscopic... Department and a CSIR National Laser Centre rental pool programme grant-holder, is involved in research into a novel method of monitoring atmospheric PAHs. The rental pool programme gives South African tertiary education institutions access to an array...

Environmental controls of C, N and P biogeochemistry in peatland pools.

Science.gov (United States)

Arsenault, Julien; Talbot, Julie; Moore, Tim R

2018-08-01

Pools are common in northern peatlands but studies have seldom focused on their nutrient biogeochemistry, especially in relation to their morphological characteristics and through seasons. We determined the environmental characteristics controlling carbon (C), nitrogen (N) and phosphorus (P) biogeochemistry in pools and assessed their evolution over the course of the 2016 growing season in a subboreal ombrotrophic peatland of eastern Canada. We showed that water chemistry variations in 62 pools were significantly explained by depth (81.9%) and the surrounding vegetation type (14.8%), but not by pool area or shape. Shallow pools had larger dissolved organic carbon (DOC) and total nitrogen (TN) concentrations and lower pH than deep pools, while pools surrounded by coniferous trees had more recalcitrant DOC than pools where vegetation was dominated by mosses. The influence of depth on pool biogeochemistry was confirmed by the seasonal survey of pools of different sizes with 47.1% of the variation in pool water chemistry over time significantly explained. Of this, 67.3% was explained by the interaction between time and pool size and 32.7% by pool size alone. P concentrations were small in all pools all summer long and combined with high N:P ratios, are indicative of P-limitation. Our results show that pool biogeochemistry is influenced by internal processes and highlight the spatial and temporal heterogeneity of nutrient biogeochemistry in ombrotrophic peatlands. Copyright © 2018 Elsevier B.V. All rights reserved.

Controls on the size and occurrence of pools in coarse-grained forest rivers

Science.gov (United States)

John M. Buffington; Thomas E. Lisle; Richard D. Woodsmith; Sue Hilton

2002-01-01

Controls on pool formation are examined in gravel- and cobble-bed rivers in forest mountain drainage basins of northern California, southern Oregon, and southeastern Alaska. We demonstrate that the majority of pools at our study sites are formed by flow obstructions and that pool geometry and frequency largely depend on obstruction characteristics (size, type, and...

Improvements in the model of neutron calculations for research reactors

International Nuclear Information System (INIS)

Calzetta, Osvaldo; Leszczynski, Francisco

1987-01-01

Within the research program in the field of neutron physics calculations being carried out in the Nuclear Engineering Division at the Centro Atomico Bariloche, the errors which due to some typical approximations appear in the final results are researched. For research MTR type reactors, two approximations, for high and low enrichment are investigated: the treatment of the geometry and the method of few-group cell cross-sections calculation, particularly in the resonance energy region. Commonly, the cell constants used for the entire reactor calculation are obtained making an homogenization of the full fuel elements, by one-dimensional calculations. An improvement is made that explicitly includes the fuel element frames in the core calculation geometry. Besides, a detailed treatment-in energy and space- is used to find the resonance few-group cross sections, and a comparison of the results with detailed and approximated calculations is made. The least number and the best mesh of energy groups needed for cell calculations is fixed too. (Author) [es

Convection in molten pool created by a concentrated energy flux on a solid metal target

International Nuclear Information System (INIS)

Dikshit, B.; Zende, G. R.; Bhatia, M. S.; Suri, B. M.

2009-01-01

During surface evaporation of metals by use of a concentrated energy flux such as electron beam or lasers, a liquid metal pool having a very high temperature gradient is formed around the hot zone created by the beam. Due to temperature dependence of surface tension, density, and depression of the evaporating surface caused by back pressure of the emitted vapor in this molten pool, a strong convective current sets in the molten pool. A proposition is made that this convection may pass through three different stages during increase in the electron beam power depending upon dominance of the various driving forces. To confirm this, convective heat transfer is quantified in terms of dimensionless Nusselt number and its evolution with power is studied in an experiment using aluminum, copper, and zirconium as targets. These experimentally determined values are also compared to the theoretical values predicted by earlier researchers to test the validity of their assumptions and to know about the type of flow in the melt pool. Thus, conclusion about the physical characteristics of flow in the molten pool of metals could be drawn by considering the roles of surface tension and curvature of the evaporating surface on the evolution of convective heat transfer.

Non-destructive evaluation methods to improve quality control in low enrichment MTR fuel plate production

International Nuclear Information System (INIS)

Milne, J.M.; Lidington, B.; Hawker, B.M.

1991-01-01

This paper summarises some preliminary non-destructive measurements made recently at the Harwell Laboratory on a prototype low enrichment MTR fuel plate. The measurements were intended to indicate the potential of two different techniques for improving quality control in plate production. Pulse Video Thermography (PVT) is being considered as an alternative to ultrasound transmission measurements for the detection and sizing of lack of thermal bonding between the fuel and the clad layers, either to verify the indications from the established ultrasonic methods before destroying the plate or as a replacement method of inspection. High frequency pulse-echo ultrasonics is being considered for providing maps of clad layer thickness on each side of the plate. The measurements have indicated the potential for both methods, but more work is required, using a test plate containing controlled defects, to establish their capability. (orig.)

Models for Pooled Time-Series Cross-Section Data

Directory of Open Access Journals (Sweden)

Lawrence E Raffalovich

2015-07-01

Full Text Available Several models are available for the analysis of pooled time-series cross-section (TSCS data, defined as “repeated observations on fixed units” (Beck and Katz 1995. In this paper, we run the following models: (1 a completely pooled model, (2 fixed effects models, and (3 multi-level/hierarchical linear models. To illustrate these models, we use a Generalized Least Squares (GLS estimator with cross-section weights and panel-corrected standard errors (with EViews 8 on the cross-national homicide trends data of forty countries from 1950 to 2005, which we source from published research (Messner et al. 2011. We describe and discuss the similarities and differences between the models, and what information each can contribute to help answer substantive research questions. We conclude with a discussion of how the models we present may help to mitigate validity threats inherent in pooled time-series cross-section data analysis.

Trophic interactions among the heterotrophic components of plankton in man-made peat pools

Directory of Open Access Journals (Sweden)

Michał Niedźwiecki

2017-03-01

Full Text Available Man-made peat pools are permanent freshwater habitats developed due to non-commercial man-made peat extraction. Yet, they have not been widely surveyed in terms of ecosystem functioning, mainly regarding the complexity of heterotrophic components of the plankton. In this study we analysed distribution and trophic interrelations among heterotrophic plankton in man-made peat pools located in different types of peatbogs. We found that peat pools showed extreme differences in environmental conditions that occurred to be important drivers of distribution of microplankton and metazooplankton. Abundance of bacteria and protozoa showed significant differences, whereas metazooplankton was less differentiated in density among peat pools. In all peat pools stress-tolerant species of protozoa and metazoa were dominant. In each peat pool five trophic functional groups were distinguished. The abundance of lower functional trophic groups (bacteria, heterotrophic nanoflagellates (HNF and ciliates feeding on bacteria and HNF was weakly influenced by environmental drivers and was highly stable in all peat pool types. Higher functional trophic groups (naupli, omnivorous and carnivorous ciliates, cladocerans, adult copepods and copepodites were strongly influenced by environmental variables and exhibited lower stability. Our study contributes to comprehensive knowledge of the functioning of peat bogs, as our results have shown that peat pools are characterized by high stability of the lowest trophic levels, which can be crucial for energy transfer and carbon flux through food webs.

Exciting Pools

Science.gov (United States)

Wright, Bradford L.

1975-01-01

Advocates the creation of swimming pool oscillations as part of a general investigation of mechanical oscillations. Presents the equations, procedure for deriving the slosh modes, and methods of period estimation for exciting swimming pool oscillations. (GS)

Radiation shielding considerations for the repair and maintenance of a swimming pool-type tokamak reactor

International Nuclear Information System (INIS)

Seki, Y.; Mori, S.

1984-01-01

The radiation shielding relevant to the repair and maintenance of a swimming pool-type tokamak reactor is considered. The dose rate during the reactor operation can be made low enough for personnel access into the reactor room if a 2m thick water layer is installed above the magnet cryostat. The dose rate 24 h after shutdown is such that the human access is allowed above the magnet cryostat. Sufficient water layer thickness is provided in the inboard space for the operation of automatic welder/cutter while retaining the magnet shielding capability. Some forced cooling is required for the decay heat removal in the first wall. The penetration shield thickness around the neutral beam injector port is estimated to be barely sufficient in terms of the magnet radiation damage. (orig.)

Mathematical modeling of the energy consumption of heated swimming pools

Energy Technology Data Exchange (ETDEWEB)

Le Bel, C.; Millette, J. [LTE Shawinigan, Shawinigan, PQ (Canada)

2007-07-01

A mathematical model was developed to estimate the water temperature of a residential swimming pool. The model can compare 2 different situations and, if local climatic conditions are known, it can accurately predict energy costs of the pool relative to the total energy consumption of the house. When used with the appropriate energy transfer coefficient and weather file, the model can estimate the water temperature of a residential swimming pool having specific characteristics, such as in-ground, above-ground, heated or non-heated. The model is suitable for determining residential loads. It can be applied to different pool types and sizes, for different water heating scenarios and different climatic regions. Data obtained from the monitoring of water temperature and electricity use of 57 residential swimming pools was used to validate the model. In addition, 5 above-ground pools were installed on the property of LTE Shawinigan to allow for a more detailed study of the parameters involved in the thermal balance of a pool. The mathematical model, based on a global heat transfer coefficient, can determine the effect of a solar blanket and the effect of water volume. 14 refs., 5 tabs., 11 figs.

Monte Carlo simulation of irradiation of MTR fuel plates in the BR2 reactor using a full-scale 3-d model with inclined channels

International Nuclear Information System (INIS)

Kuzminov, V. V; Koonen, E.; Ponsard, B.

2002-01-01

A three-dimensional full-scale Monte Carlo model of the BR2 reactor has been developed for simulation of irradiation conditions of materials and fuel loaded in various irradiation devices. This new reactor model includes a detailed geometrical description of the inclined reactor channels, the irradiation devices loaded in these channels including the materials to be tested/loaded in these devices, the burn-up of the BR2 fuel elements and the poisoning of the beryllium matrix. Recently a benchmark irradiation of new irradiation device for testing and qualification of MTR fuel plates has been performed. For this purpose the detailed irradiation conditions of fuel plates had to be predetermined. Monte Carlo calculations of neutron fluxes and heat load distributions in irradiated MTR fuel plates were performed taking into account the contents of all loaded experimental devices in the reactor channels. A comparison of the calculated and measured values of neutron fluxes and of heat loads in the BR2 reactor is presented in this paper. The comparison is part of the validation process of the new reactor model. It also serves to establish the capability to conduct a fuel plate irradiation program under requested and well- known irradiation conditions. (author)

Cryptosporidium and Giardia in Swimming Pools, Atlanta, Georgia

Centers for Disease Control (CDC) Podcasts

In this podcast, Dan Rutz speaks with Dr. Joan Shields, a guest researcher with the Healthy Swimming Program at CDC, about an article in June 2008 issue of Emerging Infectious Diseases reporting on the results of a test of swimming pools in the greater Atlanta, Georgia area. Dr. Shields tested 160 pools in metro Atlanta last year for Cryptosporidium and Giardia. These germs cause most recreational water associated outbreaks.

Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

International Nuclear Information System (INIS)

Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

2013-01-01

Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

System seismic analysis of an innovative primary system for a large pool type LMFBR plant

International Nuclear Information System (INIS)

Pan, Y.C.; Wu, T.S.; Cha, B.K.; Burelbach, J.; Seidensticker, R.

1984-01-01

The system seismic analysis of an innovative primary system for a large pool type liquid metal fast breeder reactor (LMFBR) plant is presented. In this primary system, the reactor core is supported in a way which differs significantly from that used in previous designs. The analytical model developed for this study is a three-dimensional finite element model including one-half of the primary system cut along the plane of symmetry. The model includes the deck and deck mounted components,the reactor vessel, the core support structure, the core barrel, the radial neutron shield, the redan, and the conical support skirt. The sodium contained in the primary system is treated as a lumped mass appropriately distributed among various components. The significant seismic behavior as well as the advantages of this primary system design are discussed in detail

Pressure suppression pool hydrodynamic studies for horizontal vent exit of Indian PHWR containment

International Nuclear Information System (INIS)

Mohan, N.; Bajaj, S.S.; Saha, P.

1994-01-01

The standard Indian PHWR incorporates a pressure suppression type of containment system with a suppression pool.The design of KAPS (Kakrapar Atomic Power Station) suppression pool system adopts a modified system of downcomers having horizontal vents as compared to vertical vents of NAPS (Narora Atomic Power Station). Hydrodynamic studies for vertical vents have been reported earlier. This paper presents hydrodynamic studies for horizontal type vent system during LOCA. These studies include the phenomenon of vent clearing (where the water slug standing in downcomer initially is injected to wetwell due to rapid pressurization of drywell) followed by pool swell (elevation of pool water due to formation of bubbles due to air mass entering pool at the exit of horizontal vents from drywell). The analysis performed for vent clearing and pool swell is based on rigorous thermal hydraulic calculation consisting of conservation of air-steam mixture mass, momentum and thermal energy and mass of air. Horizontal vent of downcomer is modelled in such a way that during steam-air flow, variation of flow area due to oscillating water surface in downcomer could be considered. Calculation predicts that the vent gets cleared in about 1.0 second and the corresponding downward slug velocity in the downcomer is 4.61 m/sec. The maximum pool swell for a conservative lateral expansion is calculated to be 0.56 m. (author). 3 refs., 12 figs

Sustainability of common pool resources.

Science.gov (United States)

Timilsina, Raja Rajendra; Kotani, Koji; Kamijo, Yoshio

2017-01-01

Sustainability has become a key issue in managing natural resources together with growing concerns for capitalism, environmental and resource problems. We hypothesize that the ongoing modernization of competitive societies, which we refer to as "capitalism," affects human nature for utilizing common pool resources, thus compromising sustainability. To test this hypothesis, we design and implement a set of dynamic common pool resource games and experiments in the following two types of Nepalese areas: (i) rural (non-capitalistic) and (ii) urban (capitalistic) areas. We find that a proportion of prosocial individuals in urban areas is lower than that in rural areas, and urban residents deplete resources more quickly than rural residents. The composition of proself and prosocial individuals in a group and the degree of capitalism are crucial in that an increase in prosocial members in a group and the rural dummy positively affect resource sustainability by 65% and 63%, respectively. Overall, this paper shows that when societies move toward more capitalistic environments, the sustainability of common pool resources tends to decrease with the changes in individual preferences, social norms, customs and views to others through human interactions. This result implies that individuals may be losing their coordination abilities for social dilemmas of resource sustainability in capitalistic societies.

The role of risk management in decrease of lawsuits of swimming pools

Directory of Open Access Journals (Sweden)

Behzad Izadi

2012-01-01

Full Text Available The purpose of this research is to study of risk management practices in decrease of lawsuits in public and private swimming pools in Tehran. The statistical population of the research included 310 managers of public and private swimming pools which 119 were selected as statistical samples by means of random sampling. The research method was descriptive and survey, and in measurement form. 2 questionnaires were used, on relating to demographic data and general information and the other to risk management practices and their validity was determined by alpha Cronbach method. The required information was collected by personal interviews during the time acting of managers in pools gathered and the data was analyzed by using person correlation coefficient. The result of this study indicated that: Significant relationship existed between incidents of accidents/injuries and lawsuits in swimming pools in Tehran. Significant relationship existed between risk management practice and accidents/injuries and lawsuits. Significant relationship existed between risk management practice and lawsuits and lawsuits.

Research reactors - an overview

International Nuclear Information System (INIS)

West, C.D.

1997-01-01

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

PIV measurements of turbulent jet and pool mixing produced by a steam jet discharge in a subcooled water pool

International Nuclear Information System (INIS)

Choo, Yeon Jun; Song, Chul-Hwa

2010-01-01

This experimental research is on the fluid-dynamic features produced by a steam injection into a subcooled water pool. The relevant phenomena could often be encountered in water cooled nuclear power plants. Two major topics, a turbulent jet and the internal circulation produced by a steam injection, were investigated separately using a particle image velocimetry (PIV) as a non-intrusive optical measurement technique. Physical domains of both experiments have a two-dimensional axi-symmetric geometry of which the boundary and initial conditions can be readily and well defined. The turbulent jet experiments with the upward discharging configuration provide the parametric values for quantitatively describing a turbulent jet such as the self-similar velocity profile, central velocity decay, spreading rate, etc. And in the internal circulation experiments with the downward discharging configuration, typical flow patterns in a whole pool region are measured in detail, which reveals both the local and macroscopic characteristics of the mixing behavior in a pool. This quantitative data on the condensing jet-induced mixing behavior in a pool could be utilized as benchmarking for a CFD simulation of relevant phenomena.

Spent fuel storage pool

International Nuclear Information System (INIS)

Murakami, Naoshi.

