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Sample records for in-pile thermocouple performance

  1. Long duration performance of high temperature irradiation resistant thermocouples

    International Nuclear Information System (INIS)

    Rempe, J.; Knudson, D.; Condie, K.; Cole, J.; Wilkins, S.C.

    2007-01-01

    Many advanced nuclear reactor designs require new fuel, cladding, and structural materials. Data are needed to characterize the performance of these new materials in high temperature, radiation conditions. However, traditional methods for measuring temperature in-pile degrade at temperatures above 1100 C degrees. To address this instrumentation need, the Idaho National Laboratory (INL) developed and evaluated the performance of a high temperature irradiation-resistant thermocouple that contains alloys of molybdenum and niobium. To verify the performance of INL's recommended thermocouple design, a series of high temperature (from 1200 to 1800 C) long duration (up to six months) tests has been initiated. This paper summarizes results from the tests that have been completed. Data are presented from 4000 hour tests conducted at 1200 and 1400 C that demonstrate the stability of this thermocouple (less than 2% drift). In addition, post test metallographic examinations are discussed which confirm the compatibility of thermocouple materials throughout these long duration, high temperature tests. (authors)

  2. Thermocouple

    International Nuclear Information System (INIS)

    Charlesworth, F.D.W.

    1983-01-01

    A thermocouple is provided by a cable of coaxial form with inner and outer conductors of thermocouple forming materials and with the conductors electrically joined together at one end of the cable to form the thermocouple junction. The inner and outer conductors are preferably of chromel and stainless steel respectively. (author)

  3. A method to eliminate the effect of radiation on thermocouple performance

    International Nuclear Information System (INIS)

    Ali, Fawaz; Lu Lixuan

    2007-01-01

    In-core temperature measurements are pivotal in maintaining nuclear reactors in a safe state of operation. Thermocouples serve as the liaison in ensuring this safe state. The realization of the thermocouple's full potential is hindered by the fact that thermocouples cannot be situated in areas with high radiation fields. Radiation has the potential of generating voltages in the thermocouple wires, hence producing an error in the temperature transmitter output. In this paper, a mathematical model is developed to quantify the effect that radiation from the Canada Deuterium Uranium (CANDU) Nuclear Power Plants (NPPs) has on the thermocouple temperature reading. Subsequently, a method to offset the effect of radiation on the thermocouple is proposed. Simulation is performed to verify the effectiveness of the proposed system

  4. Design and performance evaluation of a cryogenic condenser for an in-pile experiment

    Science.gov (United States)

    Graham, R. W.; Crum, R. J.; Hsu, Y.

    1972-01-01

    An apparatus was designed to enable in-pile irradiation of materials in liquid hydrogen at cryogenic temperatures. One of the principal components of this apparatus was a horizontal tube condenser. The performance of the condenser was evaluated by running a liquid-nitrogen prototype of the apparatus at heat loads comparable to or greater than those expected during the irradiation. The test showed that the condenser was capable of handling the design heat load and that the design procedure was sound.

  5. Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, A. J.; Haggard, DC; Herter, J. W.; Swank, W. D.; Knudson, D. L.; Cherry, R. S. [Idaho National Laboratory, P.O. Box 1625, MS 4112, Idaho Falls, ID, (United States); Scervini, M. [University of Cambridge, Department of Material Science and Metallurgy, 27 Charles Babbage Road, CB3 0FS, Cambridge, (United Kingdom)

    2015-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to be only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly

  6. Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

    Energy Technology Data Exchange (ETDEWEB)

    A. J. Palmer; DC Haggard; J. W. Herter; M. Scervini; W. D. Swank; D. L. Knudson; R. S. Cherry

    2011-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina

  7. Manufacturing and performance tests of in-pile creep measuring machine of zirconium alloys

    International Nuclear Information System (INIS)

    Choi, Y.; Kim, B. G.; Kang, Y. H.

    2000-01-01

    A mock-up of the in-pile creep test machine of zirconium alloys for HANARO was designed and manufactured, which performance tests were carried. The dimension of the in-pile creep machine is 55 mm in diameter and 700 mm in length for HANARO, respectively. Load is transferred to specimen by through the working mechanisms in which the contraction of bellows by gas pressure moves a yoke and an upper grip connected to a specimen, simultaneously. It was observed that the extension of the specimen mounted in grips was transferred to a linear voltage differential transformer perfectly by a yoke and a push rod in a bearing. The displacement of specimen with applied pressure was determined with the LVDT and a pressure gauge, respectively. Resultant stress-strain behaviors of the specimen was determined by the displacement-applied gas pressure curve, which showed similar values obtained with a standard tensile test machine

  8. Analyses with the FSTATE code: fuel performance in destructive in-pile experiments

    International Nuclear Information System (INIS)

    Bauer, T.H.; Meek, C.C.

    1982-01-01

    Thermal-mechanical analysis of a fuel pin is an essential part of the evaluation of fuel behavior during hypothetical accident transients. The FSTATE code has been developed to provide this required computational ability in situations lacking azimuthal symmetry about the fuel-pin axis by performing 2-dimensional thermal, mechanical, and fission gas release and redistribution computations for a wide range of possible transient conditions. In this paper recent code developments are described and application is made to in-pile experiments undertaken to study fast-reactor fuel under accident conditions. Three accident simulations, including a fast and slow ramp-rate overpower as well as a loss-of-cooling accident sequence, are used as representative examples, and the interpretation of STATE computations relative to experimental observations is made

  9. Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

    International Nuclear Information System (INIS)

    Obara, Toru; Sekimoto, Hiroshi

    1997-01-01

    The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors. (author)

  10. Study of In-Pile test facility for fast reactor safety research: performance requirements and design features

    Energy Technology Data Exchange (ETDEWEB)

    Nonaka, N.; Kawatta, N.; Niwa, H.; Kondo, S.; Maeda, K

    1996-12-31

    This paper describes a program and the main design features of a new in-pile safety facility SERAPH planned for future fast reactor safety research. The current status of R and D on technical developments in relation to the research objectives and performance requirements to the facility design is given.

  11. Thermocouple modeling

    International Nuclear Information System (INIS)

    Fryer, M.O.

    1984-01-01

    The temperature measurements provided by thermocouples (TCs) are important for the operation of pressurized water reactors. During severe inadequate core cooling incidents, extreme temperatures may cause type K thermocouples (TCs) used for core exit temperature monitoring to perform poorly. A model of TC electrical behavior has been developed to determine how TCs react under extreme temperatures. The model predicts the voltage output of the TC and its impedance. A series of experiments were conducted on a length of type K thermocouple to validate the model. Impedance was measured at several temperatures between 22 0 C and 1100 0 C and at frequencies between dc and 10 MHz. The model was able to accurately predict impedance over this wide range of conditions. The average percentage difference between experimental data and the model was less than 6.5%. Experimental accuracy was +-2.5%. There is a sriking difference between impedance versus frequency plots at 300 0 C and at higher temperatures. This may be useful in validating TC data during accident conditions

  12. Thermocouples for conditions of aggressive environments

    International Nuclear Information System (INIS)

    Blanc, J.Y.

    1988-01-01

    Two new kinds of thermocouples have been chosen for temperature measurements in the in-pile safety program for light water reactors performed in France. They must give fuel centerline or roc cladding temperatures and withstand steam oxidation between 1000 0 C and 1800 0 C or higher, under severe fuel damage conditions. We describe briefly both types, then we emphasize on improvements under way concerning the tungsten-rhenium legs, the hafnia insulation and the sheaths materials. Oxidation resistance is achieved mainly by silicides layers, but other possibilities are considered, such as iridium coatings. Some details of insulators manufacturing or sensor assembly are given, as well as other high temperature applications for these thermocouples

  13. Fabrication, characteristics, and in-pile performance of UO2 pellets prepared from dry route powder

    International Nuclear Information System (INIS)

    Chotard, A.; Ledac, A.; Bernardin, M.

    1991-01-01

    The dry route conversion process of UF 6 to sinterable UO 2 powder has been used in France on a large scale for more than 10 years for the fabrication of PWR fuels. Thus, our fabrication and irradiation experience relates to more than 10,000 tons of fuel. As everyone knows, the dry route conversion process only involves gas-gas and gas-solid reactions which present the advantage of producing very little contaminated wastes and no liquid effluents. Powders obtained by this process are characterized by: - a very high purity, - a low specific surface area (around 2 m 2 /g), therefore a high resistance to spontaneous oxidation, - a good compressibility, - a very high sinterability (.98% T.D.), - a very high reproducibility. This powder also shows a high fineness which leads to very homogeneous blends with additives like pore former, U 3 O 8 or Gd 2 O 3 . On the other hand this fineness requires a granulation step which is actually not a disadvantage since it allows to adjust the granulate size to optimize the filling of press dies and so as to guarantee a good stability of the pellet dimensions and density. This pelletizing process leads to pellets characterized by: - a good thermal stability (0.5% T.D. after 34 hours at 1700degC), - no open porosity, - low H 2 content (0,3 ppm), - an homogeneous microstructure (grain size and porosity). Such characteristics mean that the UO 2 pellets from dry route conversion present an excellent in pile behaviour for high burnup up to 58,000 MWd/MtU in commercial plant, with: - low fission gas release, - good dimensional stability (densification, swelling), of which examples and results of PIE are described in the paper. The qualities of the dry route conversion powder and its flexibility of use make it possible to consider adjustment of the pellet characteristics, mainly: density, grain size and pore size distribution for specific uses or performance upgrade. (orig.)

  14. W-1 Sodium Loop Safety Facility experiment centerline fuel thermocouple performance

    International Nuclear Information System (INIS)

    Meyers, S.C.; Henderson, J.M.

    1980-05-01

    The W-1 Sodium Loop Safety Facility (SLSF) experiment is the fifth in a series of experiments sponsored by the Department of Energy (DOE) as part of the National Fast Breeder Reactor (FBR) Safety Assurance Program. The experiments are being conducted under the direction of Argonne National Laboratory (ANL) and Hanford Engineering Development Laboratory (HEDL). The irradiation phase of the W-1 SLSF experiment was conducted between May 27 and July 20, 1979, and terminated with incipient fuel pin cladding failure during the final boiling transient. Experimental hardware and facility performed as designed, allowing completion of all planned tests and test objectives. This paper focuses on high temperature in-fuel thermocouples and discusses their development, fabrication, and performance in the W-1 experiment

  15. Evaluation of the in pile performance of boron containing fuel pellets

    International Nuclear Information System (INIS)

    Jeong, Gwanyoon; Sohn, Dongseong

    2012-01-01

    The world rare earth resource are heavily concentrated in certain area and if these natural resources are weaponized by a country, we may confront serious difficulty because rare earth element gadolinium(Gd) is used as burnable poison material in some nuclear power plants (NPP) in Korea. Gd is used as a neutron absorbing material in Gd 2 O 3 form and mixed with UO 2 When boron is used as burnable poison in nuclear fuel, in fuel pellets. The burnable poison mixed in the fuel pellets is called integral burnable absorber (BA) design which differentiates it from the old separate BA design. In the old separate BA design, boron(B) was used in borosilicate glass (PYREX) form and placed in guide tubes. With the development of the concern over the availability of rare earth material Gd, B is considered as a candidate material replacing Gd for the case when the rare earth material is weaponized. However the idea for new boron BA design is integral type because the integral type BA design has several benefits over the separate BA design, such as reduction of radioactive waste, more positions for BA location, etc. 10 B absorbs a neutron and produces helium by the following reaction: 10 B + n → 7 Li + 4 He The helium produced by the nuclear reaction may cause the increase of rod internal pressure and change the gap conductivity if the significant amount of helium gas is released to the gap between the pellet and the cladding. Thus, it is necessary to investigate the in-pile behaviors of B containing pellet. However, few experiment have been carried out so far on the behavior of in-pile produced helium in UO 2 fuel pellets, especially for the cases boron compound is mixed with UO 2 In this paper, we will evaluate the production and the release of helium depending on fuel. 10 B concentration in the fuel

  16. Cladding temperature measurement by thermocouples at preirradiated LWR fuel rod samples

    International Nuclear Information System (INIS)

    Leiling, W.

    1981-12-01

    This report describes the technique to measure cladding temperatures of test fuel rod samples, applied during the in-pile tests on fuel rod failure in the steam loop of the FR2 reactor. NiCr/Ni thermocouples with stainless steel and Inconel sheaths, respectively,of 1 mm diameter were resistance spot weld to the outside of the fuel rod cladding. For the pre-irradiated test specimens, welding had to be done under hot-cell conditions, i.e. under remote handling. In order to prevent the formation of eutectics between zirconium and the chemical elements of the thermocouple sheath at elevated temperatures, the thermocouples were covered with a platinum jacket of 1.4 mm outside diameter swaged onto the sheath in the area of the measuring junction. This thermocouple design has worked satisfactorily in the in-pile experiments performed in a steam atmosphere. Even in the heatup phase, in which cladding temperatures up to 1050 0 C were reached, only very few failures occured. This good performance is to a great part due to a careful control and a thorough inspection of the thermocouples. (orig.) [de

  17. Composite thermocouples

    International Nuclear Information System (INIS)

    Debeir, R.P.

    1975-01-01

    As a rule, a composite thermocouple is a thermocouple where one or more components (wires, sheath, insulation) differ in kind between the hot junction measurement point and the cold termination with ordinary cables going on to measurement instrumentation. Three categories of such thermocouples are discussed: composite thermocouples having in common the continuity of the thermoelement wires over complete length, and different sheaths and insulation for the high temperature and intermediate temperature parts; those with different thermoelement wires, sheaths, and insulators for the high and intermediate temperature parts; a third category includes the high temperature thermoelements insulated by Al 2 O 3 or BeO and sheathed with a refractory metal, and with the intermediate temperature part made of 2Cr-Al couples, MgO insulated, and stainless steel or inconel sheathed

  18. Studies of Nuclear Fuel Performance Using On-site Gamma-ray Spectroscopy and In-pile Measurements

    International Nuclear Information System (INIS)

    Matsson, Ingvar

    2006-01-01

    Presently there is a clear trend of increasing demands on in-pile performance of nuclear fuel. Higher target burnups, part length rods and various fuel additives are some examples of this trend. Together with an increasing demand from the public for even safer nuclear power utilisation, this implies an increased focus on various experimental, preferably non-destructive, methods to characterise the fuel. This thesis focuses on the development and experimental evaluation of such methods. In its first part, the thesis presents a method based on gamma-ray spectroscopy with germanium detectors that have been used at various power reactors in Europe. The aim with these measurements is to provide information about the thermal power distribution within fuel assemblies in order to validate core physics production codes. The early closure of the Barsebaeck 1 BWR offered a unique opportunity to perform such validations before complete depletion of burnable absorbers in Gd-rods had taken place. To facilitate the measurements, a completely submersible measuring system, LOKET, was developed allowing for convenient in-pool measurements to be performed. In its second part, the thesis describes methods that utilise in-pile measurements. These methods have been used in the Halden test-reactor for determination of fission gas release, pellet-cladding interaction studies and fuel development studies. Apart from the power measurements, the LOKET device has been used for fission gas release (FGR) measurements on single fuel rods. The significant reduction in fission gas release in the modern fuel designs, in comparison with older designs, has been demonstrated in a series of experiments. A FGR database covering a wide range of burnup, power histories and fuel designs has been compiled and used for fuel performance analysis. The fission gas release has been measured on fuel rods with average burnups well above 60 MWd/kgU. The comparison between core physics calculations (PHOENIX-4/POLCA

  19. Low drift type N thermocouples in out-of-pile advanced gas reactor mock-up test: metallurgical analysis

    International Nuclear Information System (INIS)

    Scervini, M.; Palmer, J.; Haggard, D.C.; Swank, W.D.

    2015-01-01

    Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. They are crucial for the control of current nuclear reactors and for the development of GEN IV reactors. In nuclear applications thermocouples are strongly affected by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. Previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of Nickel based thermocouples is limited to temperatures lower than 1000 deg. C due to drift related to phenomena other than nuclear irradiation. As part of a collaboration between Idaho National Laboratory (INL) and the University of Cambridge a variety of Type N thermocouples have been exposed at INL in an Advanced Gas Reactor mock-up test at 1150 deg. C for 2000 h, 1200 deg. C for 2000 h, 125 deg. C for 200 h and 1300 deg. C for 200 h, and later analysed metallurgically at the University of Cambridge. The use of electron microscopy allows to identify the metallurgical changes occurring in the thermocouples during high temperature exposure and correlate the time dependent thermocouple drift with the microscopic changes experienced by the thermoelements of different thermocouple designs. In this paper conventional Inconel 600 sheathed type N thermocouples and a type N using a customized sheath developed at the University of

  20. Low drift type N thermocouples in out-of-pile advanced gas reactor mock-up test: metallurgical analysis

    Energy Technology Data Exchange (ETDEWEB)

    Scervini, M. [University of Cambridge, Department of Materials Science and Metallurgy, 27 Charles Babbage Road, CB30FS Cambridge, (United Kingdom); Palmer, J.; Haggard, D.C.; Swank, W.D. [Idaho National Laboratory, Idaho Falls, ID 83415-3840, (United States)

    2015-07-01

    Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. They are crucial for the control of current nuclear reactors and for the development of GEN IV reactors. In nuclear applications thermocouples are strongly affected by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. Previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of Nickel based thermocouples is limited to temperatures lower than 1000 deg. C due to drift related to phenomena other than nuclear irradiation. As part of a collaboration between Idaho National Laboratory (INL) and the University of Cambridge a variety of Type N thermocouples have been exposed at INL in an Advanced Gas Reactor mock-up test at 1150 deg. C for 2000 h, 1200 deg. C for 2000 h, 125 deg. C for 200 h and 1300 deg. C for 200 h, and later analysed metallurgically at the University of Cambridge. The use of electron microscopy allows to identify the metallurgical changes occurring in the thermocouples during high temperature exposure and correlate the time dependent thermocouple drift with the microscopic changes experienced by the thermoelements of different thermocouple designs. In this paper conventional Inconel 600 sheathed type N thermocouples and a type N using a customized sheath developed at the University of

  1. Uniformity index measurement technology using thermocouples to improve performance in urea-selective catalytic reduction systems

    Science.gov (United States)

    Park, Sangki; Oh, Jungmo

    2018-05-01

    The current commonly used nitrogen oxides (NOx) emission reduction techniques employ hydrocarbons (HCs), urea solutions, and exhaust gas emissions as the reductants. Two of the primary denitrification NOx (DeNOx) catalyst systems are the HC-lean NOx trap (HC-LNT) catalyst and urea-selective catalytic reduction (urea-SCR) catalyst. The secondary injection method depends on the type of injector, injection pressure, atomization, and spraying technique. In addition, the catalyst reaction efficiency is directly affected by the distribution of injectors; hence, the uniformity index (UI) of the reductant is very important and is the basis for system optimization. The UI of the reductant is an indicator of the NOx conversion efficiency (NCE), and good UI values can reduce the need for a catalyst. Therefore, improving the UI can reduce the cost of producing a catalytic converter, which are expensive due to the high prices of the precious metals contained therein. Accordingly, measurement of the UI is an important process in the development of catalytic systems. Two of the commonly used methods for measuring the reductant UI are (i) measuring the exhaust emissions at many points located upstream/downstream of the catalytic converter and (ii) acquisition of a reductant distribution image on a section of the exhaust pipe upstream of the catalytic converter. The purpose of this study is to develop a system and measurement algorithms to measure the exothermic response distribution in the exhaust gas as the reductant passes through the catalytic converter of the SCR catalyst system using a set of thermocouples downstream of the SCR catalyst. The system is used to measure the reductant UI, which is applied in real-time to the actual SCR system, and the results are compared for various types of mixtures for various engine operating conditions and mixer types in terms of NCE.

  2. Study of thermocouples for control of high temperatures

    International Nuclear Information System (INIS)

    Villamayor, M.

    1966-12-01

    Previous works have shown that the tungsten-rhenium alloys thermocouples were a good instrument for control of high temperatures. From its, the author has studied the W/W 26 per cent and W 5 per cent Re/W 26 per cent Re french manufactured thermocouples and intended for control of temperatures in nuclear reactors until 2300 deg. C. In 'out-pile' study he determines the general characteristics of these thermocouples: average calibration curves, thermal shocks influence, response times, and alloys allowing the cold source compensation. The evolution of these thermocouples under thermal neutron flux has been determined by 'in-pile' study. The observations have led the author to propose a new type of thermocouples settled of molybdenum-columbium alloys. (author) [fr

  3. Out-of-pile and in-pile temperature noise investigations: a survey of methods results and models

    International Nuclear Information System (INIS)

    Dentico, G.; Giovannini, R.; Marseguerra, M.; Pacilio, N.; Taglienti, S.; Tosi, V.; Vigo, A.; Oguma, R.

    1982-01-01

    A review is given of the main results obtained from temperature noise measurements performed in out-of-pile sodium loops on fast fuel element mock-ups. Sources of data were thermocouples placed in the central axis of the channel downstream from the bundle end. Autoregressive moving average (ARMA) models have been applied to several temperature time series; the analysis shows that a simple ARMA (3, 2) model adequately accounts for the observed fluctuations. Finally, highlights of a heat transfer stochastic model are also reported together with a preliminary validation against in-pile experimental data. (author)

  4. In-pile Instrumentation Development

    International Nuclear Information System (INIS)

    Vermeeren, L.

    2005-01-01

    Advanced irradiations in research reactors require the on-line monitoring of crucial parameters like neutron fluxes, gamma dose rates, central fuel rod temperatures, fission gas release pressures and small geometry changes. Our activities in this field aim at a detailed understanding of the sensor behaviour in the irradiation conditions in order to extract reliable real-time information. The objectives of work performed by SCK-CEN are to study of the on-line in-pile measurement of gamma and neutron fluxes in real time and to investigate parasitic radiation-induced signals in instrumentation cables

  5. Thin film ceramic thermocouples

    Science.gov (United States)

    Gregory, Otto (Inventor); Fralick, Gustave (Inventor); Wrbanek, John (Inventor); You, Tao (Inventor)

    2011-01-01

    A thin film ceramic thermocouple (10) having two ceramic thermocouple (12, 14) that are in contact with each other in at least on point to form a junction, and wherein each element was prepared in a different oxygen/nitrogen/argon plasma. Since each element is prepared under different plasma conditions, they have different electrical conductivity and different charge carrier concentration. The thin film thermocouple (10) can be transparent. A versatile ceramic sensor system having an RTD heat flux sensor can be combined with a thermocouple and a strain sensor to yield a multifunctional ceramic sensor array. The transparent ceramic temperature sensor that could ultimately be used for calibration of optical sensors.

  6. Transmutation of Thermocouples in Thermal and Fast Nuclear Reactors

    International Nuclear Information System (INIS)

    Scervini, M.; Rae, C.; Lindley, B.

    2013-06-01

    Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. Their role is fundamental for the control of current nuclear reactors and for the development of the nuclear technology needed for the implementation of GEN IV nuclear reactors. When used for in-core measurements thermocouples are strongly affected not only by high temperatures, but also by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition in the thermoelements and, as a consequence, a time dependent drift in the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. In this work, undertaken as part of the European project METROFISSION, the change in composition occurring in irradiated thermocouples has been calculated using the software ORIGEN 2.2. Several thermocouples have been considered, including Nickel based thermocouples (type K and type N), Tungsten based thermocouples (W-5%Re vs W-26%Re and W- 3%Re vs W-25%Re), Platinum based thermocouples (type S and Platinum vs Palladium) and Molybdenum vs Niobium thermocouples. The transmutation induced by both thermal flux and fast flux has been calculated. Thermocouples undergo more pronounced transmutation in thermal fluxes rather than in fast fluxes, as the neutron cross section of an element is higher for thermal energies. Nickel based thermocouples have a minimal change in composition, while Platinum based and Tungsten based thermocouples experience a very significant transmutation. The use of coatings deposited on the sheath of a thermocouple has been considered as a mean to reduce the neutron flux the thermoelements inside the thermocouple sheath

  7. In-pile critical heat flux and post-dryout heat transfer measurements – A historical perspective

    Energy Technology Data Exchange (ETDEWEB)

    Groeneveld, D.C., E-mail: degroeneveld@gmail.com

    2017-06-15

    In the 1960s’ and 1970s’ Canada was a world leader in performing in-reactor heat transfer experiments on fuel bundles instrumented with miniature sheath thermocouples. Several Critical Heat Flux (CHF) and Post-CHF experiments were performed in Chalk River’s NRU and NRX reactors on water-cooled 3-, 18-, 19-, 21-, and 36-element fuel bundles. Most experiments were obtained at steady-state conditions, where the power was raised gradually from single-phase conditions up to the CHF and beyond. Occasionally, post-dryout temperatures up to 600 °C were maintained for several hours. In some tests, the fuel behaviour during loss-of-flow and blowdown transients was investigated – during these transients sheath temperatures could exceed 2000 °C. Because of the increasingly more stringent licensing requirements for in-pile heat transfer tests on instrumented fuel bundles, no in-pile CHF and post-dryout tests on fuel bundles have been performed anywhere in the world for the past 40 years. This paper provides details of these unique in-pile experiments and describes some of their heat transfer results.

  8. In pile AISI 316L. Low cycle fatigue. Final report

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.; Moons, F.

    1994-12-01

    In pile testing of the effect of neutron irradiation on the fatigue life of the reference material AISI 316L was performed in the framework of the European fusion technology program. The overall programme, carried out at SCK CEN (Mol,Belgium), exists of two instrumented rigs for low cycle fatigue testing, which were consecutively loaded in the BR-2 reactor during periods Jan (94) June (94) and Aug (94)-Dec(94). In each experiment, two identical samples were loaded by means of a pneumatically driven system. The samples were instrumented with thermocouples, strain gages, linear variable displacement transducers, and activation monitors. The experimental conditions are given. Type of fatigue test: load controlled, symmetric, uniaxial, triangular wave shape; stress range: about 580 MPa; sample shape: hourglass, diameter 3.2 mm, radius 12.5 mm; environment: NaK (peritectic); temperature: 250 C; maximum dpa value up to fracture: 1.7. Two of four samples were broken (one in each experiment) after having experienced 17 419 respectively 11 870 stress cycles. These new data points confirm earlier results from pile fatigue tests: irradiation causes no degradation of fatigue life of AISI 316L steel, at least for the parameters corresponding to these experiments

  9. Low Drift Type N Thermocouples for Nuclear Applications

    International Nuclear Information System (INIS)

    Scervini, M.; Rae, C.

    2013-06-01

    Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. They are crucial for the control of current nuclear reactors and for the development of GEN IV reactors. In nuclear applications thermocouples are strongly affected by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. Previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of Nickel based thermocouples is limited to temperatures lower than 1000 deg. C due to drift related to phenomena other than nuclear irradiation. In this work, undertaken as part of the European project METROFISSION, the drift of type N thermocouples has been investigated in the temperature range 600-1300 deg. C. The approach of this study is based on the attempt to separate the contributions of each thermo-element to drift. In order to identify the dominant thermo-element for drift, the contributions of both positive (NP) and negative (NN) thermo-elements to the total drift of 3.2 mm diameter MIMS thermocouples have been measured in each drift test using a pure Pt thermo-element as a reference. Conventional Inconel-600 sheathed type N thermocouples have been compared with type N thermocouples sheathed in a new alloy. At temperatures higher than 1000 deg. C conventional Inconel600 sheathed type N thermocouples can experience a

  10. Technological improvements to high temperature thermocouples for nuclear reactor applications

    International Nuclear Information System (INIS)

    Schley, R.; Leveque, J.P.

    1980-07-01

    The specific operating conditions of thermocouples in nuclear reactors have provided an incentive for further advances in high temperature thermocouple applications and performance. This work covers the manufacture and improvement of existing alloys, the technology of clad thermocouples, calibration drift during heat treatment, resistance to thermal shock and the compatibility of insulating materials with thermo-electric alloys. The results lead to specifying improved operating conditions for thermocouples in nuclear reactor media (pressurized water, sodium, uranium oxide) [fr

  11. Study of thermocouples for control of high temperatures; Etude de thermocouples pour le reperage des hautes temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Villamayor, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Faculte des Sciences de l' Universite de Lyon - 69 (France)

    1967-07-01

    Previous works have shown that the tungsten-rhenium alloys thermocouples were a good instrument for control of high temperatures. From its, the author has studied the W/W 26 per cent and W 5 per cent Re/W 26 per cent Re french manufactured thermocouples and intended for control of temperatures in nuclear reactors until 2300 deg. C. In 'out-pile' study he determines the general characteristics of these thermocouples: average calibration curves, thermal shocks influence, response times, and alloys allowing the cold source compensation. The evolution of these thermocouples under thermal neutron flux has been determined by 'in-pile' study. The observations have led the author to propose a new type of thermocouples settled of molybdenum-columbium alloys. (author) [French] Des travaux anterieurs ont montre que les thermocouples des alliages tungstene-rhenium etaient susceptibles de reperer avec precision des hautes temperatures. A partir de la, l'auteur a etudie las thermocouples W/W 26 pour cent Re et W 5 pour cent Re/W 26 pour cent Re de fabrication francaise et destines au controle des temperatures dans les reacteurs nucleaires, jusqu'a 2300 deg. C Dans l'etude 'hors-pile' il a determine les caracteristiques generales de ces thermocouples: courbes d'etalonnage moyen, influence des chocs thermiques, temps de reponse, et alliages assurant la compensation de soudure froide. L'etude 'en-pile' a permis de rendre compte de l'evolution de ces thermocouples sous flux neutroniques. Les phenomenes observes ont conduit l'auteur a proposer un nouveau type de thermocouples constitues d'alliages molybdene-niobium. (auteur)

  12. Travelling gradient thermocouple calibration

    International Nuclear Information System (INIS)

    Broomfield, G.H.

    1975-01-01

    A short discussion of the origins of the thermocouple EMF is used to re-introduce the idea that the Peltier and Thompson effects are indistinguishable from one another. Thermocouples may be viewed as devices which generate an EMF at junctions or as integrators of EMF's developed in thermal gradients. The thermal gradient view is considered the more appropriate, because of its better accord with theory and behaviour, the correct approach to calibration, and investigation of service effects is immediately obvious. Inhomogeneities arise in thermocouples during manufacture and in service. The results of travelling gradient measurements are used to show that such effects are revealed with a resolution which depends on the length of the gradient although they may be masked during simple immersion calibration. Proposed tests on thermocouples irradiated in a nuclear reactor are discussed

  13. Operating problems of the thermocouples in VVER

    International Nuclear Information System (INIS)

    Timonin, A.S.

    1997-01-01

    In WWER reactors, the coolant temperature at the outlet of the majority of fuel assemblies is measured with chromel-alumel cable thermocouples. The components of systematic errors in temperature measurements are discussed. Errors due to calibration drift can be avoided by periodical calibrations performed during the heating and hot test runs after reactor refueling. Errors due to radiation heating and response time can be estimated and thus eliminated. Errors due to flow stratification of the coolant can also be eliminated by an estimation of correction factors. The effects of the aging of the thermocouples are also discussed. The removal of thermocouples from their coverings for replacement presents some difficulties, which thus determine the service life of the thermocouples. (A.K.)

  14. An experimental study of the effect of external thermocouples on rewetting during reflood

    International Nuclear Information System (INIS)

    Shires, G.L.; Butcher, A.A.; Carpenter, B.G.; McCune, D.S.; Pearson, K.G.

    1980-04-01

    The validation of computer codes used for PWR safety assessment often depends upon experiments carried out with either real fuel pins or electrically heated fuel pin simulators. In some cases, and this applies particularly to in-pile tests, temperatures are measured by means of sheathed thermocouples attached externally to the pins and this raises the question of the possible effect of such thermocouples on the two phase hydraulics and heat transfer which are being studied. This paper describes the experiments which subjected two realistic fuel pin simulators, one with and one without external thermocouples, to identical bottom flooding conditions. They demonstrate very clearly that external thermocouples act as preferential rewetting sites and thereby increase the rate of propagation of the quench front. In the view of the authors of this paper the facts described raise serious doubts about the validity of rewetting data obtained from experiments employing external thermocouples. (U.K.)

  15. Temperature measurements by thermocouples

    International Nuclear Information System (INIS)

    Liermann, J.

    1975-01-01

    The measurement of a temperature (whatever the type of transducer used) raises three problems: the choice of transducer; where it should be placed; how it should be fixed and protected. These are the three main points examined, after a brief description of the most commonly used thermocouples [fr

  16. Nuclear heating measurements by in-pile calorimetry: prospective works for a microsensor design

    Energy Technology Data Exchange (ETDEWEB)

    Reynard-Carette, C.; Carette, M.; Aguir, K.; Bendahan, M.; Fiorido, T. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Lyoussi, A.; Fourmentel, D.; Villard, J.F. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 (France); Barthes, M.; Lanzetta, F.; Layes, G.; Vives, S. [FEMTO-ST, UMR 6174, Departement ENERGIE, Universite de Franche-Comte, 90000, Belfort (France)

    2015-07-01

    Since 2009 works have been performed in the framework of joint research programs between CEA and Aix-Marseille University. The main aim of these programs is to design and develop in-pile instrumentations, advanced calibration procedure and accurate measurement methods in particular for the new Material Testing Reactor (MTR) under construction in the South of France: Jules Horowitz Reactor (JHR). One major sensor is a specific radiometric calorimeter, which was studied out-of-pile from a thermal point of view and in-pile during irradiation campaigns. This sensor type is dedicated to measurements of nuclear heating (energy deposition rate per mass unit induced by interactions between nuclear rays and matter) inside experimental channels of MTRs. This kind of in-pile calorimeter corresponds to heat flux calorimeter exchanging with the external cooling fluid. This thermal running mode allows the establishment of steady thermal conditions inside the sensor to carry out online continuous measurements inside the reactor (core or reflector). Two main types of calorimeters exist. The first type consists of a single cell calorimeter. It is divided into a sample of material to be tested and a jacket instrumented with two thermocouples or a single thermocouple (Gamma Thermometer). The second, called a differential calorimeter, is composed of two superposed twin cells (a measurement cell containing a sample of material, and a reference cell to remove the heating of the cell body) instrumented with four thermocouples and two electrical heaters. Contrary to a single-cell calorimeter, a differential calorimeter allows the compensation of the parasite nuclear heating of the sensor body or jacket. Moreover, it possesses interesting advantages: thanks to the heaters embedded in the cells, three different measurement methods can be applied during irradiations to quantify nuclear heating. The first one is based on the use of out-of-pile calibration curves obtained by generating a heat

  17. Acoustic sensor for in-pile fuel rod fission gas release measurement

    International Nuclear Information System (INIS)

    Fourmentel, D.; Villard, J. F.; Ferrandis, J. Y.; Augereau, F.; Rosenkrantz, E.; Dierckx, M.

    2009-01-01

    We have developed a specific acoustic sensor to improve the knowledge of fission gas release in Pressurized Water Reactor (PWR) fuel rods when irradiated in materials testing reactors. In order to perform experimental programs related to the study of the fission gas release kinetics, the CEA (French Nuclear Energy Commission) acquired the ability to equip a pre-irradiated PWR fuel rod with three sensors, allowing the simultaneous on-line measurements of the following parameters: - fuel temperature with a centre-line thermocouple type C, - internal pressure with a specific counter-pressure sensor, - fraction of fission gas released in the fuel rod with an innovative acoustic sensor. The third detector is the subject of this paper. This original acoustic sensor has been designed to measure the molar mass and pressure of the gas contained in the fuel rod plenum. For in-pile instrumentation, the fraction of fission gas, such as Krypton and Xenon, in Helium, can be deduced online from this measurement. The principle of this acoustical sensor is the following: a piezoelectric transducer generates acoustic waves in a cavity connected to the fuel rod plenum. The acoustic waves are propagated and reflected in this cavity and then detected by the transducer. The data processing of the signal gives the velocity of the acoustic waves and their amplitude, which can be related respectively to the molar mass and to the pressure of the gas. The piezoelectric material of this sensor has been qualified in nuclear conditions (gamma and neutron radiations). The complete sensor has also been specifically designed to be implemented in materials testing reactors conditions. For this purpose some technical points have been studied in details: - fixing of the piezoelectric sample in a reliable way with a suitable signal transmission, - size of the gas cavity to avoid any perturbation of the acoustic waves, - miniaturization of the sensor because of narrow in-pile experimental devices

  18. Performance of core exit thermocouple for PWR accident management action in vessel top break LOCA simulation experiment at OECD/NEA ROSA project

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    2009-01-01

    Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary side in case of core temperature excursion. Test 6-1 is first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reason of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection. (author)

  19. Thermocouple evaluation model and evaluation of chromel--alumel thermocouples for High-Temperature Gas-Cooled Reactor applications

    International Nuclear Information System (INIS)

    Washburn, B.W.

    1977-03-01

    Factors affecting the performance and reliability of thermocouples for temperature measurements in High-Temperature Gas-Cooled Reactors are investigated. A model of an inhomogeneous thermocouple, associated experimental technique, and a method of predicting measurement errors are described. Error drifts for Type K materials are predicted and compared with published stability measurements. 60 references

  20. Noncontacting Measurement With a Thermocouple

    Science.gov (United States)

    Weatherill, W. T.; Schoreder, C. J.; Freitag, H. J.

    1986-01-01

    Tentlike covering brings thermocouple to within few degrees of surface temperature. Technique originally developed for measuring surface temperature of quartz fabric under radiant heating requires no direct contact with heated surface. Technique particularly useful when measuring surface temperatures of materials damaged if thermocouple or other temperature sensor attached.

  1. New trends in pile safety instrumentation

    International Nuclear Information System (INIS)

    Furet, J.

    1961-01-01

    This report addresses the protection of nuclear piles against damages due to operation incidents. The author discusses the current trends in the philosophy of safety of atomic power piles, identifies the parameters which define safety systems, presents tests to be performed on safety chains, comments the relationship between safety and the decrease of the number of pile inadvertent shutdowns, discusses the issues of instrument failures and chain multiplicity, comments the possible improvement of the operation of elements which build up safety chains (design simplification, development of semiconductors, replacement of electromechanical relays by static relays), the role of safety logical computers and the development of automatics in pile safety, presents automatic control as a safety factor (example of automatic start-up), and finally comments the use of fuses

  2. PWR thermocouple mechanical sealing structure

    International Nuclear Information System (INIS)

    Shen Qiuping; He Youguang

    1991-08-01

    The PWR in-core temperature detection device, which is one of measures to insure reactor safety operation, is to monitor and diagnose reactor thermal power output and in-core power distribution. The temperature detection device system uses thermocouples as measuring elements with stainless steel protecting sleeves. The thermocouple has a limited service time and should be replaced after its service time has reached. A new sealing device for the thermocouples of reactor in-core temperature detection system has been developed to facilitate replacement. The structure is complete tight under high temperature and pressure without any leakage and seepage, and easy to be assembled or disassembled in radioactive environment. The device is designed to make it possible to replace the thermocouple one by one if necessary. This is a new, simple and practical structure

  3. AGR-1 Thermocouple Data Analysis

    International Nuclear Information System (INIS)

    Einerson, Jeff

    2012-01-01

    simulation results (Chapter 3). The statistics-based simulation-aided experimental control procedure described for the future AGR tests is developed and demonstrated in Chapter 4. The procedure for controlling the target fuel temperature (capsule peak or average) is based on regression functions of thermocouple readings and other relevant parameters and accounting for possible changes in both physical and thermal conditions and in instrument performance.

  4. AGR-1 Thermocouple Data Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jeff Einerson

    2012-05-01

    simulation results (Chapter 3). The statistics-based simulation-aided experimental control procedure described for the future AGR tests is developed and demonstrated in Chapter 4. The procedure for controlling the target fuel temperature (capsule peak or average) is based on regression functions of thermocouple readings and other relevant parameters and accounting for possible changes in both physical and thermal conditions and in instrument performance.

  5. Program of in-pile IASCC testing under the simulated actual plant condition. Development of technique for in-pile IASCC initiation test in JMTR

    International Nuclear Information System (INIS)

    Ugachi, Hirokazu; Tsukada, Takashi; Kaji, Yoshiyuki; Nagata, Nobuaki; Dozaki, Koji; Takiguchi, Hideki

    2003-01-01

    Irradiation assisted stress corrosion cracking (IASCC) is caused by the synergistic effects of neutron irradiation, stress and corrosion by high temperature water. It is, therefore, essential to perform in-pile SCC tests, which are material tests under the conditions simulating those of actual LWR operation, in order to clarify the precise mechanism of the phenomenon, though mainly out-of-pile SCC tests for irradiated materials have been carried out in this research field. There are, however, many difficulties to perform in-pile SCC tests. Performing in-pile SCC tests, essential key techniques must be developed. Hence as a part of development of the key techniques for in-pile SCC tests, we have embarked on development of the test technique which enables us to obtain the information concerning the effect of such parameters as applied stress level, water chemistry, irradiation conditions, etc. on the crack initiation behavior. Although it is difficult to detect the crack initiation in in-pile SCC tests, the crack initiation can be evaluated by the detection of specimen rupture if the cross section area of the specimen is small enough. Therefore, we adopted the uniaxial constant loading (UCL) test with small tensile specimens. This paper will describe the current status of the development of several techniques for in-pile SCC initiation tests in JMTR and the results of the performance tests of the designed testing unit using the out-of-pile loop facility. (author)

  6. Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

    International Nuclear Information System (INIS)

    Daw, J.E.; Knudson, D.L.; Villard, J.F.; Liothin, J.; Destouches, C.; Rempe, J.L.; Matheron, P.; Lambert, T.

    2015-01-01

    Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physical property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were

  7. Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

    Energy Technology Data Exchange (ETDEWEB)

    Daw, J.E.; Knudson, D.L. [Idaho National Laboratory, Idaho Falls, ID 83415, (United States); Villard, J.F.; Liothin, J.; Destouches, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Rempe, J.L. [Rempe and Associates, LLC, Idaho Falls, ID, 83404 (United States); Matheron, P. [CEA, DEN, DEC, Uranium Fuels Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Lambert, T. [CEA, DEN, DEC, Innovative Fuel Design and Irradiation Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France)

    2015-07-01

    Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physical property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were

  8. Temperature monitoring device and thermocouple assembly therefor

    Science.gov (United States)

    Grimm, Noel P.; Bauer, Frank I.; Bengel, Thomas G.; Kothmann, Richard E.; Mavretish, Robert S.; Miller, Phillip E.; Nath, Raymond J.; Salton, Robert B.

    1991-01-01

    A temperature monitoring device for measuring the temperature at a surface of a body, composed of: at least one first thermocouple and a second thermocouple; support members supporting the thermocouples for placing the first thermocouple in contact with the body surface and for maintaining the second thermocouple at a defined spacing from the body surface; and a calculating circuit connected to the thermocouples for receiving individual signals each representative of the temperature reading produced by a respective one of the first and second thermocouples and for producing a corrected temperature signal having a value which represents the temperature of the body surface and is a function of the difference between the temperature reading produced by the first thermocouple and a selected fraction of the temperature reading provided by the second thermocouple.

  9. Development of in-pile test and evaluation technology

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Yung Hwan; Park, Jong Man; Joo, Kee Nam; Park, Duk Keun; Park, Se Jin; Oh, Jong Myung; Kim, Tae Ryong; Park Jin Suk; Lee, Jae Han [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-08-01

    To develop the in-pile test and evaluation technologies using KMRR, basic design of instrumented capsule and auxiliary system for material irradiation test and the related studies are performed. First, reactor and test hole characteristics are summarized, and conceptual design requirements of capsule to KMRR are reviewed. And fundamental principles and criteria for the instrumented capsule design are summarized. Basic design and analysis of instrumented capsule are performed, and design of capsule supporting system are also performed and structural integrity of the system is analyzed. Based on the prior studies, test mock-ups are designed and manufactured, and thermohydraulic and vibration tests are prepared. And, as in-pile test evaluation technologies, KMRR neutron dosimetry and mechanical tests related to material irradiation are investigated. 67 figs, 30 tabs, 41 refs. (Author).

  10. Spatial filtring and thermocouple spatial filter

    International Nuclear Information System (INIS)

    Han Bing; Tong Yunxian

    1989-12-01

    The design and study on thermocouple spatial filter have been conducted for the flow measurement of integrated reactor coolant. The fundamental principle of spatial filtring, mathematical descriptions and analyses of thermocouple spatial filter are given

  11. Stability Studies of a New Design Au/Pt Thermocouple Without a Strain Relieving Coil

    Science.gov (United States)

    Jahan, Ferdouse; Ballico, Mark

    2007-12-01

    The performance of a simple, new design Au/Pt thermocouple developed by NMIA is assessed. This thermocouple is proposed as a more accurate replacement, over the temperature range from 0 to 1,000°C, for the commonly used Type R and S industrial transfer standards, in a robust form familiar to industrial calibration laboratories. Due to the significantly different thermal expansions of the Au and Pt thermoelements, reported designs of the Au/Pt thermocouple incorporate a strain-relieving coil or bridge at the thermocouple junction. As the strain relieving coil is mechanically delicate, these thermocouples are usually mounted in a protective quartz tube assembly, like a standard platinum resistance thermometer (SPRT). Although providing uncertainties at the mK level, they are more delicate than the commonly used Type R and S thermocouples. A new and simple design of the Au/Pt thermocouple was developed in which the differential thermal expansion between Au and Pt is accommodated in the thermocouple leads, facilitated by a special head design. The resulting thermocouple has the appearance and robustness of the traditional Type R and S thermocouples, while retaining stability better than 10 mK up to 961°C. Three thermocouples of this design were calibrated at fixed points and by comparison to SPRTs in a stirred salt bath. In order to assess possible impurity migration, strain effects, and mechanical robustness, sequences of heat treatment up to a total of 500 h together with over 50 thermal cycles from 900°C to ambient were performed. The effect of these treatments on the calibration was assessed, demonstrating the sensors to be robust and stable to better than 10 mK. The effects on the measured inhomogeneity of the thermocouple were assessed using the NMIA thermocouple scanning bath.

  12. A comprehensive survey of thermoelectric homogeneity of commonly used thermocouple types

    Science.gov (United States)

    Machin, Jonathan; Tucker, Declan; Pearce, Jonathan V.

    2018-06-01

    Thermocouples are widely used as temperature sensors in industry. The electromotive force generated by a thermocouple is produced in a temperature gradient and not at the thermocouple tip. This means that the thermoelectric inhomogeneity represents one of the most important contributions to the overall measurement uncertainty associated with thermocouples. To characterise this effect, and to provide some general recommendations concerning the magnitude of this contribution to use when formulating uncertainty analyses, a comprehensive literature survey has been performed. Significant information was found for Types K, N, R, S, B, Pt/Pd, Au/Pt and various other Pt/Rh thermocouples. In the case of Type K and N thermocouples, the survey has been augmented by a substantial amount of data based on calibrations of mineral-insulated, metal-sheathed thermocouple cable reels from thermocouple manufacturers. Some general conclusions are drawn and outline recommendations given concerning typical values for the uncertainty arising from thermoelectric inhomogeneity for the most widely used thermocouple types in the as-new state. It is stressed that these recommendations should only be heeded when individual homogeneity measurements are not possible. It is also stressed that the homogeneity can deteriorate rapidly during use, particularly for base metal thermocouples.

  13. Characterization of a Method for Inverse Heat Conduction Using Real and Simulated Thermocouple Data

    Science.gov (United States)

    Pizzo, Michelle E.; Glass, David E.

    2017-01-01

    It is often impractical to instrument the external surface of high-speed vehicles due to the aerothermodynamic heating. Temperatures can instead be measured internal to the structure using embedded thermocouples, and direct and inverse methods can then be used to estimate temperature and heat flux on the external surface. Two thermocouples embedded at different depths are required to solve direct and inverse problems, and filtering schemes are used to reduce noise in the measured data. Accuracy in the estimated surface temperature and heat flux is dependent on several factors. Factors include the thermocouple location through the thickness of a material, the sensitivity of the surface solution to the error in the specified location of the embedded thermocouples, and the sensitivity to the error in thermocouple data. The effect of these factors on solution accuracy is studied using the methodology discussed in the work of Pizzo, et. al.1 A numerical study is performed to determine if there is an optimal depth at which to embed one thermocouple through the thickness of a material assuming that a second thermocouple is installed on the back face. Solution accuracy will be discussed for a range of embedded thermocouple depths. Moreover, the sensitivity of the surface solution to (a) the error in the specified location of the embedded thermocouple and to (b) the error in the thermocouple data are quantified using numerical simulation, and the results are discussed.

  14. NEET In-Pile Ultrasonic Sensor Enablement-Final Report

    Energy Technology Data Exchange (ETDEWEB)

    J. Daw; J. Rempe; J. Palmer; P. Ramuhalli; R. Montgomery; H.T. Chien; B. Tittmann; B. Reinhardt; P. Keller

    2014-09-01

    Ultrasonic technologies offer the potential to measure a range of parameters during irradiation of fuels and materials, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes under harsh irradiation test conditions. There are two primary issues that currently limit in-pile deployment of ultrasonic sensors. The first is transducer survivability. The ability of ultrasonic transducer materials to maintain their useful properties during an irradiation must be demonstrated. The second issue is signal processing. Ultrasonic testing is typically performed in a lab or field environment, where the sensor and sample are accessible. The harsh nature of in-pile testing and the variety of desired measurements demand that an enhanced signal processing capability be developed to make in-pile ultrasonic sensors viable. To address these issues, the NEET ASI program funded a three year Ultrasonic Transducer Irradiation and Signal Processing Enhancements project, which is a collaborative effort between the Idaho National Laboratory, the Pacific Northwest National Laboratory, the Argonne National Laboratory, and the Pennsylvania State University. The objective of this report is to document the objectives and accomplishments from this three year project. As summarized within this document, significant work has been accomplished during this three year project.

  15. Relative humidity measurements with thermocouple psychrometer and capacitance sensors

    International Nuclear Information System (INIS)

    Mao, Naihsien.

    1991-01-01

    The relative humidity is one of the important hydrological parameters affecting waste package performance. Water potential of a system is defined as the amount of work required to reversibly and isothermally move an infinitesimal quantity of water from a pool of pure water to that system at the same elevation. The thermocouple psychrometer, which acts as a wet-dry bulb instrument based on the Peltier effect, is used to measure water potential. The thermocouple psychrometer works only for relative humidity greater than 94 percent. Other sensors must be used for drier conditions. Hence, the author also uses a Vaisala Humicap, which measures the capacitance change due to relative humidity change. The operation range of the Humicap (Model HMP 135Y) is from 0 to 100 percent relative humidity and up to 160C (320F) in temperature. A psychrometer has three thermocouple junctions. Two copper-constantan junctions serve as reference temperature junctions and the constantan-chromel junction is the sensing junction. Current is passed through the thermocouple causing cooling of the sensing junction by the Peltier effect. When the temperature of the junction is below the dew point, water will condense upon the junction from the air. The Peltier current is discontinued and the thermocouple output is recorded as the temperature of the thermocouple returns to ambient. The temperature changes rapidly toward the ambient temperature until it reaches the wet bulb depression temperature. At this point, evaporation of the water from the junction produces a cooling effect upon the junction that offsets the heat absorbed from the ambient surroundings. This continues until the water is depleted and the thermocouple temperature returns to the ambient temperature (Briscoe, 1984). The datalogger starts to take data roughly at the wet bulb depression temperature

  16. LOFT small break test thermocouple installation

    International Nuclear Information System (INIS)

    Fors, R.M.

    1980-01-01

    The subject thermocouple design has been analyzed for maximum expected hydraulic loading and found to be adequate. The natural frequency of the thermocouple was found to be between the vortex shedding frequencies for the gas and liquid phase so that a tendency for resonance will exist. However, since the thermocouple support will have a restricted displacement, stresses found are below the endurance limit and, thus, are acceptable in respect to fatigue life as well as primary stress due to pressure loading

  17. Temperature measurement: Development work on noise thermometry and improvement of conventional thermocouples for applications in nuclear process heat (PNP)

    International Nuclear Information System (INIS)

    Brixy, H.; Hecker, R.; Oehmen, J.; Barbonus, P.; Hans, R.

    1982-06-01

    The behaviour was studied of NiCr-Ni sheathed thermocouples (sheath Inconel 600 or Incoloy 800, insulation MgO) in a helium and carbon atmosphere at temperatures of 950-1150 deg. C. All the thermocouples used retained their functional performance. The insulation resistance tended towards a limit value which is dependent on the temperature and quality of the thermocouple. Temperature measurements were loaded with great uncertainty in the temperature range of 950-1150 deg. C. Recalibrations at the temperature of 950 deg. C showed errors of up to 6%. Measuring sensors were developed which consist of a sheathed double thermocouple with a noise resistor positioned between the two hot junctions. Using the noise thermometer it is possible to recalibrate the thermocouple at any time in situ. A helium system with a high temperature experimental area was developed to test the thermocouples and the combined thermocouple-noise thermometer sensors under true experimental conditions

  18. In-pile gamma spectrometry and irradiation control at Osiris

    International Nuclear Information System (INIS)

    Farny, G.; Destot, M.; Corre, J.; Texier, D.; Faugere, J.L.; Mouchnino, M.

    1975-01-01

    A new gamma spectrometry facility is available near Osiris reactor core, at Saclay. This device enables nuclear fuels to be examined in loops or capsules all along their irradiation, a few minutes being sufficient to transfer the fuel from the irradiation place to the measurement bench. So, spacelike and timelike history of a lot of fission products, especially short-lived radionuclides, can be observed. Using such in-pile spectrometry device, of original design, allows to avoid radioactive decay corrections and the risks of any information less. Performance of the device is given together with some preliminary results and their interpretation [fr

  19. MCNP calculations for the HCPB submodules in-pile test

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J. [Section Nuclear and Reactor Physics, ECN Nuclear Research, Petten (Netherlands)

    1998-11-01

    This report describes the MCNP calculations that have been performed for the Helium Cooled Pebble Bed (HCPB) Submodules In-pile Test that has been planned for irradiation in the materials testing High Flux Reactor (HFR) at Petten. In this test, four HSM-8 submodules will be placed at core position H4. The report presents the neutron flux and power density profiles to be expected in the submodules. For the gamma induced heating only a rough estimation could be made. In the HCPB submodules the total specific heating does not exceed (36.7 {+-} 2.9)[W/cc]. 8 refs.

  20. SAS validation and analysis of in-pile TUCOP experiments

    International Nuclear Information System (INIS)

    Morman, J.A.; Tentner, A.M.; Dever, D.J.

    1985-01-01

    The validation of the SAS4A accident analysis code centers on its capability to calculate the wide range of tests performed in the TREAT (Transient Reactor Test Facility) in-pile experiments program. This paper presents the SAS4A analysis of a simulated TUCOP (Transient-Under-Cooled-Over-Power) experiment using seven full-length PFR mixed oxide fuel pins in a flowing sodium loop. Calculations agree well with measured thermal-hydraulic, pin failure time and post-failure fuel motion data. The extent of the agreement confirms the validity of the models used in the SAS4A code to describe TUCOP accidents

  1. Development of thermocouple re-instrumentation technique for irradiated fuel rod. Techniques for making center hole into UO2 pellets and thermocouple re-instrumentation to fuel rod

    International Nuclear Information System (INIS)

    Shimizu, Michio; Saito, Junichi; Oshima, Kunio

    1995-07-01

    The information on FP gas pressure and centerline temperature of fuel pellets during power transient is important to study the pellet clad interaction (PCI) mechanism of high burnup LWR fuel rods. At the Department of JMTR, a re-instrumentation technique of FP gas pressure gage for an irradiated fuel rod was developed in 1990. Furthermore, a thermocouple re-instrumentation technique was successfully developed in 1994. Two steps were taken to carry out the development program of the thermocouple re-instrumentation technique. In the first step, a drilling technique was developed for making a center hole of the irradiated fuel pellets. Various drilling tests were carried out using dummy of fuel rods consisted of Ba 2 FeO 3 pellets and Zry-2 cladding. On this work it is important to keep the pellets just the state cracked at a power reactor. In these tests, the technique to fix the pellets by frozen CO 2 was used during the drilling work. Also, diamond drills were used to make the center hole. These tests were completed successfully. A center hole, 54mm depth and 2.5mm diameter, was realized by these methods. The second step of this program is the in-pile demonstration test on an irradiated fuel rod instrumented dually a thermocouple and FP gas pressure gage. The demonstration test was carried out at the JMTR in 1995. (author)

  2. In pile helium loop ''Comedie''

    International Nuclear Information System (INIS)

    Blanchard, R.J.