1996-01-01

Fences are disposed to a fuel exchange floor surrounding the upper surface of a fuel pool for preventing overflow of pool water. The fences comprise a plurality of flat boards arranged in parallel with each other in the longitudinal direction while being vertically inclined, and slits are disposed between the boards for looking down the pool. Further, the fences comprise wide boards and are constituted so as to be laid horizontally on the fuel exchange floor in a normal state and uprisen by means of the signals from an earthquake sensing device. Even if pool water is overflow from the fuel pool by the vibrations occurred upon earthquake and flown out to the floor of the fuel exchange floor, the overflow from the fuel exchange floor is prevented by the fences. An operator who monitors the fuel pool can observe the inside of the fuel pool through the slits formed to the fences during normal operation. The fences act as resistance against overflowing water upon occurrence of an earthquake thereby capable of reducing the overflowing amount of water due to the vibrations of pool water. The effect of preventing overflowing water can be enhanced. (N.H.)

Economical analysis to utilize MTR fuel elements using silicides in research reactors; Analisis economico sobre el uso de elementos combustibles MTR a base de siliciuros en reactores de investigacion

Energy Technology Data Exchange (ETDEWEB)

Bergallo, Juan E; Novara, Oscar E; Adelfang, Pablo [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Combustibles Nucleares

2000-07-01

According to international programs on reducing enrichment in research reactors and the necessity to maintain their operation, new fuel elements have been developed in order to meet both objectives. Thus, U-Si alloy fuel elements for research reactors are becoming of greater interest for the international markets. It became necessary to make an economic study about the convenience of introducing this type of fuel elements in the RA-3 reactor and to know the potentiality of this fuel. The economical behavior of the reactor operation has been evaluated comparing the actual U{sub 3}O{sub 8} nuclear fuel cycle with U{sub 3}Si{sub 2} nuclear fuels. Results obtained show that the main economical factor to determine the change of fuels is the cost of fabrication, and the change is advisable up to an 80% difference. The other factors related to the cost of nuclear fuel cycle are not relevant or have real minor impacts. (author)

GABA and glutamate levels correlate with MTR and clinical disability: Insights from multiple sclerosis.

Science.gov (United States)

Nantes, Julia C; Proulx, Sébastien; Zhong, Jidan; Holmes, Scott A; Narayanan, Sridar; Brown, Robert A; Hoge, Richard D; Koski, Lisa

2017-08-15

Converging areas of research have implicated glutamate and γ-aminobutyric acid (GABA) as key players in neuronal signalling and other central functions. Further research is needed, however, to identify microstructural and behavioral links to regional variability in levels of these neurometabolites, particularly in the presence of demyelinating disease. Thus, we sought to investigate the extent to which regional glutamate and GABA levels are related to a neuroimaging marker of microstructural damage and to motor and cognitive performance. Twenty-one healthy volunteers and 47 people with multiple sclerosis (all right-handed) participated in this study. Motor and cognitive abilities were assessed with standard tests used in the study of multiple sclerosis. Proton magnetic resonance spectroscopy data were acquired from sensorimotor and parietal regions of the brains' left cerebral hemisphere using a MEGA-PRESS sequence. Our analysis protocol for the spectroscopy data was designed to account for confounding factors that could contaminate the measurement of neurometabolite levels due to disease, such as the macromolecule signal, partial volume effects, and relaxation effects. Glutamate levels in both regions of interest were lower in people with multiple sclerosis. In the sensorimotor (though not the parietal) region, GABA concentration was higher in the multiple sclerosis group compared to controls. Lower magnetization transfer ratio within grey and white matter regions from which spectroscopy data were acquired was linked to neurometabolite levels. When adjusting for age, normalized brain volume, MTR, total N-acetylaspartate level, and glutamate level, significant relationships were found between lower sensorimotor GABA level and worse performance on several tests, including one of upper limb motor function. This work highlights important methodological considerations relevant to analysis of spectroscopy data, particularly in the afflicted human brain. These findings

Modeling of condensation, stratification, and mixing phenomena in a pool of water

Energy Technology Data Exchange (ETDEWEB)

Li, H.; Kudinov, P.; Villanueva, W. (Royal Institute of Technology (KTH). Div. of Nuclear Power Safety, Stockholm (Sweden))

2010-12-15

This work pertains to the research program on Containment Thermal-Hydraulics at KTH. The objective is to evaluate and improve performance of methods, which are used to analyze thermal-hydraulics of steam suppression pools in a BWR plant under different abnormal transient and accident conditions. As a passive safety system, the function of steam pressure suppression pools is paramount to the containment performance. In the present work, the focus is on apparently-benign but intricate and potentially risk-significant scenarios in which thermal stratification could significantly impede the pool's pressure suppression capacity. For the case of small flow rates of steam influx, the steam condenses rapidly in the pool and the hot condensate rises in a narrow plume above the steam injection plane and spreads into a thin layer at the pool's free surface. When the steam flow rate increases significantly, momentum introduced by the steam injection and/or periodic expansion and shrink of large steam bubbles due to direct contact condensation can cause breakdown of the stratified layers and lead to mixing of the pool water. Accurate prediction of the pool thermal-hydraulics in such scenarios presents a computational challenge. Lumped-parameter models have no capability to predict temperature distribution of water pool during thermal stratification development. While high-order-accurate CFD (RANS, LES) methods are not practical due to excessive computing power needed to calculate 3D high-Rayleighnumber natural circulation flow in long transients. In the present work, a middleground approach is used, namely CFD-like model of the general purpose thermalhydraulic code GOTHIC. Each cell of 3D GOTHIC grid uses lumped parameter volume type closures for modeling of various heat and mass transfer processes at subgrid scale. We use GOTHIC to simulate POOLEX/PPOOLEX experiment, in order to (a) quantify errors due to GOTHIC's physical models and numerical schemes, and (b

Phase analysis in gated blood pool tomography

International Nuclear Information System (INIS)

Nakajima, Kenichi; Bunko, Hisashi; Tada, Akira; Taki, Junichi; Nanbu, Ichiro

1984-01-01

Phase analysis of gated blood pool study has been applied to detect the site of accessory conduction pathway (ACP) in the Wolff-Parkinson-White (WPW) syndrome; however, there was a limitation to detect the precise location of ACP by phase analysis alone. In this study, we applied phase analysis to gated blood pool tomography using seven pin hole tomography (7PT) and gated emission computed tomography (GECT) in 21 patients with WPW syndrome and 3 normal subjects. In 17 patients, the sites of ACPs were confirmed by epicardial mapping and the result of the surgical division of ACP. In 7PT, the site of ACP grossly agreed to the abnormal initial phase in phase image in 5 out of 6 patients with left cardiac type. In GECT, phase images were generated in short axial, vertical and horizontal long axial sections. In 8 out of 9 patients, the site of ACP was correctly identified by phase images, and in a patient who had two ACPs, initial phase corresponded to one of the two locations. Phase analysis of gated blood pool tomography has advantages for avoiding overlap of blood pools and for estimating three-dimensional propagation of the contraction, and can be a good adjunctive method in patients with WPW syndrome. (author)

Structural analysis of the reactor pool for the RRRP

International Nuclear Information System (INIS)

Alberro, J.G.; Abbate, A.D.

2005-01-01

The purpose of the present document is to describe the structural design of the Reactor Pool relevant to the RRRP (Replacement Research Reactor Project) for the Australian Nuclear Science and Technology Organisation. The structural analysis required coordinated design, engineering, analysis, and fabrication efforts. The pool has been designed, manufactured, and inspected following as guideline the ASME Boiler and Pressure Vessel Code, which defines the requirements for the pool to withstand hydrostatic and mechanical forces, ensuring its integrity throughout its lifetime. Standard off-the-shelf finite element programs (Nastran and Ansys codes) were used to evaluate the pool and further qualify the design and its construction. Both global and local effect analyses were carried out. The global analysis covers the structural integrity of the pool wall (6 mm thick) considering the different load states acting on it, namely hydrostatic pressure, thermal expansion, and seismic event. The local analysis evaluates the structural behaviour of the pool at specific points resulting from the interaction among components. It is confirmed that maximum stresses and displacements fall below the allowable values required by the ASME Boiler and Pressure Vessel Code. The water pressure analysis was validated by means of a hydrostatic test. (authors)

Concepts for the interim storage of spent fuel elements from research reactors in the Federal Republic of Germany

International Nuclear Information System (INIS)

Niephaus, D.; Bensch, D.; Quaassdorff, P.; Plaetzer, S.

1997-01-01

Research reactors have been operated in the Federal Republic of Germany since the late fifties. These are Material Test Reactors (MTR) and training, Research and Isotope Facilities of General Atomic (TRIGA). A total of seven research reactors, i.e. three TRIGA and four MTR facilities were still in operation at the beginning of 1996. Provisions to apply to the back-end of the fuel cycle are required for their continued operation and for already decommissioned plants. This was ensured until the end of the eighties by the reprocessing of spent fuel elements abroad. In view of impeding uncertainties in connection with waste management through reprocessing abroad, the development of a national back-end fuel cycle concept was commissioned by the Federal Minister of Education, Science, Research and Technology in early 1990. Development work was oriented along the lines of the disposal concept for irradiated light-water reactor fuel elements from nuclear power plants. Analogously, the fuel elements from research reactors are to be interim-stored on a long-term basis in adequately designed transport and storage casks and then be directly finally disposed without reprocessing after up to forty years of interim storage. As a first step in the development of a concept for interim storage, several sites with nuclear infrastructure were examined and assessed with respect to their suitability for interim storage. A reasonably feasible reference concept for storing the research reactor fuel elements in CASTOR MTR 2 transport and storage casks at the Ahaus interim storage facility (BZA) was evaluated and the hot cell facility and AVR store of Forschungszentrum Juelich (KFA) were proposed as an optional contingency concept for casks that cannot be repaired at Ahaus. Development work was continued with detailed studies on these two conceptual variants and the results are presented in this paper. (author)

Can Human Subject Pool Participation Benefit Sociology Students?

Science.gov (United States)

Chin, Lynn Gencianeo; Gibbs Stayte, Patricia

2015-01-01

Instructors at non-research institutions are less able to expose their students to research firsthand. Utilizing human subject pools (HSPs) in class may be a solution. Given that HSPs tend to be used in introduction to psychology classes at research institutions, we examine a community college HSP to answer three questions: (1) Do community…

Association study of folate-related enzymes (MTHFR, MTR, MTRR genetic variants with non-obstructive male infertility in a Polish population

Directory of Open Access Journals (Sweden)

Mateusz Kurzawski

2015-03-01

Full Text Available Spermatogenesis is a process where an important contribution of genes involved in folate-mediated one-carbon metabolism is observed. The aim of the present study was to investigate the association between male infertility and the MTHFR (677C > T; 1298A > C, MTR (2756A > G and MTRR (66A > G polymorphisms in a Polish population. No significant differences in genotype or allele frequencies were detected between the groups of 284 infertile men and of 352 fertile controls. These results demonstrate that common polymorphisms in folate pathway genes are not major risk factors for non-obstructive male infertility in the Polish population.

Analysis of small sample size studies using nonparametric bootstrap test with pooled resampling method.

Science.gov (United States)

Dwivedi, Alok Kumar; Mallawaarachchi, Indika; Alvarado, Luis A

2017-06-30

Experimental studies in biomedical research frequently pose analytical problems related to small sample size. In such studies, there are conflicting findings regarding the choice of parametric and nonparametric analysis, especially with non-normal data. In such instances, some methodologists questioned the validity of parametric tests and suggested nonparametric tests. In contrast, other methodologists found nonparametric tests to be too conservative and less powerful and thus preferred using parametric tests. Some researchers have recommended using a bootstrap test; however, this method also has small sample size limitation. We used a pooled method in nonparametric bootstrap test that may overcome the problem related with small samples in hypothesis testing. The present study compared nonparametric bootstrap test with pooled resampling method corresponding to parametric, nonparametric, and permutation tests through extensive simulations under various conditions and using real data examples. The nonparametric pooled bootstrap t-test provided equal or greater power for comparing two means as compared with unpaired t-test, Welch t-test, Wilcoxon rank sum test, and permutation test while maintaining type I error probability for any conditions except for Cauchy and extreme variable lognormal distributions. In such cases, we suggest using an exact Wilcoxon rank sum test. Nonparametric bootstrap paired t-test also provided better performance than other alternatives. Nonparametric bootstrap test provided benefit over exact Kruskal-Wallis test. We suggest using nonparametric bootstrap test with pooled resampling method for comparing paired or unpaired means and for validating the one way analysis of variance test results for non-normal data in small sample size studies. Copyright © 2017 John Wiley & Sons, Ltd. Copyright © 2017 John Wiley & Sons, Ltd.

Peatland Open-water Pool Biogeochemistry: The Influence of Hydrology and Vegetation

Science.gov (United States)

Arsenault, J.; Talbot, J.; Moore, T. R.

2017-12-01

Peatland open-water pools are net sources of carbon to the atmosphere. However, their interaction with the surrounding peat remains poorly known. In a previous study, we showed that shallow pools are richer in nutrients than deep pools. While depth was the main driver of biogeochemistry variations across time and space, analyses also showed that pool's adjacent vegetation may have an influence on water chemistry. Our goal is to understand the relationship between the biogeochemistry of open-water pools and their surroundings in a subboreal ombrotrophic peatland of southern Quebec (Canada). To assess the influence of vegetation on pool water chemistry, we compare two areas covered with different types of vegetation: a forested zone dominated by spruce trees and an open area mostly covered by Sphagnum spp. To evaluate the direction of water (in or out of the pools), we installed capacitance water level probes in transects linking pools in the two zones. Wells were also installed next to each probe to collect peat pore water samples. Samples were taken every month during summer 2017 and analyzed for dissolved organic carbon, nitrogen and phosphorus, pH and specific UV absorbance. Preliminary results show differences in peat water chemistry depending on the dominant vegetation. In both zones, water levels fluctuations are disconnected between peat and the pools, suggesting poor horizontal water movement. Pool water chemistry may be mostly influenced by the immediate surrounding vegetation than by the local vegetation pattern. Climate and land-use change may affect the vegetation structure of peatlands, thus affecting pool biogeochemistry. Considering the impact of pools on the overall peatland capacity to accumulate carbon, our results show that more focus must be placed on pools to better understand peatland stability over time.

Who Says: "No Fair!"? What Personality and an Experiment in Educational Value Tell Us about Perceptions of Costs and Benefits of Research Pool Requirements

Science.gov (United States)

Cromer, Lisa DeMarni; Reynolds, Shannon M.; Johnson, Mitchell D.

2013-01-01

Human subject pools (HSPs) are the basis for much psychological research. There is an explicit assumption that participants receive benefits from their participation, however there is little empirical research about the costs/benefits of participation. We conducted two studies with undergraduate psychology students to evaluate factors that can…

Evaluation of aluminum pit corrosion in oak ridge research reactor pool by quantitative imaging and thermodynamic modeling

International Nuclear Information System (INIS)

Jang, Ping-Rey; Arunkumar, Rangaswami; Lindner, Jeffrey S.; Long, Zhiling; Mott, Melissa A.; Okhuysen, Walter P.; Monts, David L.; Su, Yi; Kirk, Paula G.; Ettien, John

2007-01-01

The Oak Ridge Research Reactor (ORRR) was operated as an isotope production and irradiation facility from March 1958 until March 1987. The US Department of Energy permanently shut down and removed the fuel from the ORRR in 1987. The water level must be maintained in the ORRR pool as shielding for radioactive components still located in the pool. The U.S. Department of Energy's Office of Environmental Management (DOE EM) needs to decontaminate and demolish the ORRR as part of the Oak Ridge cleanup program. In February 2004, increased pit corrosion was noted in the pool's 6 mm (1/4'')-thick aluminum liner in the section nearest where the radioactive components are stored. If pit corrosion has significantly penetrated the aluminum liner, then DOE EM must accelerate its decontaminating and decommissioning (D and D) efforts or look for alternatives for shielding the irradiated components. The goal of Mississippi State University's Institute for Clean Energy Technology (ICET) was to provide a determination of the extent and depth of corrosion and to conduct thermodynamic modeling to determine how further corrosion can be inhibited. Results from the work will facilitate ORNL in making reliable disposition decisions. ICET's inspection approach was to quantitatively estimate the amount of corrosion by using Fourier - transform profilometry (FTP). FTP is a non-contact 3- D shape measurement technique. By projecting a fringe pattern onto a target surface and observing its deformation due to surface irregularities from a different view angle, the system is capable of determining the height (depth) distribution of the target surface, thus reproducing the profile of the target accurately. ICET has previously demonstrated that its FTP system can quantitatively estimate the volume and depth of removed and residual material to high accuracy. The results of our successful initial deployment of a submergible FTP system into the ORRR pool are reported here as are initial thermodynamic

Operation and maintenance techniques of pool and pool water purification system in IMEF

Energy Technology Data Exchange (ETDEWEB)

Soong, Woong Sup

1999-03-01

IMEF pool is used pass way between pool and hot cell in order to inlet and outlet of fuel pin in cask. All operation is performed conforming with naked eyes. Therefore floating matter is filtered so as to easy under water handling. Also radioactivity in pool water is controlled according to the nuclear law, radioactivity ration maintained less than 15mR/hr on pool side. Perfect operation and maintenance can be achieved well trained operator. Result obtained from the perfection can give more influence over restrain, spreading contamination of radioactivity materials. This report describes operation and maintenance technique of pool water purification system in IMEF. (Author). 7 refs., 13 figs.