    1985-01-01

    The loop is located in the SILOE reactor at Centre d'Etudes Nucleaires de Grenoble. The purpose and objectives are divided into two groups, principal and secondary. The primary objective was to provide basic data on the deposition behavior of important condensable fission products on a variety of steel surfaces, i.e. temperature (sorption isotherms) and mass transfer (physical adsorption) dependencies; to provide information concerning the degree of penetration of important fission products into the metals comprising the heat exchanger-recuperator tubes as a function of alloy type and/or metal temperature; to provide complementary information on the reentrainment (liftoff) of important fission and activation products by performing out-of-pile blowdown experiments on tube samples representative of the alloy types used in the heat exchanger-recuperator and of the surface temperatures experienced during plateout. The secondary objective was to provide information concerning the migration of important fission products through graphite. To this end, concentration profiles in the web between the fuel rods containing the fission product source and the coolant channels and in the graphite diffusion sample will be measured to study the corrosion of metallic specimens placed in the conditions of high temperature gas cooled reactor. The first experiment SRO enables to determine the loop characteristics and possibilities related to thermal, thermodynamic, chemical and neutronic properties. The second experiment has been carried out in high temperature gas cooled reactor operating conditions. It enables to determine in particular the deposition axial profile of activation and fission products in the plateout section constituting the heat exchanger, the fission products balance trapped in the different filter components, and the cumulated released fraction of solid fission products. The SR1 test permits to demonstrate in particular the Comedie loop operation reliability, either

  3. In-pile study of the reaction between breeder fuel and sodium

    International Nuclear Information System (INIS)

    Hugot, J.P.

    1982-10-01

    Studies carried out until now show that the determinant parameter of fuel can failure evolution is the development of the reaction between mixed uranium and plutonium dioxide and sodium. The parameters of the reaction are presented from results of an out of pile study, as also results obtained from examination on pins failed in reactors. The best way to study in pile the development of the reaction was to irradiate at a constant power a fuel pin containing sodium. In the experiment, the pin was equipped with a central thermocouple. It shows, that the reaction is developing intergranularly, from cracks and interpellet spaces, in an internal fringe of the fuel before spreading to the periphery. An overheating of the pin is associated to the development of the reaction as also a modification of the fuel pin geometry and a reduction of the oxide [fr

  4. Novel thermocouples for automotive applications

    Directory of Open Access Journals (Sweden)

    P. Gierth

    2018-02-01

    Full Text Available Measurement of temperatures in engine and exhaust systems in automotive applications is necessary for thermal protection of the parts and optimizing of the combustion process. State-of-the-art temperature sensors are very limited in their response characteristic and installation space requirement. Miniaturized sensor concepts with a customizable geometry are needed. The basic idea of this novel sensor concept is to use thick-film technology on component surfaces. Different standardized and especially nonstandard material combinations of thermocouples have been produced for the validation of this technology concept. Application-oriented measurements took place in the exhaust system of a test vehicle and were compared to standard laboratory conditions.

  5. Attaching Thermocouples by Peening or Crimping

    Science.gov (United States)

    Murtland, Kevin; Cox, Robert; Immer, Christopher

    2006-01-01

    Two simple, effective techniques for attaching thermocouples to metal substrates have been devised for high-temperature applications in which attachment by such conventional means as welding, screws, epoxy, or tape would not be effective. The techniques have been used successfully to attach 0.005- in. (0.127-mm)-diameter type-S thermocouples to substrates of niobium alloy C-103 and stainless steel 416 for measuring temperatures up to 2,600 F (1,427 C). The techniques are equally applicable to other thermocouple and substrate materials. In the first technique, illustrated in the upper part of the figure, a hole slightly wider than twice the diameter of one thermocouple wire is drilled in the substrate. The thermocouple is placed in the hole, then the edge of the hole is peened in one or more places by use of a punch (see figure). The deformed material at the edge secures the thermocouple in the hole. In the second technique a hole is drilled as in the first technique, then an annular relief area is machined around the hole, resulting in structure reminiscent of a volcano in a crater. The thermocouple is placed in the hole as in the first technique, then the "volcano" material is either peened by use of a punch or crimped by use of sidecutters to secure the thermocouple in place. This second technique is preferable for very thin thermocouples [wire diameter .0.005 in. (.0.127 mm)] because standard peening poses a greater risk of clipping one or both of the thermocouple wires. These techniques offer the following advantages over prior thermocouple-attachment techniques: . Because these techniques involve drilling of very small holes, they are minimally invasive . an important advantage in that, to a first approximation, the thermal properties of surrounding areas are not appreciably affected. . These techniques do not involve introduction of any material, other than the substrate and thermocouple materials, that could cause contamination, could decompose, or oxidize

  6. A Modified Design of a Thermocouple Based Digital Temperature Indicator with Opto-Isolation

    Directory of Open Access Journals (Sweden)

    S. C. BERA

    2008-01-01

    Full Text Available In the conventional thermocouple based digital temperature indicator the millivolt signal obtained from a thermocouple is first amplified and then converted into a digital signal by using analog-to-digital converter (ADC. This digital signal is then indicated as digital display of temperature using digital counter circuit or microprocessor/microcontroller based circuitry. In the present paper a modified AD conversion technique along with opto-isolation is used to indicate digitally the temperature without using any conventional analog-to-digital converter. The theory and design of the measuring technique are described in the paper. The non-linearity of thermocouple is eliminated by using look-up table within software program. The performance of the circuit has been experimentally tested by using mV input signal instead of a thermocouple as well as using a K-type thermocouple. The experimental results are reported in the paper.

  7. Visual in-pile fuel disruption experiments

    International Nuclear Information System (INIS)

    Cano, G.L.; Ostensen, R.W.; Young, M.F.

    1978-01-01

    In a loss-of-flow (LOF) accident in an LMFBR, the mode of disruption of fuel may determine the probability of a subsequent energetic excursion. To investigate these phenomena, in-pile disruption of fission-heated irradiated fuel pellets was recorded by high speed cinematography. Instead of fuel frothing or dust-cloud breakup (as used in the SAS code) massive and very rapid fuel swelling, not predicted by analytical models, occurred. These tests support massive fuel swelling as the initial mode of fuel disruption in a LOF accident. (author)

  8. Thermal conductivity of sintered UO2 under in-pile conditions

    International Nuclear Information System (INIS)

    Stora, J.P.; Bernardy De Sigoyer, B.; Delmas, R.; Deschamps, P.; Lavaud, B.; Ringot, C.

    1964-01-01

    The temperature distribution in a stack of sintered UO 2 cylinders has been studied both in the laboratory where the heat energy is produced by an axial heating element, and in-pile, where the heating is due solely to nuclear effects. Under a high thermal gradient the UO 2 cracks both along radial planes and along pseudo-cylindrical surfaces: these latter act as thermal barriers to the heat flow, It is therefore an apparent thermal conductivity k a (T), lower than the intrinsic value k(T) of this parameter which is measured. The efficiency of these barriers decreases when the gap decreases and when the external pressure acting on the cracked stack increases: in the limiting case, for high values of the binding strain, k a (T) ≅ k(T). In the domain of phonon conduction (T ≤ 1350 deg C), the expression kw.cm -1 .C -1 =1/(11+0.024*T) accounts for the real thermal conductivity. Above 1350 deg C the thermal conductivity increases. Two in-pile measurements up to 1250 deg C carried out using cartridges fitted with thermocouples confirm, within the limits of experimental error, the above expression and the qualitative effects of the binding strains. Similar tests have been carried out-of-pile and in-pile on the real shape of the EL-4 fuel 'pencils'. Out-of-pile, the influence of the initial free gap, of the nature of the gas filing the 'pencil' and of the external pressure have been studied; the results are compatible with the above interpretation; It appears that an external pressure of 60 kg/cm 2 is insufficient to restore completely the thermal conductivity of the fuel. (authors) [fr

  9. The MOX fuel behaviour test IFA-597.4/.5/.6/.7; Summary of in-pile fuel temperature and gas release data

    Energy Technology Data Exchange (ETDEWEB)

    Koike, Hisashi

    2003-11-15

    It is considered important to study the in-reactor behaviour of MOX fuel in order to enhance the database on such fuel. For this reason, IFA-597.4/.5/.6/.7 were included in the joint research programme of the Halden Project. The series of tests, containing two MIMAS-MOX fuel rods, both equipped with a fuel centre thermocouple and a pressure bellows transducer, has been irradiated in the Halden Reactor since July 1997 under HBWR conditions. The objectives of the test series were to study the thermal and fission gas release (FGR) behaviour of MOX fuel and to explore potential differences in behaviour between solid and hollow pellets. One of the rods had mainly solid pellets, while the other contained only hollow pellets. Both rods had an initial Pu-fissile enrichment of 6.07%. The cladding outside diameter was 9.50 mm, and the initial fuel-clad gap was 180 mum. In the course of the test, power upratings for FGR studies of the MOX fuel were planned at burnup intervals of about 10 MWd/kg MOX. The power uprating was successfully performed at approx10 MWd/kg MOX, where the estimated fuel peak temperature of the solid pellets exceeded the FGR threshold temperature for UO{sub 2} fuel, while that of the hollow pellets remained below the threshold. For the solid fuel, the temperature at onset of FGR was consistent with the empirical threshold temperature for UO{sub 2} fuel. For the hollow fuel, gas release was observed at temperatures below the threshold. FGRs at the end-of-life were approx17% for the solid pellet rod and approx14% for the hollow pellet rod, respectively. As a result of discussions in HPG meetings, IFA-597.7 was unloaded in January 2002. PIE was carried out to check in-pile pressure measurements and examine fuel structural characteristics. The discharge burn-up of the MOX fuel was 32 MWd/kg MOX as determined from in-pile power data. This report supersedes HWR-712 (June 2002) previously issued on in-pile data from IFA-597.4/5/6/7. (Author)

  10. In-pile experiments on fuel rod behaviour during a LOCA

    International Nuclear Information System (INIS)

    Sepold, E.H.; Karb, E.H.; Pruessmann, M.

    1981-07-01

    This report describes the results of the Test Series G2/3 within the in-pile experimental program for the investigation of LWR fuel rod behavior. The results were obtained with single rods of a PWR design in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechanisms of fuel rod failure were being performed with irradiated and unirradiated rods. The main parameter of the test program ist the burnup, ranging from 2500 to 35000 MWd/t. The results of test series G2/3 (35000 MWd/t) with respect to the burst data, i.e. burst temperature, burst pressure, and burst strain, do not indicate major differences from the in-pile tests with unirradiated test specimens. (orig.) [de

  11. In-pile experiemts on fuel rod behavior during a LOCA

    International Nuclear Information System (INIS)

    Pruessmann, M.; Karb, E.H.; Sepold, L.

    1981-02-01

    This report describes the results of the Test Series G1 within the in-pile experimental program for the investigation of LWR fuel rod behavior. The results were obtained with single rods of a PWR design in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechansims of fuel rod failure were being performed with irradiated and unirradiated rods. The main parameter of the test program is the burnup ranging from 2500 to 35 000 MWd/t. The results of test series G1 (35 000 MWd/t) with respect to the burst data, i.e. burst temperature, burst pressure, and burst strain, do not indicate major differences from the in-pile tests with unirradiated test specimens. (orig.) [de

  12. Thermocouple pressure bushing in suspension rod

    International Nuclear Information System (INIS)

    Pasek, J.; Ondreicka, K.

    1975-01-01

    The seal is described of jacket thermocouples located in the pressure reducer in the fuel element suspension rod. The thermocouples are sealed in the pressure reducer with a silicon sealing compound. The sealing compound is compressed between the two reducers with a Bellevile spring and a pressure washer secured in position with a spring. The axial pressure of the inner parts of the reducer on the compound is adjustable by means of a thrust screw. The tightness and alignment of the thermocouples in the pressure reducer is achieved by tightening the thrust screw to the stop of the top reducer and the subsequent setting of the sealing compound. (J.B.)

  13. Contact Thermocouple Methodology and Evaluation for Temperature Measurement in the Laboratory

    Science.gov (United States)

    Brewer, Ethan J.; Pawlik, Ralph J.; Krause, David L.

    2013-01-01

    Laboratory testing of advanced aerospace components very often requires highly accurate temperature measurement and control devices, as well as methods to precisely analyze and predict the performance of such components. Analysis of test articles depends on accurate measurements of temperature across the specimen. Where possible, this task is accomplished using many thermocouples welded directly to the test specimen, which can produce results with great precision. However, it is known that thermocouple spot welds can initiate deleterious cracks in some materials, prohibiting the use of welded thermocouples. Such is the case for the nickel-based superalloy MarM-247, which is used in the high temperature, high pressure heater heads for the Advanced Stirling Converter component of the Advanced Stirling Radioisotope Generator space power system. To overcome this limitation, a method was developed that uses small diameter contact thermocouples to measure the temperature of heater head test articles with the same level of accuracy as welded thermocouples. This paper includes a brief introduction and a background describing the circumstances that compelled the development of the contact thermocouple measurement method. Next, the paper describes studies performed on contact thermocouple readings to determine the accuracy of results. It continues on to describe in detail the developed measurement method and the evaluation of results produced. A further study that evaluates the performance of different measurement output devices is also described. Finally, a brief conclusion and summary of results is provided.

  14. Characteristics of metal sheathed thermocouples in thermowell

    International Nuclear Information System (INIS)

    Okuda, Takehiro; Nakase, Tsuyoshi; Tanabe, Yutaka; Yamada, Kunitaka; Yoshizaki, Akio; Roko, Kiyokazu

    1987-01-01

    Static and dynamic characteristics of thermowell type thermocouples which are planned to be used for the High-Temperature engineering Test Reactor (HTTR) have been investigated. A mock-up test section was installed in Kawasaki's Helium Test Loop (KH-200). Thermal characteristics tests were carried out under the 600 ∼ 1000 deg C temperature conditions. The test section was equipped with four types sheathed thermocouples; the well type, the non well type, and ones with and without the thermal radiation shielding plate. The measured temperature by the well type thermocouples with the shielding plate was only about 1.3 deg C higher than the one without the shielding plate at gas temperature 990 deg C. The measured time constant of the well type thermocouples was about 7 seconds in the condition of the heat transfer coefficient 1600 Kcal/m 2 h deg C on the well surface, and coincided with the calculated one by ''TRUMP'' code. (author)

  15. Estimation of radiation losses from sheathed thermocouples

    International Nuclear Information System (INIS)

    Roberts, I.L.; Coney, J.E.R.; Gibbs, B.M.

    2011-01-01

    Thermocouples are often used for temperature measurements in heat exchangers. However if the radiation losses from a thermocouple in a high temperature gas flow to colder surroundings are ignored significant errors can occur. Even at moderate temperature differences, these can be significant. Prediction of radiation losses from theory can be problematic, especially in situations where there are large variations in the measured temperatures as the emissivity and radiative heat transfer coefficient of the thermocouple are not constant. The following approach combines experimental results with established empirical relationships to estimate losses due to radiation in an annular heat exchanger at temperatures up to 950 o C. - Highlights: → Sheathed thermocouples are often used to measure temperatures in heat exchangers. → Errors are introduced if radiation losses are ignored. → Radiation losses are environment specific and may be significant. → Experimental and theoretical methods are used to estimate losses. → Hot side maximum temperature 950 o C.

  16. Self-adapted thermocouple-diagnostic complex

    International Nuclear Information System (INIS)

    Alekseev, S.V.; Grankovskij, K.Eh.; Olejnikov, P.P.; Prijmak, S.V.; Shikalov, V.F.

    2003-01-01

    A self-adapted thermocouple-diagnostic complex (STDC) for obtaining the reliable data on the coolant temperature in the reactors of NPP is described. The STDC in based on the thermal pulse monitoring of a thermocouple in the measuring channel of a reactor. Measurement method and STDC composition are substantiated. It is shown that introduction of the developed STDC ensures realization of precise and reliable temperature monitoring in the reactors of all types [ru

  17. Embedded cladding surface thermocouples on Zircaloy-sheathed heater rods

    International Nuclear Information System (INIS)

    Wilkins, S.C.

    1977-06-01

    Titanium-sheathed Type K thermocouples embedded in the cladding wall of zircaloy-sheathed heater rods are described. These thermocouples constitute part of a program intended to characterize the uncertainty of measurements made by surface-mounted cladding thermocouples on nuclear fuel rods. Fabrication and installation detail, and laboratory testing of sample thermocouple installations are included

  18. Adhesive-Bonded Tab Attaches Thermocouples to Titanium

    Science.gov (United States)

    Cook, C. F.

    1982-01-01

    Mechanical strength of titanium-alloy structures that support thermocouples is preserved by first spotwelding thermocouples to titanium tabs and then attaching tabs to titanium with a thermosetting adhesive. In contrast to spot welding, a technique previously used for thermocouples, fatigue strength of the titanium is unaffected by adhesive bonding. Technique is also gentler than soldering or attaching thermocouples with a tap screw.

  19. Thermoelectric properties of currently available Au/Pt thermocouples related to the valid reference function

    Directory of Open Access Journals (Sweden)

    Edler F.

    2015-01-01

    Full Text Available Au/Pt thermocouples are considered to be an alternative to High Temperature Standard Platinum Resistance Thermometers (HTSPRTs for realizing temperatures according to the International Temperature Scale of 1990 (ITS-90 in the temperature range between aluminium (660.323 °C and silver (961.78 °C. The original aim of this work was to develop and to validate a new reference function for Au/Pt thermocouples which reflects the properties of presently commercially available Au and Pt wires. The thermoelectric properties of 16 Au/Pt thermocouples constructed at different National Metrological Institutes by using wires from different suppliers and 4 commercially available Au/Pt thermocouples were investigated. Most of them exhibit significant deviations from the current reference function of Au/Pt thermocouples caused by the poor performance of the Au-wires available. Thermoelectric homogeneity was investigated by measuring immersion profiles during freezes at the freezing point of silver and in liquid baths. The thermoelectric inhomogeneities were found to be one order of magnitude larger than those of Au/Pt thermocouples of the Standard Reference Material® (SRM® 1749. The improvement of the annealing procedure of the gold wires is a key process to achieve thermoelectric homogeneities in the order of only about (2–3 mK, sufficient to replace the impracticable HTSPRTs as interpolation instruments of the ITS-90. Comparison measurements of some of the Au/Pt thermocouples against a HTSPRT and an absolutely calibrated radiation thermometer were performed and exhibit agreements within the expanded measurement uncertainties. It has been found that the current reference function of Au/Pt thermocouples reflects adequately the thermoelectric properties of currently available Au/Pt thermocouples.

  20. Recent improvements on micro-thermocouple based SThM

    OpenAIRE

    Nguyen, T. P.; Thiery, L.; Teyssieux, D.; Briand, Danick; Vairac, P.

    2017-01-01

    The scanning thermal microscope (SThM) has become a versatile tool for local surface temperature mapping or measuring thermal properties of solid materials. In this article, we present recent improvements in a SThM system, based on a micro-wire thermocouple probe associated with a quartz tuning fork for contact strength detection. Some results obtained on an electrothermal micro-hotplate device, operated in active and passive modes, allow demonstrating its performance as a coupled force detec...

  1. Remote-welding technique for assembling in-pile IASCC capsule in hot cell

    International Nuclear Information System (INIS)

    Kawamata, Kazuo; Ishii, Toshimitsu; Kanazawa, Yoshiharu; Iwamatsu, Shigemi; Ohmi, Masao; Shimizu, Michio; Matsui, Yoshinori; Saito, Jun-ichi; Ugachi, Hirokazu; Kaji, Yoshiyuki; Tsukada, Takashi

    2006-01-01

    In order to investigate behavior of the irradiation assisted stress corrosion cracking (IASCC) caused by the simultaneous effects of neutron irradiation and high temperature water environment in such a light water reactor (LWR), it is necessary to perform crack growth tests in an in-pile IASCC capsule irradiated in the Japan Materials Testing Reactor (JMTR). The development of the remote-welding technique is essential for remotely assembling the in-pile IASCC capsule installing the pre-irradiated CT specimens. This report describes a new remote-welding machine developed for assembling the in-pile IASCC capsule. The remote-welding technique that the capsule tube is rotated light under the fixed torch was applied to the machine for the welding of thick and large-diameter tubes. The assembly work of four in-pile IASCC capsules having pre-irradiated CT specimens in the hot cell was succeeded for performing the crack growth test under the neutron irradiation in JMTR. The irradiation test of two capsules has been already finished in JMTR without problems. (author)

  2. NEET In-Pile Ultrasonic Sensor Enablement-FY 2012 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    JE Daw; JL Rempe; BR Tittmann; B Reinhardt; P Ramuhalli; R Montgomery; HT Chien

    2012-09-01

    Several Department Of Energy-Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development, Advanced Reactor Concepts, Light Water Reactor Sustainability, and Next Generation Nuclear Plant programs, are investigating new fuels and materials for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials when irradiated. The Nuclear Energy Enabling Technology (NEET) Advanced Sensors and Instrumentation (ASI) in-pile instrumentation development activities are focused upon addressing cross-cutting needs for DOE-NE irradiation testing by providing higher fidelity, real-time data, with increased accuracy and resolution from smaller, compact sensors that are less intrusive. Ultrasonic technologies offer the potential to measure a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes, under harsh irradiation test conditions. There are two primary issues associated with in-pile deployment of ultrasonic sensors. The first is transducer survivability. The ability of ultrasonic transducer materials to maintain their useful properties during an irradiation must be demonstrated. The second issue is signal processing. Ultrasonic testing is typically performed in a lab or field environment, where the sensor and sample are accessible. Due to the harsh nature of in-pile testing, and the range of measurements that are desired, an enhanced signal processing capability is needed to make in-pile ultrasonic sensors viable. This project addresses these technology deployment issues.

  3. Thermocouple module halt failure acceptance test procedure for Tank 241-SY-101 DACS-1

    International Nuclear Information System (INIS)

    Ermi, A.M.

    1997-01-01

    The readiness of the Tank 241-SY-101 Data Acquisition and Control System (DACS-1) to provide monitoring and alarms for a halt failure of any thermocouple module will be tested during the performance of this procedure. Updated DACS-1 ''1/0 MODULE HEALTH STATUS'', ''MININ1'', and ''MININ2'' screens, which now provide indication of thermocouple module failure, will also be tested as part of this procedure

  4. Thermocouple calibration facility for 2900 deg C high temperature and its applications

    International Nuclear Information System (INIS)

    Chen Daolong

    1991-01-01

    The construction and the performance characteristic of a 2900 deg C high temperature thermocouple calibration facility are described. The calibration error analysis is made. The test results of the calibration characteristics of high temperature thermocouples Mo/Nb, W-3Re/W-25Re, and W-1Mo/W-25Mo are given. The test result of temperature dependent resistivity of BeO made by this facility is given

  5. Experiences with W3Re/W25Re thermocouples in fuel pins of NS Otto Hahn's two cores

    International Nuclear Information System (INIS)

    Kolb, M.

    1975-01-01

    Applications and performance of thermocouples in the Otto Hahn reactor are presented. The measurement of effective thermocouple time constants and of fuel rod heat transfer time constants utilizing the reactor noise and the resulting small temperature fluctuations which has become practical by the advent of modern noise analysis systems, is dealt with

  6. A preliminary study of oxidation-resistant coatings on refractory-metal thermocouple sheaths

    International Nuclear Information System (INIS)

    Wilkins, S.C.

    1985-01-01

    The need to make reliable temperature measurements up to 2200 0 C or higher in steam environments during in-pile nuclear fuel damage tests led to a search for oxidation-resistant coatings for the refractory-metal sheaths used to enclose and protect thermocouples used for such measurements. Iridium, thoria, and thoria-over-iridium coatings were separately sputter-deposited on molybdenum-rhenium alloy protection tubes for evaluation. The coated samples were individually heated in flowing steam in an induction furnace. An extension tube welded to each sample was connected to a vacuum pump and gauge; failure of the sample was detected by noting the degradation of the vacuum maintained in the sample. Relatively heavy coatings of iridium provided a modest degree of oxidation protection at the temperatures of interest. Thoria coatings provided no significant protection at those temperatures, compared to uncoated control samples

  7. Energy deposition measurements in fast reactor safety experiments with fission thermocouple detectors

    International Nuclear Information System (INIS)

    Wright, S.A.; Scott, H.L.

    1979-01-01

    The investigation of phenomena occurring in in-pile fast reactor safety experiments requires an accurate measurement of the time dependent energy depositions within the fissile material. At Sandia Laboratories thin-film fission thermocouples are being developed for this purpose. These detectors have high temperature capabilities (400 to 500 0 C), are sodium compatible, and have milli-second time response. A significant advantage of these detectors for use as energy deposition monitors is that they produce an output voltage which is directly dependent on the temperature of a small chip of fissile material within the detectors. However, heat losses within the detector make it necessary to correct the response of the detector to determine the energy deposition. A method of correcting the detector response which uses an inverse convolution procedure has been developed and successfully tested with experimental data obtained in the Sandia Pulse Reactor (SPR-II) and in the Annular Core Research Reactor

  8. Study on thermocouple attachment in reflood experiments

    International Nuclear Information System (INIS)

    Sugimoto, Jun

    1977-03-01

    The method of thermocouple attachment to a heater rods has been studied for surface temperature measurement in reflood experiments. The method used as far in JAERI's reflood experiments had some possibilities of not estimating exactly the quench times. Various attachment method have been tested and some proved to be effective in the respect. (auth.)

  9. A thermocouple thermometry system for ultrasound hyperthermia

    International Nuclear Information System (INIS)

    Ozarka, M.; Gharakhani, A.; Magin, R.; Cain, C.

    1984-01-01

    A thermometry system designed to be used in the treatment of cancer by ultrasound hyperthermia is described. The system monitors tumor temperatures using 16 type T (copper-constantan) thermocouples and is controlled by a 12 MHz Intel 8031 microcomputer. An analog circuit board contains the thermocouple amplifiers, an analog multiplexer, scaling circuitry, and an analog to digital converter. A digital board contains the Intel 8031, program memory, data memory, as well as circuitry for control and data communications. Communication with the hyperthermia system control computer is serially by RS-232 with selectable baud rate. Since the thermocouple amplifiers may have slight differences in gain and offset, a calibrated offset is added to a lookup table value to obtain the proper display temperature to within +- 0.1 0 C. The calibration routine, implemented in software, loads a nonvolatile random access memory chip with the proper offset values based on the outputs of each thermocouple channel at known temperatures which bracket a range of interest

  10. Thermal and mechanical analyses for the HCPB Submodules in-pile test

    Energy Technology Data Exchange (ETDEWEB)

    Bakker, K. [Fuels, Actinides and Isotopes, Nuclear Research and Consultancy Group NRG, Netherlands Energy Research Foundation ECN, Petten (Netherlands)

    1998-12-01

    A description is given of the Finite Element Method (FEM) and thermal and mechanical computations that have been performed for the Helium Cooled Pebble Bed (HCPB) Submodules in-pile tests, which have been planned for irradiation in the High Flux Reactor (HFR) at Petten. In this test, four submodules will be placed at core position H4. The report presents the temperature and stress distribution for the highest powered submodule of these four submodules. 9 refs

  11. Recent improvements on micro-thermocouple based SThM

    Science.gov (United States)

    Nguyen, TP; Thiery, L.; Teyssieux, D.; Briand, D.; Vairac, P.

    2017-01-01

    The scanning thermal microscope (SThM) has become a versatile tool for local surface temperature mapping or measuring thermal properties of solid materials. In this article, we present recent improvements in a SThM system, based on a micro-wire thermocouple probe associated with a quartz tuning fork for contact strength detection. Some results obtained on an electrothermal micro-hotplate device, operated in active and passive modes, allow demonstrating its performance as a coupled force detection and thermal measurement system.

  12. Pneumatic capsule with a measuring system for in-pile irradiation

    International Nuclear Information System (INIS)

    Oshima, Keiichi; Yamazaki, Yasaburo; Hirata, Mitsuho; Ishii, Toshio; Shimozawa, Ryohei.

    1967-01-01

    A prior-art in-pile irradiation apparatus wherein a rabbit containing an irradiation specimen therein is inserted into and removed from a pile by a pneumatic system does not include means for measuring the temperature and pressure of the specimen under irradiation. When the rabbit is deformed during irradiation, it cannot be reliably recovered. A pneumatic capsule assembly with a measuring system according to this invention has a double structure which consists of an inner capsule containing the specimen therein and an outer capsule evacuated or filled with a gas. A thermocouple lace wire and a strain gauge are welded on the outside surface of the inner capsule as detection terminals for measuring the temperature and pressure. A rupture plate which bursts when the pressure in the inner capsule reaches a predetermined value is provided at a part of the inner capsule, and a fin for heat transmission is provided between the inner and outer capsules to thus prevent the deformation of the pneumatic capsule assembly as a whole. (Takasuka, S.)

  13. Fuel disruption mechanisms determined in-pile in the ACRR

    International Nuclear Information System (INIS)

    Wright, S.A.; Fischer, E.A.

    1984-09-01

    Over thirty in-pile experiments were performed to investigate fuel disruption behavior for LMFBR loss of flow (LOF) accidents. These experiments reproduced the heating transients for a variety of accidents ranging from slow LOF accidents to rapid LOF-driven-TOP accidents. In all experiments the timing and mode of the fuel disruption were observed with a high speed camera, enabling detailed comparisons with a fuel pin code, SANDPIN. This code transient intra- and inter-granular fission gas behavior to predict the macroscopic fuel behavior, such as fission gas induced swelling and frothing, cracking and breakup of solid fuel, and fuel vapor pressure driven dispersal. This report reviews the different modes of fuel disruption as seen in the experiments and then describes the mechanism responsible for the disruption. An analysis is presented that describes a set of conditions specifying the mode of fuel disruption and the heating conditions required to produce the disruption. The heating conditions are described in terms of heating rate (K/s), temperature gradient, and fuel temperature. A fuel disruption map is presented which plots heating rate as a function of fuel temperature to illustrate the different criteria for disruption. Although this approach to describing fuel disruption oversimplifies the fission gas processes modeled by SANDPIN, it does illustrate the criteria used to determine which fuel disruption mechanism is dominant and on what major fission gas parameters it depends

  14. Core exit thermocouple upgrade at Zion station

    International Nuclear Information System (INIS)

    Ulinski, T.M.; Ferg, D.A.

    1989-01-01

    Following the Three Mile Island accident, the ability of the core exit thermocouple (CET) system to monitor reactor core conditions and core cooling status became a requirement of the U.S. Nuclear Regulatory Commission (NRC). Since the thermocouple system at Zion station was not originally required for postaccident monitoring, Commonwealth Edison Company (CECo) committed to upgrading the CET system and to installing a subcooling margin monitoring (SMM) system. The significance of this commitment was that CECo proposed to accomplish the upgrade effort using internal resources and by developing the required in-house expertise instead of procuring integrated packages from several nuclear steam supply system vendors. The result was that CECo was able to demonstrate a number of new capabilities and unique design features with a significant cost savings. These included a qualified connector with an integral thermocouple cold-reference junction temperature compensation; the design, assembly, testing, and installation of a seismically qualified class 1E microprocessor; a commercial-grade dedication/upgrade process for safety-related hardware; a human factors review capability, and a verification and validation program for safety-related software. A discussion of these new capabilities and details of the design features is presented in this paper

  15. Results from In-pile experiments on LWR fuel rod behavior under LOCA conditions with unirradiated rods

    International Nuclear Information System (INIS)

    Sepold, L.; Karb, E.H.; Pruessmann, M.

    1981-06-01

    This report summarizes the results of the FR2-in-pile tests at KfK (Kernforschungszentrum Karlsruhe) with unirradiated test rods. The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechanisms of fuel rod failure were being performed with irradiated and unirradiated single rods of a PWR design in the DK loop of the FR2 reactor. The main parameter of the test program was the burnup, ranging from 2.500 to 35.000 MWd/t. The program with unirradiated specimens comprised the series A and B with a total of 14 tests. (orig.) [de

  16. In-pile irradiation test program and safety analysis report of the KAERI fuel for HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Wan; Ryu, Woo Suck; Byun, Taek Sang; Park, Jong Man; Lee, Byung Chul; Kim, Hack No; Park, Hee Tae; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-05-01

    Localization of HANARO fuel has been carried out successfully, and design and fabrication technologies of the fuel are recently arrived the final stage of development. The performance of the fuel which has been fabricated in KAERI is confirmed through out-of-pile characterization, and the quality assurance procedure and assessment criteria are described. In order to verify the KAERI fuel, thus, in-pile irradiation test program of the KAERI fuel is scheduled in HANARO. This report summarizes the in-pile testing schedule, design documents of test rods and assemblies, fabrication history and out-of-pile characteristics of test rods, irradiation test condition and power history, post-irradiation examination scheme, linear power generation distribution, and safety analysis results. The design code for HANARO fuel is used to analyze the centerline temperature and swelling of the KAERI fuels. The results show that at 120 kW/m of linear power the maximum centerline temperature is 267 deg C which is much lower than the limitation temperature of 350 deg C, and that the swelling is 9.3 % at 95 at% lower than criterion of 20 %. Therefore, the KAERI fuels of this in-pile irradiation test is assessed to show good performance of integrity and safety in HANARO. 10 tabs., 7 figs., 3 refs. (Author).

  17. Evaluation of results from an in-pile creep test in the Studsvik R2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pettersson, Kjell [Entropy Materials, Stockholm (Sweden)

    2002-01-01

    An in-pile creep test with bowing of cladding tubes has been performed in a hot water loop in the Studsvik R2 reactor . One test was performed in the core and one outside the core. The out-of-pile sample showed some minor primary creep strain while the in-pile specimen deformed at a steady rate of 5x10{sup -7}/h . However, when the results were compared to a broader data base of Zircaloy in-pile creep it became clear that the creep deformation observed is a primary creep which occurs before the irradiation creep in Zircaloy reaches a constant steady state creep rate. This primary stage is interpreted as a consequence of the development of an irradiation induced microstructure in Zircaloy which does not reach a steady state until a dose of about 10{sup 21} n/cm{sup 2} . At this stage the steady state irradiation creep starts. From this interpretation it is concluded that it is quite feasible to use the test method on pre-irradiated material in which it can be expected that the steady state will be reached already after short irradiation times.

  18. Development of a micro-thermal flow sensor with thin-film thermocouples

    Science.gov (United States)

    Kim, Tae Hoon; Kim, Sung Jin

    2006-11-01

    A micro-thermal flow sensor is developed using thin-film thermocouples as temperature sensors. A micro-thermal flow sensor consists of a heater and thin-film thermocouples which are deposited on a quartz wafer using stainless steel masks. Thin-film thermocouples are made of standard K-type thermocouple materials. The mass flow rate is measured by detecting the temperature difference of the thin-film thermocouples located in the upstream and downstream sections relative to a heater. The performance of the micro-thermal flow sensor is experimentally evaluated. In addition, a numerical model is presented and verified by experimental results. The effects of mass flow rate, input power, and position of temperature sensors on the performance of the micro-thermal flow sensor are experimentally investigated. At low values, the mass flow rate varies linearly with the temperature difference. The linearity of the micro-thermal flow sensor is shown to be independent of the input power. Finally, the position of the temperature sensors is shown to affect both the sensitivity and the linearity of the micro-thermal flow sensor.

  19. Gas reactor in-pile safety test project (GRIST-2)

    International Nuclear Information System (INIS)

    Kelley, A.P. Jr.; Arbtin, E.; St Pierre, R.

    1979-01-01

    Although out-of-pile tests may be expected to confirm individual phenomena models in core disruptive accident analysis codes, only in-pile tests are capable of verifying the extremely complex integrated model effects within the appropriate time phase for these accidents. For this reason, the GRIST-2 project, the purpose of which is to design and construct an in-pile helium loop capable of transient safety testing in the TREAT facility in Idaho, forms a cornerstone of the US GCFR safety program. The project organization, experiment program, facility, helium system design, and schedule which have been selected to meet the objectives are described

  20. Capsule development and utilization for material irradiation tests; study on the in-pile creep measuring method of zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong; Lee, Byung Kee; Lee, Jong Jea; Kim, Chang Sik; Kim, B. Hun; Cho, I. Sik [Sunmoon University, Asan (Korea)

    2002-02-01

    The final objective of this project is to obtain a design and fabrication technology of an in-pile creep test machine of zirconium alloys. First, design concepts of the in-pile creep test machines of various foreign countries were reviewed and a preliminary design of the equipment was carried. Second, the mock-up of the in-pile creep test machine was fabricated based on the preliminary design. The mock-up consisted of upper and lower grips, a yoke, a pressure chamber including a bellows, a push rod and LVDT. Each part was made of 304 L stainless steel. The average surface roughness of the parts was 1.0-14.7 {mu}m. The mock-up precisely determined an extension of a specimen by gas pressure. Finally, in-pile creep capsule was designed, fabricated and modified. High pure aluminum blocks were put in the capsule. Considering heat transfer coefficients of helium and nitrogen gases, the cooling efficiency is about 4 .deg. C at the condition of 300 .deg. C creep test. Yield strength, ultimate tensile strength and elongation at 300 .deg. C were 335 MPa, 591 MPa, 19.8%, respectively. which were lower than the values at room temperature, 353 MPa, 740 MPa, 12.5%. This study gave an important technology related to design, fabrication and performance tests of the in-pile creep test machine, which is applied to the fabrication of a special capsule and also used for the fundamental data for the fabrication of various in-pile creep capsules. 6 refs., 45 figs., 5 tabs. (Author)

  1. Interpretation of the TRADE In-Pile source multiplication experiments

    International Nuclear Information System (INIS)

    Mercatali, Luigi; Carta, Mario; Peluso, Vincenzo

    2006-01-01

    Within the framework of the neutronic characterization of the TRIGA RC-1 reactor in support to the TRADE (TRiga Accelerator Driven Experiment) program, the interpretation of the subcriticality level measurements performed in static regime during the TRADE In-Pile experimental program is presented. Different levels of subcriticality have been measured using the MSA (Modified Source Approximated) method by the insertion of a standard fixed radioactive source into different core positions. Starting from a reference configuration, fuel elements were removed: control rods were moved outward as required for the coupling experiments envisioned with the proton accelerator and fission chambers were inserted in order to measure subcritical count rates. A neutron-physics analysis based on the modified formulation of the source multiplication method (MSM) has been carried out, which requires the systematic solution for each experimental configuration of the homogeneous, both in the forward and adjoint forms, and inhomogeneous Boltzmann equations. By means of such a methodology calculated correction factors to be applied to the MSA measured reactivities were produced in order to take into account spatial and energetic effects creating changes in the detector efficiencies and effective source with respect to the calibration configuration. The methodology presented has been tested against a large number of experimental states. The measurements have underlined the sensitivity of the MSA measured reactivities to core geometry changes and control rod perturbations; the efficiency of MSM factors to dramatically correct for this sensitivity is underlined, making of this technique a relevant methodology in view of the incoming US RACE program to be performed in TRIGA reactors

  2. Seismic analysis of the in-pile test section

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. M.; Park, K. N.; Chi, D. Y.; Park, S. K.; Sim, B. S.; Ahn, S. H.; Lee, C. Y.; Kim, Y. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    This study gives the results of the seismic analysis of the IPS (In Pile Section) with lower bracket support. The results cover the natural frequency and seismic response of the IPS for the SSE and OBE events. An FE (Finite Element) model which includes the two vessels of the IPS and its support structure were analyzed by ABAQUS.

  3. Difference equation approach to two-thermocouple sensor characterization in constant velocity flow environments

    International Nuclear Information System (INIS)

    Hung, P.C.; Irwin, G.; Kee, R.; McLoone, S.

    2005-01-01

    Thermocouples are one of the most popular devices for temperature measurement due to their robustness, ease of manufacture and installation, and low cost. However, when used in certain harsh environments, for example, in combustion systems and engine exhausts, large wire diameters are required, and consequently the measurement bandwidth is reduced. This article discusses a software compensation technique to address the loss of high frequency fluctuations based on measurements from two thermocouples. In particular, a difference equation (DE) approach is proposed and compared with existing methods both in simulation and on experimental test rig data with constant flow velocity. It is found that the DE algorithm, combined with the use of generalized total least squares for parameter identification, provides better performance in terms of time constant estimation without any a priori assumption on the time constant ratios of the thermocouples

  4. Zircaloy sheathed thermocouples for PWR fuel rod temperature measurements

    International Nuclear Information System (INIS)

    Anderson, J.V.; Wesley, R.D.; Wilkins, S.C.

    1979-01-01

    Small diameter zircaloy sheathed thermocouples have been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory. Surface mounted thermocouples were developed to measure the temperature of zircaloy clad fuel rods used in the Thermal Fuels Behavior Program (TFBP), and embedded thermocouples were developed for use by the Loss-of-Fluid Test (LOFT) Program for support tests using zircaloy clad electrically heated nuclear fuel rod simulators. The first objective of this developmental effort was to produce zircaloy sheathed thermocouples to replace titanium sheathed thermocouples and thereby eliminate the long-term corrosion of the titanium-to-zircaloy attachment weld. The second objective was to reduce the sheath diameter to obtain faster thermal response and minimize cladding temperature disturbance due to thermocouple attachment

  5. Analysis of heat transfer from fuel rods with externally attached thermocouples

    International Nuclear Information System (INIS)

    Gill, C.R.; Coddington, P.

    1988-05-01

    This paper describes the development of 2 and 3 dimensional finite element heat conduction models to simulate the behaviour of the external thermocouples attached to the LOFT fuel rods during the blowdown phase of a large break loss-of-coolant accident. To establish the model and determine the thermal coupling between the thermocouple and the fuel rod extensive use was made of two series of experiments performed at INEL in the LOFT Test Support Facility (LTSF). These experiments were high pressure reflood experiments with fluid conditions 'typical' of those seen during the bottom-up flow period of the LOFT experiments. (author)

  6. Error analysis of thermocouple measurements in the Radiant Heat Facility

    International Nuclear Information System (INIS)

    Nakos, J.T.; Strait, B.G.

    1980-12-01

    The measurement most frequently made in the Radiant Heat Facility is temperature, and the transducer which is used almost exclusively is the thermocouple. Other methods, such as resistance thermometers and thermistors, are used but very rarely. Since a majority of the information gathered at Radiant Heat is from thermocouples, a reasonable measure of the quality of the measurements made at the facility is the accuracy of the thermocouple temperature data

  7. Fabrication and use of zircaloy/tantalum-sheathed cladding thermocouples and molybdenum/rhenium-sheathed fuel centerline thermocouples

    International Nuclear Information System (INIS)

    Wilkins, S.C.; Sepold, L.K.

    1985-01-01

    The thermocouples described in this report are zircaloy/tantalum-sheathed and molybdenum/rhenium alloy-sheathed instruments intended for fuel rod cladding and fuel centerline temperature measurements, respectively. Both types incorporate beryllium oxide insulation and tungsten/rhenium alloy thermoelements. These thermocouples, operated at temperatures of 2000 0 C and above, were developed for use in the internationally sponsored Severe Fuel Damage test series in the Power Burst Facility. The fabrication steps for both thermocouple types are described in detail. A laser-welding attachment technique for the cladding-type thermocouple is presented, and experience with alternate materials for cladding and fuel therocouples is discussed

  8. A Study of the Behavior Characteristics for K-type Thermocouple

    International Nuclear Information System (INIS)

    Ye, Songhae; Kim, Yongsik; Lee, Sooill; Kim, Sungjin; Lyou, Jooon

    2014-01-01

    K-type thermocouple is widely used in nuclear power plants (NPP) and they provide reliable service. Generally, the thermocouple assembly is the finished product and usually only nondestructive tests are performed on the assembly, whereas destructive tests are confined to selected bulk cable specimens. This K-type thermocouple has been used representatively in the In-Core Instrument Assembly (ICI) in the nuclear power plants. The ICI consists of five rhodium emitter detectors that provide information on the thermal power for the core and one K-type thermocouple made with two cables (Chromel-Alumel) that provides the temperature of core exit (CET). Generally, the quantity of the ICI is absolutely different according to the number of fuel assemblies in the NPP. In the case of SKN 3 and 4, they were designed to the 61 ICI to provide information on the core cooling to the inadequate core cooling monitoring system (ICCMS). This measured temperature could be also used to check the entry condition of severe accidents. The technology of the TFDR is a generic skill to detect the fault position of the cable. In-core Instruments (ICIs) were used to detect the Core Exit Temperature (CET) in a reactor. This measured temperature was also used to check the entry condition of severe accidents. However, if a serious accident occurs, the upper portion of the core is damaged. This instrument has not been available. This paper illustrates the estimation possibility for the status of molten core through the high-temperature characteristics test of k-type thermocouple. It turns out that it is possible to measure the k-type thermocouple up to 1350 .deg. C degrees before melting during insertion into the melting furnace. Additionally, in order to measure a high temperature of 2000 .deg. C or more, the replacement possibility of k-type thermocouple was evaluated. However the tungsten-rhenium thermocouple is impossible to use in the detection of temperature at the in-core because of the

  9. Sputtered type s thermocouples on quartz glass substrates

    International Nuclear Information System (INIS)

    Sopko, B.; Vlk, J.; Chren, D.; Sopko, V.; Dammer, J.; Mengler, J.; Hynek, V.

    2011-01-01

    The work deals with the development of thin film thermocouples and their practical use. The principle of measuring planar thin film thermocouples is the same as for conventional thermocouples and is based on the thermoelectric effect, which named after its discoverer, Seebeck. Seebeck effect is direct conversion of temperature differences to electric voltage. In different applications it is necessary to use temperature sensors with high spatial resolution (with the placement of several measured points on the segment of length 1 mm) and short response time. For this application are currently used planar thermocouples with important advantage in production price and reproducible production. The innovative potential of thin-film thermocouples are to be found mainly in: 1 st use of technology in thin layers, unlike the already mature technologies applied in the production of conventional thermocouple probes are capable of further improvement with the usage of new substrate materials, modified methods for creating electrical contacts to the new thermocouple configuration and adhesive and protective layers, 2 nd in saving precious and rare metals, 3 rd decreasing the thickness of the layers and reducing the overall size of thermo probe. Measuring the temperature of molten steel, leading to a general loss of strength and the subsequent destruction of the probe. Here exhibited the highest resistance of quartz plates used in thin film substrates thermocouples. (authors)

  10. Thermocouple design for measuring temperatures of small insects

    Science.gov (United States)

    A.A. Hanson; R.C. Venette

    2013-01-01

    Contact thermocouples often are used to measure surface body temperature changes of insects during cold exposure. However, small temperature changes of minute insects can be difficult to detect, particularly during the measurement of supercooling points. We developed two thermocouple designs, which use 0.51 mm diameter or 0.127 mm diameter copper-constantan wires, to...

  11. Thermocouple correlation transit time flowmeter tests at WCL

    International Nuclear Information System (INIS)

    Lassahn, G.D.

    1976-11-01

    Scoping tests indicate the feasibility for using transit time flowmeters with thermocouple sensors in steam-water steady state flow. Conclusive results were not obtained. More conclusive results are expected from tests to be conducted in the semiscale facility with a redesigned transit time thermocouple sensor

  12. Base metal thermocouples drift rate dependence from thermoelement diameter

    International Nuclear Information System (INIS)

    Pavlasek, P; Duris, S; Palencar, R

    2015-01-01

    Temperature measurements are one of the key factors in many industrial applications that directly affect the quality, effectiveness and safety of manufacturing processes. In many industrial applications these temperature measurements are realized by thermocouples. Accuracy of thermocouples directly affects the quality of the final product of manufacturing and their durability determines the safety margins required. One of the significant effects that affect the precision of the thermocouples is short and long term stability of their voltage output. This stability issue occurs in every type of thermocouples and is caused by multiple factors. In general these factors affect the Seebeck coefficient which is a material constant, which determines the level of generated voltage when exposed to a temperature gradient. Changes of this constant result in the change of the thermocouples voltage output thus indicated temperature which can result in production quality issues, safety and health hazards. These alternations can be caused by physical and chemical changes within the thermocouple lead material. Modification of this material constant can be of temporary nature or permanent. This paper concentrates on the permanent, or irreversible changes of the Seebeck coefficient that occur in commonly used swaged MIMS Type N thermocouples. These permanent changes can be seen as systematic change of the EMF of the thermocouple when it is exposed to a high temperature over a period of time. This change of EMF by time is commonly known as the drift of the thermocouple. This work deals with the time instability of thermocouples EMF at temperatures above 1200 °C. Instability of the output voltage was taken into relation with the lead diameter of the tested thermocouples. This paper concentrates in detail on the change of voltage output of thermocouples of different diameters which were tested at high temperatures for the overall period of more than 210 hours. The gather data from this

  13. Base metal thermocouples drift rate dependence from thermoelement diameter

    Science.gov (United States)

    Pavlasek, P.; Duris, S.; Palencar, R.

    2015-02-01

    Temperature measurements are one of the key factors in many industrial applications that directly affect the quality, effectiveness and safety of manufacturing processes. In many industrial applications these temperature measurements are realized by thermocouples. Accuracy of thermocouples directly affects the quality of the final product of manufacturing and their durability determines the safety margins required. One of the significant effects that affect the precision of the thermocouples is short and long term stability of their voltage output. This stability issue occurs in every type of thermocouples and is caused by multiple factors. In general these factors affect the Seebeck coefficient which is a material constant, which determines the level of generated voltage when exposed to a temperature gradient. Changes of this constant result in the change of the thermocouples voltage output thus indicated temperature which can result in production quality issues, safety and health hazards. These alternations can be caused by physical and chemical changes within the thermocouple lead material. Modification of this material constant can be of temporary nature or permanent. This paper concentrates on the permanent, or irreversible changes of the Seebeck coefficient that occur in commonly used swaged MIMS Type N thermocouples. These permanent changes can be seen as systematic change of the EMF of the thermocouple when it is exposed to a high temperature over a period of time. This change of EMF by time is commonly known as the drift of the thermocouple. This work deals with the time instability of thermocouples EMF at temperatures above 1200 °C. Instability of the output voltage was taken into relation with the lead diameter of the tested thermocouples. This paper concentrates in detail on the change of voltage output of thermocouples of different diameters which were tested at high temperatures for the overall period of more than 210 hours. The gather data from this

  14. Heated junction thermocouple level measurement apparatus

    International Nuclear Information System (INIS)

    Bevilacqua, F.; Burger, J.M.

    1984-01-01

    A liquid level sensing apparatus senses the level of liquid surrounding the apparatus. A plurality of axially spaced sensors are enclosed in a separator tube. The separator tube tends to collapse the level of a two-phase fluid within the separator tube into essentially a liquid phase and a gaseous phase where the collapsed level bears a relationship to the coolant inventory outside the separator tube. The level of the liquid phase is sensed by level sensing apparatus. The separator tube contains inlet-outlet ports near the top and bottom thereof to equalize the liquid level inside and outside the separator tube when the level fluctuates or the water within the separator tube flashes to steam. Each sensor is comprised of a heater, a heated thermocouple junction and an unheated thermocouple junction within an elongated heat conductive housing. The heated portion of housing is enclosed in a splash guard with inlet-outlet ports near the top and bottom to equalize the liquid level inside and outside the splash guardand to eliminate the spurious indications of liquid level change which may arise if water droplets contact the housing in the region of the heater. To prevent steam bubbles entrained in a two-phase fluid cross flow from entering the lateral inlet-outlet ports of the separator tube, the separator tube is enclosed in support tube which may in turn be enclosed in an otherwise unused control element assembly shroud. The lateral inlet-outlet ports of separator tube are axially offset from lateral inlet-outlet ports of support tube at least where support tube is subjected to cross flow. The shroud is open on the bottom and has lateral inlet-outlet ports to facilitate liquid level fluctuations to equalize inside and outside shroud

  15. LOFT ECC Pitot Tube and Thermocouple Rake Penetration thermal analysis

    International Nuclear Information System (INIS)

    Tolan, B.J.

    1977-01-01

    A thermal analysis of the LOFT ECC Pitot Tube and Thermocouple Rake Penetration was performed using COUPLE, a two-dimensional finite element computer code. Four transients which conservatively cover all transients the rake will be exposed to were included in this analysis in order to comply with the ASME Code Section III requirements. The transients conservatively cover hot and cold leg operation, and nuclear and nonnuclear operation. The four transients include the LOCE with ECC injection transient, the single control rod drop transient, the scram transient, and the heatup with 0 to 100% load change transient. Temperature distributions in the rake were obtained for each of the four transients and several plots of node temperatures vs. time are given

  16. Analysis of Sandia in-pile EOS experiments

    International Nuclear Information System (INIS)

    Breitung, W.; Gorham-Bergeron, E.; Murata, K.K.

    1979-01-01

    Preliminary analysis has been carried out of the dynamic in-pile equation-of-state measurements for UO 2 , conducted at Sandia Laboratories, aimed at reducing the uncertainties in the effective UO 2 enthalpy corresponding to the measured pressures. Of the remaining width of the p-H band of some 350 J/g, about 200 J/g originate in the uncertainties of the analytical modelling and about 150 J/g result from the scatter in the experimental data

  17. In-pile experimental facility needs for LMFR safety research

    International Nuclear Information System (INIS)

    Kawata, Norio; Niwa, Hajime

    1994-01-01

    Although the achievement of the safety research during the past years has been significant, there still exists a strong need for future research, especially when there is prospect for future LMFR commercialization. In this paper, our current views are described on future research needs especially with a new in-pile experimental facility. The basic ideas and progress are outlined of a preliminary feasibility study. (author)

  18. The MOZART in-pile tritium extraction experiment

    International Nuclear Information System (INIS)

    Briec, M.

    1990-01-01

    In-pile tritium extraction behavior of various ceramics was compared in the MOZART experiment. The influence of temperature and purge gas composition was studied. The experimental results are analyzed by taking into account the processes of diffusion in the grain and desorption at grain surface. This analysis confirms that a better knowledge of the desorption process is necessary for a satisfactory explanation of the experimental data

  19. Theoretical interpretation of SCARABEE single pin in-pile boiling experiments

    International Nuclear Information System (INIS)

    Struwe, D.; Bottoni, M.; Fries, W.; Elbel, H.; Angerer, G.

    1977-01-01

    In the framework of LMFBR safety analysis a theoretical interpretation of some of the most representative of the single pin experiments of the in-pile SCARABEE project has been performed from both viewpoints of thermohydraulic and fuel behaviour using the computer codes CAPRI-2 and SATURN-1. The analysis is aimed at investigating the pin behavior from the preirradiation history, through the observed sequence of events following a coolant mass flow reduction from boiling inception up to pin breakdown. A comparison of theoretical results with experimentally recorded data has allowed a deeper insight into the peculiar features of the experiments and enabled a valuable code verification. (Auth.)

  20. Analytical out-of-pile and in-pile experiments on gadolinia bearing fuels

    International Nuclear Information System (INIS)

    Bruet, M.; Francois, B.; Do, Q.; Bergeron, J.; Trotabas, M.

    1986-06-01

    New fuel management schemes in PWRs can be achieved through the use of burnable poisons like gadolinia bearing fuel rods. However, the introduction of such a design has required a qualification program, which has been performed in collaboration between CEA, FRAGEMA and/or FRAMATOME by specialized teams in CEA facilities. The main scoops of this program concern: the fabrication process; the out of pile physical properties determination: the in pile thermomechanical behaviour and fission product release; the neutronic studies in view to validate the Computed Gd efficiency and the LBP depletion calculation schemes and to analyse and assess various schemes of core calculations

  1. Stress analysis of primary pipe rigid support of the in pile loop

    International Nuclear Information System (INIS)

    Hasibuan, Dj.

    1998-01-01

    Base on requirement of the safety analysis report and operation planning preparation on the in pile loop by using the fuel bundle in the test section, the stress analysis of primary pipe support has been done. The analysis was performed for the 3 (three) points of pipe support, which are chosen by random selection, i.e.: GU 2001, GU 2002, and GU 2331. The analysis result showed that the maximum allowable stress was greater then the actual stress. It is concluded that the existing supports fulfil the safety requirement

  2. Enhanced in-pile instrumentation at the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2011-07-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

  3. Enhanced In-Pile Instrumentation at the Advanced Test Reactor

    Science.gov (United States)

    Rempe, Joy L.; Knudson, Darrell L.; Daw, Joshua E.; Unruh, Troy; Chase, Benjamin M.; Palmer, Joe; Condie, Keith G.; Davis, Kurt L.

    2012-08-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

  4. Boron carbide in pile behaviour Rapsodie experience

    International Nuclear Information System (INIS)

    Kryger, B.; Colin, M.