Operation and maintenance techniques of pool and pool water purification system in IMEF

International Nuclear Information System (INIS)

Soong, Woong Sup

1999-03-01

IMEF pool is used pass way between pool and hot cell in order to inlet and outlet of fuel pin in cask. All operation is performed conforming with naked eyes. Therefore floating matter is filtered so as to easy under water handling. Also radioactivity in pool water is controlled according to the nuclear law, radioactivity ration maintained less than 15mR/hr on pool side. Perfect operation and maintenance can be achieved well trained operator. Result obtained from the perfection can give more influence over restrain, spreading contamination of radioactivity materials. This report describes operation and maintenance technique of pool water purification system in IMEF. (Author). 7 refs., 13 figs

Pool scrubbing and hydrodynamic experiments on jet injection regime

International Nuclear Information System (INIS)

Peyres, V.; Espigares, M.M.; Polo, J.; Escudero, M.J.; Herranz, L.E.; Lopez-Jimenez, J.

1995-01-01

Plant analyses have shown that pool scrubbing can play an important role in source term during PWR risk dominant sequences. An examination ofboundary conditions governing fission products and aerosols transport through aqueous beds revealed that most of radioactivity is discharged into the pool under jet injection regime. This fact and the lack of experimental data under such conditions pointed the need of setting out an experimental programme which provided reliable experimental data to validate code models. In this report the major results of a pool scrubbing experimental programme carried out in PECA facility are presented. One of the major findings was that a remarkable fraction of particle absorption was not a function of the residence time of bubbles rising through the pool. Such a contribution was assumed to be associated to aerosol removal mechanisms acting at the pool entrance. As a consequence, a hydrodynamic experimental plan was launched to examine the gas behaviour during the initial stages in the pool. Size and shape of gas nuclei the pool were measured and fitted to a long normal distribution. Particularly, size was found to be quite sensitive to inletgas flow and at minor extent to gas composition and pool temperature. SPARC90 and BUSCA-AUG92 were used to simulate the retention tests. Whereas SPARC90 showed a pretty good agreement with experimental data, BUSCA-AUG92 results were far away from measurements in all the cases. SPARC90consistency apparently pointed out the important role of fission products and aerosols retention at the injection zone; nonetheless, a peer examination of pool scrubbing phenomenology at the pool entrance should be carried out to test both hydrodynamic and removal models. Hence, one of the major highlights drawn from this work was the need of further research under representative severe accident conditions (i.e., saturated pools, jet injection regimes, etc.), as well as separate effect tests to validate, improve and

Pool scrubbing and hydrodynamic experiment on jet injection regime

Energy Technology Data Exchange (ETDEWEB)

Peyres, V.; Espigares, M.M.; Polo, J.; Escudero, M.J.; Herranz, L.E.; Lopez, J.

1995-07-01

Plant analyses nave shown that pool scrubbing can play an important role in source term during PWR risk dominant sequences. An examination of boundary conditions governing fission products and aerosols transport through aqueous beds revealed that most of radioactivity is discharged into the pool under jet injection regime. This fact and the lack of experimental data under such conditions pointed the need of setting out an experimental programme which provided reliable experimental data to validate code models. In this report the major results of a pool scrubbing experimental programme carried out in PECA facility are presented. One of the major findings was that a remarkable fraction of particle absorption was not a function of the residence time of bubbles rising through the pool. Such a contribution was assumed to be associated to aerosol removal mechanism acting at the pool entrance. As a consequence. a hydrodynamic experimental plan was launched to examine the gas behaviour during the initial stages in the pool. Size and shape of gas nuclei in the pool were measured and fitted to a lognormal distribution. Particularly, size was found to be quite sensitive to inlet gas flow and at minor extent to gas composition and pool temperature. SPARC90 and BUSCA-AUG92 were used to simulate the retention tests. Whereas SPARC90 showed a pretty good agreement with experimental data, BUSCA-AUG92 results were far away from measurements in all the cases. SPARC90 consistency apparently pointed out the important role of fission products and aerosols retention at the injection zone; nonetheless, a peer examination of pool scrubbing phenomenology at the pool entrance should be carried out to test both hydrodynamic and removal models. Hence, one of the major high lights drawn from this work was the need of further research under representative severe accident conditions (i.e., saturated pools, jet injection regimes, etc.), as well as separate effect tests to validate, improve and

Pool scrubbing and hydrodynamic experiment on jet injection regime

International Nuclear Information System (INIS)

Peyres, V.; Espigares, M.M.; Polo, J.; Escudero, M.J.; Herranz, L.E.; Lopez, J.

1995-01-01

Plant analyses nave shown that pool scrubbing can play an important role in source term during PWR risk dominant sequences. An examination of boundary conditions governing fission products and aerosols transport through aqueous beds revealed that most of radioactivity is discharged into the pool under jet injection regime. This fact and the lack of experimental data under such conditions pointed the need of setting out an experimental programme which provided reliable experimental data to validate code models. In this report the major results of a pool scrubbing experimental programme carried out in PECA facility are presented. One of the major findings was that a remarkable fraction of particle absorption was not a function of the residence time of bubbles rising through the pool. Such a contribution was assumed to be associated to aerosol removal mechanism acting at the pool entrance. As a consequence. a hydrodynamic experimental plan was launched to examine the gas behaviour during the initial stages in the pool. Size and shape of gas nuclei in the pool were measured and fitted to a lognormal distribution. Particularly, size was found to be quite sensitive to inlet gas flow and at minor extent to gas composition and pool temperature. SPARC90 and BUSCA-AUG92 were used to simulate the retention tests. Whereas SPARC90 showed a pretty good agreement with experimental data, BUSCA-AUG92 results were far away from measurements in all the cases. SPARC90 consistency apparently pointed out the important role of fission products and aerosols retention at the injection zone; nonetheless, a peer examination of pool scrubbing phenomenology at the pool entrance should be carried out to test both hydrodynamic and removal models. Hence, one of the major high lights drawn from this work was the need of further research under representative severe accident conditions (i.e., saturated pools, jet injection regimes, etc.), as well as separate effect tests to validate, improve and

Heat transfer performance of multi-layer insulation structure under roof-slab of pool-type LMFBR

International Nuclear Information System (INIS)

Kinoshita, I.; Yoshida, K.; Uotani, M.; Fukada, T.

1988-01-01

At the normal operation of the pool-type LMFBR, the free surface of liquid sodium at about 500 0 C is present below the roof-slab, separated by a space of the argon cover gas. The temperature of the roof-slab has to be maintained low and uniform in the horizontal direction for sufficient strength of the structure. Therefore, thermal insulation structures must be installed on the lower surface of the roof-slab. In addition to the installation of thermal insulator, forced cooling of the roof-slab is required for assured structural integrity of the roof-slab. The capacity of cooling equipment can be reduced by installation of structures with high thermal insulating performance. The objective of this study is to evaluate the thermal insulation characteristics of multi-layer type insulator installed below the roof-slab by analytically and experimentally. The analytical study is intended to evaluate the effect of number, distance and emissivity of layers on the heat transfer performances. This is treated as the one-dimensional heat transfer with natural convection, conduction and thermal radiation. In the experiments, we have evaluated effects of gap distances between adjacent thermal insulators placed below the roof-slab on the thermal insulation performances

Birth order and childhood type 1 diabetes risk: a pooled analysis of 31 observational studies.

Science.gov (United States)

Cardwell, Chris R; Stene, Lars C; Joner, Geir; Bulsara, Max K; Cinek, Ondrej; Rosenbauer, Joachim; Ludvigsson, Johnny; Svensson, Jannet; Goldacre, Michael J; Waldhoer, Thomas; Jarosz-Chobot, Przemyslawa; Gimeno, Suely Ga; Chuang, Lee-Ming; Roberts, Christine L; Parslow, Roger C; Wadsworth, Emma Jk; Chetwynd, Amanda; Brigis, Girts; Urbonaite, Brone; Sipetic, Sandra; Schober, Edith; Devoti, Gabriele; Ionescu-Tirgoviste, Constantin; de Beaufort, Carine E; Stoyanov, Denka; Buschard, Karsten; Radon, Katja; Glatthaar, Christopher; Patterson, Chris C

2011-04-01

The incidence rates of childhood onset type 1 diabetes are almost universally increasing across the globe but the aetiology of the disease remains largely unknown. We investigated whether birth order is associated with the risk of childhood diabetes by performing a pooled analysis of previous studies. Relevant studies published before January 2010 were identified from MEDLINE, Web of Science and EMBASE. Authors of studies provided individual patient data or conducted pre-specified analyses. Meta-analysis techniques were used to derive combined odds ratios (ORs), before and after adjustment for confounders, and investigate heterogeneity. Data were available for 6 cohort and 25 case-control studies, including 11,955 cases of type 1 diabetes. Overall, there was no evidence of an association prior to adjustment for confounders. After adjustment for maternal age at birth and other confounders, a reduction in the risk of diabetes in second- or later born children became apparent [fully adjusted OR = 0.90 95% confidence interval (CI) 0.83-0.98; P = 0.02] but this association varied markedly between studies (I² = 67%). An a priori subgroup analysis showed that the association was stronger and more consistent in children birth order, particularly in children aged < 5 years. This finding could reflect increased exposure to infections in early life in later born children.

Cells Lacking mtDNA Display Increased dNTP Pools upon DNA Damage

DEFF Research Database (Denmark)

Skovgaard, Tine; Rasmussen, Lene Juel; Munch-Petersen, Birgitte

Imbalanced dNTP pools are highly mutagenic due to a deleterious effect on DNA polymerase fidelity. Mitochondrial DNA defects, including mutations and deletions, are commonly found in a wide variety of different cancer types. In order to further study the interconnection between dNTP pools...... and mitochondrial function we have examined the effect of DNA damage on dNTP pools in cells deficient of mtDNA. We show that DNA damage induced by UV irradiation, in a dose corresponding to LD50, induces an S phase delay in different human osteosarcoma cell lines. The UV pulse also has a destabilizing effect...... shows that normal mitochondrial function is prerequisite for retaining stable dNTP pools upon DNA damage. Therefore it is likely that mitochondrial deficiency defects may cause an increase in DNA mutations by disrupting dNTP pool balance....

German research reactor back-end provisions

International Nuclear Information System (INIS)

Koester, Siegfried; Gruber, Gerhard

2002-01-01

Germany has several types of Research Reactors in operation. These reactors use fuel containing uranium of U.S. origin. Basically all the fuel which will be spent until May 2006 will be returned to the U.S. under existing contracts with the U.S. Department of Energy. The contracts are based on the U.S. FRR SNF (Foreign Research Reactor Spent Nuclear Fuel) Program which started in May 1996 and which will last for 10 years. In 1990, the German Federal Government started a program to long-term store (approx. 40 years) and finally dispose of spent fuel in Germany after the so-called U.S. fuel return window will be closed. In order to long-term store the fuel, a special container was designed which covers all different types of spent fuel from the Research Reactors. The container called 'CASTOR MTR 2' is basically licensed and is already in use for the spent fuel of Russian origin from the 'Research Reactor Rossendorf' in the eastern part of Germany. All that fuel is expected to be stored in the existing intermediate storage facility, the so-called BZA (Brennelemente Zwischenlager Ahaus). BZA already accomodates spent fuel from the former THTR-300 high temperature reactor. A final repository does not yet exist in Germany. Alternative provisions to close the back-end of the Research Reactor fuel cycle are reprocessing at COGEMA (France) or in Russian facilities, perspectively. Waste return in a form to be agreed will be mandatory, at least in France. (author)

Investigating signs of recent evolution in the pool of proviral HIV type 1 DNA during years of successful HAART

DEFF Research Database (Denmark)

Mens, Helene; Pedersen, Anders G; Jørgensen, Louise B

2007-01-01

In order to shed light on the nature of the persistent reservoir of human immunodeficiency virus type 1 (HIV-1), we investigated signs of recent evolution in the pool of proviral DNA in patients on successful HAART. Pro-viral DNA, corresponding to the C2-V3-C3 region of the HIV-1 env gene...... there were temporal trends indicating ongoing replication and evolution. In summary, it was not possible to detect definitive signs of ongoing evolution in either the bulk-sequenced or the clonal data with the methods employed here, but our results could be consistent with localized expression of archival...

Swimming pool granuloma

Science.gov (United States)

... this page: //medlineplus.gov/ency/article/001357.htm Swimming pool granuloma To use the sharing features on this page, please enable JavaScript. A swimming pool granuloma is a long-term (chronic) skin ...

Evaluation of Total Coliform, Fecal Coliform and Residual Chlorine in Swimming Pools in Kermanshah on the Season, the type of Pool, Disinfection System and Source of Water Supply in the during of three years (2010-2012

Directory of Open Access Journals (Sweden)

K SHarafi

2014-11-01

From the results , although the pools of water quality parameters has been studied in almost ideal But in summer, especially on a female pools and pools with wells water supply source than other pools , to be more oversight .

Swimming pool hydraulics and their significance for public pools. Bedeutung der Beckenhydraulik in oeffentlichen Schwimmbaedern

Energy Technology Data Exchange (ETDEWEB)

Gansloser, G

1989-11-01

The term of swimming pool hydraulics means the process of letting in and drawing off water to and from the pool while ensuring that no inadmissible water-borne contaminant concentrations will occur anywhere within the pool. Measurements were performed on a pool to study the significance of correct pool hydraulics. The author points out that a wrong water recirculation design will bring to nought the effects of an elaborate water treatment system; by contrast, poor pool water quality can be greatly improved by redesigning the pool water hydraulics approach. In principle, systems with with water inlet at one side and water outlet at the far side will fall short of hygienic requirements. (BWI).

Clean-up system for pool water in pressure suppression chamber and operation method therefor

Energy Technology Data Exchange (ETDEWEB)

Hirabayashi, Kentaro; Kinoshita, Shoichiro

1996-09-17

Pool water in a pressure suppression chamber of a BWR type reactor is sucked by a pump of an after-heat removing system. The pool water pressurized here is sent to the pressure suppression chamber by way of a heat exchanger and a test line backwarding pipeline to stir the pool water in the pressure suppression chamber. Further, the pool water pressurized by the pump is sent to the pressure suppression chamber by way of a filtration desalting device and an exit pipe to purify the pool water. Upon cleaning of pipelines before the start of a periodical test, the pool water sucked by the pump is sent to the filtration desalting device and recovered to the pressure suppression chamber. This can reduce the amount of impurities carried to the suppression chamber. After the cleaning of the pipelines, pool water is passed through the test line backwarding pipeline, so that the pool water can be stirred at the same time. (I.N.)

Cryptosporidium and Giardia in Swimming Pools, Atlanta, Georgia

Centers for Disease Control (CDC) Podcasts

2008-05-29

In this podcast, Dan Rutz speaks with Dr. Joan Shields, a guest researcher with the Healthy Swimming Program at CDC, about an article in June 2008 issue of Emerging Infectious Diseases reporting on the results of a test of swimming pools in the greater Atlanta, Georgia area. Dr. Shields tested 160 pools in metro Atlanta last year for Cryptosporidium and Giardia. These germs cause most recreational water associated outbreaks.  Created: 5/29/2008 by Emerging Infectious Diseases.   Date Released: 5/29/2008.