    1983-04-01

    Results concerning boron carbide irradiation experiments performed in RAPSODIE up to 10 22 .cm - 3 capture density in the temperature range 600-1100 0 lead to the following main conclusions: initial density and grain size lowering contribute to swelling decrease but density is the major parameter for swelling limitation; swelling rate can vary in a wide range (ratio 1 to 3) according to combinations of density (1.8 to 2.3) and grain size (10 to 50 μm) values; a swelling balance reveals that the most important contribution to swelling should be a high density of helium small bubbles (<400 A); helium retention increases with density and grain size and decreases with temperature elevation. A diffusion law is proposed to describe the rate of helium release

  5. Metallic and Ceramic Thin Film Thermocouples for Gas Turbine Engines

    Directory of Open Access Journals (Sweden)

    Otto J. Gregory

    2013-11-01

    Full Text Available Temperatures of hot section components in today’s gas turbine engines reach as high as 1,500 °C, making in situ monitoring of the severe temperature gradients within the engine rather difficult. Therefore, there is a need to develop instrumentation (i.e., thermocouples and strain gauges for these turbine engines that can survive these harsh environments. Refractory metal and ceramic thin film thermocouples are well suited for this task since they have excellent chemical and electrical stability at high temperatures in oxidizing atmospheres, they are compatible with thermal barrier coatings commonly employed in today’s engines, they have greater sensitivity than conventional wire thermocouples, and they are non-invasive to combustion aerodynamics in the engine. Thin film thermocouples based on platinum:palladium and indium oxynitride:indium tin oxynitride as well as their oxide counterparts have been developed for this purpose and have proven to be more stable than conventional type-S and type-K thin film thermocouples. The metallic and ceramic thin film thermocouples described within this paper exhibited remarkable stability and drift rates similar to bulk (wire thermocouples.

  6. In-pile experimental device for Sirene thermionic converters

    International Nuclear Information System (INIS)

    Bliaux, J.; Durand, J.; Lazare-Chopard, G.

    1969-01-01

    The irradiation device described here, was built for in pile life tests of 100 We SIRENE converters. The nuclear converter is located in a sealed vacuum chamber, which is plugged at the lower end of a coaxial tubing acting as electrical leads. The output power is available on a variable resistive load on the bank of the reactor pool. Thermal, electrical and neutronic parameters of the converter are recorded. Since 1967, two permanent devices allowed five experiments in the swimming pool TRITON (CEN-FAR) and the results, obtained till now, are presented. (authors) [fr

  7. Evaluation of RTD and thermocouple for PID temperature control in ...

    African Journals Online (AJOL)

    Evaluation of RTD and thermocouple for PID temperature control in distributed control system laboratory. D. A. A. Nazarudin, M. K. Nordin, A. Ahmad, M. Masrie, M. F. Saaid, N. M. Thamrin, M. S. A. M. Ali ...

  8. ECP measurements under neutron and gamma ray in in-pile loop and their data evaluation by water radiolysis calculations

    Energy Technology Data Exchange (ETDEWEB)

    Hanawa, S.; Nakamura, T.; Uchida, S. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan); Kus, P.; Vsolak, R.; Kysela, J. [Nuclear Research Inst. Rez plc, Husinec - Rez (Czech Republic)

    2010-07-01

    In order to establish reliable electrochemical corrosion potential (ECP) sensors for applying in reactor core peripherals of power plants, performance tests of sensors under irradiation were carried out in the in-pile loop of the experimental reactor, LVR-15, at the Nuclear Research Institute (NRI) in Czech Republic. Responses of different kinds of sensors under neutron and gamma irradiation conditions have been compared each other. Corrosive conditions along the in-pile loop were calculated by water radiolysis calculation code, WRAC-J and calculated corrosive conditions were compared with the measured results. As a result of the evaluation, it was confirmed that the ECP sensors could be applied to irradiation conditions of reactor peripherals, while the water radiolysis model could be also applied for evaluation of corrosive conditions of reactor peripherals. (author)

  9. In-pile Thermal Conductivity Characterization with Time Resolved Raman

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xinwei [Iowa State Univ., Ames, IA (United States). Dept. of Mechanical Engineering; Hurley, David H. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2018-03-19

    The project is designed to achieve three objectives: (1) Develop a novel time resolved Raman technology for direct measurement of fuel and cladding thermal conductivity. (2) Validate and improve the technology development by measuring ceramic materials germane to the nuclear industry. (3) Conduct instrumentation development to integrate optical fiber into our sensing system for eventual in-pile measurement. We have developed three new techniques: time-domain differential Raman (TD-Raman), frequency-resolved Raman (FR-Raman), and energy transport state-resolved Raman (ET-Raman). The TD-Raman varies the laser heating time and does simultaneous Raman thermal probing, the FR-Raman probes the material’s thermal response under periodical laser heating of different frequencies, and the ET-Raman probes the thermal response under steady and pulsed laser heating. The measurement capacity of these techniques have been fully assessed and verified by measuring micro/nanoscale materials. All these techniques do not need the data of laser absorption and absolute material temperature rise, yet still be able to measure the thermal conductivity and thermal diffusivity with unprecedented accuracy. It is expected they will have broad applications for in-pile thermal characterization of nuclear materials based on pure optical heating and sensing.

  10. The Phebus fission products in pile test programme

    International Nuclear Information System (INIS)

    Bussac, J.; Holtbecker, H.

    1988-01-01

    The need for quantifying the radioactive materials escaping from an LWR Nuclear Power Plant following a melt-down accident has arisen relatively late in the nuclear reactor technology development process. The TMI-2 accident in 1979 and the Chernobyl accident in 1986 have confirmed the importance of a good knowledge of phenomena which take place in a plant undergoing extreme accident conditions. After an extensive resarch effort which has involved the major nuclear countries for several years, we are now at the stage where a selective and converging attitude should be taken towards the wide range of problems underlying severe accidents. Selective, because we must understand what is important and what could be neglected. Converging, because we must arrive at a consensus at international level on the methods to treat these problems and a common understanding of the main scientific phenomena and the models to correctly represent them. After a large amount of separate effects tests and semi-integral in-pile and out-of-pile experiments, the Phebus FP project is being started as an experimental effort to quantify the relative importance of complicated processes and to give an insight into the interconnection of various mechanisms. The overall objective of this programme is to provide a qualified data base of integral in-pile experiments to validate codes dealing with FP transport in reactor core, primary cooling system and containment. This paper describes mainly the motivations and objectives of the Phebus PF programme

  11. In-pile loop experiments in water chemistry and corrosion

    International Nuclear Information System (INIS)

    Kysela, J.

    1986-09-01

    Results on the study of Zr-1% Nb alloy corrosion, in out-of and in-pile loops simulating the working conditions of the VVER-440 reactor (Soviet, PWR type), covered the time period May 1982-April 1986 were reported, as well as, results on transport and filtration of corrosion products. Methods and techniques used in the study included remote measurement of corrosion rate by polarizing resistance, out-of-pile loop at the temperature 350 deg. C, pressure 19 MPa, circulation 20 kgs/h and in-pile water loop with constant flow rate 10,000 kgs/h, pressure 16 MPa, temperature 330 deg. C and neutron flux 7x10 13 n/cm 2 .s. It was shown that solid suspended particles with chemical composition corresponding most frequently to magnetite or nickelous ferrite, though with non-stoichiometric composition Me x 2+ Fe 3- x 3+ O 4 were found. Continuous filtration of water by means of electromagnetic filter leads to a decrease of radioactivity of the outer epitactic layer only. Effect of filtration on the inner topotactic layer is negligible. The corrosion rates for the above-mentioned parameters are given

  12. External attachment of titanium sheathed thermocouples to zirconium nuclear fuel rods for the LOFT reactor

    International Nuclear Information System (INIS)

    Welty, R.K.

    1980-01-01

    The Exxon Nuclear Company, Inc., acting as a Subcontractor to EG and G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho, has developed a welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods. The fuel rods and thermocouples are used to test simulated loss-of-coolant accident (LOCA) conditions in a pressurized water reactor (LOFT Reactor, Idaho National Laboratory). A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A commercial pulsed laser and energy control system was installed along with specialized welding fixtures. Laser room facility requirements and tolerances were established. Performance qualifications, and detailed welding procedures were also developed. Product performance tests were conducted to assure that engineering design requirements could be met on a production basis

  13. In-pile IASCC growth tests of irradiated stainless steels in JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko; Nakano, Junichi; Nishiyama, Yutaka [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Shibata, Akira; Ohmi, Masao [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation-assisted stress corrosion cracking (IASCC) test plan to evaluate in-situ effects of neutron/{gamma}-ray irradiation on crack growth of irradiated stainless steels under high-temperature water conditions for commercial boiling water reactors (BWRs) using the Japan Materials Testing Reactor (JMTR). Crack growth rate and its electrochemical corrosion potential (ECP) dependence are different between in-pile test and post irradiation examination (PIE), but these differences are not fully understood. The objectives of the present study are to understand the difference between in-pile and out-of-pile IASCC growth and to confirm the effectiveness of mitigation due to lowering ECP on in-pile crack growth rates. For in-pile crack growth tests, we have selected a large compact tension specimen such as 0.5T-CT because of validity of SCC growth test at a high stress intensity factor (K-value). For loading a 0.5T-CT specimen up to K - 30 MPa {radical}m, we have adopted a lever type loading unit for in-pile crack growth tests in the JMTR. In this report, an in-pile test plan for crack growth of irradiated SUS316L stainless steels under simulated BWR conditions in the JMTR and current status of development of in-pile crack growth test techniques are presented. (author)

  14. Flow measurements using noise signals of axially displaced thermocouples

    Energy Technology Data Exchange (ETDEWEB)

    Kozma, R.; Hoogenboom, J.E. (Interuniversitair Reactor Inst., Delft (Netherlands))

    1990-01-01

    Determination of the flow rate of the coolant in the cooling channels of nuclear reactors is an important aspect of core monitoring. It is usually impossible to measure the flow by flowmeters in the individual channels due to the lack of space and safety reasons. An alternative method is based on the analysis of noise signals of the available in-core detectors. In such a noise method, a transit time which characterises the propagation of thermohydraulic fluctuations (density or temperature fluctuations) in the coolant is determined from the correlation between the noise signals of axially displaced detectors. In this paper, the results of flow measurements using axially displaced thermocouples in the channel wall will be presented. The experiments have been performed in a simulated MRT-type fuel assembly located in the research reactor HOR of the Interfaculty Reactor Institute, Delft. It was found that the velocities obtained via temperature noise correlation methods are significantly larger than the area-averaged velocity in the single-phase coolant flow. Model calculations show that the observed phenomenon can be explained by effects due to the radial velocity distribution in the channel. (author).

  15. In-pile loop experiments in water chemistry and corrosion

    International Nuclear Information System (INIS)

    Kysela, J.; Jindrich, K.; Masarik, V.; Fric, Z.; Chotivka, V.; Hamerska, H.; Vsolak, R.; Erben, O.

    1986-08-01

    Methods and techniques used were as follows: (a) Method of polarizing resistance for remote monitoring of instantaneous rate of uniform corrosion. (b) Out-of-pile loop at the temperature 350 degC, pressure 19 MPa, circulation 20 kgs/h, testing time 1000 h. (c) High temperature electromagnetic filter with classical solenoid and ball matrix for high pressure filtration tests. (d) High pressure and high temperature in-pile water loop with coolant flow rate 10 000 kgs/h, neutron flux in active channel 7x10 13 n/cm 2 .s, 16 MPa, 330 degC. (e) Evaluation of experimental results by chemical and radiochemical analysis of coolant, corrosion products and corrosion layer on surface. The results of measurements carried out in loop facilities can be summarized into the following conclusions: (a) In-pile and out-of-pile loops are suitable means of investigating corrosion processes and mass transport in the nuclear power plant primary circuit. (b) In studying transport phenomena in the loop, it is necessary to consider the differences in geometry of the loop and the primary circuit, mainly the ratio of irradiated and non-irradiated surfaces and volumes. (c) In the experimental facility simulating the WWER-type nuclear power plant primary circuit, solid suspended particles of a chemical composition corresponding most frequently to magnetite or nickel ferrite, though with non-stoichiometric composition Me x 2+ Fe 3-x 3+ O 4 , were found. (d) Continuous filtration of water by means of an electromagnetic filter removing large particles of corrosion products leads to a decrease in radioactivity of the outer epitactic layer only. The effect of filtration on the inner topotactic layer is negligible

  16. New in-pile water loop facility for IASCC studies at JMTR

    International Nuclear Information System (INIS)

    Tsukada, T.; Tsuji, H.; Nakajima, H.; Komori, Y.; Ito, H.

    2002-01-01

    Irradiation assisted stress corrosion cracking (IASCC) is caused by the synergistic effects of neutron and gamma radiation, residual and applied stresses and high temperature water environment on the structural materials of vessel internals. IASCC has been studied since the beginning of the 1980's and the phenomenological knowledge on IASCC is accrued extensively. However, mainly due to the experimental difficulties, data for the mechanistic understanding and prediction of failures of the specific in-vessel components are still insufficient and further well-controlled experiments are needed [1]. In recent years, efforts to perform the in-pile materials test for IASCC study have been made at some research reactors [2-4]. At JAERI, a high temperature water loop facility was designed to install at the Japan Materials Testing Reactor (JMTR) to carry out the in-core IASCC testing. This report describes an overview of design and specification of the loop facility. (authors)

  17. In-pile Hydrothermal Corrosion Evaluation of Coated SiC Ceramics and Composites

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, David [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Ang, Caen [Univ. of Tennessee, Knoxville, TN (United States); Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    Hydrothermal corrosion accelerated by water radiolysis during normal operation is among the most critical technical feasibility issues remaining for silicon carbide (SiC) composite-based cladding that could provide enhanced accident-tolerance fuel technology for light water reactors. An integrated in-pile test was developed and performed to determine the synergistic effects of neutron irradiation, radiolysis, and pressurized water flow, all of which are relevant to a typical pressurized water reactor (PWR). The test specimens were chosen to cover a range of SiC materials and a variety of potential options for environmental barrier coatings. This document provides a summary of the irradiation vehicle design, operations of the experiment, and the specimen loading into the irradiation vehicle.

  18. Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

    Energy Technology Data Exchange (ETDEWEB)

    Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-09-01

    An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

  19. Mineral insulated thermocouples - installation in steam generating plant

    International Nuclear Information System (INIS)

    Bridges, W.J.; Brown, J.F.

    1980-01-01

    The main areas of interest considered are Central Station Fossil Fuel fired boilers of around 500 MW capacity, AGR Boilers, and Industrial and Research Development projects. While the requirement for temperature measurement in each of these areas may vary the techniques adopted to overcome installation and protection problems created by thermal, chemical and mechanical hazards remain basically the same. The reasons for temperature measurement are described together with methods of attachment development and procedures for protection of the thermocouple along its route length until its exit from the hazardous environment. These relative accuracies of the different attachments are discussed along with factors influencing the life of the thermocouple. In many instances thermocouple installation is either a once only opportunity and/or an expensive exercise. It is therefore essential to develop and apply an effective quality control system during the installation phase. An effective system is described. Finally, a brief outline of possible future trends is given. (author)

  20. An Innovative Flow-Measuring Device: Thermocouple Boundary Layer Rake

    Science.gov (United States)

    Hwang, Danny P.; Fralick, Gustave C.; Martin, Lisa C.; Wrbanek, John D.; Blaha, Charles A.

    2001-01-01

    An innovative flow-measuring device, a thermocouple boundary layer rake, was developed. The sensor detects the flow by using a thin-film thermocouple (TC) array to measure the temperature difference across a heater strip. The heater and TC arrays are microfabricated on a constant-thickness quartz strut with low heat conductivity. The device can measure the velocity profile well into the boundary layer, about 65 gm from the surface, which is almost four times closer to the surface than has been possible with the previously used total pressure tube.

  1. Magnetic tunnel junction thermocouple for thermoelectric power harvesting

    Science.gov (United States)

    Böhnert, T.; Paz, E.; Ferreira, R.; Freitas, P. P.

    2018-05-01

    The thermoelectric power generated in magnetic tunnel junctions (MTJs) is determined as a function of the tunnel barrier thickness for a matched electric circuit. This study suggests that lower resistance area product and higher tunnel magnetoresistance will maximize the thermoelectric power output of the MTJ structures. Further, the thermoelectric behavior of a series of two MTJs, a MTJ thermocouple, is investigated as a function of its magnetic configurations. In an alternating magnetic configurations the thermovoltages cancel each other, while the magnetic contribution remains. A large array of MTJ thermocouples could amplify the magnetic thermovoltage signal significantly.

  2. Failure of sheathed thermocouples due to thermal cycling

    International Nuclear Information System (INIS)

    Anderson, R.L.; Ludwig, R.L.

    1982-03-01

    Open circuit failures (up to 100%) in small-diameter thermocouples used in electrically heated nuclear fuel rod simulator prototypes during thermal cycling tests were investigated to determine the cause(s) of the failures. The experiments conducted to determine the relative effects of differential thermal expansion, wire size, grain size, and manufacturing technology are described. It was concluded that the large grain size and embrittlement which result from certain common manufacturing annealing and drawing procedures were a major contributing factor in the breakage of the thermocouple wires

  3. Use of standard reliability levels in design and safety assessment of in-pile loops

    International Nuclear Information System (INIS)

    Bogani, G.; Verre, A.; Balestreri, S.; Colombo, A.G.; Luisi, T.

    1975-01-01

    This paper describes a logic-probabilistic analysis technique for a critical design review and safety assessment of in-pile loops. The examples in this paper refer to the analysis performed for the experimental loops already constructed or under construction in the ESSOR reactor of the Joint Research Centre of Ispra, as irradiation facilities for fuel element research and development tests. The proposed technique is based on the classification into categories of components and protective device malfunctions. Such subdivision into categories was agreed upon by the Italian Safety Authority and Euratom JRC, and adopted for the safety assessment of the ESSOR reactor in-pile loops. For each category, the method makes a link with a corresponding malfunction probability range (probability level). This probability level is defined taking into account design, construction, inspection and maintenance criteria as well as periodic controls; therefore the quality level and consequently the reliability level are thus also defined. The analysis is developed in the following stages: (1) definition of the analysis object (top event) and drawing of the relative fault-tree; (2) loop design analysis and preliminary optimization based on logic criteria; (3) classification into categories of the fault-tree primary events; (4) final loop design analysis and optimization based on defined component quality requirements. Stages 2 and 4 are quite different since stage 2 mainly consists of a redundance optimization, while stage 4 acts on the component quality level in such a way that each minimum cut-set leading to the top has an acceptable probability level. During analysis development, use is made of computer codes which, among other things enable the verification of fault-tree logic makeup, the listing of the minimum cut-sets with and without event categorization, and the evaluation of each cut-set order. (author)

  4. Heat penetration and thermocouple location in home canning.

    Science.gov (United States)

    Etzel, Mark R; Willmore, Paola; Ingham, Barbara H

    2015-01-01

    We processed applesauce, tomato juice, and cranberries in pint jars in a boiling water canner to test thermal processing theories against home canning of high-acid foods. For each product, thermocouples were placed at various heights in the jar. Values for f h (heating), f cl (cooling), and F 82.2°C (lethality) were determined for each thermocouple location, and did not depend substantially on thermocouple location in accordance with heat transfer theory. There was a cold spot in the jar, but the cold spot during heating became the hot spot during cooling. During heating, the geometric center was the last to heat, and remained coldest the longest, but during coooling, it was also the last to cool, and remained hottest the longest. The net effect was that calculated lethality in home canning was not affected by thermocouple location. Most of the lethality during home canning occurred during air cooling, making cooling of home canned foods of great importance. Calculated lethality was far greater than the required 5-log reduction of spores in tomato juice and vegetative cells in cranberries, suggesting a wide margin of safety for approved home-canning processes for high-acid foods.

  5. Realization of Copper Melting Point for Thermocouple Calibrations

    Directory of Open Access Journals (Sweden)

    Y. A. ABDELAZIZ

    2011-08-01

    Full Text Available Although the temperature stability and uncertainty of the freezing plateau is better than that of the melting plateau in most of the thermometry fixed points, but realization of melting plateaus are easier than that of freezing plateaus for metal fixed points. It will be convenient if the melting points can be used instead of the freezing points in calibration of standard noble metal thermocouples because of easier realization and longer plateau duration of melting plateaus. In this work a comparison between the melting and freezing points of copper (Cu was carried out using standard noble metal thermocouples. Platinum - platinum 10 % rhodium (type S, platinum – 30 % rhodium / platinum 6 % rhodium (type B and platinum - palladium (Pt/Pd thermocouples are used in this study. Uncertainty budget analysis of the melting points and freezing points is presented. The experimental results show that it is possible to replace the freezing point with the melting point of copper cell in the calibration of standard noble metal thermocouples in secondary-level laboratories if the optimal methods of realization of melting points are used.

  6. The development of a fast response thermocouple for use in liquid metals

    International Nuclear Information System (INIS)

    Morss, A.G.; Vincent, B.

    1987-03-01

    Work carried out at Berkeley Nuclear Laboratories to develop a fast-response thermocouple for use in liquid metals is described. This thermocouple because of its unique construction, has a junction mass approaching zero and hence its frequency response should be very high. Some of the problems of manufacture are discussed, in particular the high quality of seal required to avoid ingress of liquid metal. A comparison of results obtained with the fast-response thermocouple and with conventional stainless-steel-sheathed thermocouples is made. The improved response of the new thermocouple is clearly visible, hence confirming that measurements made with sheathed thermocouples suffer attenuation. It is concluded that results obtained with the fast-response thermocouple are close to the real magnitude of temperature fluctuations present in turbulent flow. It is also demonstrated that, with suitable corrections, results obtained with sheathed thermocouples can be used to estimate the real signals present in the flow. (author)

  7. Advanced In-pile Instrumentation for Material and Test Reactors

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Daw, J.E.; Unruh, T.C.; Chase, B.M.; Davis, K.L.; Palmer, A.J.; Schley, R.S.

    2013-06-01

    The US Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified; and the progress of other development efforts is summarized. As reported in this paper, INL staff is currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating 'advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors. (authors)

  8. A comprehensive in-pile test of PWR fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang Rixin; Zhang Shucheng; Chen Dianshan (Academia Sinica, Beijing (China). Inst. of Atomic Energy)

    1991-02-01

    An in-pile test of PWR fuel bundle has been conducted in HWRR at IAE of China. This paper describes the structure of the test bundle (3x3-2), fabrication process and quality control of the fuel rod, irradiation conditions and the main Post Irradiation Examination (PIE) results. The test fuel bundle was irradiated under the PWR operation and water chemistry conditions with an average linear power of 381 W/cm and reached an average burnup of 25010 MWd/tU of the fuel bundle. After the test, destructive and non-destructive examination of the fuel rods was conducted at hot laboratories. The fission gas release was 10.4-23%. The ridge height of cladding was 3 to 8 {mu}m. The hydrogen content of the cladding was 80 to 140 ppm. The fuel stack height was increased by 2.9 to 3.3 mm. The relative irradiation growth was about 0.11 to 0.17% of the fuel rod length. During the irradiation test, no fuel rod failure or other abnormal phenomena had been found by the on-line fuel failure monitoring system of the test loop and water sampling analysis. The structure of the test fuel assembly was left undamaged without twist and detectable deformation. (orig.).

  9. Advanced In-Pile Instrumentation for Materials Testing Reactors

    Science.gov (United States)

    Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T. C.; Chase, B. M.; Davis, K. L.; Palmer, A. J.; Schley, R. S.

    2014-08-01

    The U.S. Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified, and the progress of other development efforts is summarized. As reported in this paper, INL researchers are currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating `advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers, are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors.

  10. Preliminary results from initial in-pile debris bed experiments

    International Nuclear Information System (INIS)

    Rivard, J.B.

    1977-01-01

    An accident in a liquid metal fast breeder reactor (LMFBR) in which molten core material is suddenly quenched with subcooled liquid sodium could result in extensive fragmentation and dispersal of fuel as subcritical beds of frozen particulate debris within the reactor vessel. Since this debris will continue to generate power due to decay of retained fission products, containment of the debris is threatened if the generated heat is not removed. Therefore, the initial safety question is the capacity which debris beds may have for transfer of the decay heat to overlying liquid sodium by natural processes--i.e., without the aid of forced circulation of the coolant. Up to the present time, all experiments on debris bed behavior either have used substitute materials (e.g., sand and water) or have employed actual materials, but atypical heating methods. Increased confidence in the applicability of debris bed simulations is afforded if the heat is generated within the fuel component of the appropriate fast reactor materials. The initial series of in-pile tests reported on herein constitutes the first experiments in which the internal heating mode has been produced in particulate oxide fuel immersed in liquid sodium. Fission heating of the fully-enriched UO 2 in the experiment while it is contained within Sandia Laboratories Annular Core Pulse Reactor (ACPR), operating in its steady-state mode, approximates the decay heating of debris. Preliminary results are discussed

  11. Innovations for In-Pile Measurements in the Framework of the CEA-SCK•CEN Joint Instrumentation Laboratory

    Science.gov (United States)

    Villard, Jean-Francois; Schyns, Marc

    2010-12-01

    Optimizing the life cycle of nuclear systems under safety constraints requires high-performance experimental programs to reduce uncertainties on margins and limits. In addition to improvement in modeling and simulation, innovation in instrumentation is crucial for analytical and integral experiments conducted in research reactors. The quality of nuclear research programs relies obviously on an excellent knowledge of their experimental environment which constantly calls for better online determination of neutron and gamma flux. But the combination of continuously increasing scientific requirements and new experimental domains -brought for example by Generation IV programsnecessitates also major innovations for in-pile measurements of temperature, dimensions, pressure or chemical analysis in innovative mediums. At the same time, the recent arising of a European platform around the building of the Jules Horowitz Reactor offers new opportunities for research institutes and organizations to pool their resources in order to face these technical challenges. In this situation, CEA (French Nuclear Energy Commission) and SCK'CEN (Belgian Nuclear Research Centre) have combined their efforts and now share common developments through a Joint Instrumentation Laboratory. Significant progresses have thus been obtained recently in the field of in-pile measurements, on one hand by improvement of existing measurement methods, and on the other hand by introduction in research reactors of original measurement techniques. This paper highlights the state-of-the-art and the main requirements regarding in-pile measurements, particularly for the needs of current and future irradiation programs performed in material testing reactors. Some of the main on-going developments performed in the framework of the Joint Instrumentation Laboratory are also described, such as: - a unique fast neutron flux measurement system using fission chambers with 242Pu deposit and a specific online data processing

  12. In-pile and out-of-pile testing of a molybdenum-uranium dioxide cermet fueled themionic diode

    Science.gov (United States)

    Diianni, D. C.

    1972-01-01

    The behavior of Mo-UO2 cermet fuel in a diode for thermionic reactor application was studied. The diode had a Mo-0.5 Ti emitter and niobium collector. Output power ranged from 1.4 to 2.8 W/cm squared at emitter and collector temperatures of 1500 deg and 540 C. Thermionic performance was stable within the limits of the instrumentation sensitivity. Through 1000 hours of in-pile operation the emitter was dimensionally stable. However, some fission gases (15 percent) leaked through an inner clad imperfection that occurred during fuel fabrication.

  13. In-pile observations of fuel and clad relocation during LMFBR initiation phase accident experiments - the STAR experiments

    International Nuclear Information System (INIS)

    Wright, S.A.; Schumacher, G.; Henkel, P.R.; Royl, P.

    1987-01-01

    A series of seven in-pile experiments (the STAR experiments) were performed in which clad motion and fuel dispersal were observed in small pin bundles with high-speed cinematography. The experimental heating conditions reproduced a range of Loss of Flow (LOF) accident scenarios for the lead subassemblies in LMFBRs. The experiments show strong tendencies for limited clad motion in multiple pin bundles, early fuel disruption and dispersal (prior to fuel melting) in moderate power transients having simultaneous clad melting and fuel disruption. The more recent experiments indicate a possibility of steel vapor driven fuel dispersal after fuel breakup and intimate fuel/steel mixing. (author)

  14. Design criteria and fabrication in-pile test section of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1997-10-01

    Safety state fuel test loop will be equipped in HANARO to obtain the development and betterments of advanced fuel and materials through the irradiation tests. The objective of this study is to determine the design criteria and technical specification of in-pile test section and to specify the manufacturing requirements of in-pile test section. HANARO fuel test loop was designed to meet the CANDU and PWR fuel testing and in-pile section will be manufactured and installed in HANARO. The design criteria and technical specification of in-pile test section could be used the fuel and materials design with for irradiation testing IPS of HANARO fuel test loop. This results will become guidances for the planning and programming of irradiation testing. (author). 12 refs., tabs., figs.

  15. Development of In-pile Plug Assembly and Primary Shutter for Cold Neutron Guide System

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Jin Won; Cho, Yeong Garp; Ryu, Jeong Soo; Lee, Jung Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    The HANARO, a 30 MW multi-purpose research reactor in Korea, will be equipped with a neutron guide system, in order to transport cold neutrons from the neutron source to the neutron scattering instruments in the neutron guide hall near the reactor building. The neutron guide system of HANARO consists of the in-pile plug assembly with in-pile guides, the primary shutter with in-shutter guides, the neutron guides in the guide shielding room with dedicated secondary shutters, and the neutron guides connected to the instruments in the neutron guide hall. The functions of the in-pile plug assembly are to shield the reactor environment from a nuclear radiation and to support the neutron guides and maintain them precisely oriented. The primary shutter is a mechanical device to be installed just after the in-pile plug assembly, which stops neutron flux on demand. This report describes the mechanical design, fabrication, and installation procedure of the in-pile plug assembly and the primary shutter for the neutron guide system at HANARO. A special tool and procedure for a replacement of in-pile plug and guide cassette is also presented with the interface condition in the reactor hall.

  16. Zircaloy-sheathed element rods fitted with thermo-couples

    International Nuclear Information System (INIS)

    Bernardy de Sigoyer, B.; Jacques, F.; Thome, P.

    1963-01-01

    In order to carry out thermal conductivity measurements on UO 2 in conditions similar to those under which fuel rods are used, it was necessary to measure the temperature at the interior of a fuel element sheathed in zircaloy. The temperatures are taken with Thermocoax type thermocouples, that is to say fitted with a very thin sheath of stainless steel or Inconel. It is known also that fusion welding of zircaloy onto stainless steel is impossible and that high temperature welded joints are very difficult because of their aggressiveness. The technique used consists in brazing the thermocouples to relatively large stainless steel parts and then joining these plugs by electron bombardment welding to diffused stainless steel-zircaloy couplings. The properties of these diffused couplings and of the brazed joints were studied; the various stages in the fabrication of the containers are also described. (authors) [fr

  17. Classification of Unknown Thermocouple Types Using Similarity Factor Measurement

    Directory of Open Access Journals (Sweden)

    Seshu K. DAMARLA

    2011-01-01

    Full Text Available In contrast to classification using PCA method, a new methodology is proposed for type identification of unknown thermocouple. The new methodology is based on calculating the degree of similarity between two multivariate datasets using two types of similarity factors. One similarity factor is based on principle component analysis and the angles between the principle component subspaces while the other is based on the Mahalanobis distance between the datasets. Datasets containing thermo-emfs against given temperature ranges are formed for each type of thermocouple (e.g. J, K, S, T, R, E, B and N type by experimentation are considered as reference datasets. Datasets corresponding to unknown type are captured. Similarity factor between the datasets one of which being the unknown type and the other being each known type are compared. When maximum similarity factor occurs, then the class of unknown type is allocated to that of known type.

  18. Thermocouple placement and hot spots in radioactive waste tanks

    International Nuclear Information System (INIS)

    Barker, J.J.

    1991-06-01

    Analytical solutions available in Carslaw and Jaeger's Conduction of Heat in Solids for continuous point sources and for continuous finite sources are used to demonstrate that placement of thermocouples on a fine enough grid to detect a hot spot is impracticable for existing waste tanks but fortunately not necessary. Graphs covering ranges of diffusivities, times, temperatures and heat generation rates are included. 2 refs., 8 figs., 5 tabs

  19. Recommendations for the specification of thermocouples for nuclear applications

    International Nuclear Information System (INIS)

    1977-05-01

    This Code of Practice is a guide for use in the preparation of individual specifications to cover, as fully as possible the conditions governing the supply of raw materials and the ordering, manufacture, testing, inspection, handling and installation of thermocouples for use in nuclear environments in order that reliable, consistent and generally acceptable results can be obtained. The insulation resistance values quoted in this document apply to magnesium oxide. If other insulants are called for, appropriate values must be specified. (author)

  20. In-pile test results of HANA claddings in Halden research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Choi, Byoung Kwon; Jeong, Yong Hwan; Jung, Yun Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    It is a kind of facing tasks in the nuclear industry to develop advanced claddings for high burn-up fuel which is safer and more economical than the existing conventional ones. Since 1997, taking an initiative in KAERI, the Zr cladding development team has carried out the R and D activities for the development of the advanced claddings to be used in the high burn-up fuel (>70,000 MWD.MTU). The team had produced the advanced claddings (HANA, High-performance Alloy for Nuclear Application) from the patented composition and manufacturing process in the international collaboration with U.S. and Japan. Now, the HANA claddings have being demonstrated their good performances from the out-of-pile tests including the corrosion, creep, burst, tensile, microstructures LOCA, RIA, wear, and so on. In parallel to the out-of-pile performance tests, the HANA claddings are being undertaken to evaluate their in-pile properties in Halden research reactor. In this study, it is included the test overviews, conditions, and results of the HANA claddings in the Halden reactor.

  1. Feasibility Study of Thin Film Thermocouple Piles

    Science.gov (United States)

    Sisk, R. C.

    2001-01-01

    Historically, thermopile detectors, generators, and refrigerators based on bulk materials have been used to measure temperature, generate power for spacecraft, and cool sensors for scientific investigations. New potential uses of small, low-power, thin film thermopiles are in the area of microelectromechanical systems since power requirements decrease as electrical and mechanical machines shrink in size. In this research activity, thin film thermopile devices are fabricated utilizing radio frequency sputter coating and photoresist lift-off techniques. Electrical characterizations are performed on two designs in order to investigate the feasibility of generating small amounts of power, utilizing any available waste heat as the energy source.

  2. Effects of thermocouple installation and location on fuel rod temperature measurements

    International Nuclear Information System (INIS)

    McCormick, R.D.

    1983-01-01

    This paper describes the results of analyses of nuclear fuel rod cladding temperature data obtained during in-reactor experiments under steady state and transient (simulated loss-of-coolant accident) operating conditions. The objective of the analyses was to determine the effect of thermocouple attachment method and location on measured thermal response. The use of external thermocouples increased the time to critical heat flux (CHF), reduced the blowdown peak temperature, and enhanced rod quench. A comparison of laser welded and resistance welded external thermocouple responses showed that the laser welding technique reduced the indicated cladding steady state temperatures and provided shorter time-to-CHF. A comparison of internal welded and embedded thermocouples indicated that the welded technique gave generally unsatisfactory cladding temperature measurements. The embedded thermocouple gave good, consistent results, but was possibly more fragile than the welded thermocouples. Detailed descriptions of the thermocouple designs, attachment methods and locations, and test conditions are provided

  3. Calibration Technique of the Irradiated Thermocouple using Artificial Neural Network

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jin Tae; Joung, Chang Young; Ahn, Sung Ho; Yang, Tae Ho; Heo, Sung Ho; Jang, Seo Yoon [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    To correct the signals, the degradation rate of sensors needs to be analyzed, and re-calibration of sensors should be followed periodically. In particular, because thermocouples instrumented in the nuclear fuel rod are degraded owing to the high neutron fluence generated from the nuclear fuel, the periodic re-calibration process is necessary. However, despite the re-calibration of the thermocouple, the measurement error will be increased until next re-calibration. In this study, based on the periodically calibrated temperature - voltage data, an interpolation technique using the artificial neural network will be introduced to minimize the calibration error of the C-type thermocouple under the irradiation test. The test result shows that the calculated voltages derived from the interpolation function have good agreement with the experimental sampling data, and they also accurately interpolate the voltages at arbitrary temperature and neutron fluence. That is, once the reference data is obtained by experiments, it is possible to accurately calibrate the voltage signal at a certain neutron fluence and temperature using an artificial neural network.

  4. Investigation of special capsule technologies for material in-pile irradiation test and development plan in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Cho, M. S.; Son, J. M.; Kim, D. S.; Park, S. J.; Cho, Y. G.; Seo, C. K.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    In-pile test for several materials such as Zr alloy, stainless steel, Cr-Ni steel etc. which are used as structural material of the advanced reactor and KNGR(Korea Next Generation Reactor) like SMART, is necessary to produce the design data for developing new reactor materials. Advanced countries like USA, Europe and Japan etc. are not only performing the simple irradiation test for materials, but developing many kinds of special capsule to perform in-pile test having special purpose. For the special test items of fuel rod, fission products, total heat generation, swelling, deformation, sweep gas, temperature ramping and BOCA etc. are being actively concerned. There are capsules measuring creep, fatigue, crack growth, and controlling fluence etc. for special irradiation test of materials. In addition, the advanced countries are developing several instrument technologies suitable for the special capsules. In HANARO, non-instrumented, instrumented material capsules and non-instrumented fuel capsule have been developed and they have been utilized in the irradiation test for users, and creep capsule loading single specimen was made and is planned to test in the reactor soon. For some forthcoming years, special capsules not only measuring creep deformation with multi-specimens, fatigue, controlling fluence but crack propagation and gas sweep considering the requirements of users will be developed in HANARO.

  5. Zircaloy-sheathed element rods fitted with thermo-couples; Barre combustible a thermocouple gainee de zircaloy

    Energy Technology Data Exchange (ETDEWEB)

    Bernardy de Sigoyer, B; Jacques, F; Thome, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    In order to carry out thermal conductivity measurements on UO{sub 2} in conditions similar to those under which fuel rods are used, it was necessary to measure the temperature at the interior of a fuel element sheathed in zircaloy. The temperatures are taken with Thermocoax type thermocouples, that is to say fitted with a very thin sheath of stainless steel or Inconel. It is known also that fusion welding of zircaloy onto stainless steel is impossible and that high temperature welded joints are very difficult because of their aggressiveness. The technique used consists in brazing the thermocouples to relatively large stainless steel parts and then joining these plugs by electron bombardment welding to diffused stainless steel-zircaloy couplings. The properties of these diffused couplings and of the brazed joints were studied; the various stages in the fabrication of the containers are also described. (authors) [French] Pour des mesures de conductivite thermique de l'UO{sub 2} dans des conditions voisines du fonctionnement des barres combustibles, il s'agissait de mesurer la temperature a l'interieur d'un element combustible gaine de zircaloy. Les prises de temperature sont faites par thermocouples du type Thermocoax, c'est-a-dire pourvu d'une gaine tres mince en inox ou inconel. Par ailleurs on sait que le soudage par fusion du zircaloy sur l'inox est impossible et que les brasures a haute temperature sont difficiles car tres agressives. La technique utilisee consiste a braser les thermocouples sur des pieces en inox relativement massives et de rapporter par soudage au bombardement electronique ces bouchons sur des raccords diffuses zircaloy-inox. Les proprietes de ces raccords diffuses et celles de joints brases ont ete etudiees; on expose egalement les diverses etapes de fabrication des containers. (auteurs)

  6. In-Pile thermal fatigue of First Wall mock-ups under ITER relevant conditions

    International Nuclear Information System (INIS)

    Blom, F.; Schmalz, F.; Kamer, S.; Ketema, D.J.

    2006-01-01

    The objective of this study is to perform in-pile thermal fatigue testing of three actively cooled First Wall (FW) mock-ups to check the effect of neutron irradiation on the Be/CuCrZr joints under representative FW operation conditions. Three FW mock-ups with Beryllium armor tiles will be neutron irradiated at 1 dpa (in Be) with parallel thermal fatigue testing for 30,000 cycles. The temperatures, stress distributions and stress amplitudes at the Be/CuCrZr interface of the mock-ups will be as close as possible to the values calculated for ITER FW panels. For this objective the PWM mocks-up subjected to thermal fatigue will be integrated with high density (W) plates on the Be-side to provide heat flux by nuclear heating. The assembly will be placed in the pool-side facility of the HFR and thermal cycling is then arranged by mechanical movement towards and from the core box. As the thermal design of the irradiation rig is very critical a pilot-irradiation will be performed to cross check the models used in the thermal design of the rig. The project is currently in the design phase of both the pilot and actual irradiation rig. The irradiation of the actual rig is planned to start at mid 2007 and last for two years. (author)

  7. In-pile creep strain and failure of CW 316 Ti pressurized tubes

    International Nuclear Information System (INIS)

    Boutard, J.L.; Maillard, A.; Carteret, Y.; Levy, V.; Meny, L.

    1984-06-01

    The in-pile creep and failure behavior of CW 316 Ti pressurized tubes irradiated in the same rig at 660-680 0 C and 81.4 dpaF max in Phenix is presented and compared to monitors of the same heat. The in-pile plastic strains are of the same order of what is expected from the monitors and are rather independent of the dose rate in the range 4 to 9 x 10 -3 dpaF/h. Such a behavior supports the assumption that the out-of-pile deformation mechanisms are operative in pile and a certain balance occurs between modification of the microstructure, dynamic hardening and deformation mechanisms due to irradiation. Examinations by fractography and optical micrography, show that the failures are intergranular either in-pile or out-of-pile. In both cases the damage consists in intergranular wedge cracks, and no cavitation can be observed by transmission electron microscopy. Then the in-pile embrittlement which gives lower failure strain and time is to be associated with a decrease of the surface energy of grain-boundaries rather then growth and coalescence of cavities

  8. Description of an identification method of thermocouple time constant based on application of recursive numerical filtering to temperature fluctuation

    International Nuclear Information System (INIS)

    Bernardin, B.; Le Guillou, G.; Parcy, JP.

    1981-04-01

    Usual spectral methods, based on temperature fluctuation analysis, aiming at thermocouple time constant identification are using an equipment too much sophisticated for on-line application. It is shown that numerical filtering is optimal for this application, the equipment is simpler than for spectral methods and less samples of signals are needed for the same accuracy. The method is described and a parametric study was performed using a temperature noise simulator [fr

  9. Design modification of the in-pile test section for increase of sealing capability

    Energy Technology Data Exchange (ETDEWEB)

    Hong, J T; Ahn, S H; Joung, C Y; Jeong, H Y; Lee, J M; Sim, B S [Department of Research Reactor Utilization and Development, Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2012-03-15

    Since KAERI established the fuel test loop (FTL) at HANARO in 2009, KAERI has carried out several experiments to verify the performances of the equipment. Based on the experiments, the design modification of the In-Pile test Section (IPS) has been processed to improve some difficulties such as difficulty in ejecting the inner assembly of the IPS from the pressure vessel, difficulty of the sealing process of the cooling water, etc. At first, because the cooling water of HANARO in KAERI consists of an open-pool type, if a certain shock is generated during the disassembly process, the cooling water can be spattered out of the pool. Therefore, two jacking bolts will be added on the top flange part of the inner assembly to decrease the shock. Second, at the pressure boundary of the IPS where MI-cables go through, the brazing process has been used to seal out the cooling water. However, because the length of the IPS is up to 5.5 meters, it is too difficult and time consuming to carry out the brazing process at the end part of the IPS. Therefore, the brazing process will be replaced with the mechanical sealing structure to simplify the assembly process. (author)

  10. Design Improvement of Double Pressure Vessel in the In-pile Test Section

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Heo, Sung-Ho; Joung, Chang-Young; Kim, Ka-Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    To carry out an irradiation test of nuclear fuels, a nuclear fuel test rig should be fabricated and installed in the in-pile test section (IPS), which is installed in the reactor hall. While carrying out an irradiation test, sealing out coolant which passes through the test rig is one of the most important issues. In particular, although the double pressure vessel is assembled with the IPS head by two o-rings and six bolts, 15.5 MPa of highly pressurized coolant leaks through the gap between the vessel and IPS head. Because the temperature of the coolant in the test loop is 300 .deg. C , and the pool of HANARO is 40 .deg. C, the double pressure vessel is necessary to insulate them. Therefore, a new design to prevent the leakage of coolant needs to be developed. In this study, EB welding technique is considered to assemble the double pressure vessel and the IPS head, and their mechanical design is modified to enable the welding process. In this study, an improved design for sealing out the coolant at the pressure boundary between the double pressure vessel and the IPS head has been developed. An EB weld is applied to seal out the pressure boundary, and its sealing performance is verified by NDE, a cross section test, and a hydraulic pressure test. From the verification test results, the improved design can be used in fabricating the IPS for a nuclear fuel irradiation test.

  11. The design of in-pile test section for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, K. N.; Lee, J. M.; Shim, B. S.; Zee, D. Y.; Park, S. H.; Ahn, S. H.; Lee, J. Y.; Kim, Y. J. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    As an equipment for nuclear fuel's general performance irradiation test in HANARO, Fuel Test Loop(FTL) has been developed that can irradiate the pin to the maximum number of 3 at the core irradiation hole(IR1 hole) by considering for it's utility and user's irradiation requirement. 3-Pin FTL consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). IPS consists for IPS Vessel assembly, In-Pool Piping, IPS Support, In-Pool Piping Support etc. Design that such IPS considers interference item consisted to do not bear in existing facilities by one. IVA that is connected to the OPS are controlled and regulated by means of system pressure, system temperature and the water quality. IPS Vessel assembly is consisted of outer pressure vessel, inner pressure vessel, IPS head, inner assembly and test fuel carrier. After 3-Pin FTL development which is expected to be finished by the 2006, FTL will be used for the irradiation test of the new PWR-type fuel and can maximize the usage of HANARO.

  12. In-pile Creep Tests of Zircaloy Tubing in the Studsvik R2 Reactor. Final Report

    International Nuclear Information System (INIS)

    Tomani, Hans; Lindeloew, Ulf

    2000-12-01

    In this report are presented the findings of a prototype creep test on Zr4 guide tube specimens exposed in-pile and out-of-pile and stressed by constant bending moments. The calculated initial deflection curvature caused by the applied bending moment agrees very well with the measured initial values. Furthermore, the measurement results show excellent consistency. The dominant impact of neutron irradiation is clearly demonstrated. After 3 cycles (∼1300 hours) the irradiation creep is 4 times as large as the thermal creep. This is the case at least when fresh tube material is used. Irradiation creep progresses steadily, but the creep rate is not quite constant during the 3 irradiation cycles. The thermal creep, on the other hand, quickly saturates and there is hardly any further deflection after the second cycle for the specimen situated above the core. A limitation with the rig has been that the tube deflection became limited by the rig carrier body of the rig in the neutron flux (core) that disqualified the results of a fourth irradiation cycle actually performed in the fall of 1998. The test method appears to be suitable for testing the bending creep of different guide tube materials or designs under PWR conditions

  13. Ultrasonic Thermometry for In-Pile Temperature Detection

    International Nuclear Information System (INIS)

    Daw, J.E.; Rempe, J.L.; Wilkins, S.C.

    2002-01-01

    example, signal processing can be very complicated, as multiple echoes may overlap. Contact between the sensor and solid materials can cause extraneous echoes. If a sheath is required, contact bonding at high temperatures may cause extraneous echoes or attenuation of primary echoes. The most successful materials used in previous studies, tungsten and rhenium, are unattractive for nuclear applications due to material transmutation. Clearly, in order for ultrasonic thermometers to be viable for an in-pile sensor, these issues must be resolved through the use of modern signal processing and materials technologies. As part of the INL feasibility study, all of the issues associated with UT use and proposed resolution options will be identified and evaluated. Once most promising options are proven, it is planned to produce one or more prototype ultrasonic temperature sensors for evaluation. Ultimately, a full test should include a long term installation in a high temperature test assembly installed in a high neutron flux environment, such as that found in the Idaho National Laboratory's Advanced Test Reactor.

  14. The disposition of can thermocouples in a nuclear reactor

    International Nuclear Information System (INIS)

    Wilkie, D.

    1978-01-01

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

  15. Development of enclosure technique of tag gas for in-pile creep test

    International Nuclear Information System (INIS)

    Izaki, Toru; Ichikawa, Shoichi; Soroi, Masatoshi; Ito, Chikara

    2004-01-01

    Outline of the enclosure technique of tag gas for in-pile creep test is stated. In order to carry out in-pile creep test, the sample can enclose tag gas before the test and then the sample is inserted into MARICO-2 (Material Testing Rig with Temperature Control) in FBR 'JOYO' MK-III for the irradiation test. Outline of in-pile creep test using tag gas, enclosure system of tag gas, detection of a part of broken sample and identification of sample are explained. 126-, 128-, 129-, 131-, 132-, and 134-Xe are used as tag gases. The samples are identified by RIMS (Laser Resonance Ionization Mass Spectroscopy) in ppt order. ODS ferritic steel will be tested by the method in the next step. (S.Y.)

  16. Acoustic Emission Signal Processing Technique to Characterize Reactor In-Pile Phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Vivek Agarwal; Magdy Samy Tawfik; James A Smith

    2014-07-01

    Existing and developing advanced sensor technologies and instrumentation will allow non-intrusive in-pile measurement of temperature, extension, and fission gases when coupled with advanced signal processing algorithms. The transmitted measured sensor signals from inside to the outside of containment structure are corrupted by noise and are attenuated, thereby reducing the signal strength and signal-to-noise ratio. Identification and extraction of actual signal (representative of an in-pile phenomenon) is a challenging and complicated process. In this paper, empirical mode decomposition technique is proposed to reconstruct actual sensor signal by partially combining intrinsic mode functions. Reconstructed signal corresponds to phenomena and/or failure modes occurring inside the reactor. In addition, it allows accurate non-intrusive monitoring and trending of in-pile phenomena.

  17. Comparative evaluation of corrosion behaviour of type K thin film thermocouple and its bulk counterpart

    International Nuclear Information System (INIS)

    Mukherjee, S.K.; Barhai, P.K.; Srikanth, S.

    2011-01-01

    Highlights: → Anodic vacuum arc deposited chromel and alumel films are more 'noble' in 5% NaCl solution than their respective wires. → Chromel undergoes localised corrosion while alumel shows uniform corrosion. → Virgin samples of chromel-alumel TFTCs exhibit good thermoelectric response. → Their thermoelectric outputs remain largely unaffected when shelved under normal atmospheric conditions. → After 288 h of exposure in salt spray environment, their thermoelectric outputs show noticeable change due to size effects. - Abstract: This paper investigates the corrosion behaviour of type K thermoelements and their thin films, and compares the performance of chromel-alumel thin film thermocouple with its wire counterpart before and after exposure to 5% NaCl medium. Potentiodynamic polarisation tests reveal that chromel and alumel films are more 'noble' than their respective wires. Alumel corrodes faster when coupled with chromel in films than as wires. Secondary electron micrographs and electrochemical impedance spectroscopy measurements suggest that chromel shows localised corrosion while alumel undergoes uniform corrosion. Corrosion adversely affects the thermocouple output and introduces an uncertainty in the measurement.

  18. Measurement errors for thermocouples attached to thin plates

    International Nuclear Information System (INIS)

    Sobolik, K.B.; Keltner, N.R.; Beck, J.V.

    1989-01-01

    This paper discusses Unsteady Surface Element (USE) methods which are applied to a model of a thermocouple wire attached to a thin disk. Green's functions are used to develop the integral equations for the wire and the disk. The model can be used to evaluate transient and steady state responses for many types of heat flux measurement devices including thin skin calorimeters and circular foil (Gardon) head flux gauges. The model can accommodate either surface or volumetric heating of the disk. The boundary condition at the outer radius of the disk can be either insulated or constant temperature. Effect on the errors of geometrical and thermal factors can be assessed. Examples are given

  19. Temperature Control System for Chromel-Alumel Thermocouple

    International Nuclear Information System (INIS)

    Piping Supriatna; Nurhanan; Riswan DJ; Heru K, B.; Edi Karyanta

    2003-01-01

    Nuclear Power Plan Operation Safety needs serious handling on temperature measurement and control. In this report has been done manufacturing Temperature Control System for Chromel-Alumel Thermocouple, accordance to material, equipment and human resource ability in the laboratory. Basic component for the Temperature Control System is LM-741 type of Operation Amplifier, which is functionalized as summer for voltage comparator. Function test for this Control System shown its ability for damping on temperature reference. The Temperature Control System will be implemented on PCB Processing Machine. (author)

  20. Progress In Developing An In-Pile Acoustically Telemetered Sensor Infrastructure

    Energy Technology Data Exchange (ETDEWEB)

    Smith, James A.; Garrett, Steven L.; Heibel, Michael D.; Agarwal, Vivek; Heidrich, Brenden J.

    2016-09-01

    A salient grand challenge for a number of Department of Energy programs such as Fuels Cycle Research and Development ( includes Accident Tolerant Fuel research and the Transient Reactor Test Facility Restart experiments), Light Water Sustainability, and Advanced Reactor Technologies is to enhance our fundamental understanding of fuel and materials behavior under irradiation. Robust and accurate in-pile measurements will be instrumental to develop and validate a computationally predictive multi-scale understanding of nuclear fuel and materials. This sensing technology will enable the linking of fundamental micro-structural evolution mechanisms to the macroscopic degradation of fuels and materials. The in situ sensors and measurement systems will monitor local environmental parameters as well as characterize microstructure evolution during irradiation. One of the major road blocks in developing practical robust, and cost effective in-pile sensor systems, are instrument leads. If a wireless telemetry infrastructure can be developed for in-pile use, in-core measurements would become more attractive and effective. Thus to be successful in accomplishing effective in-pile sensing and microstructure characterization an interdisciplinary measurement infrastructure needs to be developed in parallel with key sensing technology. For the discussion in this research, infrastructure is defined as systems, technology, techniques, and algorithms that may be necessary in the delivery of beneficial and robust data from in-pile devices. The architecture of a system’s infrastructure determines how well it operates and how flexible it is to meet future requirements. The limiting path for the effective deployment of the salient sensing technology will not be the sensors themselves but the infrastructure that is necessary to communicate data from in-pile to the outside world in a non-intrusive and reliable manner. This article gives a high level overview of a promising telemetry

  1. Nuclear Power Plant Thermocouple Sensor-Fault Detection and Classification Using Deep Learning and Generalized Likelihood Ratio Test

    Science.gov (United States)

    Mandal, Shyamapada; Santhi, B.; Sridhar, S.; Vinolia, K.; Swaminathan, P.

    2017-06-01

    In this paper, an online fault detection and classification method is proposed for thermocouples used in nuclear power plants. In the proposed method, the fault data are detected by the classification method, which classifies the fault data from the normal data. Deep belief network (DBN), a technique for deep learning, is applied to classify the fault data. The DBN has a multilayer feature extraction scheme, which is highly sensitive to a small variation of data. Since the classification method is unable to detect the faulty sensor; therefore, a technique is proposed to identify the faulty sensor from the fault data. Finally, the composite statistical hypothesis test, namely generalized likelihood ratio test, is applied to compute the fault pattern of the faulty sensor signal based on the magnitude of the fault. The performance of the proposed method is validated by field data obtained from thermocouple sensors of the fast breeder test reactor.

  2. Summary on out-of-pile and in-pile properties of M5 alloy

    International Nuclear Information System (INIS)

    Zhao Wenjin

    2001-01-01

    The out-of-pile and in-pile corrosion, mechanical properties, microstructure,hydrogen absorption, creep and growth resistances of M5 alloy using as PWR fuel rod cladding materials developed by FRAMATOME in France has been summarized with reference to the literatures. The results obtained from in-pile irradiation tests show that the corrosion and hydrogen absorption resistances, creep and irradiation growth resistances of M5 alloy cladding are superior to that of the optimized Zircaloy-4. It could be estimated that the M5 alloy enables rod burnups close to 65GWd/tU to be reached

  3. In-Pile Loop Safety in Integrated with the Multipurpose Reactor in the case of in-Pile Loop Leakage at the Core Position

    International Nuclear Information System (INIS)

    Suharno; Sugianto; Giarno; Aliq; Widodo, Surip; Aji, Bintoro; Purba, Julwan Hendry; Karyanta, Edy

    1999-01-01

    In-Pile Loop Safety analysis in integrated with the multipurpose reactor in the case of In-Pile Loop leakage at the core position has been conducted which intended to evaluate the failure of fuel element. By considering design of In-Pile Loop and the highest possibility position of of leakage, the failure of fuel element is emphasized on mechanical aspect. The thermal hydraulic aspect is not taken into account due to the condition that when the leakage occurred the reactor has been in shut down condition. It is determined that the spray attacks the top position of fuel element, and to be calculated the force, of spray that produces 1 cm deflection on the single fuel element. Using that four (4) fuel elements is calculated because in the real condition 4 fuel elements will undergo deflection of 43.8 kg is obtained that producing 1 cm deflection and the force of 1228 kg that causes failure on the bottom of fuel element as shear force is also obtained. Whatever the force, high or low, the damage of fuel element occurred at the bottom part or at the position of grid plate. Therefore there is no damage on the fuel part (uranium meat) and the releasing of radioactive material from fuel plate is not happened

  4. In-pile creep behaviour of Zry-4 and ZrNb3Sn1 cladding under uniaxial and biaxial stress

    International Nuclear Information System (INIS)

    Boehner, G.; Wildhagen, B.; Wilhelm, H.