Research into basic rocks types

International Nuclear Information System (INIS)

1993-06-01

Teollisuuden Voima Oy (TVO) has carried out research into basic rock types in Finland. The research programme has been implemented in parallel with the preliminary site investigations for radioactive waste disposal in 1991-1993. The program contained two main objectives: firstly, to study the properties of the basic rock types and compare those with the other rock types under the investigation; secondly, to carry out an inventory of rock formations consisting of basic rock types and suitable in question for final disposal. A study of environmental factors important to know regarding the final disposal was made of formations identified. In total 159 formations exceeding the size of 4 km 2 were identified in the inventory. Of these formations 97 were intrusive igneous rock types and 62 originally extrusive volcanic rock types. Deposits consisting of ore minerals, industrial minerals or building stones related to these formations were studied. Environmental factors like natural resources, protected areas or potential for restrictions in land use were also studied

HAMLET -Human Model MATROSHKA for Radiation Exposure Determination of Astronauts -Current status and results

Science.gov (United States)

Reitz, Guenther; Berger, Thomas; Bilski, Pawel; Burmeister, Soenke; Labrenz, Johannes; Hager, Luke; Palfalvi, Jozsef K.; Hajek, Michael; Puchalska, Monika; Sihver, Lembit

The exploration of space as seen in specific projects from the European Space Agency (ESA) acts as groundwork for human long duration space missions. One of the main constraints for long duration human missions is radiation. The radiation load on astronauts and cosmonauts in space (as for the ISS) is a factor of 100 higher than the natural radiation on Earth and will further increase should humans travel to Mars. In preparation for long duration space missions it is important to evaluate the impact of space radiation in order to secure the safety of the astronauts and minimize their radiation risks. To determine the radiation risk on humans one has to measure the radiation doses to radiosensitive organs within the human body. One way to approach this is the ESA facility MATROSHKA (MTR), under the scientific and project lead of DLR. It is dedicated to determining the radiation load on astronauts within and outside the International Space Station (ISS), and was launched in January 2004. MTR is currently preparing for its fourth experimental phase inside the Japanese Experimental Module (JEM) in summer 2010. MTR, which mimics a human head and torso, is an anthropomorphic phantom containing over 6000 radiation detectors to determine the depth dose and organ dose distribution in the body. It is the largest international research initiative ever performed in the field of space dosimetry and combines the expertise of leading research institutions around the world, thereby generating a huge pool of data of potentially immense value for research. Aiming at optimal scientific exploitation, the FP7 project HAMLET aims to process and compile the data acquired individually by the participating laboratories of the MATROSHKA experiment. Based on experimental input from the MATROSHKA experiment phases as well as on radiation transport calculations, a three-dimensional model for the distribution of radiation dose in an astronaut's body will be built up. The scientific achievements

PDA: Pooled DNA analyzer

Directory of Open Access Journals (Sweden)

Lin Chin-Yu

2006-04-01

Full Text Available Abstract Background Association mapping using abundant single nucleotide polymorphisms is a powerful tool for identifying disease susceptibility genes for complex traits and exploring possible genetic diversity. Genotyping large numbers of SNPs individually is performed routinely but is cost prohibitive for large-scale genetic studies. DNA pooling is a reliable and cost-saving alternative genotyping method. However, no software has been developed for complete pooled-DNA analyses, including data standardization, allele frequency estimation, and single/multipoint DNA pooling association tests. This motivated the development of the software, 'PDA' (Pooled DNA Analyzer, to analyze pooled DNA data. Results We develop the software, PDA, for the analysis of pooled-DNA data. PDA is originally implemented with the MATLAB® language, but it can also be executed on a Windows system without installing the MATLAB®. PDA provides estimates of the coefficient of preferential amplification and allele frequency. PDA considers an extended single-point association test, which can compare allele frequencies between two DNA pools constructed under different experimental conditions. Moreover, PDA also provides novel chromosome-wide multipoint association tests based on p-value combinations and a sliding-window concept. This new multipoint testing procedure overcomes a computational bottleneck of conventional haplotype-oriented multipoint methods in DNA pooling analyses and can handle data sets having a large pool size and/or large numbers of polymorphic markers. All of the PDA functions are illustrated in the four bona fide examples. Conclusion PDA is simple to operate and does not require that users have a strong statistical background. The software is available at http://www.ibms.sinica.edu.tw/%7Ecsjfann/first%20flow/pda.htm.

13 CFR 120.1706 - Pool Originator's retained interest in Pool.

Science.gov (United States)

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Pool Originator's retained interest in Pool. 120.1706 Section 120.1706 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION BUSINESS LOANS Establishment of SBA Secondary Market Guarantee Program for First Lien Position 504 Loan...

Evaluation of LOCA in a swimming-pool type reactor using the 3D-AIRLOCA code

International Nuclear Information System (INIS)

Nagler, A.; Gilat, J.; Hirshfeld, H.

1991-01-01

The 3D-AIRLOCA code was used to calculate core temperature evolution curves in the wake of a full LOCA in a swimming pool type reactor, resulting in complete core exposure and dryout within about 1000 sec of the initiating event. The results show that fuel integrity loss thresholds (450 C for softening and 650 C for melting) are reached and exceeded over large fractions of the core at powr levels as low as 2 MW. At 4.5 MW, the softening threshold is reached even when the accident occurs up to 12 hours after reactor shutdown for continuous operation, and up to 2 hrs after shutdown for intermittent (6 hrs/day, 4 days a week) operation. The situation is even more severe in blockage cases, when the air flow through the core is blocked by residual water at the grid plate level. It is concluded that substantial fission product releases are quite likely in this class of accidents. (orig.)

Preliminary Calculation on a Spent Fuel Pool Accident using GOTHIC

Energy Technology Data Exchange (ETDEWEB)

Park, Jaehwan; Choi, Yu Jung; Hong, Tae Hyub; Kim, Hyeong-Taek [KHNP-CRI, Daejeon (Korea, Republic of)

2015-10-15

The probability of an accident happening at the spent fuel pool was believed to be quite low until the 2011 Fukushima accident occurred. Notably, large amount of spent fuel are normally stored in the spent fuel pool for a long time compared to the amount of fuel in the reactor core and the total heat released from the spent fuel is high enough to boil the water of the spent fuel pool when the cooling system does not operate. In addition, the enrichment and the burnup of the fuel have both increased in the past decade and heat generation from the spent fuel thereby has also increased. The failure of the cooling system at the spent fuel pool (hereafter, a loss-of-cooling accident) is one of the principal hypothetical causes of an accident that could occur at the spent fuel pool. In this paper, the preliminary calculation of a loss-of-cooling accident was performed. In this paper, the preliminary calculation of a loss-of cooling accident was performed with GOTHIC. The calculation results show boiling away of water in the spent fuel pool due to the loss-of-cooling accident and similar thermal performance of the spent fuel pool with previous research results.

Feedback stabilisation of pool-boiling systems : for application in thermal management schemes

NARCIS (Netherlands)

Gils, van R.W.

2012-01-01

The research scope of this thesis is the stabilisation of unstable states in a pool-boiling system. Thereto, a compact mathematical model is employed. Pool-boiling systems serve as physical model for practical applications of boiling heat transfer in industry. Boiling has advantages over

Water inventory management in condenser pool of boiling water reactor

International Nuclear Information System (INIS)

Gluntz, D.M.

1996-01-01

An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs

Benthic assemblages of rock pools in northern Portugal: seasonal and between-pool variability

Directory of Open Access Journals (Sweden)

Iacopo Bertocci

2012-11-01

Full Text Available We investigated the seasonal (winter vs summer and within season and spatial (between-pool variability of benthic assemblages of rock pools at mid-intertidal level along the shore of Viana do Castelo (North Portugal. Physical traits of rock pools, including size, depth and position along the shore, were also compared between pools. While pools did not differ for any of the examined physical traits, results indicated a clear seasonal difference in the structure of assemblages, including a total of 49 macroalgal and 13 animal taxa. This finding was driven by six taxa that are more abundant in winter (the reef-forming polychaete Sabellaria alveolata, the articulated coralline algae Corallina spp., the brown alga Bifurcaria bifurcata, the encrusting coralline alga Lithophyllum incrustans, the red alga Chondracanthus acicularis and the grazing snails Gibbula spp. and four algal taxa that are more abundant in summer (the invasive brown Sargassum muticum, the green Ulva spp., the kelp Laminaria ochroleuca and the filamentous red Ceramium spp.. These data provide a new contribution to the knowledge of rock pool systems and have potential implications for monitoring programmes aimed at assessing ecological modifications related to natural and anthropogenic disturbances and for identifying processes responsible for the variability of rock pool assemblages.

Energy Pooling Upconversion in Free Space and Optical Cavities

Science.gov (United States)

LaCount, Michael D.

The ability to efficiently convert the wavelength of light has value in a wide range of disciplines that include the fields of photovoltaics, plant growth, optics and medicine. The processes by which such transformations are carried out are known as upconversions and downconversions. There are several ways to up/down convert light, each with its own attributes, issues, and competing mechanisms. Most are associated with one-body or two-body processes. Three-body dynamics are also possible though, going by the names of quantum cutting (downconversion) and energy pooling (upconversion). These use virtual excited electronic states to mediate conversions as has been experimentally realized using lanthanide ions embedded in wide bandgap materials. The use of lanthanides to convert light is not ideal due to their relative scarcity, toxicity, and the limited range of light frequencies that can be absorbed and emitted. Organic molecules, on the other hand, are typically non-toxic, are made up of abundant elements, and can be designed with tailored spectral properties. At issue is whether or not they can be used to carry out efficient energy pooling, the central question to be answered in this thesis. The research presented here draws on a perturbative quantum electrodynamics framework previously established for generic energy pooling. It was used to develop a computational methodology for determining the rate of energy pooling and its competing processes. This, in turn, draws on a combination of time-dependent density functional theory, quantum electrodynamics, and perturbation theory to generate the requisite material property data. This computational model was applied to two test systems consisting of stilbene-fluorescein and hexabenzocoronene-oligothiophene. The stilbene-fluorescein system was found to have a maximum energy pooling rate efficiency (as compared to competing processes) of 17% and the hexabenzocoronene-oligothiophene system was found to have a maximum

The heavy water accountancy for research reactors in JAERI

International Nuclear Information System (INIS)

Yoshijima, Tetsuo; Tanaka, Sumitoshi; Nemoto, Denjirou

1998-11-01

The three research reactors have been operated by the Department of Research Reactor and used about 41 tons heavy water as coolant, moderator and reflector of research reactors. The JRR-2 is a tank type research reactor of 10MW in thermal power and its is used as moderator, coolant and reflector about 16 tons heavy water. The JRR-3M is a light water cooled and moderated pool type research reactor with a thermal power of 20MW and its is used as reflector about 7.3 tons heavy water. In the JRR-4, which is a light water cooled swimming pool type research reactor with the maximum thermal power of 3.5MW, about 1 ton heavy water is used to supply fully thermalized neutrons with a neutron beam experiment of facility. The heavy water was imported from U.S.A., CANADA and Norway. Parts of heavy water is internationally controlled materials, therefore management of heavy water is necessary for materials accountancy. This report described the change of heavy water inventories in each research reactors, law and regulations for accounting of heavy water in JAERI. (author)

Consequences in a long time of the forced loss of coolant in a pool type reactor

International Nuclear Information System (INIS)

Botelho, D.A.

1986-01-01

The fuel and pool water temperatures are calculated as a function of time using unidimensional models of heat conduction and momentum conservation, to simulate the natural convection flow of the coolant. The reactor building pressure due to the pool water evaporation is calculated using a homogeneous model with thermal equilibrium. The heat loss from the three main components of the building volume (liquid water, air, and steam) to solid surfaces such as the building walls are taking into account. (Author) [pt

Big city consultants shut down our pool : a shocking community pool gets checked for stray voltage

Energy Technology Data Exchange (ETDEWEB)

Lynch, P. [Power Line Systems Engineering Inc., Markham, ON (Canada)

2009-12-15

This article discussed an investigation conducted at a community pool where swimmers complained of receiving electrical shocks both in the pool and on the pool's deck area. Electrical measurements taken at the pool revealed current flows from the pool water to various points around the deck area. Measured current flow in the pool area was 30 amps even when the main pool service breaker was opened to shut off power to the entire facility. Thirty amps of primary neutral current was then measured on the primary side aerial neutral in front of the pool. A 10 amp primary feeder from the pool joined up with the complex's primary neutral wire to increase the neutral current to 40 amps. The combined 40 amps current then returned to the secondary side of a nearby utility transformer substation. The study showed that the underground wet low-resistance grounded surface area of the pool was attracting the 30 amps of utility current from the surrounding ground area. The local utility disconnected the primary and secondary neutral interconnection at the pool's main 600-volt step-down transformer. The pool deck was removed in order to install additional copper bonding grounds. In order to avert serious injuries, many experts propose that all electric utilities should be required by law to reconfigure their power systems to prevent primary power neutral currents from entering private buildings. 1 tab., 2 figs.

[Soil organic carbon pools and their turnover under two different types of forest in Xiao-xing'an Mountains, Northeast China].

Science.gov (United States)

Gao, Fei; Jiang, Hang; Cui, Xiao-yang

2015-07-01

Soil samples collected from virgin Korean pine forest and broad-leaved secondary forest in Xiaoxing'an Mountains, Northeast China were incubated in laboratory at different temperatures (8, 18 and 28 °C) for 160 days, and the data from the incubation experiment were fitted to a three-compartment, first-order kinetic model which separated soil organic carbon (SOC) into active, slow, and resistant carbon pools. Results showed that the soil organic carbon mineralization rates and the cumulative amount of C mineralized (all based on per unit of dry soil mass) of the broad-leaved secondary forest were both higher than that of the virgin Korean pine forest, whereas the mineralized C accounted for a relatively smaller part of SOC in the broad-leaved secondary forest soil. Soil active and slow carbon pools decreased with soil depth, while their proportions in SOC increased. Soil resistant carbon pool and its contribution to SOC were both greater in the broad-leaved secondary forest soil than in the virgin Korean pine forest soil, suggesting that the broad-leaved secondary forest soil organic carbon was relatively more stable. The mean retention time (MRT) of soil active carbon pool ranged from 9 to 24 d, decreasing with soil depth; while the MRT of slow carbon pool varied between 7 and 24 a, increasing with soil depth. Soil active carbon pool and its proportion in SOC increased linearly with incubation temperature, and consequently, decreased the slow carbon pool. Virgin Korean pine forest soils exhibited a higher increasing rate of active carbon pool along temperature gradient than the broad-leaved secondary forest soils, indicating that the organic carbon pool of virgin Korean pine forest soil was relatively more sensitive to temperature change.

poolHiTS: A Shifted Transversal Design based pooling strategy for high-throughput drug screening

Directory of Open Access Journals (Sweden)

Woolf Peter J

2008-05-01

Full Text Available Abstract Background A key goal of drug discovery is to increase the throughput of small molecule screens without sacrificing screening accuracy. High-throughput screening (HTS in drug discovery involves testing a large number of compounds in a biological assay to identify active compounds. Normally, molecules from a large compound library are tested individually to identify the activity of each molecule. Usually a small number of compounds are found to be active, however the presence of false positive and negative testing errors suggests that this one-drug one-assay screening strategy can be significantly improved. Pooling designs are testing schemes that test mixtures of compounds in each assay, thereby generating a screen of the whole compound library in fewer tests. By repeatedly testing compounds in different combinations, pooling designs also allow for error-correction. These pooled designs, for specific experiment parameters, can be simply and efficiently created using the Shifted Transversal Design (STD pooling algorithm. However, drug screening contains a number of key constraints that require specific modifications if this pooling approach is to be useful for practical screen designs. Results In this paper, we introduce a pooling strategy called poolHiTS (Pooled High-Throughput Screening which is based on the STD algorithm. In poolHiTS, we implement a limit on the number of compounds that can be mixed in a single assay. In addition, we show that the STD-based pooling strategy is limited in the error-correction that it can achieve. Due to the mixing constraint, we show that it is more efficient to split a large library into smaller blocks of compounds, which are then tested using an optimized strategy repeated for each block. We package the optimal block selection algorithm into poolHiTS. The MATLAB codes for the poolHiTS algorithm and the corresponding decoding strategy are also provided. Conclusion We have produced a practical version

Reactor TRIGA PUSPATI (RTP) spent fuel pool conceptual design

International Nuclear Information System (INIS)

Mohd Fazli Zakaria; Tonny Lanyau; Ahmad Nabil Ab Rahim

2010-01-01

Reactor TRIGA PUSPATI (RTP) is the one and only research reactor in Malaysia that has been safely operated and maintained since 1982. In order to enhance technical capabilities and competencies especially in nuclear reactor engineering a feasibility study on RTP power upgrading was proposed to serve future needs for advance nuclear science and technology in the country with the capability of designing and develop reactor system. The need of a Spent Fuel Pool begins with the discharge of spent fuel elements from RTP for temporary storage that includes all activities related to the storage of fuel until it is either sent for reprocessed or sent for final disposal. To support RTP power upgrading there will be major RTP systems replacement such as reactor components and a new temporary storage pool for fuel elements. The spent fuel pool is needed for temporarily store the irradiated fuel elements to accommodate a new reactor core structure. Spent fuel management has always been one of the most important stages in the nuclear fuel cycle and considered among the most common problems to all countries with nuclear reactors. The output of this paper will provide sufficient information to show the Spent Fuel Pool can be design and build with the adequate and reasonable safety assurance to support newly upgraded TRIGA PUSPATI TRIGA Research Reactor. (author)

Corrosion of cermet cores of fuel plates for nuclear research reactor

International Nuclear Information System (INIS)

Durazzo, M.; Ramanathan, L.V.

1984-01-01

Materials Testing Reactor (MTR) type fuel plates containing U 3 O 8 -Al cores and clad with Al are used in various research reactor. Preliminary investigations, where in the cladding of samples was drilled to simulate conditions of rupture due to pitting attack, revealed that considerable quantities of H 2 was evolved upon exposure of the core to water. The corrosion of cermets cores of different densities was characterized as a function of H 2 evolution that revealed 3 stages. A first stage consisting of an incubation period followed by initiation of H 2 evolution, a second stage with a constant rate of H 2 evolution and a third stage with a low rate of H 2 evolution. All 3 stages were found to vary as a function of cermet density and water temperature. (Author) [pt

Pooled versus separate measurements of tree-ring stable isotopes

Energy Technology Data Exchange (ETDEWEB)

Dorado Linan, Isabel, E-mail: isabel@gfz-potsdam.de [Universitat de Barcelona, Departament d' Ecologia, Diagonal 645, 08028, Barcelona (Spain); German Centre for Geosciences, Climate Dynamics and Landscape Evolution, Dendro Laboratory, Telegrafenberg, 14473, Potsdam (Germany); Gutierrez, Emilia, E-mail: emgutierrez@ub.edu [Universitat de Barcelona, Departament d' Ecologia, Diagonal 645, 08028, Barcelona (Spain); Helle, Gerhard, E-mail: ghelle@gfz-potsdam.de [German Centre for Geosciences, Climate Dynamics and Landscape Evolution, Dendro Laboratory, Telegrafenberg, 14473, Potsdam (Germany); Heinrich, Ingo, E-mail: heinrich@gfz-potsdam.de [German Centre for Geosciences, Climate Dynamics and Landscape Evolution, Dendro Laboratory, Telegrafenberg, 14473, Potsdam (Germany); Andreu-Hayles, Laia, E-mail: laiandreu@ub.edu [Universitat de Barcelona, Departament d' Ecologia, Diagonal 645, 08028, Barcelona (Spain); Tree-Ring Laboratory, Lamont-Doherty Earth Observatory of Columbia University, Palisades NY (United States); Planells, Octavi, E-mail: leocarpus@hotmail.com [Universitat de Barcelona, Departament d' Ecologia, Diagonal 645, 08028, Barcelona (Spain); Leuenberger, Markus, E-mail: leuenberger@climate.unibe.ch [Climate and Environmental Physics, Physics Institute, University of Bern, Sidlerstrasse 5, 3012 Bern (Switzerland); Oeschger Centre of Climate Change Research, University of Bern, Zaehringerstrasse 25, 3012 Bern (Switzerland); Buerger, Carmen, E-mail: buerger@gfz-potsdam.de [German Centre for Geosciences, Climate Dynamics and Landscape Evolution, Dendro Laboratory, Telegrafenberg, 14473, Potsdam (Germany); Schleser, Gerhard, E-mail: schleser@gfz-potsdam.de [German Centre for Geosciences, Climate Dynamics and Landscape Evolution, Dendro Laboratory, Telegrafenberg, 14473, Potsdam (Germany)

2011-05-01

{delta}{sup 13}C and {delta}{sup 18}O of tree rings contain time integrated information about the environmental conditions weighted by seasonal growth dynamics and are well established as sources of palaeoclimatic and ecophysiological data. Annually resolved isotope chronologies are frequently produced by pooling dated growth rings from several trees prior to the isotopic analyses. This procedure has the advantage of saving time and resources, but precludes from defining the isotopic error or statistical uncertainty related to the inter-tree variability. Up to now only a few studies have compared isotope series from pooled tree rings with isotopic measurements from individual trees. We tested whether or not the {delta}{sup 13}C and the {delta}{sup 18}O chronologies derived from pooled and from individual tree rings display significant differences at two locations from the Iberian Peninsula to assess advantages and constraints of both methodologies. The comparisons along the period 1900-2003 reveal a good agreement between pooled chronologies and the two mean master series which were created by averaging raw individual values (Mean) or by generating a mass calibrated mean (MassC). In most of the cases, pooled chronologies show high synchronicity with averaged individual samples at interannual scale but some differences also show up especially when comparing {delta}{sup 18}O decadal to multi-decadal variations. Moreover, differences in the first order autocorrelation among individuals may be obscured by pooling strategies. The lack of replication of pooled chronologies prevents detection of a bias due to a higher mass contribution of one sample but uncertainties associated with the analytical process itself, as sample inhomogeneity, seems to account for the observed differences. - Research Highlights: {yields} Pooled {delta}{sup 13}C and {delta}{sup 18}O chronologies are expected to be similar to the mean. {yields} Empirical pooled chronologies {delta}{sup 13}C and

Pooled versus separate measurements of tree-ring stable isotopes

International Nuclear Information System (INIS)

Dorado Linan, Isabel; Gutierrez, Emilia; Helle, Gerhard; Heinrich, Ingo; Andreu-Hayles, Laia; Planells, Octavi; Leuenberger, Markus; Buerger, Carmen; Schleser, Gerhard

2011-01-01

δ 13 C and δ 18 O of tree rings contain time integrated information about the environmental conditions weighted by seasonal growth dynamics and are well established as sources of palaeoclimatic and ecophysiological data. Annually resolved isotope chronologies are frequently produced by pooling dated growth rings from several trees prior to the isotopic analyses. This procedure has the advantage of saving time and resources, but precludes from defining the isotopic error or statistical uncertainty related to the inter-tree variability. Up to now only a few studies have compared isotope series from pooled tree rings with isotopic measurements from individual trees. We tested whether or not the δ 13 C and the δ 18 O chronologies derived from pooled and from individual tree rings display significant differences at two locations from the Iberian Peninsula to assess advantages and constraints of both methodologies. The comparisons along the period 1900-2003 reveal a good agreement between pooled chronologies and the two mean master series which were created by averaging raw individual values (Mean) or by generating a mass calibrated mean (MassC). In most of the cases, pooled chronologies show high synchronicity with averaged individual samples at interannual scale but some differences also show up especially when comparing δ 18 O decadal to multi-decadal variations. Moreover, differences in the first order autocorrelation among individuals may be obscured by pooling strategies. The lack of replication of pooled chronologies prevents detection of a bias due to a higher mass contribution of one sample but uncertainties associated with the analytical process itself, as sample inhomogeneity, seems to account for the observed differences. - Research Highlights: → Pooled δ 13 C and δ 18 O chronologies are expected to be similar to the mean. → Empirical pooled chronologies δ 13 C and δ 18 O and the mean show a high synchronicity. → Pooled chronologies differ

General description and production lines of the Egyptian fuel manufacturing pilot plant

International Nuclear Information System (INIS)

Zidan, W.I.; Elseaidy, I.M.

1999-01-01

The Egyptian Fuel Manufacturing Pilot Plant, FMPP, is a new facility, producing an MTR-type fuel elements required for the Egyptian Second Research Reactor, ETRR-2, as well as other plates or elements for an external clients with the same type and enrichment percent or lower, (LEU). General description is presented. The production lines in FMPP, which begin from uranium hexaflouride (UF 6 , 19.7±0.2 % U 235 by wt), aluminum powder, and nuclear grade 6061 aluminium alloy in sheets, bars, and rods with the different heat treatments and dimensions as a raw materials, are processed through a series of the manufacturing, inspection, and quality control plan to produce the final specified MTR-type fuel elements. All these processes and the product control in each step are presented. The specifications of the final product are presented. (author)

Study on Dynamic Development of Three-dimensional Weld Pool Surface in Stationary GTAW

Science.gov (United States)

Huang, Jiankang; He, Jing; He, Xiaoying; Shi, Yu; Fan, Ding

2018-04-01

The weld pool contains abundant information about the welding process. In particular, the type of the weld pool surface shape, i. e., convex or concave, is determined by the weld penetration. To detect it, an innovative laser-vision-based sensing method is employed to observe the weld pool surface of the gas tungsten arc welding (GTAW). A low-power laser dots pattern is projected onto the entire weld pool surface. Its reflection is intercepted by a screen and captured by a camera. Then the dynamic development process of the weld pool surface can be detected. By observing and analyzing, the change of the reflected laser dots reflection pattern, for shape of the weld pool surface shape, was found to closely correlate to the penetration of weld pool in the welding process. A mathematical model was proposed to correlate the incident ray, reflected ray, screen and surface of weld pool based on structured laser specular reflection. The dynamic variation of the weld pool surface and its corresponding dots laser pattern were simulated and analyzed. By combining the experimental data and the mathematical analysis, the results show that the pattern of the reflected laser dots pattern is closely correlated to the development of weld pool, such as the weld penetration. The concavity of the pool surface was found to increase rapidly after the surface shape was changed from convex to concave during the stationary GTAW process.

The ICF Core Sets for hearing loss--researcher perspective. Part I: Systematic review of outcome measures identified in audiological research.

Science.gov (United States)

Granberg, Sarah; Dahlström, Jennie; Möller, Claes; Kähäri, Kim; Danermark, Berth

2014-02-01

To review the literature in order to identify outcome measures used in research on adults with hearing loss (HL) as part of the ICF Core Sets development project, and to describe study and population characteristics of the reviewed studies. A systematic review methodology was applied using multiple databases. A comprehensive search was conducted and two search pools were created, pool I and pool II. The study population included adults (≥ 18 years of age) with HL and oral language as the primary mode of communication. 122 studies were included. Outcome measures were distinguished by 'instrument type', and 10 types were identified. In total, 246 (pool I) and 122 (pool II) different measures were identified, and only approximately 20% were extracted twice or more. Most measures were related to speech recognition. Fifty-one different questionnaires were identified. Many studies used small sample sizes, and the sex of participants was not revealed in several studies. The low prevalence of identified measures reflects a lack of consensus regarding the optimal outcome measures to use in audiology. Reflections and discussions are made in relation to small sample sizes and the lack of sex differentiation/descriptions within the included articles.

Development of a Two-dimensional Thermohydraulic Hot Pool Model and ITS Effects on Reactivity Feedback during a UTOP in Liquid Metal Reactors

International Nuclear Information System (INIS)

Lee, Yong Bum; Jeong, Hae Yong; Cho, Chung Ho; Kwon, Young Min; Ha, Kwi Seok; Chang, Won Pyo; Suk, Soo Dong; Hahn, Do Hee

2009-01-01

The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect

Measurements in large pool fires with an actively cooled calorimeter

International Nuclear Information System (INIS)

Koski, J.A.; Wix, S.D.

1995-01-01

The pool fire thermal test described in Safety Series 6 published by the International Atomic Energy Agency (IAEA) or Title 10, Code of Federal Regulations, Part 71 (10CFR71) in the United States is one of the most difficult tests that a container for larger ''Type B'' quantities of nuclear materials must pass. If retests of a container are required, costly redesign and project delays can result. Accurate measurements and modeling of the pool fire environment will ultimately lower container costs by assuring that containers past the pool fire test on the first attempt. Experiments indicate that the object size or surface temperature of the container can play a role in determining local heat fluxes that are beyond the effects predicted from the simple radiative heat transfer laws. An analytical model described by Nicolette and Larson 1990 can be used to understand many of these effects. In this model a gray gas represents soot particles present in the flame structure. Close to the container surface, these soot particles are convectively and radiatively cooled and interact with incident energy from the surrounding fire. This cooler soot cloud effectively prevents some thermal radiation from reaching the container surface, reducing the surface heat flux below the value predicted by a transparent medium model. With some empirical constants, the model suggested by Nicolette and Larson can be used to more accurately simulate the pool fire environment. Properly formulated, the gray gas approaches also fast enough to be used with standard commercial computer codes to analyze shipping containers. To calibrate this type of model, accurate experimental measurements of radiative absorption coefficients, flame temperatures, and other parameters are necessary. A goal of the calorimeter measurements described here is to obtain such parameters so that a fast, useful design tool for large pool fires can be constructed

Status of research and modelling of water-pool scrubbing

International Nuclear Information System (INIS)

Ramsdale, S.A.; Bamford, G.J.; Fishwick, S.; Starkie, H.C.

1992-11-01

A critical review has been performed of the modelling and experimental data on aerosol and vapour retention in water pools. This involved the systematic comparison of available computer codes, and the selection of the most suitable code for further improvement and future inclusion in the Ester code. Busca was the code selected, and has now been extended to model the condensation of steam onto aerosol particles, taking into account curvature and solute effects. It has also been extended to treat the enhanced rise velocity of swarms of bubbles. Busca was then validated against the best available experimental data, namely data from ACE Phase A and the EPRI experiments. Agreement of the code with experiments was generally very satisfactory

Pool-Type Fishways: Two Different Morpho-Ecological Cyprinid Species Facing Plunging and Streaming Flows

Science.gov (United States)

Branco, Paulo; Santos, José M.; Katopodis, Christos; Pinheiro, António; Ferreira, Maria T.

2013-01-01

Fish are particularly sensitive to connectivity loss as their ability to reach spawning grounds is seriously affected. The most common way to circumvent a barrier to longitudinal connectivity, and to mitigate its impacts, is to implement a fish passage device. However, these structures are often non-effective for species with different morphological and ecological characteristics so there is a need to determine optimum dimensioning values and hydraulic parameters. The aim of this work is to study the behaviour and performance of two species with different ecological characteristics (Iberian barbel Luciobarbus bocagei–bottom oriented, and Iberian chub Squalius pyrenaicus–water column) in a full-scale experimental pool-type fishway that offers two different flow regimes–plunging and streaming. Results showed that both species passed through the surface notch more readily during streaming flow than during plunging flow. The surface oriented species used the surface notch more readily in streaming flow, and both species were more successful in moving upstream in streaming flow than in plunging flow. Streaming flow enhances upstream movement of both species, and seems the most suitable for fishways in river systems where a wide range of fish morpho-ecological traits are found. PMID:23741465

Ageing management of the BR2 research reactor

International Nuclear Information System (INIS)

Verpoortem, J. R.; Van Dyck, S.

2014-01-01

At the Belgian nuclear research centre (SCK.CEN) several test reactors are operated. Among these, Belgian Reactor 2 (BR2) is the largest Material Test Reactor (MTR). This water-cooled, beryllium moderated reactor with a maximum thermal power of 100 MW became operational in 1962. Except for two major refurbishment campaigns of one year each, this reactor has been operated continuously over the past 50 years, with a frequency of 5-12 cycles per year. At present, BR2 is used for different research activities, the production of medical isotopes, the production of n-doped silicon and various training and education activities. (Author)

Is energy pooling necessary in ultraviolet matrix-assisted laser desorption/ionization?

Science.gov (United States)

Lin, Hou-Yu; Song, Botao; Lu, I-Chung; Hsu, Kuo-Tung; Liao, Chih-Yu; Lee, Yin-Yu; Tseng, Chien-Ming; Lee, Yuan-Tseh; Ni, Chi-Kung

2014-01-15

Energy pooling has been suggested as the key process for generating the primary ions during ultraviolet matrix-assisted laser desorption/ionization (UV-MALDI). In previous studies, decreases in fluorescence quantum yields as laser fluence increased for 2-aminobenzoic acid, 2,5-dihydroxybenzoic acid (2,5-DHB), and 3-hydroxypicolinic acid were used as evidence of energy pooling. This work extends the research to other matrices and addresses whether energy pooling is a universal property in UV-MALDI. Energy pooling was investigated in a time-resolved fluorescence experiment by using a short laser pulse (355 nm, 20 ps pulse width) for excitation and a streak camera (1 ps time resolution) for fluorescence detection. The excited-state lifetime of 2,5-DHB decreased with increases in laser fluence. This suggests that a reaction occurs between two excited molecules, and that energy pooling may be one of the possible reactions. However, the excited-state lifetime of 2,4,6-trihydroxyacetophenone (THAP) did not change with increases in laser fluence. The upper limit of the energy pooling rate constant for THAP is estimated to be approximately 100-500 times smaller than that of 2,5-DHB. The small energy pooling rate constant for THAP indicates that the potential contribution of the energy pooling mechanism to the generation of THAP matrix primary ions should be reconsidered. Copyright © 2013 John Wiley & Sons, Ltd.

Experimental research on the dynamic behaviors of the keyhole and molten pool in laser deep-penetration welding

Science.gov (United States)

Zhang, Yi; Lin, Qida; Yin, Xuni; Li, Simeng; Deng, Jiquan

2018-04-01

Both the morphology and temperature are two important characteristics of the keyhole and the molten pool in laser deep-penetration welding. The modified ‘sandwich’ method was adopted to overcome the difficulty in obtaining inner information about the keyhole and the molten pool. Based on this method, experimental platforms were built for observing the variations in the surface morphology, the longitudinal keyhole profile and the internal temperature. The experimental results of three dynamic behaviors exbibit as follows. The key factor, which makes the pool width go into a quasi-steady state, lies in the balance between the vortex and the sideways flows around the keyhole. Experimental observation shows that the keyhole goes through three stages in laser welding: the rapid drilling stage, the slow drilling stage and the quasi-steady state. The time for achieving a relative fixed keyhole depth is close to the formation time of the maximum pool width. The internal temperatures inside the keyhole and the molten pool first experience a rapid increase, then a decrease and finally go into a quasi-steady state. Compared to that in the unstable stage, the liquid–metal uphill formed in the stable stage of laser welding has less influence on the internal temperature.