    1987-01-01

    An irradiation programme - started in 1977 - was performed at the research reactor FRG-2 at Geesthacht, Germany, as a joint project of GKSS and KWU in order to study the in-pile creep behaviour of zirconium alloy cladding tubes of PWR fuel rods. The test objective was to establish a data base which allows refined modelling of the in-pile creep phenomenon. A wide test matrix was realized in which each of the precisely monitored test conditions (hoop stress, temperature, fast neutron flux) was varied separately. Different cladding materials (Zircaloy-4 and Zirconium-Niob-Tin alloy ZrNb3Sn1) were subjected to those varying test conditions. Cladding tube specimens of 10.75 mm outer diameter were irradiated in test capsules under various stress conditions and levels up to approx. 6000 h, at temperatures ranging from 300 0 C to 400 0 C and fast neutron flux (E > 1 MeV) of approx. 3x10 13 cm -2 .s -1 . Diametrical and/or axial creep deformation of all tubes were measured in the Hot Cells several times in the course of the tests. In order to extract the irradiation induced creep strain some out-pile experiments were carried out under the very same test conditions as the in-pile tests concerned. (orig./GL)

  5. Viability of Pushrod Dilatometry Techniques for High Temperature In-Pile Measurements

    Energy Technology Data Exchange (ETDEWEB)

    J. E. Daw; J. L. Rempe; D. L. Knudson; K. G. Condie; J. C. Crepeau

    2008-03-01

    To evaluate the performance of new fuel, cladding, and structural materials for use in advanced and existing nuclear reactors, robust instrumentation is needed. Changes in material deformation are typically evaluated out-of-pile, where properties of materials are measured after samples were irradiated for a specified length of time. To address this problem, a series of tests were performed to examine the viability of using pushrod dilatometer techniques for in-pile instrumentation to measure deformation. The tests were performed in three phases. First, familiarity was gained in the use and accuracy of this system by testing samples with well defined thermal elongation characteristics. Second, high temperature data for steels, specifically SA533 Grade B, Class 1 (SA533B1) Low Alloy Steel and Stainless Steel 304 (SS304), found in Light Water Reactor (LWR) vessels, were aquired. Finally, data were obtained from a short pushrod in a horizontal geometry to data obtained from a longer pushrod in a vertical geometry, the configuration likely to be used for in-situ measurements. Results of testing show that previously accepted data for the structural steels tested, SA533B1 and SS304, are inaccurate at high temperatures (above 500 oC) due to extrpolation of high temperature data. This is especially true for SA533B1, as previous data do not account for the phase transformation of the material between 730 oC and 830 oC. Also, comparison of results for horizontal and vertical configurations show a maximum percent difference of 2.02% for high temperature data.

  6. In-Pile Tests for IASCC Growth Behavior of Irradiated 316L Stainless Steel under Simulated BWR Condition in JMTR

    Science.gov (United States)

    Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko; Nakano, Junichi; Nishiyama, Yutaka

    The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation test plan to evaluate in-situ effects of neutron/γ-ray irradiation on stress corrosion crack (SCC) growth of irradiated stainless steels using the Japan Materials Testing Reactor (JMTR). SCC growth rate and its dependence on electrochemical corrosion potential (ECP) are different between in-pile test and post irradiation examination (PIE). These differences are not fully understood because of a lack of in-pile data. This paper presents a systematic review on SCC growth data of irradiated stainless steels, an in-pile test plan for crack growth of irradiated SUS316L stainless steel under simulated BWR conditions in the JMTR, and the development of the in-pile test techniques.

  7. Heat transfer in the in-pile test section and penetration region of 3-pin fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Dae Young; Lee, Chung Young; Sim, Bong Shick; Park, Kook Nam; Park, Su Ki; Lee, Jong Min; Kim, Young Jin

    2003-12-01

    This report studies two types of normal heat transfer. One is the heat loss from the pressure vessel of In-Pile Test Section to HANARO pool water via IPS insulation gas gap. The other is the heat transfer of the Penetration Cooling Water System including the effect of the Foamglas insulator at the penetration region. The heat transfer from IPS insulation gas gap has been performed according to the detail design results from NUKEM. The heat loss also occurs at the concrete penetration region between the HANARO pool water and the FTL pipe gallery. The Foamglas insulator has been already installed at the MCW piping of the penetration region. This insulation effect has been reviewed. The Penetration Cooling Water System has been designed to fulfill the design requirement not to exceed the allowable temperature at the penetration concrete wall. The cooling ability and heat loss of PCW system has been reviewed with the insulation effect.

  8. New trends in pile safety instrumentation; Les tendances nouvelles dans l'instrumentation de securite des piles

    Energy Technology Data Exchange (ETDEWEB)

    Furet, J.

    1961-04-19

    This report addresses the protection of nuclear piles against damages due to operation incidents. The author discusses the current trends in the philosophy of safety of atomic power piles, identifies the parameters which define safety systems, presents tests to be performed on safety chains, comments the relationship between safety and the decrease of the number of pile inadvertent shutdowns, discusses the issues of instrument failures and chain multiplicity, comments the possible improvement of the operation of elements which build up safety chains (design simplification, development of semiconductors, replacement of electromechanical relays by static relays), the role of safety logical computers and the development of automatics in pile safety, presents automatic control as a safety factor (example of automatic start-up), and finally comments the use of fuses.

  9. Minutes of the workshop on bases of in-pile irradiation tests

    International Nuclear Information System (INIS)

    1997-03-01

    The Workshop on Bases of In-pile Irradiation Tests was held on January 29th and 30th, 1997 at the Ibarakiken Sangyo Kaikan in Mito, Ibaraki. The purpose is to discuss upgrading an in-pile irradiation test, promoting the utilization of the research and testing reactors and also activating the research potential of JAERI transversely. Main topics are the role and future plan of the research and testing reactors, a challenge to an advanced irradiation test, development of peripheral techniques for irradiation tests and future trends of the in-pile irradiation test in the 21st century. It was mainly pointed out that the in-pile irradiation test based on an analytical method using interpolation and extrapolation procedures met a turning point and that the upgrading of the irradiation and testing method should be indispensable for regaining the latest frontiers of an irradiation study using the research and testing reactors. The new concepts were also proposed on the irradiation correlation and modeling for the design of innovative materials. It was also recognized the key issues of the irradiation study in future should be an advanced irradiation testing method which can combine various types of irradiation field and control the irradiation conditions freely. In the next century in which large accelerator or new neutron source competes with research and testing reactors for neutron irradiation tests, themes of research using in-pile irradiation tests will be upgrading of the light water reactor, development of fusion reactor, basic research, biological and medical research, radioisotope production and semiconductors manufacturing, etc. It was also concluded the research and testing reactors will keep their main role in neutron irradiation research in future. This report briefly summarizes the content of 16 presentations and the discussion. The result of the questionnaires on the utilization of research and testing reactors to the participants is also attached. (J.P.N.)

  10. In-Pile Qualification of the Fast-Neutron-Detection-System

    Science.gov (United States)

    Fourmentel, D.; Villard, J.-F.; Destouches, C.; Geslot, B.; Vermeeren, L.; Schyns, M.

    2018-01-01

    In order to improve measurement techniques for neutron flux assessment, a unique system for online measurement of fast neutron flux has been developed and recently qualified in-pile by the French Alternative Energies and Atomic Energy Commission (CEA) in cooperation with the Belgian Nuclear Research Centre (SCK•ECEN). The Fast-Neutron-Detection-System (FNDS) has been designed to monitor accurately high-energy neutrons flux (E > 1 MeV) in typical Material Testing Reactor conditions, where overall neutron flux level can be as high as 1015 n.cm-2.s-1 and is generally dominated by thermal neutrons. Moreover, the neutron flux is coupled with a high gamma flux of typically a few 1015 γ.cm-2.s-1, which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors, including a miniature fission chamber with a special fissile material presenting an energy threshold near 1 MeV, which can be 242Pu for MTR conditions. Fission chambers are operated in Campbelling mode for an efficient gamma rejection. FNDS also includes a specific software that processes measurements to compensate online the fissile material depletion and to adjust the sensitivity of the detectors, in order to produce a precise evaluation of both thermal and fast neutron flux even after long term irradiation. FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCK•CEN in Belgium. Then a second test was recently completed at ISIS reactor operated by CEA in France. FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%.

  11. 78 FR 56174 - In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core

    Science.gov (United States)

    2013-09-12

    ... 52 [Docket No. PRM-50-105; NRC-2012-0056] In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core AGENCY: Nuclear Regulatory Commission. ACTION: Petition for rulemaking; denial...-core thermocouples at different elevations and radial positions throughout the reactor core to enable...

  12. 77 FR 30435 - In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core

    Science.gov (United States)

    2012-05-23

    ... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [Docket No. PRM-50-105; NRC-2012-0056] In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core AGENCY: Nuclear Regulatory Commission... of operating licenses for nuclear power plants (``NPP'') to operate NPPs with in-core thermocouples...

  13. Structural evaluation of thermocouple probes for 241-AZ-101 waste tank

    International Nuclear Information System (INIS)

    Kanjilal, S.K.

    1994-01-01

    This document reports on the structural analysis of the thermocouple probe to be installed in 241-AZ-101 waste tank. The thermocouple probe is analyzed for normal pump mixing operation and potential earthquake induced loads required by the Hanford Site Design Criteria SDC-4.1

  14. Structural evaluation of thermocouple probes for 241-AZ-101 waste tank

    Energy Technology Data Exchange (ETDEWEB)

    Kanjilal, S.K.

    1994-12-06

    This document reports on the structural analysis of the thermocouple probe to be installed in 241-AZ-101 waste tank. The thermocouple probe is analyzed for normal pump mixing operation and potential earthquake induced loads required by the Hanford Site Design Criteria SDC-4.1.

  15. Detection of thermocouple malfunction in the Beacon system

    International Nuclear Information System (INIS)

    Morita, T.; Heibel, M.D.; Congedo, T.V.

    1992-01-01

    The BEACON system uses Core Exit Thermocouples (T/C) extensively for continuous radial power distribution monitoring. The T/C's are used to adjust the reference power distribution generated by the BEACON system to match the current radial power distribution. T/C reliability, repeatability, and relative accuracy have been very satisfactory. However, it is very important to detect any T/C malfunctions during operation, since a T/C signal change caused by an undetected malfunction can lead to serious errors in the radial power distribution developed by BEACON. A simple procedure has been developed which is capable of discriminating between changes in T/C signals caused by actual changes in reactor conditions and signal changes caused by T/C malfunctions

  16. Design and research of seal structure for thermocouple column assembly

    International Nuclear Information System (INIS)

    Rao Qiqi; Li Na; Zhao Wei; Ma Zhigang

    2015-01-01

    The new seal structure was designed to satisfy the function of thermocouple column assembly and the reactor structure. This seal structure uses the packing graphite ring and adopts the self-sealing principle. Cone angle is brought to the seal face of seal structure which is conveniently to assembly and disassembly. After the sealing principle analysis and stress calculation of graphite ring which adopt the cone angle, the cone angle increases the radial force of seal structure and improves the seal effect. The stress analysis result shows the seal structure strength satisfies the regulation requirement. The cold and hot function test results shows the sealing effect is good, and the design requirement is satisfied. (authors)

  17. Tile Surface Thermocouple Measurement Challenges from the Orbiter Boundary Layer Transition Flight Experiment

    Science.gov (United States)

    Campbell, Charles H.; Berger, Karen; Anderson, Brian

    2012-01-01

    Hypersonic entry flight testing motivated by efforts seeking to characterize boundary layer transition on the Space Shuttle Orbiters have identified challenges in our ability to acquire high quality quantitative surface temperature measurements versus time. Five missions near the end of the Space Shuttle Program implemented a tile surface protuberance as a boundary layer trip together with tile surface thermocouples to capture temperature measurements during entry. Similar engineering implementations of these measurements on Discovery and Endeavor demonstrated unexpected measurement voltage response during the high heating portion of the entry trajectory. An assessment has been performed to characterize possible causes of the issues experienced during STS-119, STS-128, STS-131, STS-133 and STS-134 as well as similar issues encountered during other orbiter entries.

  18. Lifetime improvement of sheathed thermocouples for use in high-temperature and thermal transient operations

    International Nuclear Information System (INIS)

    McCulloch, R.W.; Clift, J.H.

    1982-01-01

    Premature failure of small-diameter, magnesium-oxide-insulated sheathed thermocouples occurred when they were placed within nuclear fuel rod simulators (FRSs) to measure high temperatures and to follow severe thermal transients encountered during simulation of nuclear reactor accidents in Oak Ridge National Laboratory (ORNL) thermal-hydraulic test facilities. Investigation of thermally cycled thermocouples yielded three criteria for improvement of thermocouple lifetime: (1) reduction of oxygen impurities prior to and during their fabrication, (2) refinement of thermoelement grain size during their fabrication, and (3) elimination of prestrain prior to use above their recrystallization temperature. The first and third criteria were satisfied by improved techniques of thermocouple assembly and by a recovery anneal prior to thermocouple use

  19. Blind system identification of two-thermocouple sensor based on cross-relation method

    Science.gov (United States)

    Li, Yanfeng; Zhang, Zhijie; Hao, Xiaojian

    2018-03-01

    In dynamic temperature measurement, the dynamic characteristics of the sensor affect the accuracy of the measurement results. Thermocouples are widely used for temperature measurement in harsh conditions due to their low cost, robustness, and reliability, but because of the presence of the thermal inertia, there is a dynamic error in the dynamic temperature measurement. In order to eliminate the dynamic error, two-thermocouple sensor was used to measure dynamic gas temperature in constant velocity flow environments in this paper. Blind system identification of two-thermocouple sensor based on a cross-relation method was carried out. Particle swarm optimization algorithm was used to estimate time constants of two thermocouples and compared with the grid based search method. The method was validated on the experimental equipment built by using high temperature furnace, and the input dynamic temperature was reconstructed by using the output data of the thermocouple with small time constant.

  20. The transient response for different types of erodable surface thermocouples using finite element analysis

    Directory of Open Access Journals (Sweden)

    Mohammed Hussein

    2007-01-01

    Full Text Available The transient response of erodable surface thermocouples has been numerically assessed by using a two dimensional finite element analysis. Four types of base metal erodable surface thermocouples have been examined in this study, included type-K (alumel-chromel, type-E (chromel-constantan, type-T (copper-constantan, and type-J (iron-constantan with 50 mm thick- ness for each. The practical importance of these types of thermocouples is to be used in internal combustion engine studies and aerodynamics experiments. The step heat flux was applied at the surface of the thermocouple model. The heat flux from the measurements of the surface temperature can be commonly identified by assuming that the heat transfer within these devices is one-dimensional. The surface temperature histories at different positions along the thermocouple are presented. The normalized surface temperature histories at the center of the thermocouple for different types at different response time are also depicted. The thermocouple response to different heat flux variations were considered by using a square heat flux with 2 ms width, a sinusoidal surface heat flux variation width 10 ms period and repeated heat flux variation with 2 ms width. The present results demonstrate that the two dimensional transient heat conduction effects have a significant influence on the surface temperature history measurements made with these devices. It was observed that the surface temperature history and the transient response for thermocouple type-E are higher than that for other types due to the thermal properties of this thermocouple. It was concluded that the thermal properties of the surrounding material do have an impact, but the properties of the thermocouple and the insulation materials also make an important contribution to the net response.

  1. Study of Thermocurrents in ILC cavities via measurements of the Seebeck Effect in niobium, titanium, and stainless steel thermocouples

    Energy Technology Data Exchange (ETDEWEB)

    Cooley, Victoria [Univ. of Wisconsin, Madison, WI (United States)

    2014-01-01

    The goals of Fermilab’s Superconductivity and Radio Frequency Development Department are to engineer, fabricate, and improve superconducting radio frequency (SCRF) cavities in the interest of advancing accelerator technology. Improvement includes exploring possible limitations on cavity performance and mitigating such impediments. This report focuses on investigating and measuring the Seebeck Effect observed in cavity constituents titanium, niobium, and stainless steel arranged in thermocouples. These junctions exist between cavities, helium jackets, and bellows, and their connection can produce a loop of electrical current and magnetic flux spontaneously during cooling. The experimental procedure and results are described and analyzed. Implications relating the results to cavity performance are discussed.

  2. Post-Irradiation Examination and In-Pile Measurement Techniques for Water Reactor Fuels

    International Nuclear Information System (INIS)

    2009-12-01

    in the 1960s when the construction of NPPs was being started. Evidently it can be assumed that infrastructure with basic unique equipments is old enough, both morally and physically, and needs to be up-graded or replaced. Thus, a sharp increase of the hydrocarbon fuel cost, green-house effect, necessity to construct more safe and efficient NPPs, justification of the lifetime prolongation of the existing NPPs, moral and physical ageing of the hot labs and research reactors equipment lead to the strong necessity to develop more perfect and more precise methods and equipment to examine irradiated components of nuclear reactors, first of all the most expensive one - nuclear fuel. Now the national hot laboratories and material testing reactors usually act as individual independent research establishments without any common and coordinated technical and business strategy towards the future needs and challenges. Even if there are not many joint programs for the development of nuclear power engineering in different countries, the method base and accumulated experience of the in- and post-reactor experiments should be widely shared so as to decrease the cost of this base in each country and to enforce its development. Thus, both problems and results of the application of new techniques to examine nuclear reactor components, as well as the conditions of separate labs should be discussed at the international level. The IAEA technical meetings are one of the most convenient means of arranging such discussion on the problems of the hot labs and research reactors development and application of new original techniques for examination of reactor materials properties. This publication represents a summary and proceedings of the two technical meetings (TMs) organized by IAEA on the subjects of Hot Cell Post-Irradiation Examination (PIE) Techniques and Pool Side Inspection of Water Reactor Fuel Assemblies and Fuel Rod Instrumentation and In-Pile Measurement Techniques. The first TM was

  3. Temperature measurement error due to the effects of time varying magnetic fields on thermocouples with ferromagnetic thermoelements

    International Nuclear Information System (INIS)

    McDonald, D.W.

    1977-01-01

    Thermocouples with ferromagnetic thermoelements (iron, Alumel, Nisil) are used extensively in industry. We have observed the generation of voltage spikes within ferromagnetic wires when the wires are placed in an alternating magnetic field. This effect has implications for thermocouple thermometry, where it was first observed. For example, the voltage generated by this phenomenon will contaminate the thermocouple thermal emf, resulting in temperature measurement error

  4. Thin film platinum–palladium thermocouples for gas turbine engine applications

    Energy Technology Data Exchange (ETDEWEB)

    Tougas, Ian M.; Gregory, Otto J., E-mail: gregory@egr.uri.edu

    2013-07-31

    Thin film platinum:palladium thermocouples were fabricated on alumina and mullite surfaces using radio frequency sputtering and characterized after high temperature exposure to oxidizing environments. The thermoelectric output, hysteresis, and drift of these sensors were measured at temperatures up to 1100 °C. Auger electron spectroscopy was used to follow the extent of oxidation in each thermocouple leg and interdiffusion at the metallurgical junction. Minimal oxidation of the platinum and palladium thermoelements was observed after high temperature exposure, but considerable dewetting and faceting of the films were observed in scanning electron microscopy. An Arrhenius temperature dependence on the drift rate was observed and later attributed to microstructural changes during thermal cycling. The thin film thermocouples, however, did exhibit excellent stability at 1000 °C with drift rates comparable to commercial type-K wire thermocouples. Based on these results, platinum:palladium thin film thermocouples have considerable potential for use in the hot sections of gas turbine engines. - Highlights: • Stable thin film platinum:palladium thermocouples for gas turbine engines • Little oxidation but significant microstructural changes from thermal cycling • Minimal hysteresis during repeated thermal cycling • Drift comparable to commercial wire thermocouples.

  5. Thermal and in-pile densification of MOX fuels: Some recent results

    International Nuclear Information System (INIS)

    Caillot, L.; Malgouyres, P.P.; Souchon, F.; Gotta, M.J.; Warin, D.; Chotard, A.; Couty, J.C.

    1997-01-01

    In-pile densification of PWR fuels is one of the main phenomena which determine the evolution of the pellet-clad gap during the first stage of the irradiation, and thus has consequences onto the thermo-mechanical behaviours of fuel rods. It can be predicted using the results of resintering tests and appropriate correlations. In this context, CEA, FRAMATOME and EDF have undertaken a joint research programme aiming to characterize the densification of MOX fuels. Different fuels were prepared by the MIMAS process using different UO 2 powders as matrix. After a detailed characterization, fuel pellets were submitted to isothermal resintering tests and analytical irradiations. Correlations between in-pile and thermal densification were established. This paper presents the results obtained with two types of MOX fuel: one fabricated wit the AUC UO 2 powder (ammonium uranyl carbonate conversion process) and another one fabricated with the SFEROX powder (peroxide conversion process). 8 refs, 8 figs

  6. In-pile cladding tests at NRI Rez and PIE capabilities and experience

    International Nuclear Information System (INIS)

    Zmitko, M.

    2002-01-01

    In-pile cladding corrosion test facilities and relevant post-irradiation capabilities at NRI Rez plc are overviewed. Basic information about the research rector LVR-15 and in-pile water loops is given. An experience in the field of Zr-alloy cladding corrosion testing and investigation of cladding corrosion behaviour is demonstrated for two experimental programmes conducted at NRI Rez in the past period. The first example describes results obtained at studying of corrosion behaviour of advanced Zr-alloys under PWR conditions with a special concern to a high lithium content and subcooled surface boiling. The second example informs about completion of the experimental programme supported by the IAEA which is focused on investigation of Zircaloy-4 cladding behaviour under VVER water chemistry, thermal-hydraulic and irradiation conditions with the main to obtain experimental data for an assessment of the Zircaloy-4 cladding compatibility with VVER conditions. (author)

  7. In-pile vapor pressure measurements on UO2 and (U,Pu)O2

    International Nuclear Information System (INIS)

    Breitung, W.; Reil, K.O.

    1985-08-01

    The Effective-Equation-of-State (EEOS) experiments investigated the saturation vapor pressures of ultra pure UO 2 , reactor grade UO 2 , and reactor grade (Usub(.77)Pusub(.23))O2 using newly developed in-pile heating techniques. For enthalpies between 2150 and 3700 kJ/kg (about 4700 to 8500 K) vapor pressures from 1.3 to 54 MPa were measured. The p-h curves of all three fuel types were identical within the experimental uncertainties. An assessment of all published p-h measurements showed that the p-h saturation curve of UO 2 appears now well established by the EEOS and the CEA in-pile data. Using an estimate for the heat capacity of liquid UO 2 , the in-pile results were also compared to earlier p-T measurements. The assessments lead to proposal of two equations. Equation I, which includes a factor-of-2 uncertainty band, covers all p-T equilibrium evaporation measurements. Equation I yields 3817 K for the normal boiling point, 415.4 kJ/mol for the corresponding heat of vaporization, and 1.90 MPa for the vapor pressure at 5000 K. Equations I and II, which represent a parametric form of the p-h curve (T=parameter), also give a good description of the EEOS and CEA in-pile data. Thus the proposed equations allow a consistent representation of both p-T and p-h measurements, they are sufficiently precise for CDA analyses and cover the whole range of interest (3120-8500 K, 1400-3700 kJ/kg). (orig./HP) [de

  8. Design of in-pile section monitoring system in fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Sim, B. S.; Park, K. N.; Park, S. K.; Chi, D. Y.; Lee, J. M.; Ahn, S. H.; Lee, C. Y.; Kim, Y. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    In this paper a brief summary of the monitoring system of In-Pile Section(IPS) is described. To meet the user requirements on the test fuel and irradiation conditions, various instruments are installed on the test fuel pin itself and the appropriate locations in IPS. The requirements and descriptions for instruments, gas supply system, and the data acquisition system to sample and record parameters are described.

  9. Vanadium—lithium in-pile loop for comprehensive tests of vanadium alloys and multipurpose coatings

    Science.gov (United States)

    Lyublinski, I. E.; Evtikhin, V. A.; Ivanov, V. B.; Kazakov, V. A.; Korjavin, V. M.; Markovchev, V. K.; Melder, R. R.; Revyakin, Y. L.; Shpolyanskiy, V. N.

    1996-10-01

    The reliable information on design and material properties of self-cooled Li sbnd Li blanket and liquid metal divertor under neutron radiation conditions can be obtained using the concept of combined technological and material in-pile tests in a vanadium—lithium loop. The method of in-pile loop tests includes studies of vanadium—base alloys resistance, weld resistance under mechanical stress, multipurpose coating formation processes and coatings' resistance under the following conditions: high temperature (600-700°C), lithium velocities up to 10 m/s, lithium with controlled concentration of impurities and technological additions, a neutron load of 0.4-0.5 MW/m 2 and level of irradiation doses up to 5 dpa. The design of such an in-pile loop is considered. The experimental data on corrosion and compatibility with lithium, mechanical properties and welding technology of the vanadium alloys, methods of coatings formation and its radiation tests in lithium environment in the BOR-60 reactor (fast neutron fluence up to 10 26 m -2, irradiation temperature range of 500-523°C) are presented and analyzed as a basis for such loop development.

  10. A probabilistic safety assessment of in-pile test loop in HWRR

    International Nuclear Information System (INIS)

    Cao Xuewu; Li Zhaohuan

    1991-07-01

    The PSA methodology has been applied to the in-pile test loop which is installed in the Heavy Water Research Reactor (HWRR). This loop is designed and operated for fuel assembly testing of the Qinshan PWR plant. This analysis is to assess the safety and to evaluate the design of this operating loop. The procedure and models are similar to a PSA on nuclear power plant. The major contents in the analysis consist of the familiarization of the object, the investigation and selection of accident initiators, setting events and fault trees, data collections, quantitative calculations, qualitative and result analyses and final conclusion. This analysis is only limited to the initiators of in-pile loop itself and possible errors made by operators during normal operation. The accident occurence is less than 10 -4 a -1 which may be recommended as an acceptance risk for safety operation of an in-pile test loop. Finally, suggestions have been raised to improve the design of test loop, especially in reducing operation errors by local operators

  11. New In-pile Instrumentation to Support Fuel Cycle Research and Development

    Energy Technology Data Exchange (ETDEWEB)

    J. Rempe; H. MacLean; R. Schley; D. Hurley; J. Daw; S. Taylor; J. Smith; J. Svoboda; D. Kotter; D. Knudson; M. Guers; S. C. Wilkins

    2011-01-01

    New and enhanced nuclear fuels are a key enabler for new and improved reactor technologies. For example, the goals of the next generation nuclear plant (NGNP) will not be met without irradiations successfully demonstrating the safety and reliability of new fuels. Likewise, fuel reliability has become paramount in ensuring the competitiveness of nuclear power plants. Recently, the Office of Nuclear Energy in the Department of Energy (DOE-NE) launched a new direction in fuel research and development that emphasizes an approach relying on first principle models to develop optimized fuel designs that offer significant improvements over current fuels. To facilitate this approach, high fidelity, real-time, data are essential for characterizing the performance of new fuels during irradiation testing. A three-year strategic research program is proposed for developing the required test vehicles with sensors of unprecedented accuracy and resolution for obtaining the data needed to characterize three-dimensional changes in fuel microstructure during irradiation testing. When implemented, this strategy will yield test capsule designs that are instrumented with new sensor technologies for the Advanced Test Reactor (ATR) and other irradiation locations for the Fuel Cycle Research and Development (FC R&D) program. Prior laboratory testing, and as needed, irradiation testing, of these sensors will have been completed to give sufficient confidence that the irradiation tests will yield the required data. Obtaining these sensors must draw upon the expertise of a wide-range of organizations not currently supporting nuclear fuels research. This document defines this strategic program and provides the necessary background information related to fuel irradiation testing, desired parameters for detection, and an overview of currently available in-pile instrumentation. In addition, candidate sensor technologies are identified in this document, and a list of proposed criteria for ranking

  12. In-Pile Section(IPS) Inner Assembly Manufacturing Report

    International Nuclear Information System (INIS)

    Lee, Jong Min; Shim, Bong Sik; Lee, Chung Yong

    2009-12-01

    The objective of this report is to present the manufacturing, assembling and testing process of IPS Inner Assembly used in Fuel Test Loop(FTL) pre-operation test. The majority of the manufactured components are test fuels, inner assembly structures and subsidiary tools that is needed during the assembly process. In addition, Mock-up test for the welding and brazing is included at this stage. Lower structure, such as test fuels, fuel carrier legs are assembled and following structures, such as fuel carrier stem in the middle structure, top flange in the top structure are assembled together each other. To Verify the Reactor Coolant Pressure Boundary(RCPB) function in IPS Inner Assembly helium leak test and hydraulic test is performed with its acceptance criteria. According to the ASME III code Authorized Nuclear Inspector(ANI) is required during the hydraulic test. As-built measurement and insulation resistance test are performed to the structures and instrumentations after the test process. All requirements are satisfied and the IPS Inner Assembly was loaded in HANARO IR-1 hole in September 25, 2009

  13. Studies of Behavior Melting Temperature Characteristics for Multi Thermocouple In-Core Instrument Assembly

    International Nuclear Information System (INIS)

    Shin, Donghyup; Chae, Myoungeun; Kim, Sungjin; Lee, Kyulim

    2015-01-01

    Bottom-up type in-core instruments (ICIs) are used for the pressurized water reactors of OPR-1000, APR- 1400 in order to measure neutron flux and temperature in the reactor. It is a well-known technique and a proven design using years in the nuclear field. ICI consists of one pair of K-type thermocouple, five self-powered neutron detectors (SPNDs) and one back ground detector. K-type thermocouple's purpose is to measure the core exit temperature (CET) in the reactor. The CET is a very important factor for operating nuclear power plants and it is 327 .deg. C when generally operating the reactor in the nuclear power plant(NPP) in case of OPR- 1000. If the CET will exceed 650 .deg. C, Operators in the main control room should be considered to be an accident situation in accordance with a severe accident management guidance(SAMG). The Multi Thermocouple ICI is a new designed ICI assuming severe accident conditions. It consists of four more thermocouples than the existing design, so it has five Ktype thermocouples besides the thermocouple measuring CET is located in the same elevation as the ICI. Each thermocouple is able to be located in the desired location as required. The Multi Thermocouple ICI helps to measure the temperature distribution of the entire reactor. In addition, it will measure certain point of melted core because of the in-vessel debris of nuclear fuel when an accident occurs more seriously. In this paper, to simulate a circumstance such as a nuclear reactor severe accident was examined. In this study, the K-type thermocouples of Multi Thermocouple ICI was confirmed experimentally to be able to measure up to 1370 .deg. C before the thermocouples have been melted. And after the thermocouples were melted by debris, it was able to be monitored that the signal of EMF directed the infinite value of voltage. Therefore through the results of the test, it can be assumed that if any EMF data among the Multi Thermocouple ICI will direct the infinite value

  14. Studies of Behavior Melting Temperature Characteristics for Multi Thermocouple In-Core Instrument Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Donghyup; Chae, Myoungeun; Kim, Sungjin; Lee, Kyulim [Woojin inc, Hwasung (Korea, Republic of)

    2015-05-15

    Bottom-up type in-core instruments (ICIs) are used for the pressurized water reactors of OPR-1000, APR- 1400 in order to measure neutron flux and temperature in the reactor. It is a well-known technique and a proven design using years in the nuclear field. ICI consists of one pair of K-type thermocouple, five self-powered neutron detectors (SPNDs) and one back ground detector. K-type thermocouple's purpose is to measure the core exit temperature (CET) in the reactor. The CET is a very important factor for operating nuclear power plants and it is 327 .deg. C when generally operating the reactor in the nuclear power plant(NPP) in case of OPR- 1000. If the CET will exceed 650 .deg. C, Operators in the main control room should be considered to be an accident situation in accordance with a severe accident management guidance(SAMG). The Multi Thermocouple ICI is a new designed ICI assuming severe accident conditions. It consists of four more thermocouples than the existing design, so it has five Ktype thermocouples besides the thermocouple measuring CET is located in the same elevation as the ICI. Each thermocouple is able to be located in the desired location as required. The Multi Thermocouple ICI helps to measure the temperature distribution of the entire reactor. In addition, it will measure certain point of melted core because of the in-vessel debris of nuclear fuel when an accident occurs more seriously. In this paper, to simulate a circumstance such as a nuclear reactor severe accident was examined. In this study, the K-type thermocouples of Multi Thermocouple ICI was confirmed experimentally to be able to measure up to 1370 .deg. C before the thermocouples have been melted. And after the thermocouples were melted by debris, it was able to be monitored that the signal of EMF directed the infinite value of voltage. Therefore through the results of the test, it can be assumed that if any EMF data among the Multi Thermocouple ICI will direct the infinite value

  15. Field installed brazed thermocouple feedthroughs for high vacuum experiments

    International Nuclear Information System (INIS)

    Anderson, P.; Messick, C.

    1983-01-01

    In order to reduce the occurrence of vacuum leaks and to increase the availability of the DIII vacuum vessel for experimental operation, effort was applied to developing a vacuum-tight brazed feedthrough system for sheathed thermocouples, stainless steel sheathed conductor cables and tubes for cooling fluids. This brazed technique is a replacement for elastomer ''O'' ring sealed feedthroughs that have proven vulnerable to leaks caused by thermal cycling, etc. To date, about 200 feedthroughs have been used. Up to 91 were grouped on a single conflat flange mounted in a bulkhead connector configuration which facilitates installation and removal. Investigation was required to select a suitable braze alloy, flux and installation procedure. Braze alloy selection was challenging since the alloy was required to have: 1) Melting temperature in excess of the 250 0 C (482 0 F) bakeout temperature. 2) No high vapor pressure elements. 3) Good wetting properties when used in air with acceptable flux. 4) Good wettability to 300 series stainless steel and inconel

  16. In-pile test of tritium recovery from lithium oxide

    International Nuclear Information System (INIS)

    Kurasawa, Toshimasa; Yoshida, Hiroshi; Watanabe, Hitoshi; Takeshita, Hidefumi; Miyauchi, Takejiro; Matsui, Tomoaki

    1984-05-01

    In-situ tritium recovery experiment with sintered lithium oxide pellets was performed under a high neutron fluence in the JRR-2. The irradiation hole VT-10 is the vertical one in the fuel rods region of the reactor, and the neutron flux is as follows: the thermal neutron flux with the epithermal neutron; 1.12 x 10 14 n/cm 2 . sec, the fast neutron flux; 1.0 x 10 12 n/cm 2 . sec. Irradiation material is the four pellets of cylindrical Li 2 O with the size of 11mm-OD, 1.8mm-ID, 10mm-H, and their total weight is 6.67g(the apparent bulk density 86%TD). A sweep gas capsule with a inner heater was constructed for the present study. Irradiation temperatures were regulated in the high temperature range, 470 -- 760 0 C. Four cycles of irradiation tests were carried out from May to August in 1983, and the effective thermal neutron fluence and the burnup of 6 Li were 5.9 x 10 19 nvt and 0.24% of total lithium(natural abundance of Li), respectively. The amount of generated tritium was calculated to be 31.2Ci by using a value of the depression factor of the thermal neutron flux(0.148) and the effective neutron cross section(543b) for the 6 Li(n, α) 3 H reaction. Present report describes the tritium release behavior in the in-situ tritium recovery apparatus and discuss the effects of the moisture, the hydrogen spiking, the irradiation temperature, etc.. Problems relative to a real time measurement of a comparatively high tritium concentration(10 -1 -- 10 2 μCi/cm 3 ) in the helium gas stream were also investigated. (author)

  17. Thermocouples calibration and analysis of the influence of the length of the sensor coating

    International Nuclear Information System (INIS)

    Noriega, M; Ramírez, R; López, R; Vaca, M; Morales, J; Terres, H; Lizardi, A; Chávez, S

    2015-01-01

    This paper presents the design and construction of a lab prototype, with a much lower cost compared to the ones commercially sold, enabling the manufacture of thermocouples which are then calibrated to verify their functionality and acceptance. We also analyze the influence of the external insulation over the wires, to determine whether it influences temperature measurement. The tested lengths ranged from 0.00 m up to 0.030 m. The thermocouple was compared against the behavior of a thermocouple of the same type that was purchased with a commercial supplier. The obtained measurement showed less than 1 °C difference in some points. This makes the built thermocouple reliable, since the standard allows a difference of up to 2.2 °C

  18. The use of thermocouples which transmute during service in nuclear reactors

    International Nuclear Information System (INIS)

    Martin, R.E.

    1980-06-01

    Some current nuclear fuel experiments at CRNL require the use of thermocouples to measure temperatures of up to 2200 0 C under reactor operating conditions. A literature search has shown that transient electrical effects and transmutation of the thermocouple alloys can cause temperature measurement errors of up to +-1% and +-30%, respectively. However, the error due to transient electrical effects can be corrected by making temperature measurements immediately following reactor shutdown. Furthermore it has been shown that transmutation effects can be corrected for by calibrating the high temperature tungsten-rhenium thermocouples against a chromel-alumel thermocouple in a cooler part of the experiment. The use of these techniques is expected to reduce temperature measurement errors to +-2% in the best case. (auth)

  19. Experimental measurement of the interfacial heat transfer coefficients of subcooled flow boiling using micro-thermocouple and double directional images

    International Nuclear Information System (INIS)

    Seong-Jin Kim; Goon-Cherl Park

    2005-01-01

    Full text of publication follows: Models or correlations for phase interface are needed to analyze the multi-phase flow. Interfacial heat transfer coefficients are important to constitute energy equation of multi-phase flow, specially. In subcooled boiling flow, bubble condensation at the bubble-liquid interface is a major mechanism of heat transfer within bulk subcooled liquid. Bubble collapse rates and temperatures of each phase are needed to determine the interfacial heat transfer coefficient for bubble condensation. Bubble collapse rates were calculated through image processing in single direction, generally. And in case of liquid bulk temperature, which has been obtained by general temperature sensor such as thermocouple, was used. However, multi-directional images are needed to analyze images due to limitations of single directional image processing. Also, temperature sensor, which has a fast response time, must be used to obtain more accurate interfacial heat transfer coefficient. Low pressure subcooled water flow experiments using micro-thermocouple and double directional image processing with mirrors were conducted to investigate bubble condensation phenomena and to modify interfacial heat transfer correlation. Experiments were performed in a vertical subcooled boiling flow of a rectangular channel. Bubble condensing traces with respect to time were recorded by high speed camera in double direction and bubble collapse rates were calculated by processing recorded digital images. Temperatures were measured by micro-thermocouple, which is a K-type with a 12.7 μm diameter. The liquid temperature was estimated by the developed algorithm to discriminate phases and find each phase temperature in the measured temperature including both liquid and bubble temperature. The interfacial heat transfer coefficient for bubble condensation was calculated from the bubble collapse rates and the estimated liquid temperature, and its correlation was modified. The modified

  20. Results of postirradiation examination of the in-pile blockage experiments MOL-7C/4 and MOL-7C/5

    International Nuclear Information System (INIS)

    Weimar, P.; Schleisiek, K.

    1991-01-01

    The Mol-7C in-pile local blockage experiments are performed in the BR-2 reactor at Mol, Belgium as a joint project of Kernforchungszentrum Karlsruhe (KfK) and Studiecentrum voor Kernenergie/Centre d'Etude de l'Energie Nuclearire-Mol. The main objective is to investigate the consequences of local cooling disturbances in liquid-metal-cooled reactor (LMR) fuel subassemblies. In the tests Mol-7C/4 and MOL-7C/5, fuel pins from KNK II are used with a burnup of 5 and 1.7%, respectively. An active central porous blockage is used to simulate the cooling disturbance. During irradiation, the blockage causes significant local damage, including melting of cladding and fuel. Extensive postirradiation examinations (PIE) are performed to investigate the extent of damage. In this paper a description and interpretation of results of the destructive PIE performed at the Hot Cells Laboratory at KfK is given, along with some conclusions related to LMR safety

  1. Attachment of Free Filament Thermocouples for Temperature Measurements on CMC

    Science.gov (United States)

    Lei, Jih-Fen; Cuy, Michael D.; Wnuk, Stephen P.

    1997-01-01

    Ceramic Matrix Composites (CMC) are being developed for use as enabling materials for advanced aeropropulsion engine and high speed civil transport applications. The characterization and testing of these advanced materials in hostile, high-temperature environments require accurate measurement of the material temperatures. Commonly used wire Thermo-Couples (TC) can not be attached to this ceramic based material via conventional spot-welding techniques. Attachment of wire TC's with commercially available ceramic cements fail to provide sufficient adhesion at high temperatures. While advanced thin film TC technology provides minimally intrusive surface temperature measurement and has good adhesion on the CMC, its fabrication requires sophisticated and expensive facilities and is very time consuming. In addition, the durability of lead wire attachments to both thin film TC's and the substrate materials requires further improvement. This paper presents a newly developed attachment technique for installation of free filament wire TC's with a unique convoluted design on ceramic based materials such as CMC's. Three CMC's (SiC/SiC CMC and alumina/alumina CMC) instrumented with type IC, R or S wire TC's were tested in a Mach 0.3 burner rig. The CMC temperatures measured from these wire TC's were compared to that from the facility pyrometer and thin film TC's. There was no sign of TC delamination even after several hours exposure to 1200 C. The test results proved that this new technique can successfully attach wire TC's on CMC's and provide temperature data in hostile environments. The sensor fabrication process is less expensive and requires very little time compared to that of the thin film TC's. The same installation technique/process can also be applied to attach lead wires for thin film sensor systems.

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

    International Nuclear Information System (INIS)

    Pevchikh, Yu.M.; Kruglov, A.S.; Troyanov, V.M.

    2002-01-01

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

  3. Data collection for the methods of in-pile materials testing

    International Nuclear Information System (INIS)

    Markina, N.V.; Rudkevich, A.V.; Lebedeva, E.E.

    1985-01-01

    Problems, relating to the creation of automated data banks intended for accumulatioon and systematization of data according to the methods of in-pile materials testing, taking into account their specific nature are discussed. The architecture of the bank, realized on the BESM-6 computer is described. The bank contains data on about 300 methods. The library of requests, which may be used by a user comprises about 20 descriptions. A new request description preparation and input take 5-20 min. The system answering time is from 30 sec to 3-4 min

  4. In-pile intragranular densification of oxide fuels (AWBA Development Program)

    International Nuclear Information System (INIS)

    Dollins, C.C.; Nichols, F.A.

    1977-10-01

    This report proposes a model to describe in-pile densification of oxide fuels, by both vacancy boil-off due to thermal excitation and vacancy knockout by the passage of fission fragments through the pores. The model includes the migration rates of both vacancies and interstitials to pores and the production of vacancy-rich damage cascades by fission fragments. It has been coupled with a previously reported swelling and gas release model so that it can predict the total dimensional changes of the fuel as well as predicting intragranular densification for both ThO 2 and UO 2 fuels for advanced water breeder reactor applications development effort

  5. In-pile tests of HTGR fuel particles and fuel elements

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Kolesov, V.S.; Deryugin, A.I.

    1985-01-01

    Main types of in-pile tests for specimen tightness control at the initial step, research of fuel particle radiation stability and also study of fission product release from fuel elements during irradiation are described in this paper. Schemes and main characteristics of devices used for these tests are also given. Principal results of fission gas product release measurements satisfying HTGR demands are illustrated on the example of fuel elements, manufactured by powder metallurgy methods and having TRISO fuel particles on high temperature pyrocarbon and silicon carbide base. (author)

  6. Irradiation technology (1). Development of new in-pile instrumentation at JMTR

    International Nuclear Information System (INIS)

    Shibata, Akira; Kimura, Nobuaki; Tanimoto, Masataka; Nakamura, Jinichi; Saito, Takashi; Tsuchiya, Kunihiko

    2012-01-01

    Development of instrumentation which can use under severe accident condition is important issue for the purpose to cope with severe accident at nuclear reactors. And also to improve the quality of irradiation tests data and to increase the reliability of safety management system of reactors, the development of new instrumentation is key issue. JAEA is developing several in-pile instrumentations to conduct irradiation tests at JMTR. This study includes the developments of three new instrumentations and describes the characteristics of the instrumentations. These are ECP sensor, new water level indicator and in-reactor observation system using Cherenkov light. (author)

  7. In pile measurement of creep rate of stainless steel cladding tubes for fast reactor pins

    International Nuclear Information System (INIS)

    Calza Bini, A.; Cosoli, G.; Filacchioni, G.; Lanchi, M.; Nobili, A.; Pesce, E.; Rocca, U.V.; Rotoloni, P.L.

    1975-01-01

    Results are reported of a direct in pile measurement of creep on a cladding sample of 10cm length, under tensile stress of 22.82kg/mm 2 at a temperature of 550 0 during about 500 hours, up to an integrated flux of 2.6.10 20 n/cm 2 . Two identical samples were irradiated in the same temperature and flux conditions to be submitted to out of pile creep measurements together with other unirradiated samples. The aim of this first experiment was mainly to set up the device and to evaluate the kind and the quality of the available data

  8. The in-pile proving test for fuel assembly of Qinshan nuclear power plant

    International Nuclear Information System (INIS)

    Chen Dianshan; Zhang Shucheng; Kang Rixin; Wang Huarong; Chen Guanghan

    1989-10-01

    The in-pile proving test for fuel assembly of Qinshan nuclear power plant had been conducted in the experimental loop of HWRR at IAE (Institute of Atomic Energy) in Beijing, China, from January 1985 to December 1986. Average burnup of 27000 MWd/tU and peak burnup of 34000 MWd/tU of fuel rod had already been reached. The basic status of the experiment are described, emphasis is placed on the discussion of proving test parameters and analysis of experiment results

  9. Safety analysis report for packaging: the ORNL in-pile capsule shipping cask

    International Nuclear Information System (INIS)

    Evans, J.H.; Chipley, K.K.; Haynie, C.B.; Crowley, W.K.; Just, R.A.

    1977-11-01

    The ORNL in-pile capsule shipping cask is used to transport irradiated experimental capsules and spent fuel elements. The cask was analytically evaluated to determine its compliance with the applicable regulations governing containers in which radioactive materials are transported. Computational procedures were used to determine the structural integrity and thermal behavior of the cask relative to the general standards for normal conditions of transport and the standards for the hypothetical accident conditions. The results of the evaluation show that the cask is in compliance with the applicable regulations

  10. In-Pile Testing and Instrumentation for Development of Generation-IV Fuels and Materials. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2013-12-01

    For many years, the increase in efficiency in the production of nuclear electricity has been an economic challenge in many countries which have developed this kind of energy. The increase in fuel burnup and fuel residence time leads to a reduction in the volume of fresh fuel loaded and spent fuel discharged, respectively. More demanding nuclear fuel cycle parameters are combined with a need to operate nuclear power plants with maximal availability and load factors, in load-follow mode and with longer fuel cycles. In meeting these requirements, fuel has to operate in a demanding environment of high radiation fields, high temperatures, high mechanical stresses and high coolant flow. Requirements of increased fuel reliability and minimal fuel failures also remain in force. Under such circumstances, continuous development of more radiation resistant fuel materials, especially advanced cladding, careful and incremental examinations, and improved understanding and modelling of high burnup fuel behaviour are required. Following a recommendation of the IAEA Technical Working Group on Fuel Performance and Technology, the Technical Meeting on In-pile Testing and Instrumentation for Development of Generation-IV Fuels and Materials was held in Halden, Norway, on 21-24 August 2012. The purpose of the meeting was to review the current status and the progress in methods and technologies used for the in-pile testing of nuclear fuel achieved since the previous IAEA meeting on In-core Instrumentation and Reactor Core Assessment, also held in Halden in 2007. Emphasis was placed on advanced techniques applied for the understanding of high burnup fuel behaviour of water cooled power reactors that represent the vast majority of the current nuclear reactor fleet. However, the meeting also included papers and discussion on testing techniques applied or developed specifically for new fuel and structural materials considered for Generation-IV systems. The meeting was attended by 43

  11. Degradation by radiation of the response of a thermocouple of a fuel element

    International Nuclear Information System (INIS)

    Rodriguez V, A.

    1994-01-01

    In the TRIGA Mark III Reactor of the National Institute of Nuclear Research, is necessary to use an instrumented fuel element for measurement the fuel temperature during pulses of power. This fuel element is exposed to daily temperature gradient of order to 390 Centigrade degrees in normal condition of reactor operation at 1 MW. The experience which this instrumented fuel elements is that useful life of the thermocouples is less then the fuel, because they show important changes in their chemistry composition and electrical specifications, until the point they don't give any response. So is necessary to know the factors that influenced in the shortening of the thermocouples life. The change in composition affects the thermocouple calibration depends on where the changes take place relative to the temperature gradient. The change will be dependent on the neutron flux and so the value of the neutron flux may be used as a measure or the composition change. If there is no neutron flux within the temperature gradient, there will be no composition change, and so the thermocouple calibration will no change. If the neutron flux varies within the region in which a temperature gradients exists, the composition of the thermocouple will vary and the calibration will change. But the maximum change in calibration will occur if the neutron flux is high and constant within the region of the temperature gradient. In this case, a composition change takes place which is uniform throughout the gradient and so the emf output can be expected to change. In this reactor, the thermocouples are in the second case. Then, the relative position of the thermal and neutron flux gradients are the most important factor that explain the composition change after or 2,500 times of exposing the thermocouples to the temperature gradients of order to 390 Centigrade degrees. (Author)

  12. Parametric analysis of LIBRETTO-4 and 5 in-pile tritium transport model on EcosimPro

    Energy Technology Data Exchange (ETDEWEB)

    Alcalde, Pablo Martínez, E-mail: pablomiguel.martinez@externos.ciemat.es [Universidad Nacional de Educación a Distancia (UNED), c/Juan del Rosal 12, 28040 Madrid (Spain); Moreno, Carlos; Ibarra, Ángel [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain)

    2014-10-15

    Highlights: • Introduction of a new tritium transport model of LIBRETTO-4 and 5 on EcosimPro{sup ®}. • Analysis of model input parameter and variable sensitivities and effects on tritium simulated fluxes. • Demonstrations of high tritium out-flux dependencies on lead-lithium parameters. • Rough fitting achievements proposed by Li17Pb solubility or recombination increase. - Abstract: A new model for LIBRETTO-4/1, 4/2 and 5 experiments have been developed on ECOSIMPro{sup ©} tool to simulate tritium in-pile breeding and transport into two separate purge gas channels with He + 0.1%H{sub 2}. Release from lead lithium eutectic plenum with coupled permeation through an austenitic steel wall on the first and single permeation through EUROFER-97 in the temperature ranges of 300–550 °C can be simulated tuning the transport parameters involved. A parametric study has been performed to reduce the degrees of freedom and to determine the error caused in the simulation due to the uncertainty in experimental input data. The information obtained is essential for the experimental benchmarking. The Tritium Permeation Percentage (TPP) is an output calculated parameter with low variations between 2 and 6% along the whole experimental time easy to compare (730 Full Power Days for LIBRETTO-4 and 520 for 5). Tritium transport parameter ranges verifying this output are defined herein.

  13. Thermal-hydraulic analyses for in-pile SCWR fuel qualification test loops and SCWR material loop

    Energy Technology Data Exchange (ETDEWEB)

    Vojacek, A.; Mazzini, G.; Zmitkova, J.; Ruzickova, M. [Research Centre Rez (Czech Republic)

    2014-07-01

    One of the R&D directions of Research Centre Rez is dedicated to the supercritical water-cooled reactor concept (SCWR). Among the developed experimental facilities and infrastructure in the framework of the SUSEN project (SUStainable ENergy) is construction and experimental operation of the supercritical water loop SCWL focusing on material tests. At the first phase, this SCWL loop is assembled and operated out-of-pile in the dedicated loop facilities hall. At this out-of-pile operation various operational conditions are tested and verified. After that, in the second phase, the SCWL loop will be situated in-pile, in the core of the research reactor LVR-15, operated at CVR. Furthermore, it is planned to carry out a test of a small scale fuel assembly within the SuperCritical Water Reactor Fuel Qualification Test (SCWR-FQT) loop, which is now being designed. This paper presents the results of the thermal-hydraulic analyses of SCWL loop out-of-pile operation using the RELAP5/MOD3.3. The thermal-hydraulic modeling and the performed analyses are focused on the SCWL loop model validation through a comparison of the calculation results with the experimental results obtained at various operation conditions. Further, the present paper focuses on the transient analyses for start-up and shut-down of the FQT loop, particularly to explore the ability of system codes ATHLET 3.0A to simulate the transient between subcritical conditions and supercritical conditions. (author)

  14. In-pile experimental device for Sirene thermionic converters; Dispositif d'experimentation en pile des convertisseurs thermoioniques sirene

    Energy Technology Data Exchange (ETDEWEB)

    Bliaux, J; Durand, J; Lazare-Chopard, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1969-07-01

    The irradiation device described here, was built for in pile life tests of 100 We SIRENE converters. The nuclear converter is located in a sealed vacuum chamber, which is plugged at the lower end of a coaxial tubing acting as electrical leads. The output power is available on a variable resistive load on the bank of the reactor pool. Thermal, electrical and neutronic parameters of the converter are recorded. Since 1967, two permanent devices allowed five experiments in the swimming pool TRITON (CEN-FAR) and the results, obtained till now, are presented. (authors) [French] Le dispositif d'irradiation SIRENE decrit ici a ete concu en vue d'une etude statistique de performances de convertisseurs thermoioniques nucleaires de puissance unitaire 100 We. Le dispositif doit assurer la bonne marche du convertisseur en pile, permettre le changement de la position verticale du convertisseur dans le coeur, sortir du coeur la puissance electrique convertie sans degradation notable et enregistrer les differents parametres thermiques, electriques et neutroniques du convertisseur. Depuis 1967, deux dispositifs fonctionnent en permanence et ont permis de faire cinq experiences dans le reacteur piscine TRITON du CEN-FAR. Les resultats obtenus jusqu'a present, sont presentes. (auteurs)

  15. Research on in-pile release of fission products from coated particle fuels

    International Nuclear Information System (INIS)

    Fukuda, K.; Iwamoto, K.