A comprehensive review on pool boiling of nanofluids

International Nuclear Information System (INIS)

Ciloglu, Dogan; Bolukbasi, Abdurrahim

2015-01-01

Nanofluids are nanoparticle suspensions of small particle size and low concentration dispersed in base fluids such as water, oil and ethylene glycol. These fluids have been considered by researchers as a unique heat transfer carrier because of their thermophysical properties and a great number of potential benefits in traditional thermal engineering applications, including power generation, transportation, air conditioning, electronics devices and cooling systems. Many attempts have been made in the literature on nanofluid boiling; however, data on the boiling heat transfer coefficient (HTC) and the critical heat flux (CHF) have been inconsistent. This paper presents a review of recent researches on the pool boiling heat transfer behaviour of nanofluid. First, the development of nanofluids and their potential applications are briefly given. Then, the effects of various parameters on nanofluids pool boiling are discussed in detail. - Highlights: • A review on the pool boiling heat transfer of nanofluid is presented and discussed. • Nanoparticle deposition considerably affects the boiling heat transfer. • The HTC decreases due to the low contact angle and the high adhesion energy. • The HTC increases due to the formation of the new cavities and liquid suction. • The CHF increases due to the increase in roughness, wettability and capillarity

Characterization of radioactive contaminants and water treatment trials for the Taiwan Research Reactor's spent fuel pool

International Nuclear Information System (INIS)

Huang, Chun-Ping; Lin, Tzung-Yi; Chiao, Ling-Huan; Chen, Hong-Bin

2012-01-01

Highlights: ► Deal with a practical radioactive contamination in Taiwan Research Reactor spent fuel pool water. ► Identify the properties of radioactive contaminants and performance test for water treatment materials. ► The radioactive solids were primary attributed by ruptured spent fuels, spent resins, and metal debris. ► The radioactive ions were major composed by uranium and fission products. ► Diatomite-based ceramic depth filter can simultaneously removal radioactive solids and ions. - Abstract: There were approximately 926 m 3 of water contaminated by fission products and actinides in the Taiwan Research Reactor's spent fuel pool (TRR SFP). The solid and ionic contaminants were thoroughly characterized using radiochemical analyses, scanning electron microscopy equipped with an energy dispersive spectrometer (SEM-EDS), and inductively coupled plasma optical emission spectrometry (ICP-OES) in this study. The sludge was made up of agglomerates contaminated by spent fuel particles. Suspended solids from spent ion-exchange resins interfered with the clarity of the water. In addition, the ionic radionuclides such as 137 Cs, 90 Sr, U, and α-emitters, present in the water were measured. Various filters and cation-exchange resins were employed for water treatment trials, and the results indicated that the solid and ionic contaminants could be effectively removed through the use of <0.9 μm filters and cation exchange resins, respectively. Interestingly, the removal of U was obviously efficient by cation exchange resin, and the ceramic depth filter composed of diatomite exhibited the properties of both filtration and adsorption. It was found that the ceramic depth filter could adsorb β-emitters, α-emitters, and uranium ions. The diatomite-based ceramic depth filter was able to simultaneously eliminate particles and adsorb ionic radionuclides from water.

Phase diagrams for the spatial public goods game with pool punishment

Science.gov (United States)

Szolnoki, Attila; Szabó, György; Perc, Matjaž

2011-03-01

The efficiency of institutionalized punishment is studied by evaluating the stationary states in the spatial public goods game comprising unconditional defectors, cooperators, and cooperating pool punishers as the three competing strategies. Fines and costs of pool punishment are considered as the two main parameters determining the stationary distributions of strategies on the square lattice. Each player collects a payoff from five five-person public goods games, and the evolution of strategies is subsequently governed by imitation based on pairwise comparisons at a low level of noise. The impact of pool punishment on the evolution of cooperation in structured populations is significantly different from that reported previously for peer punishment. Representative phase diagrams reveal remarkably rich behavior, depending also on the value of the synergy factor that characterizes the efficiency of investments payed into the common pool. Besides traditional single- and two-strategy stationary states, a rock-paper-scissors type of cyclic dominance can emerge in strikingly different ways.

Evaluation of Decontamination Factor of Aerosol in Pool Scrubber according to Bubble Shape and Size

Energy Technology Data Exchange (ETDEWEB)

Jo, Hyun Joung; Ha, Kwang Soon; Jang, Dong Soon [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The scrubbing pool could play an important role in the wet type FCVS because a large amount of aerosol is captured in the water pool. The pool scrubbing phenomena have been modelled and embedded in several computer codes, such as SPARC (Suppression Pool Aerosol Removal Code), BUSCA (BUbble Scrubbing Algorithm) and SUPRA (Suppression Pool Retention Analysis). These codes aim at simulating the pool scrubbing process and estimating the decontamination factors (DFs) of the radioactive aerosol and iodine gas in the water pool, which is defined as the ratio of initial mass of the specific radioactive material to final massy after passing through the water pool. The pool scrubbing models were reviewed and an aerosol scrubbing code has been prepared to calculate decontamination factor through the pool. The developed code has been verified using the experimental results and parametric studies the decontamination factor according to bubble shape and size. To evaluate the decontamination factor more accurate whole pool scrubber phenomena, the code was improved to consider the variety shape and size of bubbles. The decontamination factor were largely evaluated in ellipsoid bubble rather than in sphere bubble. The pool scrubbing models will be enhanced to apply more various model such as aerosol condensation of hygroscopic. And, it is need to experiment to measure to bubble shape and size distribution in pool to improve bubble model.

"Ripples" in an Aluminum Pool?

Science.gov (United States)

Rohr, James; Wang, Si-Yin; Nesterenko, Vitali F.

2018-05-01

Our motivation for this article is for students to realize that opportunities for discovery are all around them. Discoveries that can still puzzle present day researchers. Here we explore an observation by a middle school student concerning the production of what appears to be water-like "ripples" produced in aluminum foil when placed between two colliding spheres. We both applaud and explore the student's reasoning that the ripples were formed in a melted aluminum pool.

Survey of bacterial contamination of environment of swimming pools in Yazd city, in 2013

Directory of Open Access Journals (Sweden)

Hossein Jafari Mansoorian

2015-09-01

Full Text Available Background: Infections are readily transmitted as a result of bacterial contamination of swimming pools. Therefore, hygiene and preventing the contamination of swimming pools is of particular importance. The objective of this study was to determine the amount of bacterial contamination in indoor pools of Yazd in 2013. Methods: In this descriptive and analytical study, all indoor swimming pools of Yazd (12 pools were evaluated during the spring and summer of 2013, in terms of bacterial contamination. In order to determine contamination, a sterile cotton swab was used for sampling. On average, 45 samples were taken from different surfaces in each pool (shower, dressing room, sitting places in sauna, platforms and around the pool. In total, about 540 samples from all pools were tested for bacterial contamination. Results: The results show that from 540 samples, bacterial contamination was observed in about 93 samples (17.22%; and was seen more in showers, edges of the pool and jacuzzis, and the slippers used in swimming pools. The most important isolated bacteria types were E. coli, Actinobacteria, Pseudomonas alcaligenes, Pseudomonas aeruginosa and Klebsiella pneumonia. Conclusion: The results indicate the presence of bacterial contamination on the surface of these places. It is recommended that health authorities should pay more attention to cleaning and disinfecting surfaces around the pool, showers, dressing rooms etc, to prevent infectious disease transfer as a result of contact with contaminated swimming pool surfaces.

MTHFR C677T and MTR A2756G polymorphisms and the homocysteine lowering efficacy of different doses of folic acid in hypertensive Chinese adults

Directory of Open Access Journals (Sweden)

Qin Xianhui

2012-01-01

Full Text Available Abstract Background This study aimed to investigate if the homocysteine-lowering efficacy of two commonly used physiological doses (0.4 mg/d and 0.8 mg/d of folic acid (FA can be modified by individual methylenetetrahydrofolate reductase (MTHFR C677T and/or methionine synthase (MTR A2756G polymorphisms in hypertensive Chinese adults. Methods A total of 480 subjects with mild or moderate essential hypertension were randomly assigned to three treatment groups: 1 enalapril only (10 mg, control group; 2 enalapril-FA tablet [10:0.4 mg (10 mg enalapril combined with 0.4 mg of FA, low FA group]; and 3 enalapril-FA tablet (10:0.8 mg, high FA group, once daily for 8 weeks. Results After 4 or 8 weeks of treatment, homocysteine concentrations were reduced across all genotypes and FA dosage groups, except in subjects with MTR 2756AG /GG genotype in the low FA group at week 4. However, compared to subjects with MTHFR 677CC genotype, homocysteine concentrations remained higher in subjects with CT or TT genotype in the low FA group (P P P = 0.005, but not in the low FA group (CC 9.9% vs. TT 11.2%, P = 0.989. Conclusions This study demonstrated that MTHFR C677T polymorphism can not only affect homocysteine concentration at baseline and post-FA treatment, but also can modify therapeutic responses to various dosages of FA supplementation.

New developments in transportation for research reactors

Energy Technology Data Exchange (ETDEWEB)

Mondanel, J.L. [Transnucleaire, F-75008 Paris (France)

1998-07-01

For more than 30 years, Transnucleaire has been performing safely a large number of national and international transports of radioactive material. Transnucleaire has also designed and supplied numerous packagings for all types of nuclear fuel cycle radioactive materials: for front-end and back-end products and for power and research reactors. Since the last meeting held in Bruges, Transnucleaire has been continuously involved in transportation activities for fresh and irradiated materials for research reactors. We are pleased to take the opportunity in this meeting to share with reactor operators, official bodies and other partners, the on-going developments in transportation and associated services. Special attention will be paid to the starting of transports of MTR spent fuel elements to the La Hague reprocessing plant where COGEMA offers reprocessing services on a long-term basis to reactors operators. Detailed information is provided on regulatory issues, which may affect transport activities: evolution of the regulations, real experiences of recent transportation and development of new packaging designs. Options and solutions will be proposed by Transnucleaire to improve the situation for continuation of national and international transports at an acceptable price whilst maintaining an ultimate level of safety (author)

Thermal simulations and tests in the development of a helmet transport spent fuel elements Research Reactor

International Nuclear Information System (INIS)

Saliba, R.; Quintana, F.; Márquez Turiello, R.; Furnari, J.C.; Pimenta Mourão, R.

2013-01-01

A packaging for the transport of irradiated fuel from research reactors was designed by a group of researchers to improve the capability in the management of spent fuel elements from the reactors operated in the region. Two half-scale models for MTR fuel were constructed and tested so far and a third one for both MTR and TRIGA fuels will be constructed and tested next. Four test campaigns have been carried out, covering both normal and hypothetical accident conditions of transportation. The thermal test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. In this paper both the numerical modelling and experimental thermal tests performed are presented and discussed. The cask is briefly described as well as the finite element model developed and the main adopted hypotheses for the thermal phenomena. The results of both numerical runs and experimental tests are discussed as a tool to validate the thermal modelling. The impact limiters, attached to the cask for protection, were not modelled. (author) [es

Analysis of a Neutronic Computational Model for the Core of Material Testing Reactor MTR by Using SQUID Code

International Nuclear Information System (INIS)

Al-Taweel, M.H.

2015-01-01

It is a conventional practice in the design of nuclear reactor to introduce calculation of hot points to determine spatial variation for energy generated and then determine power distribution.The study had been carried out for core of a reactor type (MTR) by the neutronic code SQUID. In this study, we replace the reflector of the reactor by H 2 O instead of D 2 O as originally the reactor designed.From the study we conclude that the reactor can operates safely, to make sure of that we calculate the multiplication factor where their values ranged from (1.0854) when all control rods are up to (1.001)when three control rods are up.Also the values of hot points were calculated and compared with French documents results with D 2 O as a reflector where the difference is (0.19%), and with light water as reflector instead of heavy water was calculated.For different cases according to control rod position , the values of hot point ranged between (0.46) to (1.64) in case all control rods are up also the values of the average power distributed on different fuel cells were calculated in case of light water as reflector firstly with three control rods are down and the maximum value (2.13*10 -2 Μw).Secondly in case offour control rods are down, the maximum value (1.925*10 -2 Μw) we notice almost coincidence between the neutron flux distribution through the core of reactor and in different positions of control rods

Siloe, Osiris, and the future perspective of swimming-pool reactors

International Nuclear Information System (INIS)

Chatoux, J.; Denielou, G.; Lerouge, B.

1964-01-01

Siloe and Osiris are two new general purpose research reactors of the 'Commissariat a l'energie Atomique'. Siloe, located within the 'Centre d'Etudes Nucleaires' of Grenoble is a swimming pool reactor of the same type as Melusine and Triton. It operates, at a nominal power of 15 MW thermal and has reached the peak power of 20 MW thermal with two thirds of its cooling system working. The fast flux above 1 MeV, which is maximum at the center of the core at 15 MW thermal is 1,2. 10 14 . The core, quite open, is downward cooled. Average specific power is 159 kW/l. Osiris is under construction at Saclay. Designed for 50 MW thermal, this reactor is upward cooled. The fast flux at the center of the core above 1 MeV is calculated to be 2, 5.10 14 . The average designed specific power is 280 kW/l. A fixed zircaloy gamma shield makes a box round the core. Future perspectives open to non-pressurised swimming-pool reactors are examined. Ways are suggested for neutronic; thermal and shielding modifications which make possible further improvements in the performances and economy of these devices. (authors) [fr

Preparations for the shipment of RA-3 reactor irradiated fuel

International Nuclear Information System (INIS)

Goldschmidt, Adrian; Novara, Oscar; Lafuente, Jose

2002-01-01

During the last quarter of 2000, in the Radioactive Waste Management Area of the Argentine National Commission of Atomic Energy (CNEA), located at Ezeiza Atomic Center (CAE), activities associated to the shipment of 207 MTR spent fuels containing high enrichment uranium were carried out within the Foreign Research Reactor/Domestic Research Reactor Receipt Program launched by the US Department of Energy (DOE). The MTR spent fuel shipped to Savannah River Site (SRS) was fabricated in Argentina with 90% enriched uranium of US origin and it was utilized in the operation of the research and radioisotope production reactor RA-3 from 1968 until 1987. After a cooling period at the reactor, the spent fuel was transferred to the Central Storage Facility (CSF) located in the waste management area of CAE for interim storage. The spent fuel (SF) inventory consisted of 166 standard assemblies (SA) and 41 control assemblies (CA). Basically, the activities performed were the fuel conditioning operations inside the storage facility (remote transference of the assemblies to the operation pool, fuel cropping, fuel re-identification, loading in transport baskets, etc.) conducted by CNEA. The loading of the filled baskets in the transport casks (NAC-LWT) by means of intermediate transfer systems and loaded casks final preparations were conducted by NAC personnel (DOE's contractor) with the support of CNEA personnel. (author)

SAFARI-1: 30 years of operation

International Nuclear Information System (INIS)

D'Arcy, A.J.; Niebuhr, H.W.; Procter, G.I.

1995-01-01

SAFARI-1, a 20 MW tank-in-pool type MTR, was commissioned in 1965. It today enjoys some benefits from having operated for 17 years at a quarter of its design power. Ageing technology, non-availability of spares and wear and tear are, however, beginning to take their toll, while changing licensing requirements and shifting focus in its utilisation are at the same time compelling the AEC to invest capital in renovation, refurbishment, backfitting and upgrading in several different areas. A brief look at the operating history of SAFARI-1 is followed by a discussion of some of the more important upgrades recently carried out, focussing particular attention on a redesign of the Fail-Free Power System. (orig.)

Internal and external hazards inside the containment in case of an emergency situation

Energy Technology Data Exchange (ETDEWEB)

Abdelhady, Amr [Atomic Energy Authoriy (Egypt). Nuclear Research Center

2017-11-15

The objective of this paper is to estimating radionuclide concentrations in air and radiological dose consequences to indoor workers in a containment of MTR open pool type reactor during emergency situations. A postulated core degradation accident causes fission products to release and the ventilation system will be converted automatically from the normal situation into emergency situation in order to purify the contaminated air by forcing it to pass through a group of filters. The study computes internal and external worker doses from inhalation and submersion in a finite cloud of contaminated air during the emergency mode. A radiological toolbox version 2 was used to evaluate the radiation dose levels inside the containment.

Internal and external hazards inside the containment in case of an emergency situation

International Nuclear Information System (INIS)

Abdelhady, Amr

2017-01-01

The objective of this paper is to estimating radionuclide concentrations in air and radiological dose consequences to indoor workers in a containment of MTR open pool type reactor during emergency situations. A postulated core degradation accident causes fission products to release and the ventilation system will be converted automatically from the normal situation into emergency situation in order to purify the contaminated air by forcing it to pass through a group of filters. The study computes internal and external worker doses from inhalation and submersion in a finite cloud of contaminated air during the emergency mode. A radiological toolbox version 2 was used to evaluate the radiation dose levels inside the containment.

Measurement of argon concentrations in a TRIGA Mark-III pool

Energy Technology Data Exchange (ETDEWEB)

Simms, R [California State University, Northridge, CA (United States)

1974-07-01

Argon-41, the principal radioactive effluent from a pool type reactor during normal operation, is produced by the {sup 40}A (n,{gamma}) reaction. The reactant, {sup 40}A, is introduced into the pool water by contact with the air. Reduction in radioactive argon release can be accomplished by reducing the concentration of dissolved {sup 40}A and retaining the {sup 41}A within the pool. However, little data were available concerning the mechanisms of argon introduction, production, retention, and release from a reactor pool. Experiments have therefore been performed at the Torrey Pines TRIGA Mark-III Reactor to develop techniques to sample dissolved argon and to provide data on argon concentrations in the pool for release modeling studies. Significant results for argon dissolved at different pool depths can only be obtained if the water samples are sealed at the point of collection. A special handling tool was developed to perform this remote operation. Pool samples were counted for {sup 41}A soon after collection with a NaI spectrometer. After allowing one day for decay of {sup 41}A, the concentration of {sup 40}A in the water sample was determined by neutron activation analysis. In each case, the 1.29 MeV gamma-ray peak of {sup 41}A was used. Interference from the 1.37 MeV {sup 24}Na peak was considered and its effect subtracted after determining {sup 24}Na content from the 2.75 MeV {sup 24}Na peak and a sodium standard. A Ge(Li) detector was tried and found to eliminate the problem, but it introduced an unacceptable geometrical effect dependent on bubble size within the sample bottles. Samples were taken from the 27 ft deep TRIGA pool at various locations. Results were obtained for samples taken on several different days along the same vertical line about 3-1/2 ft from the reactor centerline. Temperature measurements along this vertical traverse indicated a sharp temperature gradient at about 15 ft below the surface ({approx}6 ft above the top of the reactor). The

Combined UV treatment and ozonation for the removal of by-product precursors in swimming pool water

DEFF Research Database (Denmark)

Cheema, Waqas Akram; Kaarsholm, Kamilla Marie Speht; Andersen, Henrik Rasmus

2017-01-01

Both UV treatment and ozonation are used to reduce different types of disinfection by-products (DBPs) in swimming pools. UV treatment is the most common approach, as it is particularly efficient at removing combined chlorine. However, the UV treatment of pool water increases chlorine reactivity...

Hospital hydrotherapy pools treated with ultra violet light: bad bacteriological quality and presence of thermophilic Naegleria.

Science.gov (United States)

De Jonckheere, J. F.

1982-01-01

The microbiological quality of eight halogenated and two u.v.-treated hydrotherapy pools in hospitals was investigated. The microbiological quality of halogenated hydrotherapy pools was comparable to halogenated public swimming pools, although in some Pseudomonas aeruginosa and faecal pollution indicators were more frequent due to bad management. On the other hand u.v.-treated hydrotherapy pools had very bad microbiological quality. Apart from faecal pollution indicators, P. aeruginosa was present in very high numbers. Halogenated hydrotherapy pools were not highly contaminated with amoebae, and Naegleria spp. were never detected. On the other hand u.v.-treated pools contained very high numbers of thermophilic Naegleria. The Naegleria isolated were identified as N. lovaniensis, a species commonly found in association with N. fowleri. Isoenzyme analysis showed a different type of N. lovaniensis was present in each of two u.v.-treated pools. Images Plate 1 PMID:7061835

Efficient pooling designs for library screening

OpenAIRE

Bruno, William J.; Knill, Emanuel; Balding, David J.; Bruce, D. C.; Doggett, N. A.; Sawhill, W. W.; Stallings, R. L.; Whittaker, Craig C.; Torney, David C.

1994-01-01

We describe efficient methods for screening clone libraries, based on pooling schemes which we call ``random $k$-sets designs''. In these designs, the pools in which any clone occurs are equally likely to be any possible selection of $k$ from the $v$ pools. The values of $k$ and $v$ can be chosen to optimize desirable properties. Random $k$-sets designs have substantial advantages over alternative pooling schemes: they are efficient, flexible, easy to specify, require fewer pools, and have er...

Optimization of heat pump system in indoor swimming pool using particle swarm algorithm

Energy Technology Data Exchange (ETDEWEB)

Lee, Wen-Shing; Kung, Chung-Kuan [Department of Energy and Refrigerating Air-Conditioning Engineering, National Taipei University of Technology, 1, Section 3, Chung-Hsiao East Road, Taipei (China)

2008-09-15

When it comes to indoor swimming pool facilities, a large amount of energy is required to heat up low-temperature outdoor air before it is being introduced indoors to maintain indoor humidity. Since water is evaporated from the pool surface, the exhausted air contains more water and specific enthalpy. In response to this indoor air, heat pump is generally used in heat recovery for indoor swimming pools. To reduce the cost in energy consumption, this paper utilizes particle swarm algorithm to optimize the design of heat pump system. The optimized parameters include continuous parameters and discrete parameters. The former consists of outdoor air mass flow and heat conductance of heat exchangers; the latter comprises compressor type and boiler type. In a case study, life cycle energy cost is considered as an objective function. In this regard, the optimized outdoor air flow and the optimized design for heating system can be deduced by using particle swarm algorithm. (author)

Stable isotope research pool inventory

International Nuclear Information System (INIS)

1980-12-01

This report contains a listing of electromagnetically separated stable isotopes which are available for distribution within the United States for non-destructive research use from the Oak Ridge National Laboratory on a loan basis. This inventory includes all samples of stable isotopes in the Materials Research Collection and does not designate whether a sample is out on loan or in reprocessing

Afterlife of a Drop Impacting a Liquid Pool

Science.gov (United States)

Saha, Abhishek; Wei, Yanju; Tang, Xiaoyu; Law, Chung K.

2017-11-01

Drop impact on liquid pool is ubiquitous in industrial processes, such as inkjet printing and spray coating. While merging of drop with the impacted liquid surface is essential to facilitate the printing and coating processes, it is the afterlife of this merged drop and associated mixing which control the quality of the printed or coated surface. In this talk we will report an experimental study on the structural evolution of the merged droplet inside the liquid pool. First, we will analyze the depth of the crater created on the pool surface by the impacted drop for a range of impact inertia, and we will derive a scaling relation and the associated characteristic time-scale. Next, we will focus on the toroidal vortex formed by the moving drop inside the liquid pool and assess the characteristic time and length scales of the penetration process. The geometry of the vortex structure which qualitatively indicates the degree of mixedness will also be discussed. Finally, we will present the results from experiments with various viscosities to demonstrate the role of viscous dissipation on the geometry and structure formed by the drop. This work is supported by the Army Research Office and the Xerox Corporation.

Introduction to Safety Analysis Approach for Research Reactors

International Nuclear Information System (INIS)

Park, Suki

2016-01-01

The research reactors have a wide variety in terms of thermal powers, coolants, moderators, reflectors, fuels, reactor tanks and pools, flow direction in the core, and the operating pressure and temperature of the cooling system. Around 110 research reactors have a thermal power greater than 1 MW. This paper introduces a general approach to safety analysis for research reactors and deals with the experience of safety analysis on a 10 MW research reactor with an open-pool and open-tank reactor and a downward flow in the reactor core during normal operation. The general approach to safety analysis for research reactors is described and the design features of a typical open-pool and open-tank type reactor are discussed. The representative events expected in research reactors are investigated. The reactor responses and the thermal hydraulic behavior to the events are presented and discussed. From the minimum CHFR and the maximum fuel temperature calculated, it is ensured that the fuel is not damaged in the step insertion of reactivity by 1.8 mk and the failure of all primary pumps for the reactor with a 10 MW thermal power and downward core flow

The life-extension and upgrade program of the Tsing Hua Open-pool Reactor (THOR) and its research prospectives

International Nuclear Information System (INIS)

Kai, J.-J.

1992-01-01

The Tsing Hua Open-Pool Reactor (THOR) has been operated for thirty years. It is the regulations of the ROCAEC that any reactor shall be decommissioned after forty-year operation since the first fuel loading. Therefore, for extending the lifetime of THOR, it is necessary to have a life-extension program to be approved by the ROCAEC and also completed by the year of 1997. At the same time, for proceeding new research purposes, it is planed to upgrade the thermal power of THOR from 1 Wth up to 3 Wth and hopefully to reach the maximum thermal neutron flux of 5x10 13 n/cm 2 .s and the fast flux close to that order. New research directions involve (a) boron-captured neutron cancer therapy (BNCT) (b) small-angle neutron scattering (SANS). (author)

Method development and validation for simultaneous determination of IEA-R1 reactor’s pool water uranium and silicon content by ICP OES

Science.gov (United States)

Ulrich, J. C.; Guilhen, S. N.; Cotrim, M. E. B.; Pires, M. A. F.

2018-03-01

IPEN’s research reactor, IEA-R1, an open pool type research reactor moderated and cooled by light water. High quality water is a key factor in preventing the corrosion of the spent fuel stored in the pool. Leaching of radionuclides from the corroded fuel cladding may be prevented by an efficient water treatment and purification system. However, as a safety management policy, IPEN has adopted a water chemistry control which periodically monitors the levels of uranium (U) and silicon (Si) in the pool’s reactor, since IEA-R1 employs U3Si2-Al dispersion fuel. An analytical method was developed and validated for the determination of uranium and silicon by ICP OES. This work describes the validation process, in a context of quality assurance, including the parameters selectivity, linearity, quantification limit, precision and recovery.

Optimisation of the flow path in a conceptual pool type reactor under natural circulation with lead coolant

International Nuclear Information System (INIS)

Thiele, R.; Anglart, H.

2014-01-01

This contribution investigates the effects of a bypass flow blocking bottom plate and the influence of the heat transfer between the hot and cold leg in a small pool type reactor cooled through natural convection with lead coolant. The computations are carried out using 3D computational fluid dynamics, where small-detail parts, such as the core and heat exchangers are modeled using a porous media approach. The introduction of full conjugate heat transfer shows that the heat transfer between the hot and cold leg can deteriorate flow in the cold leg and lead to recirculation zones. These zones become even more pronounced with the introduction of a bottom plate, which on the other hand also increases the flow through the core and lowers the maximum temperature in the core by approximately 150 K. Based on the results, redesign suggestions for the bottom plate and the internal wall are made. (author)

Evaluation of filters in RSPCS (Reactor Service Pool Cooling System) and HWL (Hot Water Layer) in OPAL research reactor at ANSTO (Australian Nuclear Science and Technology Organization) using Gamma Spectrometry System and Liquid Scintillation Counter

Energy Technology Data Exchange (ETDEWEB)

Lee, Jim In; Foy, Robin; Jung, Seong Moon; Park, Hyeon Suk; Ye, Sung Joon [KAERI, Daejeon (Korea, Republic of)

2016-05-15

Australian Nuclear Science and Technology Organization(ANSTO) has a research reactor, OPAL (Open Pool Australian Lightwater reactor) which is a state-of-art 20 MW reactor for various purposes. In OPAL reactor, there are many kinds of radionuclides produced from various reactions in pool water and those should be identified and quantified for the safe use of OPAL. To do that, it is essential to check the efficiency of filters which are able to remove the radioactive substance from the reactor pool water. There are two main water circuits in OPAL which are RSPCS (Reactor Service Pool Cooling System) and HWL (Hot Water Layer) water circuits. The reactor service pool is connected to the reactor pool via a transfer canal and provides a working area and storage space for the spent and other materials. Also, HWL is the upper part of the reactor pool water and it minimize radiation dose rates at the pool surface. We collected water samples from these circuits and measured the radioactivity by using Gamma Spectrometry System (GSS) and Liquid Scintillation Counter (LSC) to evaluate the filters. We could evaluate the efficiency of filters in RSPCS and HWL in OPAL research reactor. Through the measurements of radioactivity using GSS and LSC, we could conclude that there is likely to be no alpha emitter in water samples, and for beta and gamma activity, there are very big differences between inlet and outlet results, so every filter is working efficiently to remove the radioactive substance.

Present status of research reactor decommissioning programme in Indonesia

International Nuclear Information System (INIS)

Suripto, A.; Mulyanto, N.

2002-01-01

At present Indonesia has 3 research reactors, namely the 30 MW MTR-type multipurpose reactor at Serpong Site, two TRIGA-type research reactors, the first one being 1 MW located at Bandung Site and the second one a small reactor of 100 kW at Yogyakarta Site. The TRIGA Reactor at the Bandung Site reached its first criticality at 250 kW in 1964, and then was operated at 1000 kW since 1971. In October 2000 the reactor power was successfully upgraded to 2 MW. This reactor has already been operated for 38 years. There is not yet any decision for the decommissioning of this reactor. However it will surely be an object for the near future decommissioning programme and hence anticipation for the above situation becomes necessary. The regulation on decommissioning of research reactor is already issued by the independent regulatory body (BAPETEN) according to which the decommissioning permit has to be applied by the BATAN. For Indonesia, an early decommissioning strategy for research reactor dictates a restricted re-use of the site for other nuclear installation. This is based on high land price, limited availability of radwaste repository site, and other cost analysis. Spent graphite reflector from the Bandung TRIGA reactor is recommended for a direct disposal after conditioning, without any volume reduction treatment. Development of human resources, technological capability as well as information flow from and exchange with advanced countries are important factors for the future development of research reactor decommissioning programme in Indonesia. (author)

The reprocessing of irradiated MTR fuel and the nuclear material accountancy - Dounreay, UKAEA

Energy Technology Data Exchange (ETDEWEB)

Barrett, T.R.; Harrison, R. [UKAEA, Nuclear Materials Control Dep., Dounreay (United Kingdom)

1997-07-01

The reprocessing of irradiated HEU MTR fuel is a sensible part of a safeguards regime. It brings together fuel otherwise scattered around the world into a concerted accountancy and protection arrangement. From a nuclear material accountants view the overall accountancy performance has been excellent. While investigations have been required for a few individual MUFs or trends, very little effort has required to be expended by the Nuclear Materials Control Department. That is a definition of a 'good plant'; it operates, measures and records input and output streams, and then the accountancy falls into place. As identified in this paper, the accountancy of the nuclear material processed in the plant is well founded and sound. The accountancy results over several decades confirm the adequacy of the safeguards arrangements at Dounreay. The processing makes good commercial sense and meets the current philosophy of recycling valuable resource materials. The risk of operating the full fuel cycle are less than those of extended storage of irradiated fuel at disparate diverse locations. The reprocessing at Dounreay accords with all of these philosophies. The assessed risk is at a very low level, well within published UK HSE 'tolerability of risk' regulatory guidelines. The impact of the operations are similarly low within the guidelines, for the operators and for the general public. (author)

The reprocessing of irradiated MTR fuel and the nuclear material accountancy - Dounreay, UKAEA

International Nuclear Information System (INIS)

Barrett, T.R.; Harrison, R.

1997-01-01

The reprocessing of irradiated HEU MTR fuel is a sensible part of a safeguards regime. It brings together fuel otherwise scattered around the world into a concerted accountancy and protection arrangement. From a nuclear material accountants view the overall accountancy performance has been excellent. While investigations have been required for a few individual MUFs or trends, very little effort has required to be expended by the Nuclear Materials Control Department. That is a definition of a 'good plant'; it operates, measures and records input and output streams, and then the accountancy falls into place. As identified in this paper, the accountancy of the nuclear material processed in the plant is well founded and sound. The accountancy results over several decades confirm the adequacy of the safeguards arrangements at Dounreay. The processing makes good commercial sense and meets the current philosophy of recycling valuable resource materials. The risk of operating the full fuel cycle are less than those of extended storage of irradiated fuel at disparate diverse locations. The reprocessing at Dounreay accords with all of these philosophies. The assessed risk is at a very low level, well within published UK HSE 'tolerability of risk' regulatory guidelines. The impact of the operations are similarly low within the guidelines, for the operators and for the general public. (author)

Multi-Locus Next-Generation Sequence Typing of DNA Extracted From Pooled Colonies Detects Multiple Unrelated Candida albicans Strains in a Significant Proportion of Patient Samples

Directory of Open Access Journals (Sweden)

Ningxin Zhang

2018-06-01

Full Text Available The yeast Candida albicans is an important opportunistic human pathogen. For C. albicans strain typing or drug susceptibility testing, a single colony recovered from a patient sample is normally used. This is insufficient when multiple strains are present at the site sampled. How often this is the case is unclear. Previous studies, confined to oral, vaginal and vulvar samples, have yielded conflicting results and have assessed too small a number of colonies per sample to reliably detect the presence of multiple strains. We developed a next-generation sequencing (NGS modification of the highly discriminatory C. albicans MLST (multilocus sequence typing method, 100+1 NGS-MLST, for detection and typing of multiple strains in clinical samples. In 100+1 NGS-MLST, DNA is extracted from a pool of colonies from a patient sample and also from one of the colonies. MLST amplicons from both DNA preparations are analyzed by high-throughput sequencing. Using base call frequencies, our bespoke DALMATIONS software determines the MLST type of the single colony. If base call frequency differences between pool and single colony indicate the presence of an additional strain, the differences are used to computationally infer the second MLST type without the need for MLST of additional individual colonies. In mixes of previously typed pairs of strains, 100+1 NGS-MLST reliably detected a second strain. Inferred MLST types of second strains were always more similar to their real MLST types than to those of any of 59 other isolates (22 of 31 inferred types were identical to the real type. Using 100+1 NGS-MLST we found that 7/60 human samples, including three superficial candidiasis samples, contained two unrelated strains. In addition, at least one sample contained two highly similar variants of the same strain. The probability of samples containing unrelated strains appears to differ considerably between body sites. Our findings indicate the need for wider surveys to

Observations of Cold Pool Properties during GoAmazon2014/5

Science.gov (United States)

Mayne, S. L.; Schumacher, C.; MacDonald, L.; Turner, D. D.