    1985-01-01

    Coated particle fuels fabricated in accordance with VHTR (Very High Temperature gas-cooled Reactor) fuel design have been irradiated by both capsules and an in-pile gas loop (OGL-1), and data on the fission products release under irradiation were obtained for loose coated particles, fuel compacts and fuel rods in the temperature range between 800 deg. C and 1600 deg. C. For the fission gases, temperature- and time dependences of the fractional release(R/B) were measured. Relation between release and failure fraction of the coated particles was elucidated on the VHTR reference fuels. Also measured was tritium concentration in the helium coolant of OGL-1. In-pile release behavior of the metallic fission products was studied by measuring the activities of the fission products adsorbed in the graphite sleeves of the OGL-1 fuel rods and the graphite fuel container of the sweep gas capsules in the PIE. Investigation on palladium interaction with SiC coating layer was included. (author)

  16. Thermal conductivity of sintered UO{sub 2} under in-pile conditions; Conductibilite thermique de l'UO{sub 2} fritte dans les conditions d'utilisation en pile

    Energy Technology Data Exchange (ETDEWEB)

    Stora, J P; Bernardy De Sigoyer, B; Delmas, R; Deschamps, P; Lavaud, B; Ringot, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The temperature distribution in a stack of sintered UO{sub 2} cylinders has been studied both in the laboratory where the heat energy is produced by an axial heating element, and in-pile, where the heating is due solely to nuclear effects. Under a high thermal gradient the UO{sub 2} cracks both along radial planes and along pseudo-cylindrical surfaces: these latter act as thermal barriers to the heat flow, It is therefore an apparent thermal conductivity k{sub a}(T), lower than the intrinsic value k(T) of this parameter which is measured. The efficiency of these barriers decreases when the gap decreases and when the external pressure acting on the cracked stack increases: in the limiting case, for high values of the binding strain, k{sub a}(T) {approx_equal} k(T). In the domain of phonon conduction (T {<=} 1350 deg C), the expression kw.cm{sup -1}.C{sup -1}=1/(11+0.024*T) accounts for the real thermal conductivity. Above 1350 deg C the thermal conductivity increases. Two in-pile measurements up to 1250 deg C carried out using cartridges fitted with thermocouples confirm, within the limits of experimental error, the above expression and the qualitative effects of the binding strains. Similar tests have been carried out-of-pile and in-pile on the real shape of the EL-4 fuel 'pencils'. Out-of-pile, the influence of the initial free gap, of the nature of the gas filing the 'pencil' and of the external pressure have been studied; the results are compatible with the above interpretation; It appears that an external pressure of 60 kg/cm{sup 2} is insufficient to restore completely the thermal conductivity of the fuel. (authors) [French] La distribution de temperature dans un empilement de cylindres d'UO{sub 2} fritte est etudiee a la fois au laboratoire, ou l'energie calorifique est produite par un element chauffant axial, et en pile, ou l'echauffement est uniquement nucleaire. Sous gradient thermique eleve, l'UO{sub 2} se fracture a la fois suivant des plans radiaux et

  17. Comparison of two surface temperature measurement using thermocouples and infrared camera

    Directory of Open Access Journals (Sweden)

    Michalski Dariusz

    2017-01-01

    Full Text Available This paper compares two methods applied to measure surface temperatures at an experimental setup designed to analyse flow boiling heat transfer. The temperature measurements were performed in two parallel rectangular minichannels, both 1.7 mm deep, 16 mm wide and 180 mm long. The heating element for the fluid flowing in each minichannel was a thin foil made of Haynes-230. The two measurement methods employed to determine the surface temperature of the foil were: the contact method, which involved mounting thermocouples at several points in one minichannel, and the contactless method to study the other minichannel, where the results were provided with an infrared camera. Calculations were necessary to compare the temperature results. Two sets of measurement data obtained for different values of the heat flux were analysed using the basic statistical methods, the method error and the method accuracy. The experimental error and the method accuracy were taken into account. The comparative analysis showed that although the values and distributions of the surface temperatures obtained with the two methods were similar but both methods had certain limitations.

  18. Establishment of the Co-C Eutectic Fixed-Point Cell for Thermocouple Calibrations at NIMT

    Science.gov (United States)

    Ongrai, O.; Elliott, C. J.

    2017-08-01

    In 2015, NIMT first established a Co-C eutectic temperature reference (fixed-point) cell measurement capability for thermocouple calibration to support the requirements of Thailand's heavy industries and secondary laboratories. The Co-C eutectic fixed-point cell is a facility transferred from NPL, where the design was developed through European and UK national measurement system projects. In this paper, we describe the establishment of a Co-C eutectic fixed-point cell for thermocouple calibration at NIMT. This paper demonstrates achievement of the required furnace uniformity, the Co-C plateau realization and the comparison data between NIMT and NPL Co-C cells by using the same standard Pt/Pd thermocouple, demonstrating traceability. The NIMT measurement capability for noble metal type thermocouples at the new Co-C eutectic fixed point (1324.06°C) is estimated to be within ± 0.60 K (k=2). This meets the needs of Thailand's high-temperature thermocouple users—for which previously there has been no traceable calibration facility.

  19. Evaluation of thermocouple fin effect in cladding surface temperature measurement during film boiling

    International Nuclear Information System (INIS)

    Tsuruta, Takaharu; Fujishiro, Toshio

    1984-01-01

    Thermocouple fin effect on surface temperature measurement of a fuel rod has been studied at elevated wall temperatures under film boiling condition in a reactivity initiated accident (RIA) situation. This paper presents an analytical equation to evaluate temperature drops caused by the thermocouple wires attached to cladding surface. The equation yielded the local temperature drop at measuring point depending on thermocouple diameter, cladding temperature, coolant flow condition and vapor film thickness. The temperature drops by the evaluating equation were shown in cases of free and forced convection conditions. The analytical results were compared with the measured data for various thermocouple sizes, and also with the estimated maximum cladding temperature based on the oxidation layer thickness in the cladding outer surface. It was concluded that the temperature drops at above 1,000 0 C in cladding temperature were around 120 and 150 0 C for 0.2 and 0.3 mm diameter Pt-Pt.Rh thermocouples, respectively, under a stagnant coolant condition. The fin effect increases with the decrease of vapor film thickness such as under forced flow cooling or at near the quenching point. (author)

  20. R and D advances in high temperature thermocouples for nuclear utilization in severe environment

    International Nuclear Information System (INIS)

    Schley, R.; Blanc, J.Y.

    1984-09-01

    Safety experiments for water reactors in Cadarache have made necessary a research program for developing special thermocouples for use in severe fuel damage conditions (superheated steam). Standard cladding thermocouples (type K, alumina insulated, zircaloy sheathed, O.D. 0.7 mm) must be replaced by others with W3Re versus W25Re legs, Ta sheath protected by a zircaloy outer sheath, and hafnia or thoria insulation. The zircaloy sheath will be sufficient to protect correctly tantalum. Fuel centerline thermocouples have W5Re versus W26Re or W3Re versus W25Re legs, hard-fired thoria insulation and rhenium CVD sheath (O.D. 1.1 mm). A protective ReSi 2 coating is applied. This protection withstands at least 1600 0 C, 45 minutes in steam. Tests are done-concerning: a) materials compatibilities in helium between 1400 0 C and 2000 0 C, b) prototypes qualification (in Saclay or Grenoble), c) determination of errors due to degradation of insulation resistance of thermocouples cables (with magnesia, hafnia, alumina), d) Ir or Re protective coatings by CVD process, other coatings by ionic bombardment, etc... A completely new type of hot junction has been patented. Future works will include: completion of these tests, Mo-Nb alloys thermocouples legs realization withstanding heavy neutronic fluence, and use of ceramics glues

  1. Thermocouple Errors when Mounted on Cylindrical Surfaces in Abnormal Thermal Environments.

    Energy Technology Data Exchange (ETDEWEB)

    Nakos, James T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Suo-Anttila, Jill M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Zepper, Ethan T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Koenig, Jerry J [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Valdez, Vincent A. [ECI Inc., Albuquerque, NM (United States)

    2017-05-01

    Mineral-insulated, metal-sheathed, Type-K thermocouples are used to measure the temperature of various items in high-temperature environments, often exceeding 1000degC (1273 K). The thermocouple wires (chromel and alumel) are protected from the harsh environments by an Inconel sheath and magnesium oxide (MgO) insulation. The sheath and insulation are required for reliable measurements. Due to the sheath and MgO insulation, the temperature registered by the thermocouple is not the temperature of the surface of interest. In some cases, the error incurred is large enough to be of concern because these data are used for model validation, and thus the uncertainties of the data need to be well documented. This report documents the error using 0.062" and 0.040" diameter Inconel sheathed, Type-K thermocouples mounted on cylindrical surfaces (inside of a shroud, outside and inside of a mock test unit). After an initial transient, the thermocouple bias errors typically range only about +-1-2% of the reading in K. After all of the uncertainty sources have been included, the total uncertainty to 95% confidence, for shroud or test unit TCs in abnormal thermal environments, is about +-2% of the reading in K, lower than the +-3% typically used for flat shrouds. Recommendations are provided in Section 6 to facilitate interpretation and use of the results. .

  2. In-pile test results of U-silicide or U-nitride coated U-7Mo particle dispersion fuel in Al

    Science.gov (United States)

    Kim, Yeon Soo; Park, J. M.; Lee, K. H.; Yoo, B. O.; Ryu, H. J.; Ye, B.

    2014-11-01

    U-silicide or U-nitride coated U-Mo particle dispersion fuel in Al (U-Mo/Al) was in-pile tested to examine the effectiveness of the coating as a diffusion barrier between the U-7Mo fuel kernels and Al matrix. This paper reports the PIE data and analyses focusing on the effectiveness of the coating in terms of interaction layer (IL) growth and general fuel performance. The U-silicide coating showed considerable success, but it also provided evidence for additional improvement for coating process. The U-nitride coated specimen showed largely inefficient results in reducing IL growth. From the test, important observations were also made that can be utilized to improve U-Mo/Al fuel performance. The heating process for coating turned out to be beneficial to suppress fuel swelling. The use of larger fuel particles confirmed favorable effects on fuel performance.

  3. In-pile test results of U-silicide or U-nitride coated U-7Mo particle dispersion fuel in Al

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Park, J.M.; Lee, K.H.; Yoo, B.O. [Korea Atomic Energy Research Institute, 989-111 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Ryu, H.J. [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Ye, B. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2014-11-15

    U-silicide or U-nitride coated U-Mo particle dispersion fuel in Al (U-Mo/Al) was in-pile tested to examine the effectiveness of the coating as a diffusion barrier between the U-7Mo fuel kernels and Al matrix. This paper reports the PIE data and analyses focusing on the effectiveness of the coating in terms of interaction layer (IL) growth and general fuel performance. The U-silicide coating showed considerable success, but it also provided evidence for additional improvement for coating process. The U-nitride coated specimen showed largely inefficient results in reducing IL growth. From the test, important observations were also made that can be utilized to improve U-Mo/Al fuel performance. The heating process for coating turned out to be beneficial to suppress fuel swelling. The use of larger fuel particles confirmed favorable effects on fuel performance.

  4. In-pile instrumentation improvements for fuel irradiations in test reactor

    International Nuclear Information System (INIS)

    Blanc, J.Y.; Bernard, J.L.; Estrade, J.; Geoffroy, G.

    1996-01-01

    Knowledge of fuel limits and safety margins in normal and off-normal transients in nuclear power plants remains a constant preoccupation for electricity producers and fuel manufacturers. Accurate determination of such limits, through fuel irradiation testing in the OSIRIS reactor at Saclay is closely linked to the reliability of appropriate instrumentation techniques. Two paths are currently followed to obtain short experimental rods: segmented fuel coming directly from power plants, or re-fabrication of rods in hot cells with our FABRICE process. It can be associated with instrumentation such as fuel centerline thermocouple in annular pellets, pressure transducer or fission gas release measurement by gamma-spectrometry using helium sweeping, in analytic experiments. Our present development, to be implemented in 1993, is the the centerline instrumentation of a fuel column with solid pellets. Inserting the thermocouple requires a cold drilling machine, using CO 2 freezing of broken UO 2 (with liquid nitrogen). During the fuel rod irradiation itself, we try to lower the uncertainties associated to power determination, using thermal balance or neutronic calibration, or even gamma spectrometry. A description of the new test train designed for the ISABELLE water loop in OSIRIS is given, with special emphasis on instrumentation: a LVDT for measuring fuel rod elongation and eventual clad failure, and increased number and better localization of thermocouples and SPDN. The third part is devoted to the measurements by optical microdensitometry of neutron radiographs of the fuel pellet dish modification after irradiation. Dishes are generally disappearing through thermal and mechanical deformation of the pellet, and this can eventually be modelized to better understand pellet-cladding mechanical interaction. (author). 3 refs, 5 figs

  5. Relocation work of temporary thermocouples for measuring the vessel cooling system in the safety demonstration test

    International Nuclear Information System (INIS)

    Shimazaki, Yosuke; Shinohara, Masanori; Ono, Masato; Yanagi, Shunki; Tochio, Daisuke; Iigaki, Kazuhiko

    2012-05-01

    It is necessary to confirm that the temperature of water cooling panel of the vessel cooling system (VCS) is controlled under the allowable working temperature during the safety demonstration test because the water cooling panel temperature rises due to stop of cooling water circulation pumps. Therefore, several temporary thermocouples are relocated to the water cooling panel near the stabilizers of RPV and the side cooling panel outlet ring header of VCS in order to observe the temperature change of VCS. The relocated thermocouples can measure the temperature change with starting of the cooling water circulation pumps of VCS. So it is confirmed that the relocated thermocouples can observe the VCS temperature change in the safety demonstration test. (author)

  6. Current in-pile absorbed dose measurements at the Boris Kidric Institute of nuclear sciences - Vinca, Status report

    Energy Technology Data Exchange (ETDEWEB)

    Draganic, G I [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1966-11-15

    So far in-pile absorbed dose measurements have been limited only to experiments in the RA reactor at the Boris Kidric Institute of Nuclear Sciences at Vinca (6.5 D{sub 2}O moderated and 2% enriched uranium). The methods used for absorbed dose and neutron flux measurements were 1,2 discussed in some earlier reports at the IAEA meetings. The purpose of the present report is to illustrate the further development of methods of determining in-pile absorbed doses (author)

  7. Neutronic and thermal estimation of blanket in-pile mockup with Li2TiO3 pebbles

    International Nuclear Information System (INIS)

    Nagao, Y.; Nakamichi, M.; Tsuchiya, K.; Kawamura, H.

    2001-01-01

    To evaluate exactly temperature distribution in large volume of tritium breeding materials during the blanket in-pile tests with the JMTR, neutronic and thermal calculations were conducted by using Monte Carlo code 'MCNP' with nuclear cross section library of 'FSXLIBJ3R2' and the transient and steady-state distribution code 'TRUMP'. From the results of preliminary estimation of temperature distribution in the blanket in-pile mockup, the calculated values were 24-28% higher than the measured values. One of the reasons is due to overestimation of calculated thermal neutron flux

  8. Evaluation on sweep gas pressure drop in fusion blanket mock-up for in-pile test

    International Nuclear Information System (INIS)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Sagawa, Hisashi; Nagakura, Masaaki; Kanzawa, Toru.

    1993-03-01

    In the ITER/CDA (Conceptual Design Activity) of a tritium breeding blanket, Japan have proposed the pebble-typed blanket. The in-pile mock-up test will be preparing in JMTR (Japan Materials Testing Reactor) for Japanese engineering design with the pebble-typed blanket. Therefore, the He sweep gas pressure drop in the pebble bed was measured for the design of the mock-up used on in-pile test. From the results of this test, it was clear that the pressure drop was predicted on Kozeny- Carman's equation within +25 ∼ -60 %, and that the pressure drop was not affected by moisture concentration (< 100 ppm). (author)

  9. Thermocouple Rakes for Measuring Boundary Layer Flows Extremely Close to Surface

    Science.gov (United States)

    Hwang, Danny P.; Fralick, Gustave C.; Martin, Lisa C.; Blaha, Charles A.

    2001-01-01

    Of vital interest to aerodynamic researchers is precise knowledge of the flow velocity profile next to the surface. This information is needed for turbulence model development and the calculation of viscous shear force. Though many instruments can determine the flow velocity profile near the surface, none of them can make measurements closer than approximately 0.01 in. from the surface. The thermocouple boundary-layer rake can measure much closer to the surface than conventional instruments can, such as a total pressure boundary layer rake, hot wire, or hot film. By embedding the sensors (thermocouples) in the region where the velocity is equivalent to the velocity ahead of a constant thickness strut, the boundary-layer flow profile can be obtained. The present device fabricated at the NASA Glenn Research Center microsystem clean room has a heater made of platinum and thermocouples made of platinum and gold. Equal numbers of thermocouples are placed both upstream and downstream of the heater, so that the voltage generated by each pair at the same distance from the surface is indicative of the difference in temperature between the upstream and downstream thermocouple locations. This voltage differential is a function of the flow velocity, and like the conventional total pressure rake, it can provide the velocity profile. In order to measure flow extremely close to the surface, the strut is made of fused quartz with extremely low heat conductivity. A large size thermocouple boundary layer rake is shown in the following photo. The latest medium size sensors already provide smooth velocity profiles well into the boundary layer, as close as 0.0025 in. from the surface. This is about 4 times closer to the surface than the previously used total pressure rakes. This device also has the advantage of providing the flow profile of separated flow and also it is possible to measure simultaneous turbulence levels within the boundary layer.

  10. Determination of the availability of core exit thermocouples during severe accident situations

    International Nuclear Information System (INIS)

    Edson, J.L.

    1985-04-01

    This report presents the findings and recommendations of the Nuclear Power Plant Instrumentation Evaluation (NPPIE) program concerning signal validation methods to determine the on-line availability of core exit thermocouples during accident situations. Methods of selecting appropriate signal validation techniques are discussed and sources of error identified. This report shows that through the use of these techniques the existence of high-temperature-caused errors may be detected as they occur. Specific recommendations for application of selected signal validation techniques to core exit thermocouples and other measurement systems are made. 23 refs., 22 figs., 3 tabs

  11. Reply to ''Comment on 'Thermocouple temperature measurements in shock-compressed solids' ''

    International Nuclear Information System (INIS)

    Bloomquist, D.D.; Sheffield, S.A.

    1982-01-01

    We disagree with the interpretation offered in the above comment. The suggestion was made that the anomalously fast response of thin-foil thermocouples reported previously is the result of strain dependence of the thermocouple response and not shock enhanced thermal equilibration. Although the emplacement geometry has a profound effect on the response of embedded thin-foil temperature gauges as noted in the above comment, the evidence presented, along with recent results discussed in this reply, do not support the conclusions presented in the above comment

  12. Accuracy of small diameter sheathed thermocouples for the core flow test loop

    International Nuclear Information System (INIS)

    Anderson, R.L.; Kollie, T.G.

    1979-04-01

    This report summarizes the research and development on 0.5-mm-diameter, compacted, metal sheathed thermocouples. The objectives of this research effort have been: to identify and analyze the sources of temperature measurement errors in the use of 0.5-mm-diameter sheathed thermocouples to measure the surface temperature of the cladding of fuel-rod simulators in the Core Flow Test Loop (CFTL) at ORNL; to devise methods for reducing or correcting for these temperature measurement errors; to estimate the overall temperature measurement uncertainties; and to recommend modifications in the manufacture, installation, or materials used to minimize temperature measurement uncertainties in the CFTL experiments

  13. Review Report on the Design of In-Pile Test Section(IPS)

    International Nuclear Information System (INIS)

    Lee, Jong Min; Park, Kook Nam; Shim, Bong Sik; Lee, Chung Young; Chi, Dae Young; Park, Su Ki; Ahn, Sung Ho; Kim, Young Ki; Lee, Kye Hong; Kim, Kwan Hyun

    2009-01-01

    The In-Pile Test Section(IPS) accommodating fuel pins has loaded IR-1 hole in HANARO has double pressure vessel for the design conditions of 350 deg. C, 17.5 MPa and is composed of outer assembly and inner assembly. Dummy fuel, dummy fuel supports and Top flange are the main components in inner assembly and inner pressure vessel, outer pressure vessel and head are the components in outer assembly. The IPS at current status has dummy fuels and confirm the requirements for the IPS design improvements during the design, manufacturing and installation process. Head, Top Flange, Instrumentation Feed through, Lifting Eye, Fuel Carrier Leg, Retainer and Nozzle cover are the main parts that the design needs to be changed. This report suggest the needs for the IPS design modification and it would be reflected to the new IPS design which would accommodating test fuel pins

  14. Integrated, digital experiment transient control and safety protection of an in-pile test

    International Nuclear Information System (INIS)

    Thomas, R.W.; Whitacre, R.F.; Klingler, W.B.

    1982-01-01

    The Sodium Loop Safety Facility experimental program has demonstrated that in-pile loop fuel failure transient tests can be digitally controlled and protected with reliability and precision. This was done in four nuclear experiments conducted in the Engineering Test Reactor operated by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory. Loop sodium flow and reactor power transients can be programmed to sponsor requirements and verified prior to the test. Each controller has redundancy, which reduces the effect of single failures occurring during test transients. Feedback and reject criteria are included in the reactor power control. Timed sequencing integrates the initiation of the controllers, programmed safety set-points, and other experiment actions (e.g., planned scram). Off-line and on-line testing is included. Loss-of-flow, loss-of-piping-integrity, boiling-window, transient-overpower, and local fault tests have been successfully run using this system

  15. Report On Design And Preliminary Data Of Halden In-Pile Creep Rig

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, Kurt A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Karlsen, T. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    A set of in-pile creep tests is ongoing in the Halden reactor on ORNL’s candidate accident tolerant fuel cladding materials. These tests are meant to provide essential material property information that is needed for an informed analysis of these fuel concepts under normal operating conditions. These tests provide detailed information regarding swelling, thermal creep, and irradiation creep rates of these materials. The results to date have been compared with the limited set of information available in literature that is form irradiation tests in other reactors or out-of-pile tests. Most of the results are in good agreement with prior literature, except for irradiation creep rate of SiC. To elucidate the difference between the HFIR and Halden test results continued testing is necessary. The tests describe in this progress report are ongoing and will continue for at least another year.

  16. Characteristics of pressure control system on PWR/PHWR in pile loop facility

    International Nuclear Information System (INIS)

    Sarwani; Hendro, P.; Suwoto; Sutrisno

    1998-01-01

    PWR/PHWR in-pile loop facility is used for testing of fuel element bundle which is correspond to the condition of power reactor operation. So, this facility is designed at 150 bar of pressure and 350 o C of temperature. Pressure control system is one of the components of the facility and it is equipped with 6 electrical heaters (30 KW), water spray, pressure and temperature monitors. The characterization test of pressure control system has been carried out with operating of 2 electrical heaters (10 KW). The K eff calculation value is different 5.2% from pressure in the pressure control system can be increased to 160 bar within 27 hours. After the system pressure reached the nominal pressure, the pressure control system was in the steady state condition

  17. Probabilistic Modeling of Updating Epistemic Uncertainty In Pile Capacity Prediction With a Single Failure Test Result

    Directory of Open Access Journals (Sweden)

    Indra Djati Sidi

    2017-12-01

    Full Text Available The model error N has been introduced to denote the discrepancy between measured and predicted capacity of pile foundation. This model error is recognized as epistemic uncertainty in pile capacity prediction. The statistics of N have been evaluated based on data gathered from various sites and may be considered only as a eneral-error trend in capacity prediction, providing crude estimates of the model error in the absence of more specific data from the site. The results of even a single load test to failure, should provide direct evidence of the pile capacity at a given site. Bayes theorem has been used as a rational basis for combining new data with previous data to revise assessment of uncertainty and reliability. This study is devoted to the development of procedures for updating model error (N, and subsequently the predicted pile capacity with a results of single failure test.

  18. Review Report on the Design of In-Pile Test Section(IPS)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jong Min; Park, Kook Nam; Shim, Bong Sik; Lee, Chung Young; Chi, Dae Young; Park, Su Ki; Ahn, Sung Ho; Kim, Young Ki; Lee, Kye Hong; Kim, Kwan Hyun

    2009-01-15

    The In-Pile Test Section(IPS) accommodating fuel pins has loaded IR-1 hole in HANARO has double pressure vessel for the design conditions of 350 deg. C, 17.5 MPa and is composed of outer assembly and inner assembly. Dummy fuel, dummy fuel supports and Top flange are the main components in inner assembly and inner pressure vessel, outer pressure vessel and head are the components in outer assembly. The IPS at current status has dummy fuels and confirm the requirements for the IPS design improvements during the design, manufacturing and installation process. Head, Top Flange, Instrumentation Feed through, Lifting Eye, Fuel Carrier Leg, Retainer and Nozzle cover are the main parts that the design needs to be changed. This report suggest the needs for the IPS design modification and it would be reflected to the new IPS design which would accommodating test fuel pins.

  19. ACTIV, Sandwich Detector Activity from In-Pile Slowing-Down Spectra Experiment

    International Nuclear Information System (INIS)

    Bozzi, L. and others

    1978-01-01

    1 - Nature of physical problem solved: Calculates the activities of a sandwich detector, to be used for in-pile measurements in slowing-down spectra below a few keV. The effect of scattering with energy degradation in the filter and in the detectors has been included to a first approximation. 2 - Method of solution: An iterative procedure is used: the calculation starts with a flux guess in which one assumes that each measured reactivity difference depends on the principal resonance only. The secondary resonance contribution is computed through the iterative process. For self-shielded cross-section calculations the model of Pearlstein and Weinstock (ref. 3) is used. The neutron spectrum can optionally be constant or 1/E inside each finite energy group relative to the resonance considered. 3 - Restrictions on the complexity of the problem: Maximum number of energy groups : 60

  20. State-of-the-art review of some artificial intelligence applications in pile foundations

    Directory of Open Access Journals (Sweden)

    Mohamed A. Shahin

    2016-01-01

    Full Text Available Geotechnical engineering deals with materials (e.g. soil and rock that, by their very nature, exhibit varied and uncertain behavior due to the imprecise physical processes associated with the formation of these materials. Modeling the behavior of such materials in geotechnical engineering applications is complex and sometimes beyond the ability of most traditional forms of physically-based engineering methods. Artificial intelligence (AI is becoming more popular and particularly amenable to modeling the complex behavior of most geotechnical engineering applications because it has demonstrated superior predictive ability compared to traditional methods. This paper provides state-of-the-art review of some selected AI techniques and their applications in pile foundations, and presents the salient features associated with the modeling development of these AI techniques. The paper also discusses the strength and limitations of the selected AI techniques compared to other available modeling approaches.

  1. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    International Nuclear Information System (INIS)

    Agarwal, Vivek; Smith, James A.; Jewell, James Keith

    2015-01-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  2. Investigation of in-pile grown corrosion films on zirconium-based alloys

    International Nuclear Information System (INIS)

    Gebhardt, O.; Hermann, A.; Bart, G.; Blank, H.; Ray, I.L.F.

    1996-01-01

    In-pile grown corrosion films on different fuel rod claddings (standard Zircaloy-4, extra low tin Zircaloy (ELS), and Zr2.5Nb) have been studied using a variety of experimental techniques. The aim of the investigations was to find out common features and differences between the corrosion layers grown on zirconium alloys having different composition. Methods applied were scanning and transmission electron microscopy (SEM, TEM), electrochemical impedance spectroscopy (EIS), and electrochemical anodization. The morphological differences have been observed between the specimens that could explain the irradiation enhancement of corrosion of Zircaloy-4. The features of the compact oxide close to the oxide/metal interface have been characterized by electrochemical methods. The relationship between the thickness of this protective oxide and the overall oxide thickness has been investigated by EIS. It was found that this relation is dependent on the location of the oxide along the fuel rod and on the corrosion rate

  3. State-of-the-art review of some artificial intelligence applications in pile foundations

    Institute of Scientific and Technical Information of China (English)

    Mohamed A. Shahin

    2016-01-01

    Geotechnical engineering deals with materials (e.g. soil and rock) that, by their very nature, exhibit varied and uncertain behavior due to the imprecise physical processes associated with the formation of these materials. Modeling the behavior of such materials in geotechnical engineering applications is complex and sometimes beyond the ability of most traditional forms of physically-based engineering methods. Artificial intelligence (AI) is becoming more popular and particularly amenable to modeling the complex behavior of most geotechnical engineering applications because it has demonstrated superior predictive ability compared to traditional methods. This paper provides state-of-the-art review of some selected AI techniques and their applications in pile foundations, and presents the salient features associated with the modeling development of these AI techniques. The paper also discusses the strength and limitations of the selected AI techniques compared to other available modeling approaches.

  4. MCNP simulation to optimise in-pile and shielding parts of the Portuguese SANS instrument.

    Science.gov (United States)

    Gonçalves, I F; Salgado, J; Falcão, A; Margaça, F M A; Carvalho, F G

    2005-01-01

    A Small Angle Neutron Scattering instrument is being installed at one end of the tangential beam tube of the Portuguese Research Reactor. The instrument is fed using a neutron scatterer positioned in the middle of the beam tube. The scatterer consists of circulating H2O contained in a hollow disc of Al. The in-pile shielding components and the shielding installed around the neutron selector have been the object of an MCNP simulation study. The quantities calculated were the neutron and gamma-ray fluxes in different positions, the energy deposited in the material by the neutron and gamma-ray fields, the material activation resulting from the neutron field and radiation doses at the exit wall of the shutter and around the shielding. The MCNP results are presented and compared with results of an analytical approach and with experimental data collected after installation.

  5. In-pile measurement of the thermal conductivity of irradiated metallic fuel

    International Nuclear Information System (INIS)

    Bauer, T.H.; Holland, J.W.

    1995-01-01

    Transient test data and posttest measurements from recent in-pile overpower transient experiments are used for an in situ determination of metallic fuel thermal conductivity. For test pins that undergo melting but remain intact, a technique is described that relates fuel thermal conductivity to peak pin power during the transient and a posttest measured melt radius. Conductivity estimates and their uncertainty are made for a database of four irradiated Integral Fast Reactor-type metal fuel pins of relatively low burnup (<3 at.%). In the assessment of results, averages and trends of measured fuel thermal conductivity are correlated to local burnup. Emphasis is placed on the changes of conductivity that take place with burnup-induced swelling and sodium logging. Measurements are used to validate simple empirically based analytical models that describe thermal conductivity of porous media and that are recommended for general thermal analyses of irradiated metallic fuel

  6. Visual observations of fuel disruption in in-pile LMFBR accident experiments

    International Nuclear Information System (INIS)

    Wright, S.A.; Mast, P.K.

    1982-01-01

    Sandia National Laboratories has been investigating initiation phase phenomena in a series of Fuel Disruption (FD) experiments since 1977. In this program high speed cinematography is used to observe fuel disruption in in-pile experiments that simulate loss of flow accidents. Thus, these experiments provide high resolution measurements of initial fuel and clad motion with prototypic materials and prototypic heating conditions. The main objective of the FD experiment is to determine the timing (relative to fuel temperature) and the mode of fuel disruption under LOF heating conditions. Observed modes of disruption include fuel swelling, solid state breakup, cracking, ejection of a molten fuel jet, slumping, and rapid expansion of small particles. Because the temperature and character of the fuel at disruption are known, disruption can be correlated with the mechanisms driving the disruption such as fuel vapor pressure, molten fuel expansion, fission gases, and impurity gases

  7. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Vivek [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Human Factors, Controls, and Statistics; Smith, James A. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design; Jewell, James Keith [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design

    2015-02-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  8. An In-Pile Kinetic Method for Determining the Delayed Neutron Fraction βeff

    International Nuclear Information System (INIS)

    Gilad, E.; Rivin, O.; Ettedgui, H.; Yaar, I.; Geslot, B.; Pepino, A.; Di Salvo, J.; Gruel, A.; Blaise, P.

    2014-01-01

    Delayed neutrons are of fundamental importance in the field of nuclear reactor dynamics and control. Although only a small fraction of the neutrons emitted by fission are not prompt, the knowledge of the delayed neutrons parameters is essential for transient analysis, such as startup or shutdown of the reactor, as well as for accidents analysis and control system design [1]. One of the main delayed neutron parameters used in the point reactor model equations is the effective delayed neutron fraction, which incorporates both delayed neutron spectral properties and core geometrical configuration [1,2]. Additional delayed neutron parameters include the fraction of fission neutrons emitted in each delayed group, and the delayed neutron precursors decay constants . Experimental efforts aimed at determining the value ofβ, which provide experimental support for the evaluation of delayed neutron parameters, are extremely valuable. This is due to the fact that unlike other fields in reactor physics, e.g. criticality safety or shielding, the availability of experimental data and benchmark problems for validating delayed neutron parameters and its implementation in different models is highly limited. Furthermore, the existing experimental data exhibit significant discrepancies between the different sets of parameter, which lead to substantial disparity in the analysis of kinetic experiments and reactor dynamic behavior]. In this work, a method for determining the effective delayed neutron fraction using in-pile reactivity oscillation and Fourier analysis is presented. The method is based on measurements of the reactor's power response to small periodic in-pile reactivity perturbations and utilizes Fourier analysis for reconstruction of the reactor zero power transfer function. Knowledge of the reactor transfer function enables the estimation of theβ value using multi-parameter nonlinear fit. The method accounts for higher harmonics, which are excited by the trapezoidal

  9. Thermal-hydraulic calculation and analysis on helium cooled ceramic breeder pebble bed assembly for in-pile irradiation and in-situ tritium extraction

    International Nuclear Information System (INIS)

    Guo Chunqiu; Xie Jiachun; Liu Xingmin

    2013-01-01

    In-pile irradiation and in-situ tritium extraction experiment is one of associated domestic research projects in ITER special program. According to the technical requirements of in-pile irradiation experiment of helium cooled ceramic breeder (ceramic) pebble bed assembly in a research reactor, the feasibility of the design for the in-pile irradiation and in-situ tritium extraction experiment of ceramic pebble bed assembly was evaluated. By conducting thermal-hydraulic design calculation with different in-pile irradiation channels, locations and structure parameters for ceramic pebble bed assembly, a reasonable design scheme of ceramic pebble bed assembly satisfying the design requirements for in-pile irradiation was obtained. (authors)

  10. 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

  11. Manifestations of nonlinearity in fuel center thermocouple steady-state and transient data: implications for data analysis

    International Nuclear Information System (INIS)

    Lanning, D.D.; Barnes, B.O.; Williford, R.E.

    1979-01-01

    The interpretation and verification of fuel centerline thermocouple data are analyzed. Two new concepts are discussed along with their application to in-reactor data from IFA-432, a heavily instrumented six-rod Halden reactor test assembly sponsored by the Nuclear Regulatory Commission. The main ideas presented in this report are that: it is more useful to plot resistance versus power than simply to plot temperature versus power; and the response of the centerline temperature to a linear power decrease is correlated to the rod's current resistance-vs-power behavior. Thus, the resistance-vs-power measurement can be verified by performing a linear power decrease and by plotting the temperature response

  12. In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

    1983-01-01

    Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio

  13. Computer subroutines to aid analysis of experimental data from thermocouples and pressure transducers

    International Nuclear Information System (INIS)

    Durham, M.E.

    1976-08-01

    Three subroutines (CALSET, CALBR8 and PTRCAL) have been written to provide a convenient system for converting experimental measurements obtained from thermocouples and pressure transducers to temperatures and pressures. The method of operation and the application of the subroutines are described. (author)

  14. Measuring skin temperature before, during and after exercise: a comparison of thermocouples and infrared thermography

    International Nuclear Information System (INIS)

    Fernandes, Alex de Andrade; Amorim, Paulo Roberto dos Santos; De Moura, Anselmo Gomes; Moreira, Danilo Gomes; Costa, Carlos Magno Amaral; Marins, João Carlos Bouzas; Brito, Ciro José; Sillero-Quintana, Manuel

    2014-01-01

    Measuring skin temperature (T SK ) provides important information about the complex thermal control system and could be interesting when carrying out studies about thermoregulation. The most common method to record T SK  involves thermocouples at specific locations; however, the use of infrared thermal imaging (IRT) has increased. The two methods use different physical processes to measure T SK , and each has advantages and disadvantages. Therefore, the objective of this study was to compare the mean skin temperature (MT SK ) measurements using thermocouples and IRT in three different situations: pre-exercise, exercise and post-exercise. Analysis of the residual scores in Bland–Altman plots showed poor agreement between the MT SK  obtained using thermocouples and those using IRT. The averaged error was −0.75 °C during pre-exercise, 1.22 °C during exercise and −1.16 °C during post-exercise, and the reliability between the methods was low in the pre- (ICC = 0.75 [0.12 to 0.93]), during (ICC = 0.49 [−0.80 to 0.85]) and post-exercise (ICC = 0.35 [−1.22 to 0.81] conditions. Thus, there is poor correlation between the values of MT SK  measured by thermocouples and IRT pre-exercise, exercise and post-exercise, and low reliability between the two forms of measurement. (paper)

  15. Preparation of a thermal-hydraulic design method for driver core fuel pins of a new in-pile experimental reactor for FBR safety research

    International Nuclear Information System (INIS)

    Mizuno, Masahiro; Yamaguchi, Katsuhisa; Uto, Nariaki

    1999-07-01

    A design study of a new in-pile experimental reactor, SERAPH (Safety Engineering Reactor for Accident PHenomenology), for FBR safety research has progressed at JNC (Japan Nuclear Cycle Development Institute). SERAPH is intended for various in-pile experiments to be performed under quasi-steady state and various transient operation modes. In order to evaluate the driver core performance in conducting such experiments, clarify the relating design issues to be resolved and refine the experimental needs, it is indispensable to comprehend the allowable margin for the thermal-hydraulic fuel pin design since it largely affects the strategy for the driver core design. This report presents a thermal-hydraulic design method for the driver core fuel pins, which is a combination of a two-dimensional time-dependent heat transfer analysis code TAC-2D and a general non-linear finite-element structural analysis code FINAS. In TAC-2D, the allowable spatial mesh and the time step sizes are evaluated. The code is modified so as to treat time-dependent thermal properties, include an improved gap heat-transfer model and treat the change of intra-pin gap width under transient modes, for the purpose of improving the accuracy of evaluating heat transfer characteristics which gives a significant impact on the thermal-hydraulic design. As for FINAS, the number of element nodes and spatial meshes required to obtain adequate accuracy for the thermal stress characteristics of a fuel pellet during transient modes are investigated. In addition, post-processing tools are newly developed to process the calculation results obtained from these codes. The results of this work contribute to advancing the fuel pin design study for SERAPH as well with the investigation on the technique of manufacturing fuel pins. (author)

  16. Direct Measurement of Neutral/Ion Beam Power using Thermocouple Analysis

    International Nuclear Information System (INIS)

    Day, I.; Gee, S.

    2006-01-01

    Modern Neutral Beam Injection systems such as those used on JET and MAST routinely use thermocouples embedded close to the surface of beam stopping elements, such as calorimeters and ion dumps, coupled to high speed data acquisition systems to determine beam profile and position from temperature rise data. With the availability of low cost data acquisition and storage systems it is now possible to record data from all thermocouples in a fully instrumented calorimeter or ion dump on 20 ms timescales or better. This sample rate is sufficiently fast to enable the thermocouple data to be used to calculate the incident power density from 1d heat transfer theory. This power density data coupled with appropriate Gaussian fits enables the determination of the 2d beam profile and thus allows an instantaneous and direct measurement of beam power. The theory and methodology required to analyse the fast thermocouple data from the MAST calorimeter and residual ion dump thermocouples is presented and direct measurements of beam power density are demonstrated. The power of desktop computers allows such analysis to be carried out virtually instantaneously. The methods used to automate this analysis are discussed in detail. A code, utilising the theory and methodology, has been developed to allow immediate measurements of beam power on a pulse by pulse basis. The uncertainty in determining the beam power density is shown to be less than 10 %. This power density data is then fitted to a 2d Gaussian beam profile and integrated to establish the total beam power. Results of this automated analysis for the neutral beam and residual ion power of the MAST duopigatron and PINI NBI systems are presented. This technology could be applied to a beam power safety interlock system. The application to a beam shine through protection system for the inner wall of the JET Tokamak is discussed as an example. (author)

  17. Thermal Recovery from Cold-Working in Type K Bare-Wire Thermocouples

    Science.gov (United States)

    Greenen, A. D.; Webster, E. S.

    2017-12-01

    Cold-working of most thermocouples has a significant, direct impact on the Seebeck coefficient which can lead to regions of thermoelectric inhomogeneity and accelerated drift. Cold-working can occur during the wire swaging process, when winding the wire onto a bobbin, or during handling by the end user—either accidentally or deliberately. Swaging-induced cold-work in thermocouples, if uniformly applied, may result in a high level of homogeneity. However, on exposure to elevated temperatures, the subsequent recovery process from the cold-working can then result in significant drift, and this can in turn lead to erroneous temperature measurements, often in excess of the specified manufacturer tolerances. Several studies have investigated the effects of cold-work in Type K thermocouples usually by bending, or swaging. However, the amount of cold-work applied to the thermocouple is often difficult to quantify, as the mechanisms for applying the strains are typically nonlinear when applied in this fashion. A repeatable level of cold-working is applied to the different wires using a tensional loading apparatus to apply a known yield displacement to the thermoelements. The effects of thermal recovery from cold-working can then be accurately quantified as a function of temperature, using a linear gradient furnace and a high-resolution homogeneity scanner. Variation in these effects due to differing alloy compositions in Type K wire is also explored, which is obtained by sourcing wire from a selection of manufacturers. The information gathered in this way will inform users of Type K thermocouples about the potential consequences of varying levels of cold-working and its impact on the Seebeck coefficient at a range of temperatures between ˜ 70°C and 600° C. This study will also guide users on the temperatures required to rapidly alleviate the effects of cold-working using thermal annealing treatments.

  18. Risk analysis of the delayed ettringite formation in pile caps foundation in the metropolitan region of Recife - PE - Brasil

    Directory of Open Access Journals (Sweden)

    I. F. Torres

    Full Text Available ABSTRACT Currently, there is an awareness that is critical to assess the durability characteristics of concrete with as much attention as the mechanical properties. The durability of concrete structures can often be affected by chemical attacks, jeopardizing its performance and security. When concrete is subjected to high temperature at early ages, many physical and chemical changes in hardened concrete may occur. It iswidely accepted that concrete subjected to these conditions of temperature and exposed to moisture is prone to cracking due to Delayed Ettringite Formation (DEF. This work aims at providing a DEF risk analysis on foundation pile caps at the Metropolitan Region of Recife - PE. Temperature rise measurement was performed in situ at 5 different caps through datalogger and thermocouples equipments. Furthermore, the Duggan test was performed in order to assess the level of expansion of 3 cements studied: X (CP II E 40, Y (CP II F 32 and Z (CP V ARI RS. Simultaneously, the chemical compositions of these cements and their respective clinkers were quantified by analysis of X-ray fluorescence (XRF. The cement X (CP II E 40 showed the chemical characteristics favoring with more intensity DEF and, as a result, higher level of expansion in the test Duggan. It is noteworthy that incorporation of metakaolin (8% and 16% and silica fume (5% and 10% showed mitigating potential of expansions. It is important to point out that all factors related to thermal properties and chemical composition of the concrete used in the region converge to a condition of ideal susceptibility for triggering DEF. Therefore, it is essential at least minimum and basic requirements in the design specification in order to avoid high temperatures in the massive concrete elements, preventing them from delayed ettringite formation.

  19. Irradiation experiments of 3rd, 4th and 5th fuel assemblies by an in-pile gas loop, OGL-1

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kobayashi, Fumiaki; Hayashi, Kimio; Minato, Kazuo; Kikuchi, Teruo; Adachi, Mamoru; Iwamoto, Kazumi; Ikawa, Katsuichi; Itami, Hiroharu.

    1986-07-01

    Three irradiation experiments for 3rd, 4th and 5th fuel assemblies which had been composed of VHTR reference coated particle fuels and graphite components were carried out by an in-pile gas loop, OGL-1 during 1979 and 1982. The main purposes of these experiments were to study on bowing of the fuel rod by irradiation for the 3rd fuel assembly, to study on fuel behavior under relatively low burnup irradiation for the 4th fuel assembly, and to study on fuel behavior up to full burnup of VHTR design for the 5th fuel assembly. For understanding in-pile fuel behavior, fractional releases of fission gases from each fuel assembly were estimated by measuring the fission gas concentrations in the primary loop of OGL-1. The post-irradiation examination (PIE) was carried out extensively on the fuel block, the fuel rods and the fuel compacts in Tokai Hot Laboratory. Also, made were the measurements of metallic fission product distributions in the fuel assemblies and the fuel rods. The results in these experiments were given as follows ; bowing of the fuel rod in the 3rd fuel assembly was 0.7 mm, but integrity of the rod was kept under irradiation. Fractional release of the fission gas from the 4th fuel assembly remained in the order of 10 -7 during irradiation, suggesting that the fuel performance was excellent. The fractional release from the 5th fuel assembly, on the other hand, was in the order of 10 -5 which was the same level in the VHTR design. (author)

  20. Performance of MOX fuel: An overview of the experimental programme of the OECD Halden Reactor Project and review of selected results

    International Nuclear Information System (INIS)

    Wiesenack, W.; McGrath, M.

    2000-01-01

    The OECD Halden Reactor Project has defined an extensive experimental programme related to MOX fuels which is being executed with the objective to provide a performance data base similar to that available for UO 2 . In addition to utilising fresh MOX fuel and re-instrumented segments from LWR irradiations to high burnup, the concept of inert matrix fuel is being addressed. The irradiation in the Halden reactor is performed in rigs allowing steady state, power ramping and cyclic operation. In-pile data are obtained from instrumentation such as fuel centreline thermocouples, pressure transducers, fuel and cladding elongation detectors, and movable gauges for measuring the diametral deformation. Various phenomena can be assessed in this way, e.g. thermal performance, swelling and densification, PCMI and fission gas release. The paper describes the objectives of various experiments and provides examples of temperature, pressure and cladding elongation measurements performed on MOX fuel. Salient results are related to the threshold for the onset of significant fission gas release and the relaxation behaviour in a power ramp-PCMI situation. (author)

  1. In-Pile Instrumentation Multi- Parameter System Utilizing Photonic Fibers and Nanovision

    Energy Technology Data Exchange (ETDEWEB)

    Burgett, Eric [Idaho State Univ., Pocatello, ID (United States)

    2015-10-13

    An advanced in-pile multi-parameter reactor monitoring system is being proposed in this funding opportunity. The proposed effort brings cutting edge, high fidelity optical measurement systems into the reactor environment in an unprecedented fashion, including in-core, in-cladding and in-fuel pellet itself. Unlike instrumented leads, the proposed system provides a unique solution to a multi-parameter monitoring need in core while being minimally intrusive in the reactor core. Detector designs proposed herein can monitor fuel compression and expansion in both the radial and axial dimensions as well as monitor linear power profiles and fission rates during the operation of the reactor. In addition to pressure, stress, strain, compression, neutron flux, neutron spectra, and temperature can be observed inside the fuel bundle and fuel rod using the proposed system. The proposed research aims at developing radiation-hard, harsh-environment multi-parameter systems for insertion into the reactor environment. The proposed research holds the potential to drastically increase the fidelity and precision of in-core instrumentation with little or no impact in the neutron economy in the reactor environment while providing a measurement system capable of operation for entire operating cycles.

  2. In-Pile Instrumentation Multi- Parameter System Utilizing Photonic Fibers and Nanovision

    International Nuclear Information System (INIS)

    Burgett, Eric

    2015-01-01

    An advanced in-pile multi-parameter reactor monitoring system is being proposed in this funding opportunity. The proposed effort brings cutting edge, high fidelity optical measurement systems into the reactor environment in an unprecedented fashion, including in-core, in-cladding and in-fuel pellet itself. Unlike instrumented leads, the proposed system provides a unique solution to a multi-parameter monitoring need in core while being minimally intrusive in the reactor core. Detector designs proposed herein can monitor fuel compression and expansion in both the radial and axial dimensions as well as monitor linear power profiles and fission rates during the operation of the reactor. In addition to pressure, stress, strain, compression, neutron flux, neutron spectra, and temperature can be observed inside the fuel bundle and fuel rod using the proposed system. The proposed research aims at developing radiation-hard, harsh-environment multi-parameter systems for insertion into the reactor environment. The proposed research holds the potential to drastically increase the fidelity and precision of in-core instrumentation with little or no impact in the neutron economy in the reactor environment while providing a measurement system capable of operation for entire operating cycles.

  3. In situ monitored in-pile creep testing of zirconium alloys

    Science.gov (United States)

    Kozar, R. W.; Jaworski, A. W.; Webb, T. W.; Smith, R. W.

    2014-01-01

    The experiments described herein were designed to investigate the detailed irradiation creep behavior of zirconium based alloys in the HALDEN Reactor spectrum. The HALDEN Test Reactor has the unique capability to control both applied stress and temperature independently and externally for each specimen while the specimen is in-reactor and under fast neutron flux. The ability to monitor in situ the creep rates following a stress and temperature change made possible the characterization of creep behavior over a wide stress-strain-rate-temperature design space for two model experimental heats, Zircaloy-2 and Zircaloy-2 + 1 wt%Nb, with only 12 test specimens in a 100-day in-pile creep test program. Zircaloy-2 specimens with and without 1 wt% Nb additions were tested at irradiation temperatures of 561 K and 616 K and stresses ranging from 69 MPa to 455 MPa. Various steady state creep models were evaluated against the experimental results. The irradiation creep model proposed by Nichols that separates creep behavior into low, intermediate, and high stress regimes was the best model for predicting steady-state creep rates. Dislocation-based primary creep, rather than diffusion-based transient irradiation creep, was identified as the mechanism controlling deformation during the transitional period of evolving creep rate following a step change to different test conditions.

  4. A review of radionuclide release and transport in recent in-pile experiments

    International Nuclear Information System (INIS)

    Harman, N.F.; Clough, P.N.

    1992-01-01

    The experimental series, reviewed in this work, are LOFT-LP-FP-2, PBF SFD ST to 1-4, Treat/STEP I to 4, and ACRR ST-1 and ST-2. These have the common features of in-core heating of a test fuel bundle to high temperatures (usually with some fuel melting) in an experimental reactor core, and of collecting and analysing the released fission products. They were designed to provide detailed information on the release from fuel of fission products and other radionuclides under LWR severe accident conditions, and on the chemical and physical forms and transport of the fission products. The main aim of this review is to bring together, in a systematic way, information on the conduct of the tests, on their successes and failures, and particularly on the information they generated on the chemical and physical behaviour of released fission products. By examining and analysing the data from all of the tests together, patterns of fission product behaviour may become apparent and insights may be gained, which would not be arrived at from individual test results. Moreover, important lessons may be learned, and useful guidance obtained, relating to the aims and conduct of future experimental programmes of fission product release from fuel and transport behaviour. The conclusions should be particularly relevant to the imminent Phebus-FP in-pile test series at Cadarache

  5. In-pile experiments and test facilities proposed for fast reactor safety

    International Nuclear Information System (INIS)

    Grolmes, M.A.; Avery, R.; Goldman, A.J.; Fauske, H.K.; Marchaterre, J.F.; Rose, D.; Wright, A.E.

    1976-01-01

    The role of in-pile experiments in support of the resolution of fast breeder reactor safety and licensing issues has been re-examined, with emphasis on key safety issues. Experiment needs have been related to the specific characteristics of these safety issues and to realistic requirements for additional test facility capabilities which can be achieved and utilized within the next ten years. It is found that those safety issues related to the energetics of core disruptive accidents have the largest impact on new facility requirements. However, utilization of existing facilities with modifications can provide for a continuing increase in experiment capability and experiment results on a timely bases. Emphasis has been placed upon maximum utilization of existing facilities and minimum requirements for new facilities. This evaluation has concluded that a new Safety Test Facility, STF, along with major modifications to the EBR II facility, improvement in TREAT capabilities, the existing Sodium Loop Safety Facility and corresponding Support Facilities provide the essential elements of the Safety Research Experiment Facilities (SAREF) required for resolution of key issues

  6. Evaluation on sweep gas pressure drop in fusion blanket mock-up for in-pile test

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Sagawa, Hisashi (Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment); Nagakura, Masaaki; Kanzawa, Toru.

    1993-03-01

    In the ITER/CDA (Conceptual Design Activity) of a tritium breeding blanket, Japan have proposed the pebble-typed blanket. The in-pile mock-up test will be preparing in JMTR (Japan Materials Testing Reactor) for Japanese engineering design with the pebble-typed blanket. Therefore, the He sweep gas pressure drop in the pebble bed was measured for the design of the mock-up used on in-pile test. From the results of this test, it was clear that the pressure drop was predicted on Kozeny- Carman's equation within +25 [approx] -60 %, and that the pressure drop was not affected by moisture concentration (< 100 ppm). (author).

  7. Evaluation on sweep gas pressure drop in fusion blanket mock-up for in-pile test

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Sagawa, Hisashi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Nagakura, Masaaki; Kanzawa, Toru

    1993-03-01

    In the ITER/CDA (Conceptual Design Activity) of a tritium breeding blanket, Japan have proposed the pebble-typed blanket. The in-pile mock-up test will be preparing in JMTR (Japan Materials Testing Reactor) for Japanese engineering design with the pebble-typed blanket. Therefore, the He sweep gas pressure drop in the pebble bed was measured for the design of the mock-up used on in-pile test. From the results of this test, it was clear that the pressure drop was predicted on Kozeny- Carman`s equation within +25 {approx} -60 %, and that the pressure drop was not affected by moisture concentration (< 100 ppm). (author).

  8. New Sensors for In-Pile Temperature Detection at the Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Daw, J.E.; Condie, K.G.; Wilkins, S. Curtis

    2009-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. As a user facility, the ATR is supporting new users from universities, laboratories, and industry, as they conduct basic and applied nuclear research and development to advance the nation's energy security needs. A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing measurements of key parameters during irradiation. This paper describes the strategy for determining what instrumentation is needed and the program for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available and under development for in-pile detection of temperature at various irradiation locations in the ATR.

  9. Proposed algorithm for determining the delta intercept of a thermocouple psychrometer curve

    International Nuclear Information System (INIS)

    Kurzmack, M.A.

    1993-01-01

    The USGS Hydrologic Investigations Program is currently developing instrumentation to study the unsaturated zone at Yucca Mountain in Nevada. Surface-based boreholes up to 2,500 feet in depth will be drilled, and then instrumented in order to define the water potential field within the unsaturated zone. Thermocouple psychrometers will be used to monitor the in-situ water potential. An algorithm is proposed for simply and efficiently reducing a six wire thermocouple psychrometer voltage output curve to a single value, the delta intercept. The algorithm identifies a plateau region in the psychrometer curve and extrapolates a linear regression back to the initial start of relaxation. When properly conditioned for the measurements being made, the algorithm results in reasonable results even with incomplete or noisy psychrometer curves over a 1 to 60 bar range

  10. Attachment of Free Filament Thermocouples for Temperature Measurements on Ceramic Matrix Composites

    Science.gov (United States)

    Lei, Jih-Fen; Cuy, Michael D.; Wnuk, Stephen P.

    1998-01-01

    At the NASA Lewis Research Center, a new installation technique utilizing convoluted wire thermocouples (TC's) was developed and proven to produce very good adhesion on CMC's, even in a burner rig environment. Because of their unique convoluted design, such TC's of various types and sizes adhere to flat or curved CMC specimens with no sign of delamination, open circuits, or interactions-even after testing in a Mach 0.3 burner rig to 1200 C (2200 F) for several thermal cycles and at several hours at high temperatures. Large differences in thermal expansion between metal thermocouples and low-expansion materials, such as CMC's, normally generate large stresses in the wires. These stresses cause straight wires to detach, but convoluted wires that are bonded with strips of coating allow bending in the unbonded portion to relieve these expansion stresses.

  11. A Simple Test to Evaluate the Calibration Stability and Accuracy of Infrared Thermocouple Sensors

    OpenAIRE

    Pinnock, Derek R.; Bugbee, Bruce

    2002-01-01

    Accurately measuring surface temperature is not difficult when the surface, the sensor, and air temperatures are similar, but it is challenging when the surface temperature is significantly different than air and sensor temperatures. We tested three Infrared Thermocouple sensors (IRT’s) that had been used for two years in a greenhouse environment. The importance of the correction for sensor body temperature was also examined.

  12. R and D advances in high temperature thermocouples for nuclear utilization in severe environment

    International Nuclear Information System (INIS)

    Schley, R.; Blanc, J.Y.

    1985-01-01

    Safety experiments for water reactors in Cadarache have made necessary a research program for developing special thermocopules for use in severe fuel damage conditions (superheated steam). Standard cladding thermocouples (type K, alumina insulated, zircaloy sheathed, O.D. 0.7 mm) must be replaced by others with W3Re versus W25Re legs, Ta sheath protected by a zircaloy outer sheath, and hafnia or thoria insulation. The zircaloy sheath will be sufficient to protect correctly tantalum. Fuel centerline thermocouples have W5Re versus W26Re or W3Re versus W25Re legs, hard-fired thoria insulation and rhenium CVD sheath (O.D. 1.1 mm). A protective ReSi/sub 2/ coating is applied. This protection withstands at least 1600 0 C, 45 minutes in steam. Tests are done concerning: (a) materials compatibilities in helium between 1400 0 C and 2000 0 C, (b) prototypes qualification (In Saclay or Grenoble), (c) determination of errors due to degradation of insulation resistance of thermocouples cables (with magnesia, hafnia, alumina), (d) Ir or Re protective coatings by CVD process, other coatings by ionic bombardment, etc...A completely new type of hot junction has been patented

  13. A Highly Thermostable In2O3/ITO Thin Film Thermocouple Prepared via Screen Printing for High Temperature Measurements

    Directory of Open Access Journals (Sweden)

    Yantao Liu

    2018-03-01

    Full Text Available An In2O3/ITO thin film thermocouple was prepared via screen printing. Glass additives were added to improve the sintering process and to increase the density of the In2O3/ITO films. The surface and cross-sectional images indicate that both the grain size and densification of the ITO and In2O3 films increased with the increase in annealing time. The thermoelectric voltage of the In2O3/ITO thermocouple was 53.5 mV at 1270 °C at the hot junction. The average Seebeck coefficient of the thermocouple was calculated as 44.5 μV/°C. The drift rate of the In2O3/ITO thermocouple was 5.44 °C/h at a measuring time of 10 h at 1270 °C.

  14. Thermocouples used in emission systems of internal combustion engines; Thermoelemente fuer den Einsatz in Abgassystemen von Verbrennungsmotoren

    Energy Technology Data Exchange (ETDEWEB)

    Augustin, Silke; Froehlich, Thomas; Mammen, Helge [Technische Univ. Illmenau (Germany). Inst. fuer Prozessmess- und Sensortechnik; Ament, Christoph; Guether, Thomas [Technische Univ. Illmenau (Germany). Inst. fuer Automatisierungs- und Systemtechnik

    2012-11-01

    Thermocouples used in exhaust systems of combustion engines are exposed to high temperature gradients and temperature leaps ({Delta}T > 900 K), high flow speeds and pressure. When constructing these thermocouples, a compromise is needed between the resulting high demands on the mechanical-thermal stability, accuracy and the fast response time demanded by the servo-control of the motors. Additionally, a numerical correction of the measured signal may contribute to an improved sensor dynamics. (orig.)

  15. A novel method for in-situ estimation of time constant for core temperature monitoring thermocouples of operating reactors

    International Nuclear Information System (INIS)

    Sylvia, J.I.; Chandar, S. Clement Ravi; Velusamy, K.

    2014-01-01

    Highlights: • Core temperature sensor was mathematically modeled. • Ramp signal generated during reactor operating condition is used. • Procedure and methodology has been demonstrated by applying it to FBTR. • Same technique will be implemented for all fast reactors. - Abstract: Core temperature monitoring system is an important component of reactor protection system in the current generation fast reactors. In this system, multiple thermocouples are housed inside a thermowell of fuel subassemblies. Response time of the thermocouple assembly forms an important input for safety analysis of fast reactor and hence frequent calibration/time constant estimation is essential. In fast reactors the central fuel subassembly is provided with bare fast response thermocouples to detect under cooling events in reactor and take proper safety action. On the other hand, thermocouples in thermowell are mainly used for blockage detection in individual fuel subassemblies. The time constant of thermocouples in thermowell can drift due to creep, vibration and thermal fatigue of the thermowell assembly. A novel method for in-situ estimation of time constant is proposed. This method uses the Safety Control Rod Accelerated Mechanism (SCRAM) or lowering of control Rod (LOR) signals of the reactor along with response of the central subassembly thermocouples as reference data. Validation of the procedure has been demonstrated by applying it to FBTR

  16. A preliminary study of factors affecting the calibration stability of the iridium versus iridium-40 percent rhodium thermocouple

    Science.gov (United States)

    Ahmed, Shaffiq; Germain, Edward F.; Daryabeigi, Kamran; Alderfer, David W.; Wright, Robert E.

    1987-01-01

    An iridium versus iridium-40% rhodium thermocouple was studied. Problems associated with the use of this thermocouple for high temperature applications (up to 2000 C) were investigated. The metallurgical studies included X-ray, macroscopic, resistance, and metallographic studies. The thermocouples in the as-received condition from the manufacturer revealed large amounts of internal stress caused by cold working during manufacturing. The thermocouples also contained a large amount of inhomogeneities and segregations. No phase transformations were observed in the alloy up to 1100 C. It was found that annealing the thermocouple at 1800 C for two hours, and then at 1400 C for 2 to 3 hours yielded a fine grain structure, relieving some of the strains, and making the wire more ductile. It was also found that the above annealing procedure stabilized the thermal emf behavior of the thermocouple for application below 1800 C (an improvement from + or - 1% to + or - 0.02% within the range of the test parameters used).

  17. Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

    Energy Technology Data Exchange (ETDEWEB)

    Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

    2014-09-15

    Highlights: • Gas as a coolant in Gen-IV reactors, history and development. • Main physical parameters comparison of gas coolants: carbon dioxide, helium, hydrogen with water. • Forced convection in turbulent pipe flow. • Gas cooled fast reactor concept comparisons to very high temperature reactor concept. • High temperature helium loop: concept, development, mechanism, design and constraints. - Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR). The VHTR concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behavior within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Center Rez Ltd. One of the topics analyzed in this article are also physical characteristic and benefits of gas

  18. Accounting for the inertia of the thermocouples' measurements by modelling of a NPP Kalinin-3 transient with the coupled system code ATHLET-BIPR-VVER

    International Nuclear Information System (INIS)

    Nikonov, S.; Velkov, K.

    2008-01-01

    The ATHLET-BIPR-VVER coupled system code is applied for performing of safety analysis for different WWER reactors. During the last years its validation matrix is continuously being enlarged. The measurements performed during the commissioning phase of NPP Kalinin Unit 3 for the transient 'Switching-off of one Main Circulation Pump at nominal power' are very well documented and have a variety of recorded integral and local thermo-hydraulic and neutron-physic parameters including the measurements' errors. This data is being used for further validation of the coupled code system ATHLET-BIPR-VVER. In the paper are discussed the problems and our solutions by the correct interpretation of the measured thermocouples' records at NPP Kalinin-3 and the comparison with the predicted results by the coupled thermal-hydraulic/neutron-kinetic code ATHLET-BIPR-VVER. Of primary importance by such comparisons is the correct accounting of the fluid mixing process that take place in the surrounding of the measuring sensors and also the consideration of the time delay (inertia term) of the measuring devices. On the bases of previous experience and many simulations of the defined transient a method is discussed and proposed to consider correctly the inertia term of the thermocouples' measurements. The new modelling is implemented in the coupled system code ATHLET-BIPR-VVER for further validation. (Author)

  19. Seismic analysis for shroud facility in-pile tube and saturated temperature capsules

    International Nuclear Information System (INIS)

    Iimura, Koichi; Yamaura, Takayuki; Ogawa, Mitsuhiro

    2009-07-01

    At Oarai Research and Development Center, Japan Atomic Energy Agency (JAEA), the plan of repairing and refurbishing Japan Materials Testing Reactor (JMTR) has progressed in order to restart JMTR operation in the fiscal 2011. As a part of effective use of JMTR, the neutron irradiation tests of LWR fuels and materials has been planned in order to study their soundness. By using Oarai Shroud Facility (OSF-1) and Fuel Irradiation Facility with the He-3 gas control system for power lamping test using Boiling Water Capsules (BOCA Irradiation Facility), the irradiation tests with power ramping will be carried out to study the soundness of fuel under LWR Transient condition. OSF-1 is the irradiation facility of shroud type that can insert and eject the capsule under reactor operation, and is composed of 'In-pile Tube', 'Cooling system' and 'Capsule exchange system'. BOCA Irradiation Facility is the facility which simulates irradiation environment of LWR, and is composed of 'Boiling water Capsule', 'Capsule control system' and 'Power control system by He-3'. By using Saturated temperature Capsules and the water environment control system, the material irradiation tests under the water chemistry condition of LWR will be carried out to clarify the mechanism of IASCC. In JMTR, these facilities are in service at the present. However, the detailed design for renewal or remodeling was carried out based on the new design condition in order to be correspondent to the irradiation test plan after restart JMTR operation. In this seismic analysis of the detailed design, each equipment classification and operating state were arranged with 'Japanese technical standards of the structure on nuclear facility for test research' and 'Technical guidelines for seismic design of nuclear power plants on current, and then, stress calculation and evaluation were carried out by FEM piping analysis code 'SAP' and structure analysis code 'ABAQUS'. About the stress of the seismic force, it was proven

  20. In-pile behavior of controlled beta-quenched fuel channels

    Energy Technology Data Exchange (ETDEWEB)

    Moeckel, Andreas; Pflaum, Wolfgang; Cremer, Ingo [AREVA NP GmbH, Erlangen (Germany); Zbib, Ali A. [AREVA NP Inc., Richland, WA (United States)

    2011-07-01

    of a German boiling water reactor (BWR) nuclear power plant in 2004. Since then, additional beta-quenched lead fuel channels have been placed in the core of some other European BWRs to broaden the in-pile experience with such channels in different nuclear power plants. (orig.)

  1. In-pile experiments on fuel rod behavior during a LOCA

    International Nuclear Information System (INIS)

    Karb, E.; Pruessmann, M.; Sepold, L.

    1980-05-01

    This report describes the results of the Test Series F, Tests F 1 through F 5, in the in-pile experimental program with single rods in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The research is part of the Nuclear Safety Project's (PNS) fuel behavior program. The main objective of the FR2-LOCA tests is to provide information about the effects of a nuclear environment on the mechanisms of fuel rod failure in the second heatup phase of a LOCA. The test rods have a heated length of 50 cm, and their radial dimensions are identical with those of a commercial German PWR. The main parameter of the FR2-LOCA test program is the burnup. The F tests were perfomed from Oct. 25, 1977 to Nov. 22, 1977. They were the first tests in this program to use pre-irradiated fuel rods. The nominal burnup of the test rods was 20 000 MWd/t. During the transient test, the test rods were subjected to rod powers between 36 and 41 W/cm and were pressurized with He to hot internal pressures between 46 and 83 bar. The test rods during the heatup phase at pressures of 56, 53, 42, 72 and 60 bar, respectively. The burst temperatures were determined to be 890, 893, 932, 835 and 880 0 C for test F 1 through F 5. The maximum total circumferential elongations amount to 59, 38, 27, 34 and 41%, respectively. The F tests revealed a fragmentation of the fuel after the irradiation (prior to the tests) and a disintegration of the fuel pellet column after the transient tests due to cladding ballooning. The post-test results indicated a significant reduction of the pellet stack length for all five test rods. The burst data of the F tests did not reveal any difference between tests with unirradiated fuel rods and the irradiated fuel rods of this test series. (orig./HP) [de

  2. Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

    Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

    2015-01-01

    Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system

  3. Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system.

  4. An Explicit Approach Toward Modeling Thermo-Coupled Deformation Behaviors of SMPs

    Directory of Open Access Journals (Sweden)

    Hao Li

    2017-03-01

    Full Text Available A new elastoplastic J 2 -flow models with thermal effects is proposed toward simulating thermo-coupled finite deformation behaviors of shape memory polymers. In this new model, an elastic potential evolving with development of plastic flow is incorporated to characterize the stress-softening effect at unloading and, moreover, thermo-induced plastic flow is introduced to represent the strain recovery effect at heating. It is shown that any given test data for both effects may be accurately simulated by means of direct and explicit procedures. Numerical examples for model predictions compare well with test data in literature.

  5. Improvement in the technology of thermocouples for the detection of high temperatures with a view to using them in irradiation safety tests in reactor

    International Nuclear Information System (INIS)

    Schley, R.; Liermann, J.; Aujollet, J.M.; Wilkins, S.C.

    1979-01-01

    The safety tests carried out under the CABRI and PHEBUS programmes have made it possible to improve the technology of W/Re thermocouples and their reliability in particularly hard operating conditions. An element of response is provided to the problem of W/Re thermocouple drift under neutron flux by defining the new thermocouple Mo 5% Nb/Nb 10% Mo which, because of the low capture cross section of thermoelectric elements, gives one reason to hope for a less significant drift of these thermocouples under neutron flux than that found with W/Re thermocouples. Finally, determining the surface temperature of fuel element cladding with the Mo/Zircaloy thermocouple may prove worthwhile providing the temperatures do not exceed 1300 0 C and the electric insulator is aluminium oxide which up to 1300 0 C does not appear to react with thermoelectric wires [fr

  6. Pre-test prediction and post-test analysis of PWR fuel rod ballooning in the MT-3 in-pile LOCA simulation experiment in the NRU reactor

    International Nuclear Information System (INIS)

    Donaldson, A.T.; Horwood, R.A.; Healey, T.

    1983-01-01

    The USNRC and the UKAEA have jointly funded a series of in-pile LOCA simulation experiments in the Canadian NRU reactor in order to secure further information on the thermal hydraulic and clad deformation response of PWR fuel rod bundles. Test MT-3 in the series was performed using reflood rate and rod internal pressure conditions specified by the UK nuclear industry. The parameters were selected to ensure the development of a near-isothermal clad temperature history during which zircaloy was required to balloon and rupture near the alpha-alpha/beta phase transition. Specification of the reflood rate conditions was assisted by the performance of a precursor test on an unpressurised rod bundle and by complementary application of appropriate thermal hydraulic analyses. Identification of the rod internal pressure needed to cause ballooning and rupture was achieved using a creep deformation model, BALLOON, in conjunction with the clad thermal history defined by the prior thermal hydraulic test. This paper presents the basis of the BALLOON analysis and describes its application in calculating the fill gas pressure for rods MT-3, their axial ballooning profile and the clad temperature at peak radial strain elevations. (author)

  7. Evaluation of the in-pile pressure data from instrumented fuel assemblies IFA-431 and IFA-432

    International Nuclear Information System (INIS)

    Bradley, E.R.; Cunningham, M.E.; Lanning, D.D.; Williford, R.E.

    1979-10-01

    This report includes results of the examination of the in-pile pressure data from instrumented test assemblies IFA-431 and 432. The pressure data have been used to estimate the fission gas release fraction as a function of fuel burnup. Included are comparisons of the estimated release functions and those predicted by three fission gas release models using the experimental temperature histories of the fuel rods. These comparisons show that fuel temperature is the primary factor in determining fission gas release and that burnup-enhanced fission gas release is not important in UO 2 fuels irradiated to 1700 GJ/kgU

  8. In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

    International Nuclear Information System (INIS)

    Gunnerson, F.S.; MacDonald, P.E.

    1980-01-01

    The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made

  9. Study of a device for the direct measurement of the fission gas pressure inside an in-pile fuel element

    International Nuclear Information System (INIS)

    Lavaud, B.; Uschanoff, S.

    1964-01-01

    The fission gas pressure inside a fuel element made of a refractory fuel constitutes an important limiting factor for the burn-up. Although it is possible to calculate approximately the volume of gas produced outside the fuel during its life-time; it is nevertheless very difficult to evaluate the pressure since the volume allowed to the fission gases, as well as their temperature are known only very approximately. This physical value, which is essential for the technologist, can only be known by direct in-pile measurement of the pressure. The report describes the equipment which has been developed for this test. (authors) [fr

  10. Experience from replacement and check of thermocouples during reconstruction of in-reactor temperature measurements at Bohunice V-1 units 1 and 2

    International Nuclear Information System (INIS)

    Slanina, M.; Stanc, J.

    2001-01-01

    Replacement of thermocouples in the protection tube blocks was a key phase of the reconstruction of in-reactor temperature measurements at Bohunice V-1 with regard to the success, reliability and impact on safety of unit operation. The replacement consisted of reliable and safe withdrawal of 216 old thermocouples, their disposal and installation of new thermocouples into dry channels. In the material presented, this phase of reconstruction is described in details, with focus on the evaluation of replacement quality and check activities carried out at the new installed thermocouples. (Authors)

  11. Calibration of the Dodewaard downcomer thermocouple cross-correlation flow-rate measurements

    Energy Technology Data Exchange (ETDEWEB)

    Stekelenburg, A J.C. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Hagen, T.H.J.J. van der [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Akker, H.E.A. van den [Technische Univ. Delft (Netherlands). Lab. voor Fysische Technologie

    1992-12-01

    The cross-correlation flow measurement technique, applied for measuring the coolant flow rate in a nuclear reactor, was calibrated with the use of numerical simulations of turbulent flow. The three-dimensional domain was collapsed into two dimensions. With a two-dimensional calculation of steady-state flow with transient thermal characteristics the response of thermocouples to a temperature variation was calculated. By cross-correlating the calculated thermocouple responses, the link between total flow rate and measured transit times was made. Three calibration points were taken in the range of 579 kg/s to 1477 kg/s. In this range, the product of the calculated transit time and the mass flow-rate is constant up to +3.5% and -2.4%. The reliability of the calibration was estimated at {+-}4.6%. The influence of the inlet boundary conditions, and the modelling of the flow in the upper part of the downcomer channel on the calibration result is shown to be small. A measured velocity profile effect was successfully predicted. (orig.).

  12. Thermal conductivity of Na3(U/sub 1-y/Pu/sub y/)O4: A preliminary in-pile determination

    International Nuclear Information System (INIS)

    Lee, M.J.; Lambert, J.D.B.; Ukai, S.; Odo, T.

    1987-01-01

    During Run-Beyond-Cladding-Breach (RBCB) operation in an oxide LMR, the performance of a breached fuel element is intimately associated with the formation of fuel-sodium reaction product (FSRP), Na 3 (U/sub 1-y/Pu/sub y/)O 4 . In-pile experiments coupled with destructive examinations of breached fuel have consistently revealed noticeable changes in fuel structure accompanying FSRP formation at the fuel surface. Previous analyses have also indicated a significant impact of FSRP on fuel centerline temperature. Successful modeling of breached fuel thermal behavior therefore requires a reasonably accurate knowledge of the thermal properties of the FSRP, especially its thermal conductivity. But laboratory investigations have been scarce and limited to the Na/UO 2 system because of the toxicity of plutonium and hygroscopicity of the FSRP. Hence, post-irradiation observations of fuel samples remain the most amenable way of deriving the thermal conductivity of the FSRP. Such work is a spin-off of the RBCB program in the Experimental Breeder Reactor-II (EBR-II), a program jointly sponsored by the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan

  13. Simulation of the in-pile test Phebus-FPT3 using ASTEC V2 and ATHLET-CD 2.1A

    Energy Technology Data Exchange (ETDEWEB)

    Kruse, Philipp; Koch, Marco K. [Bochum Univ. (Germany). Chair of Energy Systems and Energy Economics

    2011-07-01

    The Phebus-FPT programme, initiated in 1988 by the 'Institut de Radioprotection et de Surete Nucleaire' (IRSN) and the Joint Research Centre (JRC) of the European Commission (EC), was performed in the Phebus facility operated by 'Commissariat a'Energie Atomique' (CEA). The facility represent a 900 MWe Pressurized Water Reactor (PWR) scaled down by a factor 1:5000 which objective is to study fuel degradation and the subsequent release, transport and retention of fission products, structure, control rod and fuel materials, in case of a severe accident. The Phebus-FPT programme consists of integral in-pile tests, varying the fuel burn-up and geometry, the control rod nature, the thermal hydraulic conditions in the bundle and through the experimental circuit as well as in the containment. In primary, the integral experiments should outline a detailed description of the main phenomena of core degradation, fission product release and transport as well as radionuclide interactions. Due to that it is possible to analyse the physical and chemical processes due to a severe accident. With the ascertained data, an evaluation of the accident management measures could be made as well. A secondary aim of the Phebus tests was to enable model development and evaluation of severe accident codes such like ASTEC and ATHLET-CD. (orig.)

  14. In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions in the LVR-15 nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, Jan [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Entler, Slavomir, E-mail: slavomir.entler@cvrez.cz [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Vsolak, Rudolf; Klabik, Tomas [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Zlamal, Ondrej [CEZ, Duhova 2/1444, 140 53 Praha 4 (Czech Republic); Bellin, Boris; Zacchia, Francesco [Fusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2015-10-15

    Highlights: • Irradiated thermal fatigue testing of the ITER primary first wall mock-ups. • Cyclic heat flux of 0.5 MW/m{sup 2} in the neutron field of the nuclear reactor core. • 17,040 thermal cycles. • Radiation damage in the range of 0.41–1.17 dpa depending on the material. - Abstract: The TW3 in-pile rig enabled the thermal fatigue testing of ITER primary first wall mock-ups in the core of the nuclear reactor. This experiment investigated the neutron irradiation influence on the design performance under high heat flux testing. A thermal flux of 0.5 MW/m{sup 2} in the neutron field of the core of the LVR-15 nuclear reactor was applied. Within the scope of the tests with simultaneous neutron irradiation, the TW3 rig reached a record of 17,040 thermal cycles with the radiation damage in the range of 0.41–1.17 dpa depending on the material. Even after a high number of thermal cycles, while being irradiated by neutrons, no damage of the tested mock-ups was visually observed. Further testing and analysis will follow in the Forschungszentrum Juelich.

  15. Experience with W3Re/W25Re thermocouples in fuel pins of NS Otto Hahn's two cores

    International Nuclear Information System (INIS)

    Kolb, M.

    1976-01-01

    The paper first deals with the installation of 18 and 9 high-temperature sheathed thermocouples in fuel rods of the cores FDR-1 and FDR-2, respectively. The measured fuel rod centerline temperatures could be related to the local linear rod power at any given time by means of the densities of fission products with different half-lives obtained from fuel rod γ-scans. The fuel temperatures show then already an increase with the burn-up of the FDR-1 which becomes steeper when taking into account the decrease of the EMF measured at irradiated thermocouples taken from the fuel rods. Finally, the determination of effective thermocouple time constants and of fuel rod heat transfer time constants is demonstrated by utilizing the reactor noise to measure the transfer function between neutron flux and fuel temperature signal. (orig.) [de

  16. Apparatus for spot welding sheathed thermocouples to the inside of small-diameter tubes at precise locations

    International Nuclear Information System (INIS)

    Baucum, W.E.; Dial, R.E.

    1976-01-01

    Equipment and procedures used to spot weld tantalum- or stainless-steel-sheathed thermocouples to the inside diameter of Zircaloy tubing to meet the requirements of the Multirod Burst Test (MRBT) Program at ORNL are described. Spot welding and oxide cleaning tools were fabricated to remove the oxide coating on the Zircaloy tubing at local areas and spot weld four thermocouples separated circumferentially by 90 0 at any axial distribution desired. It was found necessary to apply a nickel coating to stainless-steel-sheathed thermocouples to obtain acceptable welds. The material and shape of the inner electrode and resistance between inner and outer electrodes were found to be critical parameters in obtaining acceptable welds

  17. Status of the EXOTIC-8 programme and first in-pile results for Li{sub 2}TiO{sub 3} pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Van der Laan, J G; Stijkel, M P [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Conrad, R

    1998-03-01

    After renewal of the Tritium Measuring Station the HFR is again fully operational for in-pile breeder irradiations. The EXOTIC-8 series has started with first three experiments on June 12, 1997. First in-pile results have been obtained for Li{sub 2}TiO{sub 3}-pebbles supplied by CEA: preliminary analyses indicate satisfactory in-pile behaviour with fast recovery from transient conditions. Five further experiments have been defined which implies that in the present planning EXOTIC-8 is filled completely up to Fall`98 and 2 of 4 positions are occupied up to Spring`99. P.I.E. results will be obtained from Spring`98 onwards. (J.P.N.)

  18. Boiling detection using signals of self-powered neutron detectors and thermocouples

    International Nuclear Information System (INIS)

    Kozma, R.

    1989-01-01

    A specially-equipped simulated fuel assembly has been placed into the core of the 2 MW research reactor of the IRI, Delft. In this paper the recent results concerning the detection of coolant boiling in the simulated fuel assembly are introduced. Applying the theory of boiling temperature noise, different stages of boiling, i.e. one-phase flow, subcooled boiling, volume boiling, were identified in the measurements using the low-frequency noise components of the thermocouple signals. It has been ascertained that neutron noise spectra remained unchanged when subcooled boiling appeared, and that they changed reasonably only when developed volume boiling took place in the channels. At certain neutron detector positions neutron spectra did not vary at all, although developed volume boiling occurred at a distance of 3-4 cm from these neutron detectors. This phenomenon was applied in studying the field-of-view of neutron detectors

  19. Preparation and Thermoelectric Characteristics of ITO/PtRh:PtRh Thin Film Thermocouple.

    Science.gov (United States)

    Zhao, Xiaohui; Wang, Hongmin; Zhao, Zixiang; Zhang, Wanli; Jiang, Hongchuan

    2017-12-15

    Thin film thermocouples (TFTCs) can provide more precise in situ temperature measurement for aerospace propulsion systems without disturbance of gas flow and surface temperature distribution of the hot components. ITO/PtRh:PtRh TFTC with multilayer structure was deposited on alumina ceramic substrate by magnetron sputtering. After annealing, the TFTC was statically calibrated for multiple cycles with temperature up to 1000 °C. The TFTC with excellent stability and repeatability was realized for the negligible variation of EMF in different calibration cycles. It is believed that owing to oxygen diffusion barriers by the oxidation of top PtRh layer and Schottky barriers formed at the grain boundaries of ITO, the variation of the carrier concentration of ITO film is minimized. Meanwhile, the life time of TFTC is more than 30 h in harsh environment. This makes ITO/PtRh:PtRh TFTC a promising candidate for precise surface temperature measurement of hot components of aeroengines.

  20. Operating Temperatures of a Sodium-Cooled Exhaust Valve as Measured by a Thermocouple

    Science.gov (United States)

    Sanders, J. C.; Wilsted, H. D.; Mulcahy, B. A.

    1943-01-01

    A thermocouple was installed in the crown of a sodium-cooled exhaust valve. The valve was then tested in an air-cooled engine cylinder and valve temperatures under various engine operating conditions were determined. A temperature of 1337 F was observed at a fuel-air ratio of 0.064, a brake mean effective pressure of 179 pounds per square inch, and an engine speed of 2000 rpm. Fuel-air ratio was found to have a large influence on valve temperature, but cooling-air pressure and variation in spark advance had little effect. An increase in engine power by change of speed or mean effective pressure increased the valve temperature. It was found that the temperature of the rear spark-plug bushing was not a satisfactory indication of the temperature of the exhaust valve.

  1. Low-noise audio amplifiers and preamplifier for use with intrinsic thermocouples

    International Nuclear Information System (INIS)

    Langner, G.C.; Sachs, R.D.; Stewart, F.L.

    1979-03-01

    Two simple, low-noise audio amplifiers and one low-noise preamplifier for use with intrinsic thermocouples were designed, built, and tested. The amplifiers and the preamplifier have different front end designs. One amplifier uses ultralow-noise operational amplifiers; the other amplifier uses a hybrid component. The preamplifier uses ultralow-noise discrete components. The amplifiers' equivalent noise inputs, at maximum gain, are 4.09 nV and 50 nV; the preamplifier's input is 4.05 μV. Their bandwidths are 15 600 Hz, 550 Hz, and 174 kHz, respectively. the amplifiers' equivalent noise inputs were measured from approx. 0 to 100 Hz, whereas the preamplifier's equivalent noise input was measured from approx. 0 to 174 kHz

  2. SMORN-1: thermoelectrically generated noise in sheathed thermocouples and in other low level instrumentation cables

    International Nuclear Information System (INIS)

    Mathieu, F.; Meier, R.; Soenen, M.; Delcon, M.; Nysten, C.

    Starting from the fact that thermoelectric emfs of thermocouples are generated in the thermal gradients and not at the hot junction, it is shown how thermoelectric heterogeneity in conjunction with natural and forced convection phenomena gives rise to unwanted noise called: ''thermoelectric noise'' in the technological sense. A distinction is made between four different types of noise--i.e. uncorrelated noise, correlated noise, spectral noise and thermoelectric noise in the physical sense--each of which has its own characteristics. The experimental results presented reveal that noise amplitudes may be quite embarrassing when dealing with problems of quantitative signal fluctuation analysis. It is however emphasized that thermoelectric noise may also convey useful information which, without noise, might be lost

  3. Boiling point measurement of a small amount of brake fluid by thermocouple and its application.

    Science.gov (United States)

    Mogami, Kazunari

    2002-09-01

    This study describes a new method for measuring the boiling point of a small amount of brake fluid using a thermocouple and a pear shaped flask. The boiling point of brake fluid was directly measured with an accuracy that was within approximately 3 C of that determined by the Japanese Industrial Standards method, even though the sample volume was only a few milliliters. The method was applied to measure the boiling points of brake fluid samples from automobiles. It was clear that the boiling points of brake fluid from some automobiles dropped to approximately 140 C from about 230 C, and that one of the samples from the wheel cylinder was approximately 45 C lower than brake fluid from the reserve tank. It is essential to take samples from the wheel cylinder, as this is most easily subjected to heating.

  4. Specific features of thermocouple calorimeter application for measurements of pulsed X-ray emission from plasma

    International Nuclear Information System (INIS)

    Gavrilov, V. V.; Fasakhov, I. K.

    2012-01-01

    It is shown that the accuracy of time-integrated measurements of pulsed X-ray emission from hot plasma with calibrated thermocouple calorimeters is mainly determined by two factors. The first and the most important factor is heating of the filter by the absorbed X-rays; as a result, the calorimeter measures the thermal radiation of the filter, which causes appreciable distortion of the temporal profile and amplitude of the recorded signal. The second factor is the dependence of the effective depth of X-ray absorption in the dielectric that covers the entrance window of the calorimeter on the energy of X-ray photons, i.e., on the recorded radiation spectrum. The results of model calculations of the calorimeter signal are compared with the experimental data.

  5. Co-C and Pd-C Eutectic Fixed Points for Radiation Thermometry and Thermocouple Thermometry

    Science.gov (United States)

    Wang, L.

    2017-12-01

    Two Co-C and Pd-C eutectic fixed point cells for both radiation thermometry and thermocouple thermometry were constructed at NMC. This paper describes details of the cell design, materials used, and fabrication of the cells. The melting curves of the Co-C and Pd-C cells were measured with a reference radiation thermometer realized in both a single-zone furnace and a three-zone furnace in order to investigate furnace effect. The transition temperatures in terms of ITS-90 were determined to be 1324.18 {°}C and 1491.61 {°}C with the corresponding combined standard uncertainty of 0.44 {°}C and 0.31 {°}C for Co-C and Pd-C, respectively, taking into account of the differences of two different types of furnaces used. The determined ITS-90 temperatures are also compared with that of INRIM cells obtained using the same reference radiation thermometer and the same furnaces with the same settings during a previous bilateral comparison exercise (Battuello et al. in Int J Thermophys 35:535-546, 2014). The agreements are within k=1 uncertainty for Co-C cell and k = 2 uncertainty for Pd-C cell. Shapes of the plateaus of NMC cells and INRIM cells are compared too and furnace effects are analyzed as well. The melting curves of the Co-C and Pd-C cells realized in the single-zone furnace are also measured by a Pt/Pd thermocouple, and the preliminary results are presented as well.

  6. Investigating Microbial Habitats in Hydrothermal Chimneys using Ti-Thermocouple Arrays: Microbial Diversity

    Science.gov (United States)

    Pagé, A.; Tivey, M. K.; Stakes, D. S.; Bradley, A. M.; Seewald, J. S.; Wheat, C. G.; Reysenbach, A.

    2004-12-01

    In order to examine the changes that occur in the microbial community composition as a deep-sea hydrothermal vent chimney develops, we deployed Ti-thermocouple arrays over high temperature vents at two active sites of the Guaymas Basin Southern Trough. Chimney material that precipitated around the arrays was recovered after 4 and 72 days. Chimney material that precipitated prior to deployment of the arrays was also recovered at one of the sites (Busted Shroom). Culture-independent analysis based on the small subunit rRNA sequence (cloning and DGGE) was used to determine the microbial diversity associated with subsamples of each chimney. The original Busted Shroom chimney (BSO) was dominated by members of the Crenarchaeota Marine Group I, a group of cosmopolitan marine Archaea, ɛ -Proteobacteria, and γ -Proteobacteria, two divisions of Bacteria that are common to deep-sea vents. The 4 days old Busted Shroom chimney (BSD1) was dominated by members of the Methanocaldococcaceae, hyperthermophilic methanogens, and the 72 days old chimney (BSD2) by members of the Methanosarcinaceae, mesophilic and thermophilic methanogens. At the second site, Toadstool, the 72 days old chimney material that had precipitated around the array (TS) revealed the dominance of sequences from uncultured marine Archaea, the DHVE group I and II, and from the ɛ -Proteobacteria. Additionally, sequences belonging to the Methanocaldococcaceae and Desulfurococcaceae were recovered next to thermocouples that were at temperatures of 109° C (at Busted Shroom) and 116° C (at Toadstool), respectively. These temperatures are higher than the upper limit for growth of cultured representatives from each family.

  7. Thin film thermocouples for in situ membrane electrode assembly temperature measurements in a polybenzimidazole-based high temperature proton exchange membrane unit cell

    DEFF Research Database (Denmark)

    Ali, Syed Talat; Lebæk, Jesper; Nielsen, Lars Pleth

    2010-01-01

    m thick layer of TFTCs on 75 mu m thick Kapton foil. The Kapton foil was treated with in situ argon plasma etching to improve the adhesion between TFTCs and the Kapton substrate. The TFTCs were covered with a 7 mu m liquid Kapton layer using spin coating technique to protect them from environmental......This paper presents Type-T thin film thermocouples (TFTCs) fabricated on Kapton (polyimide) substrate for measuring the internal temperature of PBI(polybenzimidazole)-based high temperature proton exchange membrane fuel cell (HT-PEMFC). Magnetron sputtering technique was employed to deposit a 2 mu...... degradation. This Kapton foil with deposited TFTCs was used as sealing inside a PBI (polybenzimidazole)-based single cell test rig, which enabled measurements of in situ temperature variations of the working fuel cell MEA. The performance of the TFTCs was promising with minimal interference to the operation...

  8. In-pile test of Li 2TiO 3 pebble bed with neutron pulse operation

    Science.gov (United States)

    Tsuchiya, K.; Nakamichi, M.; Kikukawa, A.; Nagao, Y.; Enoeda, M.; Osaki, T.; Ioki, K.; Kawamura, H.

    2002-12-01

    Lithium titanate (Li 2TiO 3) is one of the candidate materials as tritium breeder in the breeding blanket of fusion reactors, and it is necessary to show the tritium release behavior of Li 2TiO 3 pebble beds. Therefore, a blanket in-pile mockup was developed and in situ tritium release experiments with the Li 2TiO 3 pebble bed were carried out in the Japan Materials Testing Reactor. In this study, the relationship between tritium release behavior from Li 2TiO 3 pebble beds and effects of various parameters were evaluated. The ( R/ G) ratio of tritium release ( R) and tritium generation ( G) was saturated when the temperature at the outside edge of the Li 2TiO 3 pebble bed became 300 °C. The tritium release amount increased cycle by cycle and saturated after about 20 pulse operations.

  9. IR1 flow tube and In-Pile Test Section Pressure drop test for the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. H.; Park, K. N.; Chi, D. Y.; Sim, B. S.; Park, S. K.; Lee, J. M.; Lee, C. Y.; Kim, H. N

    2006-02-15

    The in-pile Section (IPS) of 3-pin Fuel Test Loop(FTL) shall be installed in the vertical hole call IR1 of HANARO reactor core. In order to verify the pressure drop and flow rate both the inside region of IPS at the annular region between IPS and IR1 flow tube, a pressure drop was measured by varing the flow rate on both regions. The measured pressure drop in the annular region is 209kpa at 14.9kg/s which meets the limiting condition of operation of 200kpa. The measured pressure drop in side the IPS becomes 260.25kpa which is lower than the designed value of 306.65kpa. As the pressure drop is lower than the design value, it is quite conservative from the safety and operating point of view.

  10. Annealing tests of in-pile irradiated oxide coated U-Mo/Al-Si dispersed nuclear fuel

    Science.gov (United States)

    Zweifel, T.; Valot, Ch.; Pontillon, Y.; Lamontagne, J.; Vermersch, A.; Barrallier, L.; Blay, T.; Petry, W.; Palancher, H.

    2014-09-01

    U-Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800 °C of in-pile irradiated U-Mo/Al-Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500 °C and 670 °C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500 °C FGs are released from IDL/matrix interfaces. The second peak at 670 °C representing the main release of FGs originates from the interaction between U-Mo and matrix in the vicinity of the cladding.

  11. Alpha radiation and in-pile annealing effects on the fracture properties of a sintered alumino borosilicate glass

    International Nuclear Information System (INIS)

    Bevilacqua, Arturo M.; Prado, Miguel O.; Messi de Bernasconi, Norma B.; Heredia, Arturo D.; Sanfilippo, Miguel

    1999-01-01

    The alpha radiation and the in-pile during irradiation effects on the hardness, the crack nucleation and the fracture toughness of the German alumino borosilicate glass SG7 were investigated by using the Vickers indentation. Cold pressed and sintered samples were irradiated with thermal neutrons, in the Argentine nuclear reactors RA-3 and RA-6, to produce alpha particles in the whole volume of the glass by means of the (n, alpha)-reaction with B-10. The Vickers hardness, the crack nucleation, as 50 percent fracture probability load, plotted as the Weibull's fracture probability distribution function and the fracture toughness, as critical stress intensity factor K Ic , were correlated to the four cumulative disintegration values. It was ascertained that: a) the Vickers hardness decreases from 5.6 GPa for the non-irradiated sample up to 4.7 GPa for the sample irradiated 70 h at the lower neutron flux (4.0 x 10 - sup 18 - alpha disintegration per cm - sup 3 -), b) the 50 % fracture probability load increases from 1.4 N for the non-irradiated sample up to 4.7 g for the sample irradiated 22 h at the higher flux (6.8 x 10 - sup 18 - alpha disintegration per cm - sup 3 -), and c) the stress intensity factor increases from 0.80 MPa.m - sup 1/2 - for the non irradiated sample up to 0.86 MPa.m - sup 1/2 - for the sample mentioned in b). The in-pile annealing was analyzed by comparing the crack nucleation after irradiation with data obtained by heavy ion irradiation followed by thermal annealing. Results for the SG7 glass were compared to those for soda-lime and borosilicate glasses. (author)

  12. Measurement data of cesium 137 yields in primary coolant of an in-pile water loop in fission products release experiment

    International Nuclear Information System (INIS)

    Ishiwatari, Nasumi; Nagai, Hitoshi; Takeda, Tsuneo

    1979-03-01

    Series of fuel rods (UO 2 pellets sheathed with stainless steel) having an artificial pinhole were irradiated in the in-pile test section of water loop JMTR OWL-1. Presented are the results of measurements of cesium 137 yields in primary coolant of OWL-1 from 1975 to 1978. (author)

  13. Measurement of the in-pile core temperature of an EL-4 pencil element, first charge (can of type-347 stainless steel, 0.4 mm thick, UO{sub 2} fuel, 11 mm diameter). Determination of the apparent thermal conductivity integral of in-pile UO{sub 2}; Mesure de la temperature a coeur en pile d'un crayon EL-4 1er jeu (gaine acier inoxydable, nuance 347 - epaisseur 0,4 mm - combustible UO{sub 2} - diametre 11 mm). Determination de l'integrale de conductibilite thermique apparente de l'UO{sub 2} en pile

    Energy Technology Data Exchange (ETDEWEB)

    Lavaud, B; Ringot, C; Vignesoult, N [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1966-11-01

    The core temperature of a pencil fuel element depends on the thermal conductivity of the UO{sub 2}, and on the UO{sub 2}-can contact. This temperature may be known accurately only if in-pile tests using the actual geometry are carried out. The test described concerns the measurement of the core- temperature of an EL-4 fuel element, first charge, having a stainless steel can. This temperature is measured at the center of the in-pile pencil element using a high-temperature thermocouple (W-Re with Ta sheath). The element is subjected to operating conditions similar to those of EL-4, both for the specific power and the can temperature and for the pressure acting on the can. The specific power is obtained in the EL-3 reactor using a slightly higher enrichment for the UO{sub 2} than that planned for EL-4. The required can temperature and pressure are obtained using a Zircaloy-2 irradiation container filled with NaK, adapted for use in the EL-3 reactor. The core temperatures of the UO{sub 2}, and that of the can surface are measured. The power is calculated from the heat exchanges in the container calibrated in the laboratory. The temperature drop at the UO{sub 2}-can interface is deduced from laboratory measurements carried out under comparable heat flux conditions, and in a gas atmosphere corresponding to the beginning of the life-time of the fuel element. It is possible to draw an integral conductivity curve. It is also possible to check the temperature distribution in the oxide, as deduced from the thermal conductivity integral, by micro-graphic examination of the oxide structure. (authors) [French] La temperature a coeur d'un crayon combustible est fonction de la conductibilite thermique de l'UO{sub 2}, mais aussi du contact UO{sub 2}-gaine. Les essais de mesure en geometrie reelle en pile sont les seuls qui permettent d'avoir une connaissance exacte de cette valeur. L'essai dont il est question dans ce rapport a trait a la mesure de la temperature a coeur d

  14. Thermocouple and Infrared Sensor-Based Measurement of Temperature Distribution in Metal Cutting

    Directory of Open Access Journals (Sweden)

    Abdil Kus

    2015-01-01

    Full Text Available In metal cutting, the magnitude of the temperature at the tool-chip interface is a function of the cutting parameters. This temperature directly affects production; therefore, increased research on the role of cutting temperatures can lead to improved machining operations. In this study, tool temperature was estimated by simultaneous temperature measurement employing both a K-type thermocouple and an infrared radiation (IR pyrometer to measure the tool-chip interface temperature. Due to the complexity of the machining processes, the integration of different measuring techniques was necessary in order to obtain consistent temperature data. The thermal analysis results were compared via the ANSYS finite element method. Experiments were carried out in dry machining using workpiece material of AISI 4140 alloy steel that was heat treated by an induction process to a hardness of 50 HRC. A PVD TiAlN-TiN-coated WNVG 080404-IC907 carbide insert was used during the turning process. The results showed that with increasing cutting speed, feed rate and depth of cut, the tool temperature increased; the cutting speed was found to be the most effective parameter in assessing the temperature rise. The heat distribution of the cutting tool, tool-chip interface and workpiece provided effective and useful data for the optimization of selected cutting parameters during orthogonal machining.

  15. Thermocouple and infrared sensor-based measurement of temperature distribution in metal cutting.

    Science.gov (United States)

    Kus, Abdil; Isik, Yahya; Cakir, M Cemal; Coşkun, Salih; Özdemir, Kadir

    2015-01-12

    In metal cutting, the magnitude of the temperature at the tool-chip interface is a function of the cutting parameters. This temperature directly affects production; therefore, increased research on the role of cutting temperatures can lead to improved machining operations. In this study, tool temperature was estimated by simultaneous temperature measurement employing both a K-type thermocouple and an infrared radiation (IR) pyrometer to measure the tool-chip interface temperature. Due to the complexity of the machining processes, the integration of different measuring techniques was necessary in order to obtain consistent temperature data. The thermal analysis results were compared via the ANSYS finite element method. Experiments were carried out in dry machining using workpiece material of AISI 4140 alloy steel that was heat treated by an induction process to a hardness of 50 HRC. A PVD TiAlN-TiN-coated WNVG 080404-IC907 carbide insert was used during the turning process. The results showed that with increasing cutting speed, feed rate and depth of cut, the tool temperature increased; the cutting speed was found to be the most effective parameter in assessing the temperature rise. The heat distribution of the cutting tool, tool-chip interface and workpiece provided effective and useful data for the optimization of selected cutting parameters during orthogonal machining.

  16. Thermocouple and Infrared Sensor-Based Measurement of Temperature Distribution in Metal Cutting

    Science.gov (United States)

    Kus, Abdil; Isik, Yahya; Cakir, M. Cemal; Coşkun, Salih; Özdemir, Kadir

    2015-01-01

    In metal cutting, the magnitude of the temperature at the tool-chip interface is a function of the cutting parameters. This temperature directly affects production; therefore, increased research on the role of cutting temperatures can lead to improved machining operations. In this study, tool temperature was estimated by simultaneous temperature measurement employing both a K-type thermocouple and an infrared radiation (IR) pyrometer to measure the tool-chip interface temperature. Due to the complexity of the machining processes, the integration of different measuring techniques was necessary in order to obtain consistent temperature data. The thermal analysis results were compared via the ANSYS finite element method. Experiments were carried out in dry machining using workpiece material of AISI 4140 alloy steel that was heat treated by an induction process to a hardness of 50 HRC. A PVD TiAlN-TiN-coated WNVG 080404-IC907 carbide insert was used during the turning process. The results showed that with increasing cutting speed, feed rate and depth of cut, the tool temperature increased; the cutting speed was found to be the most effective parameter in assessing the temperature rise. The heat distribution of the cutting tool, tool-chip interface and workpiece provided effective and useful data for the optimization of selected cutting parameters during orthogonal machining. PMID:25587976

  17. External attachment of titanium sheathed thermocouples to zirconium nuclear fuel rods for the loss-of-fluid-test (LOFT) Reactor

    International Nuclear Information System (INIS)

    Welty, R.K.

    1980-01-01

    A welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods has been developed. A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. Irradiation tests showed no degradation of thermocouples or weld structure. Fast thermal cycle and heater rod blowdown reflood tests were made to subject the weldments to high temperatures, high pressure steam, and fast water quench cycles. From the behavior of these tests, it was concluded that the attachment welds would survive a series of reactor safety tests. 2 refs

  18. New fixed-point mini-cell to investigate thermocouple drift in a high-temperature environment under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Laurie, M.; Vlahovic, L.; Rondinella, V.V. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe, (Germany); Sadli, M.; Failleau, G. [Laboratoire Commun de Metrologie, LNE-Cnam, Saint-Denis, (France); Fuetterer, M.; Lapetite, J.M. [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten, (Netherlands); Fourrez, S. [Thermocoax, 8 rue du pre neuf, F-61100 St Georges des Groseillers, (France)

    2015-07-01

    Temperature measurements in the nuclear field require a high degree of reliability and accuracy. Despite their sheathed form, thermocouples subjected to nuclear radiations undergo changes due to radiation damage and transmutation that lead to significant EMF drift during long-term fuel irradiation experiment. For the purpose of a High Temperature Reactor fuel irradiation to take place in the High Flux Reactor Petten, a dedicated fixed-point cell was jointly developed by LNE-Cnam and JRC-IET. The developed cell to be housed in the irradiation rig was tailor made to quantify the thermocouple drift during the irradiation (about two year duration) and withstand high temperature (in the range 950 deg. C - 1100 deg. C) in the presence of contaminated helium in a graphite environment. Considering the different levels of temperature achieved in the irradiation facility and the large palette of thermocouple types aimed at surveying the HTR fuel pebble during the qualification test both copper (1084.62 deg. C) and gold (1064.18 deg. C) fixed-point materials were considered. The aim of this paper is to first describe the fixed-point mini-cell designed to be embedded in the reactor rig and to discuss the preliminary results achieved during some out of pile tests as much as some robustness tests representative of the reactor scram scenarios. (authors)

  19. Use of indexed sensitivity factors in the analysis of nickel and iron based alloys: study of the decalibration of sheathed Chromel/Alumel thermocouples

    International Nuclear Information System (INIS)

    Christie, W.H.

    1978-01-01

    Sheathed Chromel versus Alumel thermocouples decalibrate when exposed to temperatures in excess of 1100 0 C. Thermocouples sheathed in Inconel-600 and type 304 stainless steel were studied in this work. Quantified SIMS data showed that the observed decalibrations were due to significant alterations that took place in the Chromel and Alumel thermoelements. The amount of alteration was different for each thermocouple and was influenced by the particular sheath material used in the thermocouple construction. Relative sensitivity factors, indexed by a matrix ion species ratio, were used to quantify SIMS data for three nickel-based alloys, Chromel, Alumel, and Inconel-600, and an iron-based alloy, type 304 stainless steel. Oxygen pressure >2 x 10 -6 torr in the sputtering region gave enhanced sensitivity and superior quantitative results as compared to data obtained at instrumental residual pressure

  20. In-Pile Qualification of the Fast-Neutron-Detection-System

    Directory of Open Access Journals (Sweden)

    Fourmentel D.

    2018-01-01

    FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCK•CEN in Belgium. Then a second test was recently completed at ISIS reactor operated by CEA in France. FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%.

  1. In-pile test of Li{sub 2}TiO{sub 3} pebble bed with neutron pulse operation

    Energy Technology Data Exchange (ETDEWEB)

    Tsuchiya, K. E-mail: tsuchiya@oarai.jaeri.go.jp; Nakamichi, M.; Kikukawa, A.; Nagao, Y.; Enoeda, M.; Osaki, T.; Ioki, K.; Kawamura, H

    2002-12-01

    Lithium titanate (Li{sub 2}TiO{sub 3}) is one of the candidate materials as tritium breeder in the breeding blanket of fusion reactors, and it is necessary to show the tritium release behavior of Li{sub 2}TiO{sub 3} pebble beds. Therefore, a blanket in-pile mockup was developed and in situ tritium release experiments with the Li{sub 2}TiO{sub 3} pebble bed were carried out in the Japan Materials Testing Reactor. In this study, the relationship between tritium release behavior from Li{sub 2}TiO{sub 3} pebble beds and effects of various parameters were evaluated. The (R/G) ratio of tritium release (R) and tritium generation (G) was saturated when the temperature at the outside edge of the Li{sub 2}TiO{sub 3} pebble bed became 300 deg. C. The tritium release amount increased cycle by cycle and saturated after about 20 pulse operations.

  2. Some aspects of in-pile swelling of fissile materials, 1. part: non-alloyed α uranium

    International Nuclear Information System (INIS)

    Mikailoff, H.

    1964-01-01

    An examination has been carried out of non-alloyed uranium samples, having various structural states, cold-worked and recrystallized, as-cast and β-treated, and irradiated at temperatures of between 450 and 600 C and with burn-ups from 1300 to 5500 MW days/metric ton. These samples swelled because of precipitation of the fission gases the porosity thus produced has a morphology depending mainly on the type of deformation to which the metal has been subjected and which is due to in-pile growth. The most homogeneous distribution of pores, and thus that leading to the minimum swelling, is only observed in the material having a marked [010] texture in which the growth and perhaps the thermal cycling introduce little or no strain. For other materials the deformation /swelling association causes a more rapid destruction of the samples either by cracking when the deformation is due to twinning, or by pronounced swelling localized in the bands when deformation is due to slipping. Finally the fission-gas precipitation considerably facilitates, above 500 C, the germination and growth of the intergranular cracks which can then develop at low stresses. (author) [fr

  3. Analysis of ULOF accident in Monju reflecting the knowledge from CABRI in-pile experiments and others

    International Nuclear Information System (INIS)

    Sato, Ikken; Tobita, Yoshiharu; Suzuki, Tohru; Kawada, Ken-ichi; Fukano, Yoshitaka; Fujita, Satoshi; Kamiyama, Kenji; Ishikawa, Makoto; Nonaka, Nobuyuki; Usami, Shin

    2007-05-01

    In the safety evaluation in the original licensing procedure of the prototype FBR Monju, mechanical energy release during an unprotected loss-of-flow (ULOF) event, one of technically inconceivable events postulated beyond design basis, was evaluated and shown that the radiological consequence is suitably limited. Since the sodium leakage accident in the secondary heat-transport system, the Monju plant has been kept under the stand-by condition for more than ten years and the composition of fuel in the current reactor core has changed due to decay of a plutonium isotope into americium. This change in fuel composition affects the neutronic characteristics of the core, requiring assessment of its effect on safety. In this study, event sequences of ULOF were analyzed to assess the effect of the change in neutronic characteristics on the mechanical energy release during ULOF. A significant advance has been made in the safety analysis codes reflecting the knowledge obtained through extensive safety research programs in the last decades, notably the CABRI in-pile experiments. The present results with the advanced analysis codes showed that the mechanical energy release in the current Monju core with the changed neutronic characteristics would not exceed the value evaluated formerly. (author)

  4. In-pile test of tritium release from tritium breeding materials (VOM-21H experiment)

    International Nuclear Information System (INIS)

    Kurasawa, Toshimasa; Takeshita, Hidefumi; Watanabe, Hitoshi; Yoshida, Hiroshi.

    1986-10-01

    Material development and blanket design of lithium-based ceramics such as lithium oxide, lithium aluminate, lithium silicate and lithium zirconate have been performed in Japan, United State of America and Europian Communities. Lithium oxide is a most attractive candidate for tritium breeding materials because of its high lithium density, high thermal conductivity and good tritium release performance. This work has been done to clarify the characteristics of tritium release and recovery from Li 2 O by means of in-situ tritium release measurement. The effects of temperature and sweep gas composition on the tritium release were investigated in this VOM-21H Experiment. Good measurement of tritium release was achieved but there were uncertainties in reproduciblity of data. The experimental results show that the role of surface adsorption/desorption makes a significant contribution to the tritium release and tritium inventory. Also, it is necessary to define the rate limiting process either diffusion or surface adsorption/desorption. (author)

  5. The influence of lead temperature on the accuracy of various stainless-steel sheathed, mineral-inulated nickel-chromium/nickel aluminium thermocouples

    International Nuclear Information System (INIS)

    Burnett, P.; Burns, J.S.

    1977-10-01

    Samples of three types of stainless steel sheathed MI thermocouples, such as are currently used in fire and furnace tests of transport flasks, have been subjected to high lead temperatures whilst the thermojunctions were kept at a constant low temperature. Both the lead temperature and the length of lead at temperature have been varied. As the lead temperature rises from ambient to a selected value, the emf output from the thermocouple initially decreases and then increases, taking up a final value dependent on the particular conditions. Below a threshold lead temperature, no significant steady state error occurs and the negative transient is generally negligible. Each thermocouple has its own threshold temperature, the lowest found being about 600 0 C, although the average lies at about 750 0 C. Above the threshold lead temperature, the thermal emf can be in error by the equivalent of more than 100 0 C, the highest error found being nearly 230 0 C at a temperature 250 0 C above threshold. The same thermocouple showed a negative transient of 13 0 C 3 minutes after start of heating to 890 0 C. It is probable that the steady state error arises because of the degradation of the thermocouple mineral insulation at elevated temperatures and recommendations are made on the use of such thermocouples in fire and furnace tests. The cause of the initial negative transient error has not been identified, but ways of minimising any resultant errors are suggested. (author)

  6. Summarizing evaluation of the results of in-pile experiments for the transient fission gas release under accidental conditions of fast breeders

    International Nuclear Information System (INIS)

    Fischer, E.A.; Vaeth, L.

    1989-04-01

    The transient fission gas behaviour and the fission gas induced fuel motion were studied in in-pile experiments in different countries, under conditions typical for hypothetical accidents. This report summarizes first the different experiment series and the main results. Then, a comparative evaluation is given, which provides a basis for the choice of the fission gas parameters in the accident code SAS3D

  7. Thermal annealing behaviour of sulphur-35 produced in pile-irradiated mixed crystals AlCl/sub 3/-FeCl/sub 3/

    Energy Technology Data Exchange (ETDEWEB)

    Dyakovich, V; Todorovski, D S; Kostadinova, Z D [Sofia Univ. (Bulgaria). Khimicheski Fakultet

    1983-12-19

    The regression analysis of the experimental results on the thermal annealing behaviour of /sup 35/S produced in pile-irradiated mixed crystals AlCl/sub 3/-FeCl/sub 3/ confirms some suppositions made in a previous paper. The chemical state of /sup 35/S is defined by the target prehistory and the iron concentration. The influence of Fe/sup 3 +/ can be observed indirectly through its influence on the defect structure formed.

  8. Long Hole Film Cooling Dataset for CFD Development . Part 1; Infrared Thermography and Thermocouple Surveys

    Science.gov (United States)

    Shyam, Vikram; Thurman, Douglas; Poinsatte, Phillip; Ameri, Ali; Eichele, Peter; Knight, James

    2013-01-01

    An experiment investigating flow and heat transfer of long (length to diameter ratio of 18) cylindrical film cooling holes has been completed. In this paper, the thermal field in the flow and on the surface of the film cooled flat plate is presented for nominal freestream turbulence intensities of 1.5 and 8 percent. The holes are inclined at 30deg above the downstream direction, injecting chilled air of density ratio 1.0 onto the surface of a flat plate. The diameter of the hole is 0.75 in. (0.01905 m) with center to center spacing (pitch) of 3 hole diameters. Coolant was injected into the mainstream flow at nominal blowing ratios of 0.5, 1.0, 1.5, and 2.0. The Reynolds number of the freestream was approximately 11,000 based on hole diameter. Thermocouple surveys were used to characterize the thermal field. Infrared thermography was used to determine the adiabatic film effectiveness on the plate. Hotwire anemometry was used to provide flowfield physics and turbulence measurements. The results are compared to existing data in the literature. The aim of this work is to produce a benchmark dataset for Computational Fluid Dynamics (CFD) development to eliminate the effects of hole length to diameter ratio and to improve resolution in the near-hole region. In this report, a Time-Filtered Navier Stokes (TFNS), also known as Partially Resolved Navier Stokes (PRNS), method that was implemented in the Glenn-HT code is used to model coolant-mainstream interaction. This method is a high fidelity unsteady method that aims to represent large scale flow features and mixing more accurately.

  9. Thermocouple psychrometer measurements of in situ water potential changes in heated welded tuff

    International Nuclear Information System (INIS)

    Mao, Nai-hsien; Wang, H.F.