2017-12-01

Convectively generated cold pools are instrumental in both the development of the sub-cloud layer and the organization of deep convection. Despite this, analyses of cold pools in the tropics are constrained by a lack of observational data; insight into the phenomena therefore relies heavily on numerical models. GoAmazon2014/5, a 2-year DOE-sponsored field campaign centered on Manacapuru, Brazil in the central Amazon, provides a unique opportunity to characterize tropical cold pools and allows for the comparison of observational data with theoretical results from model cold pool simulations and parameterizations. This investigation analyzes radar, disdrometer, and profiler measurements at the DOE mobile facility site to study tropical cold pool characteristics. The Brazilian military (SIPAM) operational S-band radar in Manaus is used to provide a broad context of convective systems, while measurements from Parsivel disdrometers are used to assess drop-size distributions (DSDs) at the surface. A unique aspect of this research is the use of the Atmospheric Emitted Radiance Interferometer (AERI) instrument, which utilizes down-welling IR measurements to obtain vertical profiles of thermodynamic quantities such as temperature and water vapor in the lowest few km of the atmosphere. Combined with surface observations and sounding data, these datasets will result in a thorough investigation of the horizontal and vertical characteristics of cold pools over the tropical rain forest. Preliminary analyses of 20 events reveal a mean cold pool height of 220 m and a mean radius of approximately 8.5 km. The average cold pool experienced a temperature (specific humidity) decrease of approximately 1 K (0.4 g/kg) at the surface. The temperature decrease is consistent with modeling studies and limited observations from previous studies over the tropics. The small decrease in specific humidity is attributed to the high moisture content within the cold pools. AERI retrievals of

Nuclear data uncertainties propagation methods in Boltzmann/Bateman coupled problems: Application to reactivity in MTR

International Nuclear Information System (INIS)

Frosio, Thomas; Bonaccorsi, Thomas; Blaise, Patrick

2016-01-01

Highlights: • Hybrid methods are developed for uncertainty propagation. • These methods take into account the flux perturbation in the coupled problem. • We show that OAT and MC methods give coherent results, except for Pearson correlations. • Local sensitivity analysis is performed. - Abstract: A novel method has been developed to calculate sensitivity coefficients in coupled Boltzmann/Bateman problem for nuclear data (ND) uncertainties propagation on the reactivity. Different uncertainty propagation methodologies, such as One-At-a-Time (OAT) and hybrid Monte-Carlo/deterministic methods have been tested and are discussed on an actual example of ND uncertainty problem on a Material Testing Reactor (MTR) benchmark. Those methods, unlike total Monte Carlo (MC) sampling for uncertainty propagation and quantification (UQ), allow obtaining sensitivity coefficients, as well as Bravais–Pearson correlations values between Boltzmann and Bateman, during the depletion calculation for global neutronics parameters such as the effective multiplication coefficient. The methodologies are compared to a pure MC sampling method, usually considered as the “reference” method. It is shown that methodologies can seriously underestimate propagated variances, when Bravais–Pearson correlations on ND are not taken into account in the UQ process.

Selective separation of actinides and long-lived fission products from 1 AW MTR liquid waste: pilot plant tests part II

International Nuclear Information System (INIS)

Grossi, G.; Marrocchelli, A.; Pietrelli, L.; Calle, C.; Gili, M.; Luce, A.; Troiani, F.

1992-01-01

In Italy there are some 120 m 3 of liquid High-level radioactive Wastes coming from MTR, Candu and EPK River fuel elements reprocessing. These High-level radioactive wastes contain a large amount of chemicals and inert salts together with cesium, strontium and transuranium elements. Transuranium elements and strontium are separated from the inert salts by means of a selective precipitation while Cesium is adsorbed on synthetic zeolithes (AZE Process) or precipitated with sodium Tetraphenyl borate (NaTPB) (ATE process). The benchscale experiments have confirmed the feasibility of selective separation processes and have showed that decontamination efficiency for strontium, plutonium and cesium were, respectively, 100, 5000 and 1000. This second part of the CEC final report describes Searse pilot plant tests with cold experiments. 37 Refs.; 17 Figs.; 16 Tabs

Topography Battles Surface Texture: An Experimental Study of Pool-riffle Formation

Science.gov (United States)

Chartrand, S. M.; Hassan, M. A.; Jellinek, M.

2016-12-01

ingredients needed to drive pool-riffle formation, as well as formation of other types of gravel bedforms. We believe our work holds promise for application in identifying suitable conditions for pool-riffle construction, and natural maintenance over typical restoration project time frames.

Condensation of vapor bubble in subcooled pool

Science.gov (United States)

Horiuchi, K.; Koiwa, Y.; Kaneko, T.; Ueno, I.

2017-02-01

We focus on condensation process of vapor bubble exposed to a pooled liquid of subcooled conditions. Two different geometries are employed in the present research; one is the evaporation on the heated surface, that is, subcooled pool boiling, and the other the injection of vapor into the subcooled pool. The test fluid is water, and all series of the experiments are conducted under the atmospheric pressure condition. The degree of subcooling is ranged from 10 to 40 K. Through the boiling experiment, unique phenomenon known as microbubble emission boiling (MEB) is introduced; this phenomenon realizes heat flux about 10 times higher than the critical heat flux. Condensation of the vapor bubble is the key phenomenon to supply ambient cold liquid to the heated surface. In order to understand the condensing process in the MEB, we prepare vapor in the vapor generator instead of the evaporation on the heated surface, and inject the vapor to expose the vapor bubble to the subcooled liquid. Special attention is paid to the dynamics of the vapor bubble detected by the high-speed video camera, and on the enhancement of the heat transfer due to the variation of interface area driven by the condensation.

Establishment and validation of the model of molten pool in fast reactor

International Nuclear Information System (INIS)

Zhou Shufeng; Luo Rui; Wang Zhou; Shi Xiaobo; Yang Xianyong

2007-01-01

Running under the beyond design base accidental condition, sodium boiling and dry-out will soon be brought about in LMFBR. If not stopped timely, the fuel pins of the subassembly will be melt and broken to form a molten pool at the bottom of the subassembly. to present a reasonable analysis about the molten pool accident, a method of establishing model according to the mechanism is selected, by which an integral model of the molten pool is established. Validated on the three power groups of BF1 experiments which belong to the France SCARABEE series experimenters, the model shows good results. After compared with the models of GEYSER and BF2 experiments which had been validated before, some conclusions about mechanism of molten pool are derived. Moreover, through comparing the relative parameters such as the discharged heat and the increment of temperature etc., a reasonable analysis about the type of heat transfer is present, on the basis of which some conclusions are derived as well. (authors)

Stable isotope research pool inventory

International Nuclear Information System (INIS)

1984-03-01

This report contains a listing of electromagnetically separated stable isotopes which are available at the Oak Ridge National Laboratory for distribution for nondestructive research use on a loan basis. This inventory includes all samples of stable isotopes in the Research Materials Collection and does not designate whether a sample is out on loan or is in reprocessing. For some of the high abundance naturally occurring isotopes, larger amounts can be made available; for example, Ca-40 and Fe-56

Numerical modeling of sodium fire – Part II: Pool combustion and combined spray and pool combustion

International Nuclear Information System (INIS)

Sathiah, Pratap; Roelofs, Ferry

2014-01-01

Highlights: • A CFD based method is proposed for the simulation of sodium pool combustion. • A sodium evaporation based model is proposed to model sodium pool evaporation. • The proposed method is validated against sodium pool experiments of Newman and Payne. • The results obtained using the proposed method are in good agreement with the experiments. - Abstract: The risk of sodium-air reaction has received considerable attention after the sodium-fire accident in Monju reactor. The fires resulting from the sodium-air reaction can be detrimental to the safety of a sodium fast reactor. Therefore, predicting the consequences of a sodium fire is important from a safety point of view. A computational method based on CFD is proposed here to simulate sodium pool fire and understand its characteristics. The method solves the Favre-averaged Navier-Stokes equation and uses a non-premixed mixture fraction based combustion model. The mass transfer of sodium vapor from the pool surface to the flame is obtained using a sodium evaporation model. The proposed method is then validated against well-known sodium pool experiments of Newman and Payne. The flame temperature and location predicted by the model are in good agreement with experiments. Furthermore, the trends of the mean burning rate with initial pool temperature and oxygen concentration are captured well. Additionally, parametric studies have been performed to understand the effects of pool diameter and initial air temperature on the mean burning rate. Furthermore, the sodium spray and sodium pool combustion models are combined to simulate simultaneous spray and pool combustion. Simulations were performed to demonstrate that the combined code could be applied to simulate this. Once sufficiently validated, the present code can be used for safety evaluation of a sodium fast reactor

Numerical modeling of sodium fire – Part II: Pool combustion and combined spray and pool combustion

Energy Technology Data Exchange (ETDEWEB)

Sathiah, Pratap, E-mail: pratap.sathiah78@gmail.com [Shell Global Solutions Ltd., Brabazon House, Concord Business Park, Threapwood Road, Manchester M220RR (United Kingdom); Roelofs, Ferry, E-mail: roelofs@nrg.eu [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755ZG Petten (Netherlands)

2014-10-15

Highlights: • A CFD based method is proposed for the simulation of sodium pool combustion. • A sodium evaporation based model is proposed to model sodium pool evaporation. • The proposed method is validated against sodium pool experiments of Newman and Payne. • The results obtained using the proposed method are in good agreement with the experiments. - Abstract: The risk of sodium-air reaction has received considerable attention after the sodium-fire accident in Monju reactor. The fires resulting from the sodium-air reaction can be detrimental to the safety of a sodium fast reactor. Therefore, predicting the consequences of a sodium fire is important from a safety point of view. A computational method based on CFD is proposed here to simulate sodium pool fire and understand its characteristics. The method solves the Favre-averaged Navier-Stokes equation and uses a non-premixed mixture fraction based combustion model. The mass transfer of sodium vapor from the pool surface to the flame is obtained using a sodium evaporation model. The proposed method is then validated against well-known sodium pool experiments of Newman and Payne. The flame temperature and location predicted by the model are in good agreement with experiments. Furthermore, the trends of the mean burning rate with initial pool temperature and oxygen concentration are captured well. Additionally, parametric studies have been performed to understand the effects of pool diameter and initial air temperature on the mean burning rate. Furthermore, the sodium spray and sodium pool combustion models are combined to simulate simultaneous spray and pool combustion. Simulations were performed to demonstrate that the combined code could be applied to simulate this. Once sufficiently validated, the present code can be used for safety evaluation of a sodium fast reactor.

10 CFR 36.63 - Pool water purity.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Pool water purity. 36.63 Section 36.63 Energy NUCLEAR... § 36.63 Pool water purity. (a) Pool water purification system must be run sufficiently to maintain the conductivity of the pool water below 20 microsiemens per centimeter under normal circumstances. If pool water...

Stable isotope research pool inventory

International Nuclear Information System (INIS)

1982-01-01

This report contains a listing of electromagnetically separated stable isotopes which are available for distribution within the United States for nondestructive research use from the Oak Ridge National Laboratory on a loan basis. This inventory includes all samples of stable isotopes in the Material Research Collection and does not designate whether a sample is out on loan or in reprocessing. For some of the high abundance naturally occurring isotopes, larger amounts can be made available; for example, Ca-40 and Fe-56

Model of large pool fires

Energy Technology Data Exchange (ETDEWEB)

Fay, J.A. [Department of Mechanical Engineering, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)]. E-mail: jfay@mit.edu

2006-08-21

A two zone entrainment model of pool fires is proposed to depict the fluid flow and flame properties of the fire. Consisting of combustion and plume zones, it provides a consistent scheme for developing non-dimensional scaling parameters for correlating and extrapolating pool fire visible flame length, flame tilt, surface emissive power, and fuel evaporation rate. The model is extended to include grey gas thermal radiation from soot particles in the flame zone, accounting for emission and absorption in both optically thin and thick regions. A model of convective heat transfer from the combustion zone to the liquid fuel pool, and from a water substrate to cryogenic fuel pools spreading on water, provides evaporation rates for both adiabatic and non-adiabatic fires. The model is tested against field measurements of large scale pool fires, principally of LNG, and is generally in agreement with experimental values of all variables.

Model of large pool fires

International Nuclear Information System (INIS)

Fay, J.A.

2006-01-01

A two zone entrainment model of pool fires is proposed to depict the fluid flow and flame properties of the fire. Consisting of combustion and plume zones, it provides a consistent scheme for developing non-dimensional scaling parameters for correlating and extrapolating pool fire visible flame length, flame tilt, surface emissive power, and fuel evaporation rate. The model is extended to include grey gas thermal radiation from soot particles in the flame zone, accounting for emission and absorption in both optically thin and thick regions. A model of convective heat transfer from the combustion zone to the liquid fuel pool, and from a water substrate to cryogenic fuel pools spreading on water, provides evaporation rates for both adiabatic and non-adiabatic fires. The model is tested against field measurements of large scale pool fires, principally of LNG, and is generally in agreement with experimental values of all variables

Comparison of serum pools and oral fluid samples for detection of porcine circovirus type 2 by quantitative real-time PCR in finisher pigs

DEFF Research Database (Denmark)

Nielsen, Gitte Blach; Nielsen, Jens Peter; Haugegaard, John

2018-01-01

Porcine circovirus type 2 (PCV2) diagnostics in live pigs often involves pooled serum and/or oral fluid samples for group-level determination of viral load by quantitative real-time polymerase chain reaction (qPCR). The purpose of the study was to compare the PCV2 viral load determined by q......PCR of paired samples at the pen level of pools of sera (SP) from 4 to 5 pigs and the collective oral fluid (OF) from around 30 pigs corresponding to one rope put in the same pen. Pigs in pens of 2 finishing herds were sampled by cross-sectional (Herd 1) and cross-sectional with follow-up (Herd 2) study designs....... In Herd 1, 50 sample pairs consisting of SP from 4 to 5 pigs and OF from around 23 pigs were collected. In Herd 2, 65 sample pairs consisting of 4 (SP) and around 30 (OF) pigs were collected 4 times at 3-week intervals. A higher proportion of PCV2-positive pens (86% vs. 80% and 100% vs. 91%) and higher...

Neisseria gonorrhoeae Sequence Typing for Antimicrobial Resistance, a Novel Antimicrobial Resistance Multilocus Typing Scheme for Tracking Global Dissemination of N. gonorrhoeae Strains.

Science.gov (United States)

Demczuk, W; Sidhu, S; Unemo, M; Whiley, D M; Allen, V G; Dillon, J R; Cole, M; Seah, C; Trembizki, E; Trees, D L; Kersh, E N; Abrams, A J; de Vries, H J C; van Dam, A P; Medina, I; Bharat, A; Mulvey, M R; Van Domselaar, G; Martin, I

2017-05-01

A curated Web-based user-friendly sequence typing tool based on antimicrobial resistance determinants in Neisseria gonorrhoeae was developed and is publicly accessible (https://ngstar.canada.ca). The N. gonorrhoeae Sequence Typing for Antimicrobial Resistance (NG-STAR) molecular typing scheme uses the DNA sequences of 7 genes ( penA , mtrR , porB , ponA , gyrA , parC , and 23S rRNA) associated with resistance to β-lactam antimicrobials, macrolides, or fluoroquinolones. NG-STAR uses the entire penA sequence, combining the historical nomenclature for penA types I to XXXVIII with novel nucleotide sequence designations; the full mtrR sequence and a portion of its promoter region; portions of ponA , porB , gyrA , and parC ; and 23S rRNA sequences. NG-STAR grouped 768 isolates into 139 sequence types (STs) ( n = 660) consisting of 29 clonal complexes (CCs) having a maximum of a single-locus variation, and 76 NG-STAR STs ( n = 109) were identified as unrelated singletons. NG-STAR had a high Simpson's diversity index value of 96.5% (95% confidence interval [CI] = 0.959 to 0.969). The most common STs were NG-STAR ST-90 ( n = 100; 13.0%), ST-42 and ST-91 ( n = 45; 5.9%), ST-64 ( n = 44; 5.72%), and ST-139 ( n = 42; 5.5%). Decreased susceptibility to azithromycin was associated with NG-STAR ST-58, ST-61, ST-64, ST-79, ST-91, and ST-139 ( n = 156; 92.3%); decreased susceptibility to cephalosporins was associated with NG-STAR ST-90, ST-91, and ST-97 ( n = 162; 94.2%); and ciprofloxacin resistance was associated with NG-STAR ST-26, ST-90, ST-91, ST-97, ST-150, and ST-158 ( n = 196; 98.0%). All isolates of NG-STAR ST-42, ST-43, ST-63, ST-81, and ST-160 ( n = 106) were susceptible to all four antimicrobials. The standardization of nomenclature associated with antimicrobial resistance determinants through an internationally available database will facilitate the monitoring of the global dissemination of antimicrobial-resistant N. gonorrhoeae strains. © Crown copyright 2017.

Mathematical modelling and simulation of the thermal performance of a solar heated indoor swimming pool

Directory of Open Access Journals (Sweden)

Mančić Marko V.

2014-01-01

Full Text Available Buildings with indoor swimming pools have a large energy footprint. The source of major energy loss is the swimming pool hall where air humidity is increased by evaporation from the pool water surface. This increases energy consumption for heating and ventilation of the pool hall, fresh water supply loss and heat demand for pool water heating. In this paper, a mathematical model of the swimming pool was made to assess energy demands of an indoor swimming pool building. The mathematical model of the swimming pool is used with the created multi-zone building model in TRNSYS software to determine pool hall energy demand and pool losses. Energy loss for pool water and pool hall heating and ventilation are analyzed for different target pool water and air temperatures. The simulation showed that pool water heating accounts for around 22%, whereas heating and ventilation of the pool hall for around 60% of the total pool hall heat demand. With a change of preset controller air and water temperatures in simulations, evaporation loss was in the range 46-54% of the total pool losses. A solar thermal sanitary hot water system was modelled and simulated to analyze it's potential for energy savings of the presented demand side model. The simulation showed that up to 87% of water heating demands could be met by the solar thermal system, while avoiding stagnation. [Projekat Ministarstva nauke Republike Srbije, br. III 42006: Research and development of energy and environmentally highly effective polygeneration systems based on using renewable energy sources