    1991-05-01

    Ten thermocouple psychrometers (TCPs) to measure water potential (WP) were installed in three holes in G-Tunnel at the Nevada Test Site as part of the Prototype Engineered Barrier System Field Tests. These integrated tests measured several parameters as a function of location and time within a few meters of a heater emplaced in welded tuff. The primary goal of the TCP experiment was to find out whether the combination of laboratory calibration and field use of the TCP can provide useful data for determining the change of moisture condition in the field. We calibrated the TCPs in NaCl solutions up to 80 degree C(176 degree F) in the laboratory. In two holes, we used rubber sleeves and packers to house TCPs, and in the third hole, we used foam. All three holes were grouted behind the TCP assemblages. Field results of the heater test showed that small temperature gradients were present for all measurements. Nevertheless, the WP calibration made the necessary correction for the nonisothermal condition. A drying and re-wetting cycle peaked at about day 140 with a WP of -65 bar in borehole P3, located below the heater. A similar cycle but reduced in scale was found at about day 175 with a WP of -45 bar in borehole P2, above the heater. This difference in drying behavior above and below the heater was also observed from neutron data and was explained as a gravity effect. As temperatures increased, the evaporation rate of pore water increased, In unfractured rock, the gas-phase flow was primarily outward. Water condensed above the heater would drain back to keep the boiling region wet, but water condensed below the heater would drain away from the boiling region. This conceptual model explained both the time and magnitude differences for data from holes above and below the heater. 7 refs., 14 figs., 2 tabs

  10. Verification of tritium production evaluation procedure using Monte Carlo code MCNP for in-pile test of fusion blanket with JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Y. E-mail: nagao@jmtr.oarai.jaeri.go.jp; Nakamichi, K.; Tsuchiya, M.; Ishitsuka, E.; Kawamura, H

    2000-11-01

    To evaluate exactly the total amount of tritium production in tritium breeding materials during in-pile test with JMTR, the 'tritium monitor' has been produced and evaluation of total tritium generation was done by using 'tritium monitor' in preliminary in-pile mock-up, and verification of procedure concerning tritium production evaluation was conducted by using Monte Carlo code MCNP and nuclear cross section library of FSXLIBJ3R2. Li-Al alloy (Li 3.4 wt.%, 95.5% enrichment of {sup 6}Li) was selected as tritium monitor material for the evaluation on the total amount of tritium production in high {sup 6}Li enriched materials. From the results of preliminary experiment, calculated amounts of total tritium production at each 'tritium monitor', which was installed in the preliminary in-pile mock-up, were about 50-290% higher than the measured values. Concerning tritium measurement, increase of measurement error in tritium leak form measuring system to measure small amount of tritium (0.2-0.7 mCi in tritium monitor) was found in the results of present experiment. The tendency for overestimation of calculated thermal neutron flux in the range of 1-6x10{sup 13} n cm{sup -2} per s was found in JMTR and the reason may be due to the beryllium cross section data base in JENDL3.2.

  11. Verification of tritium production evaluation procedure using Monte Carlo code MCNP for in-pile test of fusion blanket with JMTR

    International Nuclear Information System (INIS)

    Nagao, Y.; Nakamichi, K.; Tsuchiya, M.; Ishitsuka, E.; Kawamura, H.

    2000-01-01

    To evaluate exactly the total amount of tritium production in tritium breeding materials during in-pile test with JMTR, the 'tritium monitor' has been produced and evaluation of total tritium generation was done by using 'tritium monitor' in preliminary in-pile mock-up, and verification of procedure concerning tritium production evaluation was conducted by using Monte Carlo code MCNP and nuclear cross section library of FSXLIBJ3R2. Li-Al alloy (Li 3.4 wt.%, 95.5% enrichment of 6 Li) was selected as tritium monitor material for the evaluation on the total amount of tritium production in high 6 Li enriched materials. From the results of preliminary experiment, calculated amounts of total tritium production at each 'tritium monitor', which was installed in the preliminary in-pile mock-up, were about 50-290% higher than the measured values. Concerning tritium measurement, increase of measurement error in tritium leak form measuring system to measure small amount of tritium (0.2-0.7 mCi in tritium monitor) was found in the results of present experiment. The tendency for overestimation of calculated thermal neutron flux in the range of 1-6x10 13 n cm -2 per s was found in JMTR and the reason may be due to the beryllium cross section data base in JENDL3.2

  12. Qualification of power determination and in-pile measurements in the UO{sub 2} Gd{sub 2} 0{sub 2} fuel irradiation test IFA 636

    Energy Technology Data Exchange (ETDEWEB)

    Tverberg, T.; Volkov, B.; Kim, J-C.

    2004-04-15

    IFA-S36 is irradiated with the main objective of extending the database on the performance of UO{sub 2}Gd{sub 2}O{sub 2} fuel (with 8% absorbing gadolinia isotopes) compared with commercial UO{sub 2}. The rig carries 6 rods in the lower cluster (including three Gd-doped fuel rods) and 3 rods in the upper cluster (one rod with Gd-doped fuel). The rods are instrumented with expansion thermometers (ETs), fuel and cladding elongation detectors (EFs and ECs) and pressure transducers (PFs). Repeated calorimetric power measurements, physics calculations by the HELIOS code and gamma scans of selected rods in both clusters enabled the power and burnup determination to be qualified and corrected. The data suggest that as of May 2004 the power ratings in both fuels are much alike and burnups are about 30 and 34 MW/kgUO{sub 2} in the Gd-doped and ordinary UO{sub 2} rods respectively. Analysis of in-pile measurements compared with calculations shows that neutron absorption affects fuel temperature, power and burnup radial distributions in Gd-doped fuel at BOL compared with UO{sub 2} fuel. Sensitivity analyses performed with the HELIOS and FTEMP3 codes show that fuel centreline temperature in Gd-doped fuel is influenced by radial power depression, depletion of fissile materials and absorbing Gd isotopes as well as thermal conductivity of the fuel matrix and its degradation during irradiation. Analysis of the fuel dimension changes revealed densification only in the UO{sub 2} fuel whereas fuel elongation measurements in the Gd-doped fuel rods indicated essentially constant swelling with burnup. At burnups above 5 MWd/kgUO{sub 2} the swelling rate was about 0.5-O.fi % DELTAV/V per 10 MWd/kgUO{sub 2} for both fuel types. Internal pressure measured in the Gd-doped rod at BOL showed slight fuel densification and possibly He gas absorption, whereas derived swelling rate was somewhat Iarger than values obtained from the fuel elongation measurements. Cladding elongation measurements

  13. Development of in-pile instruments for fuel and material irradiation tests

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Akira; Kitagishi, Shigeru; Kimura, Nobuaki; Saito, Takashi; Nakamura, Jinichi; Ohmi, Masao; Izumo, Hironobu; Tsuchiya, Kunihiko [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    To get measurement data with high accuracy for fuel and material behavior studies in irradiation tests, two kinds of measuring equipments have been developed; these are the Electrochemical Corrosion Potential (ECP) sensor and the Linear Voltage Differential Transformer (LVDT) type gas pressure gauge. The ECP sensor has been developed to determine the corrosive potential under high temperature and high pressure water conditions. The structure of the joining parts was optimized to avoid stress concentration. The ECP sensor showed enough performance at 288degC and at 9MPa conditions. The LVDT type rod inner gas pressure gauge has been developed to measure gas pressure in a fuel element during neutron irradiation. To perform stable measurements with high accuracy under high temperature, high pressure and high dosed environment, the coil material of LVDT was changed to MI cable. As a result of this development, the LVDT type gas pressure gauge showed high accuracy within 1.8% of a full scale, and good stability. (author)

  14. Development of in-pile instruments for fuel and material irradiation tests

    International Nuclear Information System (INIS)

    Shibata, Akira; Kitagishi, Shigeru; Kimura, Nobuaki; Saito, Takashi; Nakamura, Jinichi; Ohmi, Masao; Izumo, Hironobu; Tsuchiya, Kunihiko

    2012-01-01

    To get measurement data with high accuracy for fuel and material behavior studies in irradiation tests, two kinds of measuring equipments have been developed; these are the Electrochemical Corrosion Potential (ECP) sensor and the Linear Voltage Differential Transformer (LVDT) type gas pressure gauge. The ECP sensor has been developed to determine the corrosive potential under high temperature and high pressure water conditions. The structure of the joining parts was optimized to avoid stress concentration. The ECP sensor showed enough performance at 288degC and at 9MPa conditions. The LVDT type rod inner gas pressure gauge has been developed to measure gas pressure in a fuel element during neutron irradiation. To perform stable measurements with high accuracy under high temperature, high pressure and high dosed environment, the coil material of LVDT was changed to MI cable. As a result of this development, the LVDT type gas pressure gauge showed high accuracy within 1.8% of a full scale, and good stability. (author)

  15. Determination of delayed neutrons source in the frequency domain based on in-pile oscillation measurements

    International Nuclear Information System (INIS)

    Yedvab, Y.; Reiss, I.; Bettan, M.; Harari, R.; Grober, A.; Ettedgui, H.; Caspi, E. N.

    2006-01-01

    A method for determining delayed neutrons source in the frequency domain based on measuring power oscillations in a non-critical reactor is presented. This method is unique in the sense that the delayed neutrons source is derived from the dynamic behavior of the reactor, which serves as the measurement system. An algorithm for analyzing power oscillation measurements was formulated, which avoids the need for a multi-parameter non-linear fit process used by other methods. Using this algorithm results of two sets of measurements performed in IRR-I and IRR-II (Israeli Research Reactors I and II) are presented. The agreement between measured values from both reactors and calculated values based on Keepin (and JENDL-3.3) group parameters is very good. (authors)

  16. IN-PILE INSTRUMENTATION TO SUPPORT FUEL CYCLE RESEARCH AND DEVELOPMENT - FY12 STATUS REPORT

    Energy Technology Data Exchange (ETDEWEB)

    J. . Rempe; J. Daw; D. Knudson; R. Schley

    2012-09-01

    As part of the FCRD program objective to emphasize science-based, goal-oriented research, a strategic research program is underway to develop new sensors that can be used to obtain the high fidelity, real-time, data required for characterizing the performance of new fuels during irradiation testing. The overarching goal of this initiative is to develop new test vehicles with new sensors of unprecedented accuracy and resolution that can obtain the required data. Prior laboratory testing and, as needed, irradiation testing of sensors in these capsules will be completed as part of this initiative to give sufficient confidence that the irradiation tests will yield the required data. This report documents FY12 progress in this initiative.

  17. Safety analyses for an in-pile SCWR fuel qualification test loop

    Energy Technology Data Exchange (ETDEWEB)

    Schulenberg, T.; Raque, M. [Karlsruhe Inst. of Tech., Karlsruhe (Germany)

    2014-07-01

    As a nuclear facility cooled with supercritical water has never been built nor operated in the past, the planned SCWR fuel qualification test will give the first experience with supercritical water-cooled nuclear systems in general. With a fuel inventory of almost 1 kg of UO{sub 2} with almost 20% enrichment, the supercritical pressure test section inside a low pressure, pool type research reactor needs to be cooled properly even in case of a number of postulated design basis accidents. Depressurization systems and emergency cooling systems will need to be designed with similar reliability as for a prototype reactor to ensure the integrity of barriers retaining the radioactive material. The paper reports about the safety concept and summarizes the safety analyses which have been performed in this context. (author)

  18. Risk analysis for a radiolysis gas detonation in an in-pile loop with supercritical water

    International Nuclear Information System (INIS)

    Zeiger, T.; Raque, M.; Kuznetsov, M.; Redlinger, R.; Schulenberg, T.

    2012-01-01

    The SCWR (supercritical water reactor) -FQT project is a cooperation between European and Chinese partners aimed to test the fuel SCWR elements under reactor conditions. In the frame of this work the risk of radiolysis gas production in the active range of the test track was assessed. The radiolysis gas could accumulate in an emergency cooling system with stagnating coolant. The ignition of this radiolysis gas could cause pressure peaks that are able to damage the primary coolant circuit. Pressure increase and deformations in case of ignition of accumulated gas were investigated. As piping material the Ti stabilized austenitic steel 08Ch18N10T was assumed, the simulation was performed using the ANSYS code. The results show that pipes without significant wall thickness enhancement cannot withstand the radiolysis gas detonation.

  19. In-pile thermocycling testing and post-test analysis of beryllium divertor mockups

    Energy Technology Data Exchange (ETDEWEB)

    Giniatulin, R.; Mazul, I. [Efremov Inst., St. Petersburg (Russian Federation); Melder, R.; Pokrovsky, A.; Sandakov, V.; Shiuchkin, A.

    1998-01-01

    The main damaging factors which impact the ITER divertor components are neutron irradiation, cyclic surface heat loads and hydrogen environment. One of the important questions in divertor mockups development is the reliability of beryllium/copper joints and the beryllium resistance under neutron irradiation and thermal cycling. This work presents the experiment, where neutron irradiation and thermocyclic heat loads were applied simultaneously for two beryllium/copper divertor mockups in a nuclear reactor channel to simulate divertor operational conditions. Two mockups with different beryllium grades were mounted facing each other with the tantalum heater placed between them. This device was installed in the active zone of the nuclear reactor SM-2 (Dimitrovgrad, Russia) and the tantalum block was heated by neutron irradiation up to a high temperature. The main part of the heat flux from the tantalum surface was transported to the beryllium surface through hydrogen, as a result the heat flux loaded two mockups simultaneously. The mockups were cooled by reactor water. The device was lowered to the active zone so as to obtain the heating regime and to provide cooling lifted. This experiment was performed under the following conditions: tantalum heater temperature - 1950degC; hydrogen environment -1000 Pa; surface heat flux density -3.2 MW/m{sup 2}; number of thermal cycles (lowering and lifting) -101; load time in each cycle - 200-5000 s; dwell time (no heat flux, no neutrons) - 300-2000 s; cooling water parameters: v - 1 m/s, Tin - 50degC, Pin - 5 MPa; neutron fluence -2.5 x 10{sup 20} cm{sup -2} ({approx}8 years of ITER divertor operation from the start up). The metallographic analysis was performed after experiment to investigate the beryllium and beryllium/copper joint structures, the results are presented in the paper. (author)

  20. In-pile loop studies of the effect of PWR coolant pH on corrosion product radionuclide deposition

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Harling, O.K.; Kohse, G.E.

    1992-02-01

    An in-pile loop which simulates the primary coolant system of a PWR has been constructed and operated in the MIT research reactor. A total of seven one-month-long irradiations have been carried out to evaluate the effect of coolant pH controlled by variation in LiOH/H 3 BO 3 concentrations. With the exception of one run at zero boron, all employed 800 ppm B; pH 300degreesC values of 6.5, 7.0, 7.2, 7.5 were studied, and two runs each at 7.0 and 7.2 were carried out. Finally, one of the runs at a pH 300degreesC of 7.2 was conducted with special care to exclude zinc because of its potential effects on cobalt deposition. The results show the expected benefits of high pH in reducing the rate of activity deposition on plant surfaces, but pH 300degreesC = 7.2 is approximately as effective as 7.5, while pH 300degreesC = 6.5 exhibits much larger activity transport and qualitatively different deposition behavior. Significant heat flux effects not predicted by current models have been consistently observed. While not as extensively studied, the zero-boron run suggests that the presence of boron species, at fixed pH, may reduce the net amount of activity deposited on ex-core surfaces. Neutron activation analysis of a variety of samples ruled out Zircaloy as an important source of Co-60, since its cobalt content is less than one ppm, considerably less than the applicable ASTM specification of ≤ 20 ppm. Amendment of the latter has been recommended

  1. In-pile data analysis of the comparative WWER/PWR test IFA-503.1. Final report.

    Energy Technology Data Exchange (ETDEWEB)

    Volkov, B.; Devold, H.; Ryazantzev, E.; Yakovlev, V.

    1999-04-15

    The comparative WWER/PWR test in IFA-503.1 was commenced in July 1995 and successfully finished at the end of November 1998. The main objective of the test was generation of representative and comparative data of standard WWER-440 fuel fabricated at the 'MSZ' Electrostal (Russia) and PWR type fuel manufactured at IFE Kjeller (Norway). The test assembly comprised two clusters, each with 3 WWER rods and 3 PWR type rods. Eight rods with two types of fuel were instrumented with expansion thermometers, four rods were equipped with both fuel stack elongation detectors and pressure transducers. All sensors worked satisfactorily during the test. The average burnups achieved in the lower and upper clusters were around 25 and 20 MWd/kgUO{sub 2}, respectively. Some difference in densification of the two types of fuel was revealed during the first irradiation period. However, the fuel temperatures and commencement of fuel stack swelling were similar despite this fact. At the end of the test the rig was moved to a higher flux position in the HBWR core with the aim of promoting FGR and to compare the behaviour of the two types of fuel under higher power. Pressure measurements indicated a comparable low FGR (around 1 percent) in both types of rods. The centreline temperatures measured in the PWR rods were very close to the Halden FGR threshold whilst the WWER fuel temperatures were slightly lower. Despite the differences found in the behaviour of the two types of fuel during the test, the analysis of the in-pile data showed that these differences would not affect the fuel efficiency, at least, up to the burnup achieved in the test. It is supposed that these differences can be related to the fuel microstructure, in particular to the fuel grain and pore sizes (author) (ml)

  2. In-pile data analysis of the comparative WWER/PWR test IFA-503.1. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Volkov, B.; Devold, H.; Ryazantzev, E.; Yakovlev, V

    1999-04-15

    The comparative WWER/PWR test in IFA-503.1 was commenced in July 1995 and successfully finished at the end of November 1998. The main objective of the test was generation of representative and comparative data of standard WWER-440 fuel fabricated at the 'MSZ' Electrostal (Russia) and PWR type fuel manufactured at IFE Kjeller (Norway). The test assembly comprised two clusters, each with 3 WWER rods and 3 PWR type rods. Eight rods with two types of fuel were instrumented with expansion thermometers, four rods were equipped with both fuel stack elongation detectors and pressure transducers. All sensors worked satisfactorily during the test. The average burnups achieved in the lower and upper clusters were around 25 and 20 MWd/kgUO{sub 2}, respectively. Some difference in densification of the two types of fuel was revealed during the first irradiation period. However, the fuel temperatures and commencement of fuel stack swelling were similar despite this fact. At the end of the test the rig was moved to a higher flux position in the HBWR core with the aim of promoting FGR and to compare the behaviour of the two types of fuel under higher power. Pressure measurements indicated a comparable low FGR (around 1 percent) in both types of rods. The centreline temperatures measured in the PWR rods were very close to the Halden FGR threshold whilst the WWER fuel temperatures were slightly lower. Despite the differences found in the behaviour of the two types of fuel during the test, the analysis of the in-pile data showed that these differences would not affect the fuel efficiency, at least, up to the burnup achieved in the test. It is supposed that these differences can be related to the fuel microstructure, in particular to the fuel grain and pore sizes (author) (ml)

  3. Scanning thermal microscopy based on a quartz tuning fork and a micro-thermocouple in active mode (2ω method)

    International Nuclear Information System (INIS)

    Bontempi, Alexia; Nguyen, Tran Phong; Salut, Roland; Thiery, Laurent; Teyssieux, Damien; Vairac, Pascal

    2016-01-01

    A novel probe for scanning thermal microscope using a micro-thermocouple probe placed on a Quartz Tuning Fork (QTF) is presented. Instead of using an external deflection with a cantilever beam for contact detection, an original combination of piezoelectric resonator and thermal probe is employed. Due to a non-contact photothermal excitation principle, the high quality factor of the QTF allows the probe-to-surface contact detection. Topographic and thermal scanning images obtained on a specific sample points out the interest of our system as an alternative to cantilevered resistive probe systems which are the most spread.

  4. Scanning thermal microscopy based on a quartz tuning fork and a micro-thermocouple in active mode (2ω method).

    Science.gov (United States)

    Bontempi, Alexia; Nguyen, Tran Phong; Salut, Roland; Thiery, Laurent; Teyssieux, Damien; Vairac, Pascal

    2016-06-01

    A novel probe for scanning thermal microscope using a micro-thermocouple probe placed on a Quartz Tuning Fork (QTF) is presented. Instead of using an external deflection with a cantilever beam for contact detection, an original combination of piezoelectric resonator and thermal probe is employed. Due to a non-contact photothermal excitation principle, the high quality factor of the QTF allows the probe-to-surface contact detection. Topographic and thermal scanning images obtained on a specific sample points out the interest of our system as an alternative to cantilevered resistive probe systems which are the most spread.

  5. Scanning thermal microscopy based on a quartz tuning fork and a micro-thermocouple in active mode (2ω method)

    Energy Technology Data Exchange (ETDEWEB)

    Bontempi, Alexia; Nguyen, Tran Phong; Salut, Roland; Thiery, Laurent; Teyssieux, Damien; Vairac, Pascal [FEMTO-ST Institute UMR 6174, Université de Franche-Comté, CNRS, ENSMM, UTBM, 15B Avenue des Montboucons, F-25030 Besançon (France)

    2016-06-15

    A novel probe for scanning thermal microscope using a micro-thermocouple probe placed on a Quartz Tuning Fork (QTF) is presented. Instead of using an external deflection with a cantilever beam for contact detection, an original combination of piezoelectric resonator and thermal probe is employed. Due to a non-contact photothermal excitation principle, the high quality factor of the QTF allows the probe-to-surface contact detection. Topographic and thermal scanning images obtained on a specific sample points out the interest of our system as an alternative to cantilevered resistive probe systems which are the most spread.

  6. Attachment of lead wires to thin film thermocouples mounted on high temperature materials using the parallel gap welding process

    Science.gov (United States)

    Holanda, Raymond; Kim, Walter S.; Pencil, Eric; Groth, Mary; Danzey, Gerald A.

    1990-01-01

    Parallel gap resistance welding was used to attach lead wires to sputtered thin film sensors. Ranges of optimum welding parameters to produce an acceptable weld were determined. The thin film sensors were Pt13Rh/Pt thermocouples; they were mounted on substrates of MCrAlY-coated superalloys, aluminum oxide, silicon carbide and silicon nitride. The entire sensor system is designed to be used on aircraft engine parts. These sensor systems, including the thin-film-to-lead-wire connectors, were tested to 1000 C.

  7. Advanced instrumentation and analysis methods for in-pile thermal and nuclear measurements: from out-of-pile studies to irradiation campaigns

    International Nuclear Information System (INIS)

    Reynard-Carette, C.; Lyoussi, A.

    2015-01-01

    Research and development on nuclear fuel behavior under irradiations and accelerated ageing of structure materials is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. A new Material Testing Reactor (MTR), the Jules Horowitz Reactor (JHR) currently under construction in the South of France in the CEA Cadarache research centre will offer a real opportunity to perform R and D programs and hence will crucially contribute to the selection, optimization and qualification of innovative materials and fuels. To perform such programs advanced accurate and innovative experiments, irradiation devices that contain material and fuel samples are required to be set up inside or beside the reactor core. These experiments needs beforehand in situ and on line sophisticated measurements to accurately reach specific and determining parameters such as thermal and fast neutron fluxes, nuclear heating and temperature conditions to precisely monitor and control the conducted assays. Consequently, since 2009 CEA and Aix-Marseille University collaborate in order to design and develop a new multi-sensor device which will be dedicated to measuring profiles of such conditions inside the experimental channels of the JHR. These works are performed in the framework of two complementary joint research programs called MAHRI-BETHY and INCORE. These programs couple experimental studies carried out both out-of nuclear fluxes (in laboratory) and under irradiation conditions (in OSIRIS MTR reactor in France and MARIA MTR reactor in Poland) with numerical works realized by thermal simulations (CAST3M code) and Monte Carlo simulations (MCNP code). These programs deal with three main aims. The first one corresponds to the design and/or the test of new in-pile instrumentation. The second one concerns the development of advanced calibration procedures in particular in the case of one specific sensor: a differential calorimeter used to quantify

  8. Advanced instrumentation and analysis methods for in-pile thermal and nuclear measurements: from out-of-pile studies to irradiation campaigns

    Energy Technology Data Exchange (ETDEWEB)

    Reynard-Carette, C. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 (France)

    2015-07-01

    Research and development on nuclear fuel behavior under irradiations and accelerated ageing of structure materials is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. A new Material Testing Reactor (MTR), the Jules Horowitz Reactor (JHR) currently under construction in the South of France in the CEA Cadarache research centre will offer a real opportunity to perform R and D programs and hence will crucially contribute to the selection, optimization and qualification of innovative materials and fuels. To perform such programs advanced accurate and innovative experiments, irradiation devices that contain material and fuel samples are required to be set up inside or beside the reactor core. These experiments needs beforehand in situ and on line sophisticated measurements to accurately reach specific and determining parameters such as thermal and fast neutron fluxes, nuclear heating and temperature conditions to precisely monitor and control the conducted assays. Consequently, since 2009 CEA and Aix-Marseille University collaborate in order to design and develop a new multi-sensor device which will be dedicated to measuring profiles of such conditions inside the experimental channels of the JHR. These works are performed in the framework of two complementary joint research programs called MAHRI-BETHY and INCORE. These programs couple experimental studies carried out both out-of nuclear fluxes (in laboratory) and under irradiation conditions (in OSIRIS MTR reactor in France and MARIA MTR reactor in Poland) with numerical works realized by thermal simulations (CAST3M code) and Monte Carlo simulations (MCNP code). These programs deal with three main aims. The first one corresponds to the design and/or the test of new in-pile instrumentation. The second one concerns the development of advanced calibration procedures in particular in the case of one specific sensor: a differential calorimeter used to quantify

  9. Part 1: Logging residues in piles - Needle loss and fuel quality. Part 2: Nitrogen leaching under piles of logging residues

    International Nuclear Information System (INIS)

    Lehtikangas, P.; Lundkvist, H.

    1991-01-01

    Part 1: Experimental piles were built in three geographical locations during May-Sept. 1989. Logging residues consisted of 95% spruce and 5% pine. Height of the piles varied between 80 and 230 cm. Needles were collected by placing drawers under 40 randomely chosen piles. The drawers were emptied every two weeks during the storage period. Natural needle loss was between 18 and 32% of the total amount of needles after the first two months of storage. At the end of the storage period, 24-42% of the needles had fallen down to the drawers. At the end of the experiment the total needle fall was 95-100% in the shaken piles. According to the results of this study piles smaller than 150 cm had the most effective needle fall. Piles should be placed on open places where the air and sun heat penetrate and dry them. Needles were the most sensitive fraction to variations in precipitation compared to the other components, such as branches. Piles usually dried quickly, but they also rewet easily. This was especially true in the smaller piles. The lowest moisture content was measured at the end of June. The ash content in needles varied between 4 and 8%. 16 refs., 15 figs. Part 2: Three field experiments were equipped with no-tension humus lysimeters. Pairs of lysimeters with the same humus/field layer vegetation material were placed in pairs, one under a pile of felling residues and another in the open clear felling. Leaching of nitrogen as well as pH and electric conductivity in the leachate was followed through sampling of the leachate at regular intervals. The results from the investigation show that: * the amount of leachate was higher in lysimeters in the open clear felling, * pH in the leachate was initially lower under piles of felling residues, * the amount of nitrogen leached was higher in the open clear felling. Thus, storing of felling residues in piles during the summer season did not cause any increase in nitrogen leaching, which had been considered to be a risk

  10. Investigation of pool boiling dynamics on a rectangular heater using nano-thermocouples: is it chaotic or stochastic?

    Energy Technology Data Exchange (ETDEWEB)

    Sathyamurthi, Vijaykumar; Banerjee, Debjyoti [Texas A and M University, College Station, TX (United States). Dept. of Mechanical Engineering], e-mail: dbanerjee@tamu.edu

    2009-07-01

    The non-linear dynamical model of pool boiling on a horizontal rectangular heater is assessed from experimental results in this study. Pool boiling experiments are conducted over a horizontal rectangular silicon substrate measuring 63 mm x 35 mm with PF-5060 as the test fluid. Novel nano-thermocouples, micro-machined in-situ on the silicon substrate are used to measure the surface temperature fluctuations for steady state pool boiling. The acquisition frequency for temperature data from the nano-thermocouples is 1 k Hz. The surface temperature fluctuations are analyzed using the TISEAN{sup c} package. A time-delay embedding is employed to generate higher dimensional phase-space vectors from the temperature time series record. The optimal delay is determined from the first minimum of the mutual information function. Techniques such as recurrence plots, and false nearest neighbors tests are employed to assess the presence of deterministic chaotic dynamics. Chaos quantifiers such as correlation dimensions are found for various pool boiling regimes using the raw data as well as noise-reduced data. Additionally, pseudo-phase spaces are used to reconstruct the 'attractors'. The results after non-linear noise reduction shows definitive presence of low-dimensional (d {<=} 7) chaos in fully developed nucleate boiling, at critical heat flux and in film boiling. (author)

  11. Investigation of pool boiling dynamics on a rectangular heater using nano-thermocouples: is it chaotic or stochastic?

    International Nuclear Information System (INIS)

    Sathyamurthi, Vijaykumar; Banerjee, Debjyoti

    2009-01-01

    The non-linear dynamical model of pool boiling on a horizontal rectangular heater is assessed from experimental results in this study. Pool boiling experiments are conducted over a horizontal rectangular silicon substrate measuring 63 mm x 35 mm with PF-5060 as the test fluid. Novel nano-thermocouples, micro-machined in-situ on the silicon substrate are used to measure the surface temperature fluctuations for steady state pool boiling. The acquisition frequency for temperature data from the nano-thermocouples is 1 k Hz. The surface temperature fluctuations are analyzed using the TISEAN c package. A time-delay embedding is employed to generate higher dimensional phase-space vectors from the temperature time series record. The optimal delay is determined from the first minimum of the mutual information function. Techniques such as recurrence plots, and false nearest neighbors tests are employed to assess the presence of deterministic chaotic dynamics. Chaos quantifiers such as correlation dimensions are found for various pool boiling regimes using the raw data as well as noise-reduced data. Additionally, pseudo-phase spaces are used to reconstruct the 'attractors'. The results after non-linear noise reduction shows definitive presence of low-dimensional (d ≤ 7) chaos in fully developed nucleate boiling, at critical heat flux and in film boiling. (author)

  12. In-Pile Experiment of a New Hafnium Aluminide Composite Material to Enable Fast Neutron Testing in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Douglas L. Porter; James R. Parry; Heng Ban

    2010-06-01

    A new hafnium aluminide composite material is being developed as a key component in a Boosted Fast Flux Loop (BFFL) system designed to provide fast neutron flux test capability in the Advanced Test Reactor. An absorber block comprised of hafnium aluminide (Al3Hf) particles (~23% by volume) dispersed in an aluminum matrix can absorb thermal neutrons and transfer heat from the experiment to pressurized water cooling channels. However, the thermophysical properties, such as thermal conductivity, of this material and the effect of irradiation are not known. This paper describes the design of an in-pile experiment to obtain such data to enable design and optimization of the BFFL neutron filter.

  13. Internal attachment of laser beam welded stainless steel sheathed thermocouples into stainless steel upper end caps in nuclear fuel rods for the LOFT Reactor

    International Nuclear Information System (INIS)

    Welty, R.K.; Reid, R.D.

    1980-01-01

    The Exxon Nuclear Company, Inc., acting as a subcontractor to EG and G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho, conducted a laser beam welding study to attach internal stainless steel thermocouples into stainless steel upper end caps in nuclear fuel rods. The objective of this study was to determine the feasibility of laser welding a single 0.063 inch diameter stainless steel (304) sheathed thermocouple into a stainless steel (316) upper end cap for nuclear fuel rods. A laser beam was selected because of the extremely high energy input in unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A special weld fixture was designed and fabricated to hold the end cap and the thermocouple with angular and rotational adjustment under the laser beam. A commercial pulsed laser and energy control system was used to make the welds

  14. Development of remote welding techniques for in-pile IASCC capsules and evaluation of material integrity on capsules for long irradiation period

    Science.gov (United States)

    Shibata, A.; Nakano, J.; Ohmi, M.; Kawamata, K.; Nakagawa, T.; Tsukada, T.

    2012-03-01

    To simulate irradiation assisted stress corrosion cracking (IASCC) behavior by in-pile experiments, it is necessary to irradiate specimens up to a neutron fluence that is higher than the IASCC threshold fluence. Pre-irradiated specimens must be relocated from pre-irradiation capsules to in-pile capsules. Hence, a remote welding machine has been developed. And the integrity of capsule housing for a long term irradiation was evaluated by tensile tests in air and slow strain rate tests in water. Two type specimens were prepared. Specimens were obtained from the outer tubes of capsule irradiated to 1.0-3.9 × 1026 n/m2 (E > 1 MeV). And specimens were irradiated in a leaky capsule to 0.03-1.0 × 1026 n/m2. Elongation more than 15% in tensile test at 423 K was confirmed and no IGSCC fraction was shown in SSRT at 423 K which was estimated as temperature at the outer tubes of the capsule under irradiation.

  15. Development of remote welding techniques for in-pile IASCC capsules and evaluation of material integrity on capsules for long irradiation period

    International Nuclear Information System (INIS)

    Shibata, A.; Nakano, J.; Ohmi, M.; Kawamata, K.; Nakagawa, T.; Tsukada, T.

    2012-01-01

    To simulate irradiation assisted stress corrosion cracking (IASCC) behavior by in-pile experiments, it is necessary to irradiate specimens up to a neutron fluence that is higher than the IASCC threshold fluence. Pre-irradiated specimens must be relocated from pre-irradiation capsules to in-pile capsules. Hence, a remote welding machine has been developed. And the integrity of capsule housing for a long term irradiation was evaluated by tensile tests in air and slow strain rate tests in water. Two type specimens were prepared. Specimens were obtained from the outer tubes of capsule irradiated to 1.0–3.9 × 10 26 n/m 2 (E > 1 MeV). And specimens were irradiated in a leaky capsule to 0.03–1.0 × 10 26 n/m 2 . Elongation more than 15% in tensile test at 423 K was confirmed and no IGSCC fraction was shown in SSRT at 423 K which was estimated as temperature at the outer tubes of the capsule under irradiation.

  16. Home-made temperature monitoring system from four-channel K-type thermocouples via internet of thing technology platform

    Science.gov (United States)

    Detmod, Thitaporn; Özmen, Yiǧiter; Songkaitiwong, Kittiphot; Saenyot, Khanuengchat; Locharoenrat, Kitsakorn; Lekchaum, Sarai

    2018-06-01

    This paper is aimed to design and construct the home-made temperature monitoring system from four-channel K-type thermocouples in order to improve the temperature measurement based on standard evaluation measurements guidance. The temperature monitoring system was capable to record the temperature on SD card and to display the realtime temperature on Internet of Thing Technology platform. The temperature monitoring system was tested in terms of the temperature measurement accuracy and delay response time. It was found that a standard deviation was acceptable as compared to the Instrument Society of America. The response time of the microcontroller to SD card was 2 sec faster than that of the microcontroller to Thingspeak.

  17. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    International Nuclear Information System (INIS)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm 2 , 1000 0 C cladding temperature, and (2) 40 h at 40 W/cm 2 , 1200 0 C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370 0 C

  18. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  19. The Combined Use of a Gas-Controlled Heat Pipe and a Copper Point to Improve the Calibration of Thermocouples up to 1100 ˚C

    Science.gov (United States)

    Astrua, M.; Iacomini, L.; Battuello, M.

    2008-10-01

    The calibration of platinum-based thermocouples from 420 °C to 1,100 ˚C is currently carried out at INRIM making use of two different apparatus: for temperatures below 930 ˚C, a potassium gas-controlled heat pipe (GCHP) is used, whereas a metal-block furnace is adopted for higher temperatures. The standard uncertainty of the reference temperature obtained in the lower temperature range is almost one order of magnitude better than in the higher temperature range. A sealed copper cell was investigated to see if it could be used to calibrate thermocouples above 930 ˚C with a lower uncertainty than our current procedures allowed. The cell was characterized with Type S and Pt/Pd thermocouples and with an HTPRT. The freezing plateaux were flat within 0.01 ˚C and lasted up to 1 h with a repeatability of 0.02 ˚C. The temperature of the cell was determined with a standard uncertainty of 0.04 ˚C. Hence, the copper cell was found to be superior to the comparator furnace for the calibration of platinum-based thermocouples because of the significant decrease in the uncertainty that it provides. An analysis was also carried out on the calibration of Pt/Pd thermocouples, and it was found that the combined use of the potassium GCHP and the Cu fixed-point cell is adequate to exploit the potential of these sensors in the range from 420 °C to 1,084 °C. A comparison with a fixed-point calibration was also made which gave rise to agreement within 0.07 ˚C between the two approaches.

  20. Chemical effects of 13N produced by recoil protons and deuterons in pile-irradiated methanol and methanol-d4

    International Nuclear Information System (INIS)

    Sensui, Y.; Tomura, K.; Matsuura, T.

    1982-01-01

    The stabilized chemical forms of 13 N resulting from the reactions 13 C(p,n) 13 N by a recoil proton and 12 C(d,n) 13 N by a recoil deuteron, were studied in pile-irradiated methanol and methanol-d 4 in the temperature range from 77 to 295 K. Contrary to the target of benzene, cyclohexane, acetone and diethyl ether previously studied, the relative yield of 13 N-compounds did not depend on the irradiation temperature in the present media. In the yield of 13 N-compounds no marked change was observed between methanol and methanol-d 4 , differing from the results between benzene and benzene-d 6 . A mechanism is proposed to explain the results. (author)

  1. Absolute on-line in-pile measurement of neutron fluxes using self-powered neutron detectors: Monte Carlo sensitivity calculations

    Energy Technology Data Exchange (ETDEWEB)

    Vermeeren, L. [SCK/CEN, B-2400 Mol (Belgium)

    2001-07-01

    Self-powered neutron detectors (SPND) are well suited to monitor continuously the neutronic operating conditions of driver fuel of research reactors and to follow its burnup evolution. This is of particular importance when advanced or new MTR fuel designs need to be qualified. We have developed a detailed MCNP-4B based Monte Carlo approach for the calculation of neutron sensitivities of SPNDs. Results for the neutron sensitivity of a Rh SPND are in excellent agreement with experimental data recently obtained at the BR2 research reactor. A critical comparison of the Monte Carlo results with results from standard analytical methods reveals an important deficiency of the analytical methods in the description of the electron transport efficiency. Our calculation method allows a reliable on-line determination of the absolute in-pile neutron flux. (author)

  2. Absolute on-line in-pile measurement of neutron fluxes using self-powered neutron detectors: Monte Carlo sensitivity calculations

    International Nuclear Information System (INIS)

    Vermeeren, L.

    2001-01-01

    Self-powered neutron detectors (SPND) are well suited to monitor continuously the neutronic operating conditions of driver fuel of research reactors and to follow its burnup evolution. This is of particular importance when advanced or new MTR fuel designs need to be qualified. We have developed a detailed MCNP-4B based Monte Carlo approach for the calculation of neutron sensitivities of SPNDs. Results for the neutron sensitivity of a Rh SPND are in excellent agreement with experimental data recently obtained at the BR2 research reactor. A critical comparison of the Monte Carlo results with results from standard analytical methods reveals an important deficiency of the analytical methods in the description of the electron transport efficiency. Our calculation method allows a reliable on-line determination of the absolute in-pile neutron flux. (author)

  3. Use of FET in automatic scanning of measurements using thermocouples and self-powered neutron detectors

    International Nuclear Information System (INIS)

    Plaige, Yves.

    1977-01-01

    Advantages lying in using FET switches in the relays of multiplexing systems are shown with two examples of application. Their performance as regard fast reliable operation are used in temperature measurement scanning inside nuclear reactors. As for current measurements using self-powered neutron detectors, the weak leakage currents of said switches ( [fr

  4. Effect of in-pile degradation of the meat thermal conductivity on the maximum temperature of the plate-type U-Mo dispersion fuels

    International Nuclear Information System (INIS)

    Medvedev, Pavel G.

    2009-01-01

    Effect of in-pile degradation of thermal conductivity on the maximum temperature of the plate-type research reactor fuels has been assessed using the steady-state heat conduction equation and assuming convection cooling. It was found that due to very low meat thickness, characteristic for this type of fuel, the effect of thermal conductivity degradation on the maximum fuel temperature is minor. For example, the fuel plate featuring 0.635 mm thick meat operating at heat flux of 600 W/cm2 would experience only a 20 C temperature rise if the meat thermal conductivity degrades from 0.8 W/cm-s to 0.3 W/cm-s. While degradation of meat thermal conductivity in dispersion-type U-Mo fuel can be very substantial due to formation of interaction layer between the particles and the matrix, and development of fission gas filled porosity, this simple analysis demonstrates that this phenomenon is unlikely to significantly affect the temperature-based safety margin of the fuel during normal operation.

  5. Results of water chemistry control in the in-pile ''Callisto'' loop (an experimental PWR rig installed in the BR2 reactor)

    International Nuclear Information System (INIS)

    Weber, M.; Benoit, P.; Dekeyser, J.; Verwimp, A.

    1994-01-01

    Since June 1992, a new experimental facility, called CALLISTO, is being irradiated in the BR2 materials testing reactor at Mol, Belgium. The main objective of the present test campaign is to study the behaviour of advanced fuel to high burn-up rates in a realistic PWR environment. Three in-pile sections, containing each 9 fuel rods, are loaded inside the reactor vessel and are connected to a common out-of-pile pressurized water circulation loop (ref.1). The later is branched-off into a purification circuit (feed-bleed concept) and further equipped with safety and auxiliary systems. To cope with the test programme, the equipments are designed so that the guidelines of a PWR primary water chemistry can be followed (ref.2). Real steady-state conditions cannot be observed because the typical BR2 cycle (3 weeks running/3 weeks shut-down) is much shorter and because the rig is cooled down during each reactor shut-down. The purpose of this poster is to provide results of chemical parameters recorded during the cycling behaviour of the CALLISTO primary water. (authors). 4 figs., 1 tab., 2 refs

  6. Manufacturing of In-Pile Test Section(IPS) Mock-up for the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. M.; Park, K. N.; Chi, D. Y. (and others)

    2005-10-15

    Manufacturing process of IPS Mock-up was initiated in late of 2003 with DAEWOO Precision industries Company. Manufacturing drawings due to detail drawings are composed of Outer assembly and Inner assembly. Welding of IPS Mock-up was performed by the GMAW(Gas Metal Arc Welding) process. After the welding process, non-destructive examination was conducted. Leak test was performed to the Main cooling water part and Neon gas inter-space gap part by the He gas injection with the pressure of 6.0 kg{sub f}/cm{sup 2} and 30 minutes holding time. the result was shown that there was no leak at the Neon gas inter-space gap part but leak was occurred at Main cooling water part according to imperfect screw of purge plug. so, it was re-finished and test was performed to certify the leak tightness. To satisfy the HANARO Limiting Operation Condition, IPS should be tested ahead of installation at the HANARO reactor by the use of test facilities. IPS Mock-up and its test facilities will be designed and used for the test of 'HANARO flow tube pressure drop', 'IPS inner pressure drop' and 'IPS inner vibration'.

  7. In pile programme of first valutation of UO2 + PuO2 fuel produced by a new process (GSP)

    International Nuclear Information System (INIS)

    Caracchin, R.; Lanchi, M.; Marinucci, G.; Nobili, A.; Dupont, G.; Galtier, J.

    1982-01-01

    The main scope of the ENEA-AGN-CEA programme collaboration is a first valutation of fuel elements produced by GSP method. This valuation will be done by in reactor experiment which enable to compare the performance of GSP and 'standard' FBR fuels. The composition is done by means of theree experimental device: P3, Lugel and Digel. The P3 device gives a direct measurement during irradiation of fuel central temperature, power and integral conductivity. The Lugel device measures fuel stack axial variations and Digel device gives the diameter variations of the pin and PCMI

  8. Effect of phosphorus on out-of-pile and in-pile behaviour of stabilized austenitic stainless steels

    International Nuclear Information System (INIS)

    Delalande, C.

    1992-02-01

    This work deals with the improvement of swelling resistance for austenitic stainless steels used as fuel pin cladding in Fast Breeder Reactor. The effect of phosphorus addition and multistabilization by Ti and Nb or Ti, Nb and V are studied on Fe-15Cr-15/25Ni based alloys. First, different ageings are performed to verify the stability of dislocation network, main condition of swelling absence at high irradiation temperature (T>550 deg C, and to study the precipitation, especially the one being able to form during irradiation and to control swelling at lower temperature. Then, 1 MeV electron irradiations are performed to estimate the swelling resistance of these multistabilized steels. Furthermore, neutron radiation induced microstructure of phosphorus modified steels already irradiated in reactor give us fundamental informations to predict and explain the effect of phosphorus and multistabilization on the behaviour of the multistabilized steels. Our results show that niobium plays the same role as titanium on the stabilization ratio in steels, but it is present in more phases. Vanadium seems to have less effect on stability of dislocation network and chemical composition of precipitates. Phosphorus increases the stability of dislocation network of multistabilized steels and FeNbP phosphides are observed at high temperature for phosphorus level above 600 ppm. 1 MeV electron irradiations show that multistabilized steels present good swelling resistance. Phosphorus addition increases the swelling resistance of neutron irradiated steels. (Author). refs., figs., tabs

  9. Self-Powered Neutron Detector Qualification for Absolute On-Line In-Pile Neutron Flux Measurements in BR2

    Science.gov (United States)

    Vermeeren, L.; Wéber, M.

    2003-06-01

    A set of ten Self-Powered Neutron Detectors with Co, Rh and Ag emitters has been irradiated in several channels of the BR2 research reactor at SCK•CEN aiming at a comparison of their performance as thermal neutron flux detectors under various conditions. To allow for a correct interpretation of their signals, all detector sensitivity contributions (prompt and delayed) were calculated using a dedicated Monte Carlo model. The various contributions were also measured separately; the agreement between calculated and experimental data, including data from activation dosimetry, was excellent. Detailed neutron flux profiles were obtained from the SPND data, after correction for the finite detector lengths and for the slow response of delayed SPNDs.

  10. Post irradiation examination of type 316 stainless steels for in-pile Oarai water loop No.2 (OWL-2)

    International Nuclear Information System (INIS)

    Shibata, Akira; Kimura, Tadashi; Nagata, Hiroshi; Aoyama, Masashi; Kanno, Masaru; Ohmi, Masao

    2010-11-01

    The Oarai water loop No.2 (OWL-2) was installed in JMTR in 1972 for the purpose of irradiation experiments of fuel element and component material for light water reactors. Type 316 stainless steels (SSs) were used for tube material of OWL-2 in the reactor. But data of mechanical properties of highly irradiated Type 316 SSs has been insufficient since OWL-2 was installed. Therefore surveillance tests of type 316 SSs which were irradiated up to 3.4x10 25 n/m 2 in fast neutron fluence (>1 MeV) were performed. Meanwhile type 316 stainless steel (SS) is widely used in JMTR such as other irradiation apparatus and irradiation capsule, and additional data of type 316 SSs irradiated higher is required. Therefore post irradiation examinations of surveillance specimens made of type 316 SSs which were irradiated up to 1.0x10 26 n/m 2 in fast neutron fluence were performed and reported in this paper. In this result of surveillance tests of type 316 SSs irradiated up to 1.0x10 26 n/m 2 , tensile strength increase with increase of Neutron fluence and total elongation decreased with increase of Neutron fluence compared to unirradiated specimens and specimens irradiated up to 3.4x10 25 n/m 2 . This tendency has good agreement with results of 10 24 - 10 25 n/m 2 in fast neutron fluence. More than 37% in total elongation was confirmed in all test conditions. It was confirmed that type 316 SS irradiated up to 1.0x10 26 n/m 2 in fast neutron fluence has enough ductility as structure material. (author)

  11. A Grid of Fine Wire Thermocouples to Study the Spatial Coherence of Turbulence within Katabatic Flow through a Vineyard Canopy

    Science.gov (United States)

    Everard, K.; Christen, A.; Sturman, A.; Skaloud, P.

    2016-12-01

    Knowledge of the dynamics and thermodynamics of katabatic flow is relevant in vineyards, where grapevines are sensitive to temperature changes (frost protection and cooling). Basic understanding of the occurrence and evolution of, and turbulence within, katabatic flow is well known over bare slopes. However, little work has been completed to extend this understanding to mid-sized canopies and how the presence of a canopy affects the turbulent exchange of momentum and heat within the flow. Measurements were carried out over a 6° vineyard slope near Oliver, BC, Canada in the Okanagan Valley between July 5 and July 22, 2016. The set-up consisted of an array of five vertically arranged CSAT 3D (Campbell Scientific, Inc.) ultrasonic anemometers at z = 0.45 m, 0.90 m, 1.49 m, 2.34 m, and 4.73 m above ground level (AGL), and a 2-D grid of 40 Type-E (copper-constantan) fine-wire thermocouples (FWTC) arranged at the same heights as the CSAT 3D array on 8 masts extending in the upslope (flow) direction at locations x = 0.0 m (CSAT 3D tower), 0.5 m, 1.0 m, 2.0 m, 4.0 m, 8.0 m, 16.0 m, and 32.0 m. The FWTC array formed a sheet of 40 sampling points in the upslope-vertical plane. The height of the grapevine canopy (h) was approximately 2 m AGL, and rows were aligned along the local slope direction with a row spacing of 2.45 m. CSAT-3s were sampled at 60 Hz with 20 Hz data recording, the FWTCs were sampled at 2 Hz, all synchronized by a data logger. Katabatic flow was observed on several nights during the campaign, with a wind speed maximum located within the canopy. This contribution will focus on the measurement techniques, combining ultrasonic anemometer data with the spatially synchronized FWTC array using image process techniques. We identify the dynamics and structure of the katabatic flow, relevant for heat exchange, using the spatial coherence of the temperature field given by the FWTC array. Improved knowledge of the vertical structure and the dynamics of katabatic

  12. Neutronic design of pulse operation simulating device for in-pile functional test of fusion blanket by MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Yoshiharu; Nakamichi, Masaru; Kawamura, Hiroshi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan)

    2000-03-01

    The pulse operation of a fusion reactor can be simulated in a fission reactor by controlling the neutron flux entering a test section by using a rotating 'hollow cylinder with window' made of hafnium. The rotating cylinder is installed between the test section and the fixed outer neutron absorber cylinder and is also made of hafnium with an opening in the direction to the core center. For gathering engineering data for the tritium breeding blanket such as characteristics of temperature change, tritium release and recovery, etc., it is desirable that the ratio of minimum to maximum thermal neutron fluxes is greater than 1:10. Design calculations were performed for the test assembly which considered local neutronic effects and the mechanical constraints of the device. From the results of these calculations, the ratio of minimum to maximum thermal neutron flux under irradiation would be about 1:10 using a pulse operation simulating device which has a thickness of 6.5 mm and a 150deg window angle for the rotating hollow cylinder and 5.0 mm in thickness of fixed neutron absorber. (author)

  13. Program for in-pile qualification of high density silicide dispersion fuel at IPEN/CNEN-SP

    International Nuclear Information System (INIS)

    Silva, Jose E.R. da; Silva, Antonio T. e; Terremoto, Luis A.A.; Durazzo, Michelangelo

    2009-01-01

    The development of high density nuclear fuel (U 3 Si 2 -Al) with 4,8 gU/cm 3 is on going at IPEN, at this time. This fuel has been considered to be utilized at the new Brazilian Multipurpose Reactor (RMB), planned to be constructed up to 2014. As Brazil does not have hot-cell facilities available for post-irradiation analysis, an alternative qualifying program for this fuel is proposed based on the same procedures used at IPEN since 1988 for qualifying its own U 3 O 8 -Al (1,9 and 2,3 gU/cm 3 ) and U 3 Si 2 -Al (3,0 gU/cm 3 ) dispersion fuels. The fuel miniplates and full-size fuel elements irradiations should be tested at IEA-R1 core. The fuel characterization along the irradiation time should be made by means of non-destructive methods, including periodical visual inspections with an underwater video camera system, sipping tests for fuel elements suspected of leakage, and underwater dimensional measurements for swelling evaluation, performed inside the reactor pool. This work presents the program description for the qualification of the high density nuclear fuel (U 3 Si 2 -Al) with 4,8 gU/cm 3 , and describes the IPEN fuel fabrication infrastructure and some basic features of the available systems for non-destructive tests at IEA-R1 research reactor. (author)

  14. Instrument for continuous supervision of the radioactivity of CO2 coolant in piles - DCCA -CO2 (1960)

    International Nuclear Information System (INIS)

    Fitoussi, L.

    1960-01-01

    This paper describes an apparatus for continuous measurement of CO 2 activity, which can be used on piles cooled by circulation of gas. The first part is devoted mainly to describing the apparatus used and the character of the radioactivity and thermodynamic measurements carried out, and giving the general characteristics of the gas circuit required if the instrument is to be suitably gas-tight. In the second part theoretical calculations are given, particularly on the determination of the ionisation current in an ionisation chamber with circulating gas. Several parameters enter into this determination, such as the mean path of β particles in the ionisation chamber, the linear number of ion pairs formed in the gas by these β particles as a function of their energy, the temperature and pressure of the gas in the ionisation chamber. This part also evaluates the sensitivity areas of the apparatus for measuring the concentrations of radioactive gases such as argon-41 and fission gases from uranium-235 in the CO 2 coolant. In the last part are described the results of measurements performed with such an apparatus on the pile EL2, the special investigations carried out on the CO 2 coolant of this pile, and the information gained during normal operation and during accidents. The DCCA - CO 2 which has just been put in operation at G2 is briefly presented. In the conclusion the possibilities offered by this apparatus are underlined. (author) [fr

  15. Research on Elemental Technology of Advanced Nuclear Fuel Performance Verification

    International Nuclear Information System (INIS)

    Kim, Yong Soo; Lee, Dong Uk; Jean, Sang Hwan; Koo, Min

    2003-04-01

    Most of current properties models and fuel performance models used in the performance evaluation codes are based on the in-pile data up to 33,000 MWd/MtU. Therefore, international experts are investigating the properties changes and developing advanced prediction models for high burn-up application. Current research is to develop high burn-up fission gas release model for the code and to support the code development activities by collecting data and models, reviewing/assessing the data and models together, and benchmarking the selected models against the appropriate in-pile data. For high burn-up applications, two stage two step fission gas release model is developed based on the real two diffusion process in the grain lattice and grain boundaries of the fission gases and the observation of accelerated release rate in the high burn-up. It is found that the prediction of this model is in excellent agreement with the in-pile measurement results, not only in the low burn-up but also in the high burn-up. This research is found that the importance of thermal conductivity of oxide fuel, especially in the high burn-up, is focused again. It is found that even the temperature dependent models differ from one to another and most of them overestimate the conductivity in the high burn-up. An in-pile data benchmarking of high LHGR fuel rod shows that the difference can reach 30%∼40%, which predicts 400 .deg. C lower than the real fuel centerline temperature. Recent models on the thermal expansion and heat capacity of oxide fuel are found to be well-defined. Irradiation swelling of the oxide fuel are now well-understood that in most cases in LWRs solid fission product swelling is dominant. Thus, the accumulation of in-pile data can enhance the accuracy of the model prediction, rather than theoretical modeling works. Thermo-physical properties of Zircaloy cladding are also well-defined and well-understood except the thermal expansion. However, it turns out that even the

  16. Optical Fier Based System for Multiple Thermophysical Properties for Glove Box, Hot Cell and In-Pile Application

    Energy Technology Data Exchange (ETDEWEB)

    Ban, Heng

    2017-11-30

    Thermal diffusivity of materials is of interest in nuclear applications at temperatures in excess of 2000°C. Commercial laser flash apparatus (LFA) that heats samples with a furnace typically do not reach these elevated temperatures nor are they easily adapted to a glove-box or hot cell environment. In this research, we performed work on an experimental technique using single laser surface heating, i.e. heating the disk sample only at its front surface with the continuous wave (CW) laser, to allow measurement of thermal diffusivity at very high temperatures within a small chamber. Thermal diffusivity is measured using a separate pulsed laser on the front side and IR detector on the rear side. The new way of heating provides easy operation in comparison to other heating methods. The measurement of sample reference temperature is needed for the measured thermal diffusivity. A theoretical model was developed to describe transient heat transfer across the sample due to the laser pulse, starting from the steady state temperature of the sample heated by the CW laser. The experimental setup was established with a 500W CW laser and maximum 50 Joule pulse laser irradiated at the front surface of the sample. The induced temperature rise at the rear surface, along with the steady-state temperature at the front surface, was recorded for the determination of thermal diffusivity and the sample temperature. Three samples were tested in vacuum over a wide temperature range of 500°C to 2100°C, including graphite, Inconel 600 and tungsten. The latter two samples were coated with sprayed graphite on their front surfaces in order to achieve surface absorption/emission needs, i.e. high absorptivity of the front surface against relatively low emissivity of the rear surface. Thermal diffusivity of graphite determined by our system are within a 5% difference of the commercial LFA data at temperatures below 1300°C and agree well with its trend at higher temperatures. Good agreement

  17. n⁺ GaAs/AuGeNi-Au Thermocouple-Type RF MEMS Power Sensors Based on Dual Thermal Flow Paths in GaAs MMIC.

    Science.gov (United States)

    Zhang, Zhiqiang; Liao, Xiaoping

    2017-06-17

    To achieve radio frequency (RF) power detection, gain control, and circuit protection, this paper presents n⁺ GaAs/AuGeNi-Au thermocouple-type RF microelectromechanical system (MEMS) power sensors based on dual thermal flow paths. The sensors utilize a conversion principle of RF power-heat-voltage, where a thermovoltage is obtained as the RF power changes. To improve the heat transfer efficiency and the sensitivity, structures of two heat conduction paths are designed: one in which a thermal slug of Au is placed between two load resistors and hot junctions of the thermocouples, and one in which a back cavity is fabricated by the MEMS technology to form a substrate membrane underneath the resistors and the hot junctions. The improved sensors were fabricated by a GaAs monolithic microwave integrated circuit (MMIC) process. Experiments show that these sensors have reflection losses of less than -17 dB up to 12 GHz. At 1, 5, and 10 GHz, measured sensitivities are about 63.45, 53.97, and 44.14 µ V/mW for the sensor with the thermal slug, and about 111.03, 94.79, and 79.04 µ V/mW for the sensor with the thermal slug and the back cavity, respectively.

  18. n+ GaAs/AuGeNi-Au Thermocouple-Type RF MEMS Power Sensors Based on Dual Thermal Flow Paths in GaAs MMIC

    Directory of Open Access Journals (Sweden)

    Zhiqiang Zhang

    2017-06-01

    Full Text Available To achieve radio frequency (RF power detection, gain control, and circuit protection, this paper presents n+ GaAs/AuGeNi-Au thermocouple-type RF microelectromechanical system (MEMS power sensors based on dual thermal flow paths. The sensors utilize a conversion principle of RF power-heat-voltage, where a thermovoltage is obtained as the RF power changes. To improve the heat transfer efficiency and the sensitivity, structures of two heat conduction paths are designed: one in which a thermal slug of Au is placed between two load resistors and hot junctions of the thermocouples, and one in which a back cavity is fabricated by the MEMS technology to form a substrate membrane underneath the resistors and the hot junctions. The improved sensors were fabricated by a GaAs monolithic microwave integrated circuit (MMIC process. Experiments show that these sensors have reflection losses of less than −17 dB up to 12 GHz. At 1, 5, and 10 GHz, measured sensitivities are about 63.45, 53.97, and 44.14 µV/mW for the sensor with the thermal slug, and about 111.03, 94.79, and 79.04 µV/mW for the sensor with the thermal slug and the back cavity, respectively.

  19. In pile helium loop ''COMEDIE''

    International Nuclear Information System (INIS)

    Abassin, J.J.; Blanchard, R.J.; Gentil, J.

    1981-01-01

    The SR1 test in the COMEDIE loop has permitted to demonstrate particularly the device operation reliability with a fuel loading. The post-irradiation examinations have pointed out the good filter efficiency and have enabled to determine the deposition profiles either for the activation products (e.g.: 51 Cr, 60 Co) or for the fission products (e.g.: sup(110m)Ag, 131 I, 134 Cs, 137 Cs). (author)

  20. Gamma thermometer longevity test: Laguna Verde 2 instruments recent performance

    International Nuclear Information System (INIS)

    Cuevas V, G.; Avila N, A.; Calleros M, G.

    2013-10-01

    This paper is informative of the General Electric Hitachi and Global Nuclear Fuel - Americas are collaboration with Comision Federal de Electricidad in a longevity test of thermocouples as power monitoring devices. The test conclusions will serve for final engineering design in detailing the Automated Fixed In-core Probes for calibration of the Local Power Range Monitors (LPRMs) of the Economic Simplified Boiling Water Reactor. This paper introduces the collaboration description and some recent performance evaluation of the thermocouples that are sensitive to gamma radiation and are known generically as Gamma Thermometers (G T). The G Ts in Laguna Verde 2 are radially located inside six instrumentation tubes in the core and consist of seven thermocouples, four are aligned with the LPRM heights and three are axially located between LPRM heights. The Laguna Verde 2 G T test has become the longest test of thermocouples as power monitoring devices in a BWR industry history and confirms their reliability in terms of time-dependent small noise under steady state reactor conditions and good agreement against Traversing In-core Probes power measurements. (Author)

  1. Gamma thermometer longevity test: Laguna Verde 2 instruments recent performance

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas V, G. [Global Nuclear Fuel, Americas, 3901 Castle Hayne Road, Wilmington, North Carolina (United States); Avila N, A.; Calleros M, G., E-mail: Gabriel.Cuevas-Vivas@gnf.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verda, Carretera Veracruz-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

    2013-10-15

    This paper is informative of the General Electric Hitachi and Global Nuclear Fuel - Americas are collaboration with Comision Federal de Electricidad in a longevity test of thermocouples as power monitoring devices. The test conclusions will serve for final engineering design in detailing the Automated Fixed In-core Probes for calibration of the Local Power Range Monitors (LPRMs) of the Economic Simplified Boiling Water Reactor. This paper introduces the collaboration description and some recent performance evaluation of the thermocouples that are sensitive to gamma radiation and are known generically as Gamma Thermometers (G T). The G Ts in Laguna Verde 2 are radially located inside six instrumentation tubes in the core and consist of seven thermocouples, four are aligned with the LPRM heights and three are axially located between LPRM heights. The Laguna Verde 2 G T test has become the longest test of thermocouples as power monitoring devices in a BWR industry history and confirms their reliability in terms of time-dependent small noise under steady state reactor conditions and good agreement against Traversing In-core Probes power measurements. (Author)

  2. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    International Nuclear Information System (INIS)

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C.

    2015-01-01

    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

  3. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

  4. Construção e avaliação de psicrômetro aspirado de termopar Construction and evaluation of an aspirated thermocouple psychrometer

    Directory of Open Access Journals (Sweden)

    Fábio Ricardo Marin

    2001-12-01

    Full Text Available Construiu-se um psicrômetro de termopar aspirado, de baixo custo e fácil utilização em sistemas automáticos de aquisição de dados, utilizando-se tubos de PVC. A aspiração foi feita por ventiladores utilizados em microcomputadores e as temperaturas foram determinadas com junções de termopar de cobre-constantan. Para umidecimento do bulbo, utilizou-se um cordão de algodão. Os resultados da comparação com higrômetro capacitivo Vaisala Inc. e com psicrômetro aspirado tipo Assman mostraram que tanto em ambientes naturais como em controlados, a precisão e a exatidão das medidas foi muito boa, de maneira que o psicrômetro aqui descrito pode ser empregado para determinação da pressão atual de vapor e da umidade relativa sem perda de qualidade dos dados, e também em estudos que levem em conta gradientes de temperatura e umidade específica.The construction of a low cost aspirated thermocouple psychrometer made of PVC tubes is described. The instrument can easily be connected to dataloggers. The aspiration is made by fans used in microcomputers and temperatures measured with cooper-constantan thermocouples. A cotton string was used to make the wet junction. Its perfomance was evaluated in comparison to an Assman aspirated psychrometer and a Vaisala Inc. capacitive higrometer, in natural and controlled environments. The results show a good agreement between measures, allowing air vapour, relative humidity, temperature and specific humidity gradients to be determined using the proposed psychrometer.

  5. PERFORMANCE

    Directory of Open Access Journals (Sweden)

    M Cilli

    2014-10-01

    Full Text Available This study aimed to investigate the kinematic and kinetic changes when resistance is applied in horizontal and vertical directions, produced by using different percentages of body weight, caused by jumping movements during a dynamic warm-up. The group of subjects consisted of 35 voluntary male athletes (19 basketball and 16 volleyball players; age: 23.4 ± 1.4 years, training experience: 9.6 ± 2.7 years; height: 177.2 ± 5.7 cm, body weight: 69.9 ± 6.9 kg studying Physical Education, who had a jump training background and who were training for 2 hours, on 4 days in a week. A dynamic warm-up protocol containing seven specific resistance movements with specific resistance corresponding to different percentages of body weight (2%, 4%, 6%, 8%, 10% was applied randomly on non consecutive days. Effects of different warm-up protocols were assessed by pre-/post- exercise changes in jump height in the countermovement jump (CMJ and the squat jump (SJ measured using a force platform and changes in hip and knee joint angles at the end of the eccentric phase measured using a video camera. A significant increase in jump height was observed in the dynamic resistance warm-up conducted with different percentages of body weight (p 0.05. In jump movements before and after the warm-up, while no significant difference between the vertical ground reaction forces applied by athletes was observed (p>0.05, in some cases of resistance, a significant reduction was observed in hip and knee joint angles (p<0.05. The dynamic resistance warm-up method was found to cause changes in the kinematics of jumping movements, as well as an increase in jump height values. As a result, dynamic warm-up exercises could be applicable in cases of resistance corresponding to 6-10% of body weight applied in horizontal and vertical directions in order to increase the jump performance acutely.

  6. Simulation of a welding process in polyduct pipelines resolved with a finite elements computational model. Comparison with analytical solutions and tests with thermocouples

    International Nuclear Information System (INIS)

    Sanzi, H; Elvira, G; Kloster, M; Asta, E; Zalazar, M

    2006-01-01

    All welding processes induce deformations and thermal tensions, which must be evaluated correctly since they can influence a component's structural integrity. This work determines the distribution of temperatures that develop during a manual welding process with shielded electrodes (SMAW), on the circumference seam of a pipe for use in polyducts. A simplified model of Finite Elements (FEA) using three dimensional solids is proposed for the study. The analysis considers that while the welding process is underway, no heat is lost into the environment, that is, adiabatic conditions are considered, and the transformations produced in the material due to phase changes do not produce modifications in the properties of the supporting or base materials. The results of the simulation are compared with those obtained by recent analytical studies developed by different investigators, such as Nguyen, Ohta, Matsuoka, Suzuki and Taeda, where a continuously moving three dimensional double ellipsoidal source was used. The results are then compared with the experimental results by measuring with thermocouples. This study reveals the sensitivity and the validity of the proposed computer model, and in a second stage optimizes the engineering times for the resolution of a problem like the one presented in order to design the corresponding welding procedure (CW)

  7. A novel approach for fault detection and classification of the thermocouple sensor in Nuclear Power Plant using Singular Value Decomposition and Symbolic Dynamic Filter

    International Nuclear Information System (INIS)

    Mandal, Shyamapada; Santhi, B.; Sridhar, S.; Vinolia, K.; Swaminathan, P.

    2017-01-01

    Highlights: • A novel approach to classify the fault pattern using data-driven methods. • Application of robust reconstruction method (SVD) to identify the faulty sensor. • Analysing fault pattern for plenty of sensors using SDF with less time complexity. • An efficient data-driven model is designed to the false and missed alarms. - Abstract: A mathematical model with two layers is developed using data-driven methods for thermocouple sensor fault detection and classification in Nuclear Power Plants (NPP). The Singular Value Decomposition (SVD) based method is applied to detect the faulty sensor from a data set of all sensors, at the first layer. In the second layer, the Symbolic Dynamic Filter (SDF) is employed to classify the fault pattern. If SVD detects any false fault, it is also re-evaluated by the SDF, i.e., the model has two layers of checking to balance the false alarms. The proposed fault detection and classification method is compared with the Principal Component Analysis. Two case studies are taken from Fast Breeder Test Reactor (FBTR) to prove the efficiency of the proposed method.

  8. Study and modelling of the in-pile densification of the UO{sub 2} and MO{sub x} nuclear oxides; Etude et modelisation de la densification en pile des oxydes nucleaires UO{sub 2} et MO{sub x}

    Energy Technology Data Exchange (ETDEWEB)

    Boulore, A

    2001-03-01

    Amongst the many phenomena which take place in the course of the irradiation of UO{sub 2} or (U, Pu)O{sub 2} nuclear fuels, one of them involves the elimination of a fraction of the as-fabricated porosity. In-pile densification or sintering can reach 2.5%, i.e. approximately half the initial volume of pores is likely to disappear. Our literature survey indicates that the amplitude and kinetics of the phenomenon are both heavily dependent on the initial fuel microstructure. Micro-structural characterisation techniques of oxide fuels have therefore been developed in conjunction with quantitative image analysis methods. The ensuing methodology enables a quantitative comparison of micro-structural features in different fuels and has been applied to ascertaining the influence of the local fission rate and temperature on in-pile densification. It is thus revealed that in-pile operation eliminates a significant fraction of pores smaller than 3 microns in diameter. The experimental data generated has been used to set up a semi-empirical and a mechanistic model. The former is based on experimental results and is not essentially predictive. The inability of this model to predict the in-pile densification of oxide fuels is illustrated by the fact that the maximum fraction of pores that disappears is proportional to an empirical function of fission rate, and temperature. The proportionality factor appears to be difficult to correlate quantitatively to any given micro-structural feature. The model has however been applied to the interpretation of an in-pile densification experiment carried out in the Halden reactor (Norway). The latter model is mechanistic, i.e. it is based on the solution to a set of equations that describe the coupled temperature and radiation induced phenomena which occur in-pile. These can broadly be broken down into three categories: the fission fragment-pore interaction, the creation of point defects as the fission fragments slow down, and the diffusion

  9. IAMBUS, a computer code for the design and performance prediction of fast breeder fuel rods

    International Nuclear Information System (INIS)

    Toebbe, H.

    1990-05-01

    IAMBUS is a computer code for the thermal and mechanical design, in-pile performance prediction and post-irradiation analysis of fast breeder fuel rods. The code deals with steady, non-steady and transient operating conditions and enables to predict in-pile behavior of fuel rods in power reactors as well as in experimental rigs. Great effort went into the development of a realistic account of non-steady fuel rod operating conditions. The main emphasis is placed on characterizing the mechanical interaction taking place between the cladding tube and the fuel as a result of contact pressure and friction forces, with due consideration of axial and radial crack configuration within the fuel as well as the gradual transition at the elastic/plastic interface in respect to fuel behavior. IAMBUS can be readily adapted to various fuel and cladding materials. The specific models and material correlations of the reference version deal with the actual in-pile behavior and physical properties of the KNK II and SNR 300 related fuel rod design, confirmed by comparison of the fuel performance model with post-irradiation data. The comparison comprises steady, non-steady and transient irradiation experiments within the German/Belgian fuel rod irradiation program. The code is further validated by comparison of model predictions with post-irradiation data of standard fuel and breeder rods of Phenix and PFR as well as selected LWR fuel rods in non-steady operating conditions

  10. High performance fuel technology development : Development of high performance cladding materials

    International Nuclear Information System (INIS)

    Park, Jeongyong; Jeong, Y. H.; Park, S. Y.

    2012-04-01

    The superior in-pile performance of the HANA claddings have been verified by the successful irradiation test and in the Halden research reactor up to the high burn-up of 67GWD/MTU. The in-pile corrosion and creep resistances of HANA claddings were improved by 40% and 50%, respectively, over Zircaloy-4. HANA claddings have been also irradiated in the commercial reactor up to 2 reactor cycles, showing the corrosion resistance 40% better than that of ZIRLO in the same fuel assembly. Long-term out-of-pile performance tests for the candidates of the next generation cladding materials have produced the highly reliable test results. The final candidate alloys were selected and they showed the corrosion resistance 50% better than the foreign advanced claddings, which is beyond the original target. The LOCA-related properties were also improved by 20% over the foreign advanced claddings. In order to establish the optimal manufacturing process for the inner and outer claddings of the dual-cooled fuel, 18 different kinds of specimens were fabricated with various cold working and annealing conditions. Based on the performance tests and various out-of-pile test results obtained from the specimens, the optimal manufacturing process was established for the inner and outer cladding tubes of the dual-cooled fuel

  11. Effect of thermo-coupled processes on the behaviour of a clay barrier submitted to heating and hydration

    Directory of Open Access Journals (Sweden)

    Marcelo Sánchez

    2010-03-01

    Full Text Available The storage of high level radioactive waste is still an unresolved problem of the nuclear industry, being geological disposal the most favoured option and, naturally, the one requiring the strongest geo-mechanical input. Most conceptual designs for the deep geological disposal of nuclear waste envisage placing the canisters containing the waste in horizontal drifts or vertical boreholes. The empty space surrounding the canisters is filled by an engineered barrier often made up of compacted swelling clay. Inthebarrierandthenearfield,significantthermo-hydro-mechanical(THM phenomena take place that interact in a complex way. A good understanding of THM issues is, therefore, necessary to ensure a correct performance of engineered barriers and seals. The conditions of the bentonite in an engineered barrier for high-level radioactive waste disposal are being simulated in a mock-up heating test at almost scale, at the premises of CIEMAT in Madrid. The evolution of the main Thermo-Hydro-Mechanical (THM variables of this test are analysed in this paper by using a fully coupled THM formulation and the corresponding finite element code. Special emphasis has been placed on the study of the effect of thermo-osmotic flow in the hydration of the clay barrier at an advanced staged of the experiment.O armazenamento de rejeitos altamente radioativos é ainda um problema em aberto na área de engenharia nuclear sendo os sítios geológicos ainda a opção mais favorável e naturalmente aquela que demanda maior conhecimento na área de geomecânica. A maioria dos projetos conceituais de armazenamento do lixo nuclear objetiva a alocação de cilindros que contêm os rejeitos em poços verticais ou horizontais. O espaço vazio que circunda os cilindros é preenchido por uma barreirade engenharia na maioria dos casos composta por uma argila expansiva. Na barreira e na vizinhança fenômenos significativos de acoplamento termo-hidro-mecânico(THM tomam lugar e

  12. The development of an electrically compensated differential calorimeter for the measurement of in-pile heat evolution (1962); Etude et realisation d'un calorimetre differentiel a compensation electrique pour la mesure des echaufpements en pile (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Ayela, F; Derrien, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1962-07-01

    It has been possible with the calorimeter of which we give the principle and details in the present report, to determine the calorific power produced by nuclear radiation in a given sample using simple electrical measurement. It also renders unnecessary all absolute measurements of temperature, all possible calibration, and the use of the thermal or thermoelectric constants of the constituent elements. The possible uses of the apparatus are fairly varied, as shown by the first in-pile experiments which we have carried out and for which we give results. (authors) [French] Le calorimetre, dont nous decrivons le principe et les details dans le present rapport, nous permet de determiner, par de simples mesures electriques, la puissance calorifique liberee par les rayonnements nucleaires dans un echantillon donne. Il nous affranchit aussi, de toute mesure absolue de temperature, de tout etalonnage prealable, et de l'utilisation des constantes thermiques ou thermoelectriques de ses elements constitutifs. Les possibilites d'emploi de l'appareil sont assez variees comme nous le montrent les premiers essais que nous avons effectues en pile et dont nous donnons les resultats. (auteurs)

  13. Effects of fuel particle size and fission-fragment-enhanced irradiation creep on the in-pile behavior in CERCER composite pellets

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yunmei [Institute of Mechanics and Computational Engineering, Department of Aeronautics and Astronautics, Fudan University, Shanghai 200433 (China); Ding, Shurong, E-mail: dsr1971@163.com [Institute of Mechanics and Computational Engineering, Department of Aeronautics and Astronautics, Fudan University, Shanghai 200433 (China); Zhang, Xunchao; Wang, Canglong; Yang, Lei [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China)

    2016-12-15

    The micro-scale finite element models for CERCER pellets with different-sized fuel particles are developed. With consideration of a grain-scale mechanistic irradiation swelling model in the fuel particles and the irradiation creep in the matrix, numerical simulations are performed to explore the effects of the particle size and the fission-fragment-enhanced irradiation creep on the thermo-mechanical behavior of CERCER pellets. The enhanced irradiation creep effect is applied in the 10 μm-thick fission fragment damage matrix layer surrounding the fuel particles. The obtained results indicate that (1) lower maximum temperature occurs in the cases with smaller-sized particles, and the effects of particle size on the mechanical behavior in pellets are intricate; (2) the first principal stress and radial axial stress remain compressive in the fission fragment damage layer at higher burnup, thus the mechanism of radial cracking found in the experiment can be better explained. - Highlights: • A grain-scale gas swelling model considering the development of recrystallization and resolution is adopted for particles. • The influence of fission-gas-induced porosity is considered in the constitutive relations for particles. • A simulation method is developed for the multi-scale thermo-mechanical behavior. • The effects of fuel particle size and fission-fragment-enhanced irradiation creep are investigated in pellets.

  14. In-pile Tritium Permeation through F82H Steel with and without a Ceramic Coating of Cr2O3-SiO2 Including CrPO4

    International Nuclear Information System (INIS)

    Nakamichi, M.; Hayashi, K.; Kulsartov, T.V.; Afanasyev, S.E.; Shestakov, V.P.; Chikhray, Y.V.; Kenzhin, E.A.; Kolbaenkov, A.N.

    2006-01-01

    Development of coating on blanket structural materials with significant reduction capability of tritium permeation is highly required in order to realize a reasonable design of a tritium recovery and processing system of demonstration (DEMO) fusion reactors. An effective coating has been developed in Japan Atomic Energy Agency (JAEA) using a ceramic material of Cr 2 O 3 -SiO 2 including CrPO 4 . In previous out-of-pile deuterium permeation experiments at 600 o C [T.V. Kulsartov et al., Fusion Eng. Des. 81 (2006) 701], a significant permeation reduction factor (PFR) of about 300 was obtained for the coating on the inner-side surface of tubular diffusion cells made by ferritic steel (F82H). In the present study, in-pile experiments on tritium permeation were conducted for F82H steel with and without the same coating, using a testing reactor IGV-1M in Kazakhstan. The tritium source used was liquid lithium-lead eutectics, Pb17Li, which was poured into a space around a tubular diffusion cell (specimen) of F82H steel with or without the coating on the inner side the cell. The irradiation time was about 4 hours, which corresponds to a fast-neuron fluence of about 2x10 21 m -2 (E > 1.1 MeV). The permeation reduction factor (PRF) was obtained by comparison of kinetics curves of tritium permeation through the diffusion cell of F82H steel with and without the coating. The PRFs at 600 and 500 o C were 292 and 30, respectively. These values are close to corresponding PRF values of 307 and 45, which had been obtained at 600 and 500 o C, respectively, in the previous out-of-pile experiments [T.V. Kulsartov et al., Fusion Eng. Des. 81 (2006) 701]. (author)

  15. High performance fuel technology development

    Energy Technology Data Exchange (ETDEWEB)

    Koon, Yang Hyun; Kim, Keon Sik; Park, Jeong Yong; Yang, Yong Sik; In, Wang Kee; Kim, Hyung Kyu [KAERI, Daejeon (Korea, Republic of)

    2012-01-15

    {omicron} Development of High Plasticity and Annular Pellet - Development of strong candidates of ultra high burn-up fuel pellets for a PCI remedy - Development of fabrication technology of annular fuel pellet {omicron} Development of High Performance Cladding Materials - Irradiation test of HANA claddings in Halden research reactor and the evaluation of the in-pile performance - Development of the final candidates for the next generation cladding materials. - Development of the manufacturing technology for the dual-cooled fuel cladding tubes. {omicron} Irradiated Fuel Performance Evaluation Technology Development - Development of performance analysis code system for the dual-cooled fuel - Development of fuel performance-proving technology {omicron} Feasibility Studies on Dual-Cooled Annular Fuel Core - Analysis on the property of a reactor core with dual-cooled fuel - Feasibility evaluation on the dual-cooled fuel core {omicron} Development of Design Technology for Dual-Cooled Fuel Structure - Definition of technical issues and invention of concept for dual-cooled fuel structure - Basic design and development of main structure components for dual- cooled fuel - Basic design of a dual-cooled fuel rod.

  16. In-pile tritium permeation experiment

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Miller, L.G.; Watts, K.D.; Kershner, C.J.; Rogers, M.L.

    1982-01-01

    The experiments in progress are examining various aspects of the permeation of hydrogen isotopes through fusion materials. Of particular importance will be the measurement of permeation due to ion implantation in the presence of a neutron radiation field. Theoretical and early experimental results for these experiments have suggested that sufficient tritium will permeate fusion reactor interior structures that development of a permeation barrier will be needed. (orig.)

  17. In-pile tritium permeation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Longhurst, G.R.; Miller, L.G.; Watts, K.D. (Idaho National Engineering Lab., Idaho Falls (USA)); Kershner, C.J.; Rogers, M.L. (Monsanto Research Corp., Miamisburg, OH (USA). Mound Facility)

    The experiments in progress are examining various aspects of the permeation of hydrogen isotopes through fusion materials. Of particular importance will be the measurement of permeation due to ion implantation in the presence of a neutron radiation field. Theoretical and early experimental results for these experiments have suggested that sufficient tritium will permeate fusion reactor interior structures that development of a permeation barrier will be needed.

  18. Specific methods on in-pile gammametry

    International Nuclear Information System (INIS)

    Pointud, M.-L.; Michel, Francois.

    1979-01-01

    The gammametry technique in nuclear research reactors has evolved by the adequation of its means and the quality of its results since its beginnings in 1972. We do not propose here to make a detailed presentation, nor describe the kinds of well known results that can henceforth be attained with it in a conventional manner; our intention, on the other hand, is to describe a few specific methods developed for using it in the SILOE reactor of the CEN/G [fr

  19. In-pile tritium permeation experiment

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Miller, L.G.; Watts, K.D.; Kershner, C.J.; Rogers, M.L.

    1982-01-01

    To examine radiation and implantation effects simultaneously, an experiment has been designed which makes use of the Coupled Fast Reactivity Measurement Facility (CFRMF), a small pool reactor at the INEL. The neutron flux is low in this reactor, but the high cross section (5300 b) for the 3 He(n,p) 3 H reaction with thermal neutrons gives a sufficiently intense flux of protons and tritons to a simulated fusion first wall for meaningful results

  20. COMMIX analysis of four constant flow thermal upramp experiments performed in a thermal hydraulic model of an advanced LMR

    International Nuclear Information System (INIS)

    Yarlagadda, B.S.

    1989-04-01

    The three-dimensional thermal hydraulics computer code COMMIX-1AR was used to analyze four constant flow thermal upramp experiments performed in the thermal hydraulic model of an advanced LMR. An objective of these analyses was the validation of COMMIX-1AR for buoyancy affected flows. The COMMIX calculated temperature histories of some thermocouples in the model were compared with the corresponding measured data. The conclusions of this work are presented. 3 refs., 5 figs

  1. Some aspects of in-pile swelling of fissile materials, 1. part: non-alloyed {alpha} uranium; Quelques aspects du gonflement en pile des materiaux fissiles. 1. partie: uranium {alpha} non allie

    Energy Technology Data Exchange (ETDEWEB)

    Mikailoff, H [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-07-01

    An examination has been carried out of non-alloyed uranium samples, having various structural states, cold-worked and recrystallized, as-cast and {beta}-treated, and irradiated at temperatures of between 450 and 600 C and with burn-ups from 1300 to 5500 MW days/metric ton. These samples swelled because of precipitation of the fission gases the porosity thus produced has a morphology depending mainly on the type of deformation to which the metal has been subjected and which is due to in-pile growth. The most homogeneous distribution of pores, and thus that leading to the minimum swelling, is only observed in the material having a marked [010] texture in which the growth and perhaps the thermal cycling introduce little or no strain. For other materials the deformation /swelling association causes a more rapid destruction of the samples either by cracking when the deformation is due to twinning, or by pronounced swelling localized in the bands when deformation is due to slipping. Finally the fission-gas precipitation considerably facilitates, above 500 C, the germination and growth of the intergranular cracks which can then develop at low stresses. (author) [French] On a examine des echantillons d'uranium non allie, de divers etats structuraux, marteles et recristallises, bruts de coulee et traites {beta}, irradies a des temperatures comprises entre 450 et 600 C, et a des taux de combustion allant de 1300 a 5500 MWj/t. Ces echantillons ont gonfle par suite de la precipitation de gaz de fission: la porosite ainsi fournie a une morphologie qui depend principalement des modes de deformation subie par le metal et due a la croissance en pile. La repartition la plus homogene des pores, donc celle qui donnera le gonflement minimum, est observee seulement dans le materiau a forte texture [010] dans lequel la croissance et eventuellement le cyclage thermique introduisent peu ou pas de contraintes. Dans les autres materiaux l'association deformation/gonflement rend plus rapide

  2. The statistical analysis techniques to support the NGNP fuel performance experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh T., E-mail: Binh.Pham@inl.gov; Einerson, Jeffrey J.

    2013-10-15

    This paper describes the development and application of statistical analysis techniques to support the Advanced Gas Reactor (AGR) experimental program on Next Generation Nuclear Plant (NGNP) fuel performance. The experiments conducted in the Idaho National Laboratory’s Advanced Test Reactor employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule. The tests are instrumented with thermocouples embedded in graphite blocks and the target quantity (fuel temperature) is regulated by the He–Ne gas mixture that fills the gap volume. Three techniques for statistical analysis, namely control charting, correlation analysis, and regression analysis, are implemented in the NGNP Data Management and Analysis System for automated processing and qualification of the AGR measured data. The neutronic and thermal code simulation results are used for comparative scrutiny. The ultimate objective of this work includes (a) a multi-faceted system for data monitoring and data accuracy testing, (b) identification of possible modes of diagnostics deterioration and changes in experimental conditions, (c) qualification of data for use in code validation, and (d) identification and use of data trends to support effective control of test conditions with respect to the test target. Analysis results and examples given in the paper show the three statistical analysis techniques providing a complementary capability to warn of thermocouple failures. It also suggests that the regression analysis models relating calculated fuel temperatures and thermocouple readings can enable online regulation of experimental parameters (i.e. gas mixture content), to effectively maintain the fuel temperature within a given range.

  3. Postirradiation examination results for the Irradiation Effects Scoping Test 2

    International Nuclear Information System (INIS)

    Mehner, A.S.

    1977-01-01

    The postirradiation examination results are reported for two rods from the second scoping test (IE-ST-2) of the Nuclear Regulatory Commission Irradiation Effects Program. The rods were irradiated in the in-pile test loop of the Power Burst Facility at the Idaho National Engineering Laboratory. Rod IE-005 was fabricated from fresh fuel and cladding previously irradiated in the Saxton Reactor. Rod IE-006, fabricated from fresh fuel and unirradiated cladding, was equipped with six developmental cladding surface thermocouples. The rods were preconditioned, power ramped, and then subjected to film boiling operation. The performance of the rods and the developmental thermocouples are evaluated from the post irradiation examination results. The effects of prior irradiation damage in cladding are discussed in relation to fuel rod behavior during a power ramp and subsequent film boiling operation

  4. Development of high performance cladding materials

    International Nuclear Information System (INIS)

    Park, Jeong Yong; Jeong, Y. H.; Park, S. Y.

    2010-04-01

    The irradiation test for HANA claddings conducted and a series of evaluation for next-HANA claddings as well as their in-pile and out-of pile performances tests were also carried out at Halden research reactor. The 6th irradiation test have been completed successfully in Halden research reactor. As a result, HANA claddings showed high performance, such as corrosion resistance increased by 40% compared to Zircaloy-4. The high performance of HANA claddings in Halden test has enabled lead test rod program as the first step of the commercialization of HANA claddings. DB has been established for thermal and LOCA-related properties. It was confirmed from the thermal shock test that the integrity of HANA claddings was maintained in more expanded region than the criteria regulated by NRC. The manufacturing process of strips was established in order to apply HANA alloys, which were originally developed for the claddings, to the spacer grids. 250 kinds of model alloys for the next-generation claddings were designed and manufactured over 4 times and used to select the preliminary candidate alloys for the next-generation claddings. The selected candidate alloys showed 50% better corrosion resistance and 20% improved high temperature oxidation resistance compared to the foreign advanced claddings. We established the manufacturing condition controlling the performance of the dual-cooled claddings by changing the reduction rate in the cold working steps

  5. Irradiated fuel performance evaluation technology development

    International Nuclear Information System (INIS)

    Koo, Yang Hyun; Bang, J. G.; Kim, D. H.

    2012-01-01

    Alpha version performance code for dual-cooled annular fuel under steady state operation, so called 'DUOS', has been developed applying performance models and proposed methodology. Furthermore, nonlinear finite element module which could be integrated into transient/accident fuel performance code was also developed and evaluated using commercial FE code. The first/second irradiation and PIE test of annular pellet for dual-cooled annular fuel in the world have been completed. In-pile irradiation test DB of annular pellet up to burnup of 10,000 MWd/MTU through the 1st test was established and cracking behavior of annular pellet and swelling rate at low temperature were studied. To do irradiation test of dual-cooled annular fuel under PWR's simulating steady-state conditions, irradiation test rig/rod design/manufacture of mock-up/performance test have been completed through international collaboration program with Halden reactor project. The irradiation test of large grain pellets has been continued from 2002 to 2011 and completed successfully. Burnup of 70,000 MWd/MTU which is the highest burnup among irradiation test pellets in domestic was achieved

  6. Pyrexia's effect on the CBG-cortisol thermocouple, rather than CBG cleavage, elevates the acute free cortisol response to TNF-α in humans

    DEFF Research Database (Denmark)

    Nenke, Marni Anne; Nielsen, Signe Tellerup; Lehrskov, Louise Lang

    2017-01-01

    an inflammatory stimulus is unknown. Hence our aim was to determine the immediate effect of the key pro-inflammatory cytokine TNF-α on CBG levels and cleavage. We performed a crossover study of 12 healthy males receiving a TNF-α versus saline infusion, measuring total CBG, haCBG, laCBG and free and total cortisol...... hourly for 6 h. There was no change in total CBG or haCBG levels in the first 6 h of inflammation between the groups, suggesting that CBG cleavage is not activated nor is hepatic CBG production affected by TNF-α in this time frame. There was an early increase in the ratio of free:total cortisol...

  7. Reference Material Properties and Standard Problems to Verify the Fuel Performance Models Ver 1.0

    International Nuclear Information System (INIS)

    Yang, Yong Sik; Kim, Jae Yong; Koo, Yang Hyun

    2010-12-01

    All fuel performance models must be validated by in-pile and out-pile tests. However, the model validation requires much efforts and times to confirm its exactness. In many fields, new performance models and codes are confirmed by code-to-code benchmarking process under simplified standard problem analysis. At present, the DUOS, which is the steady state fuel performance analysis code for dual cooled annular fuel, development project is progressing and new FEM module is developed to analyze the fuel performance during transient period. In addition, the verification process is planning to examine the new models and module's rightness by comparing with commercial finite element analysis such as a ADINA, ABAQUS and ANSYS. This reports contains the result of unification of material properties and establishment of standard problem to verify the newly developed models with commercial FEM code

  8. Code structure for U-Mo fuel performance analysis in high performance research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Cho, Tae Won; Lee, Chul Min; Sohn, Dong Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    A performance analysis modeling applicable to research reactor fuel is being developed with available models describing fuel performance phenomena observed from in-pile tests. We established the calculation algorithm and scheme to best predict fuel performance using radio-thermo-mechanically coupled system to consider fuel swelling, interaction layer growth, pore formation in the fuel meat, and creep fuel deformation and mass relocation, etc. In this paper, we present a general structure of the performance analysis code for typical research reactor fuel and advanced features such as a model to predict fuel failure induced by combination of breakaway swelling and pore growth in the fuel meat. Thermo-mechanical code dedicated to the modeling of U-Mo dispersion fuel plates is being under development in Korea to satisfy a demand for advanced performance analysis and safe assessment of the plates. The major physical phenomena during irradiation are considered in the code such that interaction layer formation by fuel-matrix interdiffusion, fission induced swelling of fuel particle, mass relocation by fission induced stress, and pore formation at the interface between the reaction product and Al matrix.

  9. The Statistical Analysis Techniques to Support the NGNP Fuel Performance Experiments

    International Nuclear Information System (INIS)

    Pham, Bihn T.; Einerson, Jeffrey J.

    2010-01-01

    This paper describes the development and application of statistical analysis techniques to support the AGR experimental program on NGNP fuel performance. The experiments conducted in the Idaho National Laboratory's Advanced Test Reactor employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule. The tests are instrumented with thermocouples embedded in graphite blocks and the target quantity (fuel/graphite temperature) is regulated by the He-Ne gas mixture that fills the gap volume. Three techniques for statistical analysis, namely control charting, correlation analysis, and regression analysis, are implemented in the SAS-based NGNP Data Management and Analysis System (NDMAS) for automated processing and qualification of the AGR measured data. The NDMAS also stores daily neutronic (power) and thermal (heat transfer) code simulation results along with the measurement data, allowing for their combined use and comparative scrutiny. The ultimate objective of this work includes (a) a multi-faceted system for data monitoring and data accuracy testing, (b) identification of possible modes of diagnostics deterioration and changes in experimental conditions, (c) qualification of data for use in code validation, and (d) identification and use of data trends to support effective control of test conditions with respect to the test target. Analysis results and examples given in the paper show the three statistical analysis techniques providing a complementary capability to warn of thermocouple failures. It also suggests that the regression analysis models relating calculated fuel temperatures and thermocouple readings can enable online regulation of experimental parameters (i.e. gas mixture content), to effectively maintain the target quantity (fuel temperature) within a given range.

  10. Study of a thermocouple with refractory joints

    International Nuclear Information System (INIS)

    Vandsmael, C.

    1977-01-01

    After having discussed the relationship between the transverse profile, the crown of the rolls and the flatness of the strips, it is proposed to determine the profile indirectly from a measurement of crown or flatness. The crown measurement was attempted by means of contactless transducers fixed on two movable frames, placed on both sides of the rolls to be measured. The method was tested in a 4-high stand at C.R.M. Calibrations were also carried out on industrial rolls. The effect of temperature, as well as the nature and state of the industrial roll surface, did not favour the use of this method for applications requiring both precision and reliability. On the contrary, the first industrial measurements of strip flatness have proved more promising. Tests were made at the discharge from the last finishing stand of the wide-strip mill of S.A. SIDMAR by means of an optical method. By triangulation with the aid of lasers and photodiode grids, local undulations were measured, sometimes at the edges and sometimes at the centre of the strip, even when the strip was under tension. These first successes ought to be followed by the construction and operation of a gauge functioning permanently on the mill, within the framework of a new research

  11. Mechanical performance of irradiated beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Dalle-Donne, M.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik

    1998-01-01

    For the Helium Cooled Pebble Bed (HCPB) Blanket, which is one of the two reference concepts studied within the European Fusion Technology Programme, the neutron multiplier consists of a mixed bed of about 2 and 0.1-0.2 mm diameter beryllium pebbles. Beryllium has no structural function in the blanket, however microstructural and mechanical properties are important, as they might influence the material behavior under neutron irradiation. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating it. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from these irradiation experiments, emphasizing the effects of irradiation of essential material properties and trying to elucidate the processes controlling the property changes. The microstructure, the porosity distribution, the impurity content, the behavior under compression loads and the compatibility of the beryllium pebbles with lithium orthosilicate (Li{sub 4}SiO{sub 4}) during the in-pile irradiation are presented and critically discussed. Qualitative information on ductility and creep obtained by hardness-type measurements are also supplied. (author)

  12. Study of a device for the direct measurement of the fission gas pressure inside an in-pile fuel element; Etude d'un dispositif pour la mesure directe de la pression des gaz de fission a l'interieur d'un element combustible en pile

    Energy Technology Data Exchange (ETDEWEB)

    Lavaud, B; Uschanoff, S [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The fission gas pressure inside a fuel element made of a refractory fuel constitutes an important limiting factor for the burn-up. Although it is possible to calculate approximately the volume of gas produced outside the fuel during its life-time; it is nevertheless very difficult to evaluate the pressure since the volume allowed to the fission gases, as well as their temperature are known only very approximately. This physical value, which is essential for the technologist, can only be known by direct in-pile measurement of the pressure. The report describes the equipment which has been developed for this test. (authors) [French] La pression des gaz de fission a l'interieur d'un element combustible a combustible refractaire constitue une des limitations importantes du taux de combustion. Si on peut approcher par calcul la determination du volume, des gaz degages hors du combustible au cours de sa vie, il est par contre tres difficile d'evaluer la pression car le volume alloue aux gaz de fission et leur temperature sont tres mal connus. Cette donnee essentielle pour le technologue ne peut etre atteinte que par une mesure directe en pile de la pression. Le rapport decrit l'appareillage qui a ete mis au point pour cet essai. (auteurs)

  13. Performing Performance Design Anglonationally

    DEFF Research Database (Denmark)

    2016-01-01

    Video recording of pecha kucha style bricolage aural enactment of an international version of performance design......Video recording of pecha kucha style bricolage aural enactment of an international version of performance design...

  14. Improved performance of U-Mo dispersion fuel by Si addition in Al matrix.

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y S; Hofman, G L [Nuclear Engineering Division

    2011-06-01

    The purpose of this report is to collect in one publication and fit together work fragments presented in many conferences in the multi-year time span starting 2002 to the present dealing with the problem of large pore formation in U-Mo/Al dispersion fuel plates first observed in 2002. Hence, this report summarizes the excerpts from papers and reports on how we interpreted the relevant results from out-of-pile and in-pile tests and how this problem was dealt with. This report also provides a refined view to explain in detail and in a quantitative manner the underlying mechanism of the role of silicon in improving the irradiation performance of U-Mo/Al.

  15. Effects of UO2 fuel microstructure and density on fuel in-reactor performance

    International Nuclear Information System (INIS)

    Hansson, L.

    1988-02-01

    The volume changes of UO 2 fuel pellets, produced by neutron irradiation, can be characterized by two processes: fission spike induced densification through pore skrinkage and later fission produced induced swelling of UO 2 matrix. In-pile densification is controlled by the initial density and microstructure of the fuel, particularly by the pore size distribution. The extent of swelling depends mainly on the amount of fission products produced, but the fission gas release as well as the swelling may be reduced by increasing the grain size of UO 2 . Fabrication of fuel pellets having certain in-reactor properties requires detailed knowledge of the effects of individual fabrication parameters. The irradiation experience of fuels fabricated by using different conversion and pelletizing methods is extensive. Based on this experience, some general characteristics of stable/well-performing fuel microstructures have been summarized

  16. Comparison of the cladding deformation measured during the Power Burst Facility loss-of-coolant accident in-pile experiments with recent Oak Ridge National Laboratory out-of-pile results

    International Nuclear Information System (INIS)

    Broughton, J.M.; McCardell, R.K.; MacDonald, P.E.

    1981-01-01

    A series of four large break loss-of-coolant accident fuel behavior experiments have been performed in the Power Burst Facility. The results of these experiments are briefly reviewed and compared with results from the ORNL multirod burst test program. The effect of cladding burst temperature and prior irradiation were investigated. The cladding strain of the previously irradiated test rods was more uniformly distributed around the cladding circumference and larger than for similar unirradiated test rods. The ORNL out-of-pile single rod test results are in good agreement with the Power Burst Facility (PBF) test results with unirradiated test rods, and the ORNL out-of-pile, single-rod test results with heated shrouds and the PBF test results with previously irradiated test rods are comparable

  17. Prediction of the Long Term Cooling Performance for the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, H. R

    2005-12-15

    In the long term cooling phase that the emergency cooling water injection ends, the performance of the residual heat removal for the 3-pin fuel test loop has been predicted by a simplified heat transfer model. In the long term cooling phase the residual heat is 1323W for PWR fuel test mode and 1449W for CANDU fuel test mode. The each residual heat is assumed as 2% of the fission power of the test fuel used in the anticipated operational occurrence and design basis accident analyses. The each fission power used for the analyses is 105% of the rated fission power in the normal operation. In the long term cooling phase the residual heat is removed to the HANARO pool through the double pressure vessels of the in-pile test section. Saturate pooling boiling is assumed on the test fuel and condensation heat transfer is expected on the inner wall of the fuel carrier and the flow divider. Natural convection heat transfer on a heated vertical wall is also assumed on the outer wall of the outer pressure vessel. The conduction heat transfer is only considered in the gap between the double pressure vessels charged with neon gas and in the downcomer filled with coolant. The heat transfer rate between the coolant temperature of 152 .deg. C in the in-pile test section and the water temperature of 45 .deg. C in the HANARO pool is predicted as about 1666W. The 152 .deg. C is the saturate temperature of the coolant pressure predicted from the MARS code. The cooling capacity of 1666W is greater than the residual heats of 1323W and 1449W. Consequently the long term cooling performance of the 3-pin fuel test loop is sufficient for the anticipated operational occurrences and design basis accidents.

  18. Instrument response during overpower transients at TREAT

    International Nuclear Information System (INIS)

    Meek, C.C.; Bauer, T.H.; Hill, D.J.; Froehle, P.H.; Klickman, A.E.; Tylka, J.P.; Doerner, R.C.; Wright, A.E.

    1982-01-01

    A program to empirically analyze data residuals or noise to determine instrument response that occurs during in-pile transient tests is out-lined. As an example, thermocouple response in the Mark III loop during a severe overpower transient in TREAT is studied both in frequency space and in real-time. Time intervals studied included both constant power and burst portions of the power transient. Thermocouple time constants were computed. Benefits and limitations of the method are discussed

  19. Performance of Polycrystalline Photovoltaic and Thermal Collector (PVT on Serpentine-Parallel Absorbers Design

    Directory of Open Access Journals (Sweden)

    Mustofa Mustofa

    2017-03-01

    Full Text Available This paper presents the performance of an unglazed polycrystalline photovoltaic-thermal PVT on 0.045 kg/s mass flow rate. PVT combine photovoltaic modules and solar thermal collectors, forming a single device that receive solar radiation and produces heat and electricity simultaneously. The collector figures out serpentine-parallel tubes that can prolong fluid heat conductivity from morning till afternoon. During testing, cell PV, inlet and outlet fluid temperaturs were recorded by thermocouple digital LM35 Arduino Mega 2560. Panel voltage and electric current were also noted in which they were connected to computer and presented each second data recorded. But, in this performance only shows in the certain significant time data. This because the electric current was only noted by multimeter device not the digital one. Based on these testing data, average cell efficieny was about 19%, while thermal efficiency of above 50% and correspondeng cell efficiency of 11%, respectively

  20. Performance of Polycrystalline Photovoltaic and Thermal Collector (PVT on Serpentine-Parallel Absor

    Directory of Open Access Journals (Sweden)

    Mustofa

    2015-10-01

    Full Text Available This paper presents the performance of an unglazed polycrystalline photovoltaic-thermal PVT on 0.045 kg/s mass flow rate. PVT combine photovoltaic modules and solar thermal collectors, forming a single device that receive solar radiation and produces heat and electricity simultaneously. The collector figures out serpentine-parallel tubes that can prolong fluid heat conductivity from morning till afternoon. During testing, cell PV, inlet and outlet fluid temperatures were recorded by thermocouple digital LM35 Arduino Mega 2560. Panel voltage and electric current were also noted in which they were connected to computer and presented each second data recorded. But, in this performance only shows in the certain significant time data. This because the electric current was only noted by multimeter device not the digital one. Based on these testing data, average cell efficiency was about 19%, while thermal efficiency of above 50% and correspondent cell efficiency of 11%, respectively.

  1. In-pile irradiation of rock-like oxide fuels

    International Nuclear Information System (INIS)

    Nitani, N.; Kuramoto, K.; Yamashita, T.; Nakano, Y.; Akie, H.

    2001-01-01

    Five kinds of ROX fuels were prepared and irradiated using 20% enriched U instead of Pu. Non-destructive and destructive post-irradiation examinations were carried out. FP gas release rates of the particle-dispersed type fuels and homogeneously-blended type fuels were larger than that of the Yttria-stabilized zirconia containing UO 2 single phase fuel. From results of SEM and EPMA, decomposition of the spinel was observed. The decomposition of the spinel is probably avoided by lowering the irradiation temperature, less than 1700 K. The regions suffering the irradiation damage of the particle dispersed type fuels were less than those of the homogeneously-blended type fuels. (author)

  2. In-pile irradiation of rock-like oxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nitani, N.; Kuramoto, K.; Yamashita, T.; Nakano, Y.; Akie, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2001-07-01

    Five kinds of ROX fuels were prepared and irradiated using 20% enriched U instead of Pu. Non-destructive and destructive post-irradiation examinations were carried out. FP gas release rates of the particle-dispersed type fuels and homogeneously-blended type fuels were larger than that of the Yttria-stabilized zirconia containing UO{sub 2} single phase fuel. From results of SEM and EPMA, decomposition of the spinel was observed. The decomposition of the spinel is probably avoided by lowering the irradiation temperature, less than 1700 K. The regions suffering the irradiation damage of the particle dispersed type fuels were less than those of the homogeneously-blended type fuels. (author)

  3. Synergistic smart fuel for in-pile nuclear reactor measurements

    Energy Technology Data Exchange (ETDEWEB)

    Smith, J.A.; Kotter, D.K. [Idaho National Laboratories, Idaho Falls (United States); Ali, R.A.; Garrett, S.L. [Penn State University, University Park, State College, PA 16801 (United States)

    2013-07-01

    The thermo-acoustic fuel rod sensor developed in this research has demonstrated a novel technique for monitoring the temperature within the core of a nuclear reactor or the temperature of the surrounding heat-transfer fluid. It uses the heat from the nuclear fuel to generate sustained acoustic oscillations whose frequency will be indicative of the temperature. Converting a nuclear fuel rod into this type of thermo-acoustic sensor simply requires the insertion of a porous material (stack). This sensor has demonstrated a synergy with the elevated temperatures that exist within the nuclear reactor using materials that have only minimal susceptibility to high-energy particle fluxes. When the sensor is in operation, the sound waves radiated from the fuel rod resonator will propagate through the surrounding cooling fluid. The frequency of these oscillations is directly correlated with an effective temperature within the fuel rod resonator. This device is self-powered and is operational even in case of total loss of power of the reactor.

  4. Synergistic Smart Fuel For In-pile Nuclear Reactor Measurements

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; Dale K. Kotter; Randall A. Ali; Steven L . Garrett

    2013-10-01

    In March 2011, an earthquake of magnitude 9.0 on the Richter scale struck Japan with its epicenter on the northeast coast, near the Tohoku region. In addition to the immense physical destruction and casualties across the country, several nuclear power plants (NPP) were affected. It was the Fukushima Daiichi NPP that experienced the most severe and irreversible damage. The earthquake brought the reactors at Fukushima to an automatic shutdown and because the power transmission lines were damaged, emergency diesel generators (EDGs) were activated to ensure that there was continued cooling of the reactors and spent fuel pools. The situation was being successfully managed until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to the reactors.2 At this point, the situation became critical. There was a loss of the sensors and instrumentation within the reactor that could have provided valuable information to guide the operators to make informed decisions and avoid the unfortunate events that followed. In the light of these events, we have developed and tested a potential self-powered thermoacoustic system, which will have the ability to serve as a temperature sensor and can transmit data independently of electronic networks. Such a device is synergistic with the harsh environment of the nuclear reactor as it utilizes the heat from the nuclear fuel to provide the input power.

  5. European in-pile investigations on fast breeder reactor safety

    International Nuclear Information System (INIS)

    Bailly, J.; Penet, F.; Teague, H.J.

    1977-01-01

    Because of the difficulties faced by the various organisations responsible for the design and safety analysis of fast reactors, in the conception and execution of perfectly representative experiments corresponding to the various hypothetical accidents considered, the European experts in this field have chosen to adopt a more fundamental approach. By the coordination of the efforts in the different countries and pooling the results, it appears that sufficient data can be assembled to permit them to: isolate the influence of the principle parameters on the course of an accident and to have an appreciation of the possibilities for detection and propagation; and validate calculational models of the essential phenomena, and determine the values of the adjustable parameters used in accident codes. The present paper does not propose to give an exhaustive description of the European experimental programmes, nor of the possibilities for exploiting the results in safety analysis. The aim is simply to show with the aid of results already obtained or expected from the immediate programme, that it should be possible to answer certain fundamental questions concerning those accidents at present considered most important. Particular consideration is given to local blockages (SCARABEE, DFR and MOL 7C programmes), unprotected loss of pumps leading to cooling failure, and inlet blockage (SCARABEE and CABRI), and reactivity excursions (CABRI and VIPER). In addition it is planned to study, as part of the SCARABEE programme, the long-term post-accident behaviour of fuel in some simple configurations. It is shown that these various results should lead to a great improvement in our understanding of the progress of the hypothetical accidents taken into account in the design of fast breeder reactors [fr

  6. Analysis of in-pile tritium release experiments

    International Nuclear Information System (INIS)

    Kopasz, J.P.; Tam, S.W.; Johnson, C.E.

    1992-01-01

    The objective of this work is to characterize tritium release behavior from lithium ceramics and develop insight into the underlying tritium release mechanisms. Analysis of tritium release data from recent laboratory experiments with lithium aluminate has identified physical processes which were previously unaccounted for in tritium release models. A new model that incorporates the recent data and provides for release from multiple sites rather than only one site was developed. Calculations of tritium release using this model are in excellent agreement with the tritium release behavior reported for the MOZART experiment

  7. Evaluation of Candidate In-Pile Thermal Conductivity Techniques

    International Nuclear Information System (INIS)

    Fox, B.; Ban, H.; Daw, J.; Condie, K.; Knudson, D.; Rempe, J.

    2009-01-01

    Thermophysical properties of materials must be known for proper design, test, and application of new fuels and structural properties in nuclear reactors. In the case of nuclear fuels during irradiation, the physical structure and chemical composition change as a function of time and position within the rod. Typically, thermal conductivity changes, as well as other thermophysical properties being evaluated during irradiation in a materials and test reactor, are measured out-of-pile in 'hot-cells'. Repeatedly removing samples from a test reactor to make out-of-pile measurements is expensive, has the potential to disturb phenomena of interest, and only provide understanding of the sample's end state at the time each measurement is made. There are also limited thermophysical property data for advanced fuels. Such data are needed for the development of next generation reactors and advanced fuels for existing nuclear plants. Having the capacity to effectively and quickly characterize fuels and material properties during irradiation has the potential to improve the fidelity of nuclear fuel data and reduce irradiation testing costs

  8. Aluminum cladding oxidation of prefilmed in-pile fueled experiments

    Energy Technology Data Exchange (ETDEWEB)

    Marcum, W.R., E-mail: marcumw@engr.orst.edu [Oregon State University, School of Nuclear Science and Engineering, 116 Radiation Center, Corvallis, OR 97331 (United States); Wachs, D.M.; Robinson, A.B.; Lillo, M.A. [Idaho National Laboratory, Nuclear Fuels & Materials Department, 2525 Fremont Ave., Idaho Falls, ID 83415 (United States)

    2016-04-01

    A series of fueled irradiation experiments were recently completed within the Advanced Test Reactor Full size plate In center flux trap Position (AFIP) and Gas Test Loop (GTL) campaigns. The conduct of the AFIP experiments supports ongoing efforts within the global threat reduction initiative (GTRI) to qualify a new ultra-high loading density low enriched uranium-molybdenum fuel. This study details the characterization of oxide growth on the fueled AFIP experiments and cross-correlates the empirically measured oxide thickness values to existing oxide growth correlations and convective heat transfer correlations that have traditionally been utilized for such an application. This study adds new and valuable empirical data to the scientific community with respect to oxide growth measurements of highly irradiated experiments, of which there is presently very limited data. Additionally, the predicted oxide thickness values are reconstructed to produce an oxide thickness distribution across the length of each fueled experiment (a new application and presentation of information that has not previously been obtainable in open literature); the predicted distributions are compared against experimental data and in general agree well with the exception of select outliers. - Highlights: • New experimental data is presented on oxide layer thickness of irradiated aluminum fuel. • Five oxide growth correlations and four convective heat transfer correlations are used to compute the oxide layer thickness. • The oxide layer thickness distribution is predicted via correlation for each respective experiment. • The measured experiment and predicted distributions correlate well, with few outliers.

  9. Modeling and performance analysis of a concentrated photovoltaic–thermoelectric hybrid power generation system

    International Nuclear Information System (INIS)

    Lamba, Ravita; Kaushik, S.C.

    2016-01-01

    Highlights: • Thermodynamic model of concentrated photovoltaic–thermoelectric system is analysed. • Thomson effect reduces the power output of PV, TE and hybrid PV–TEG system. • Effect of thermocouple number, irradiance, PV and TE current have been studied. • The optimum concentration ratio for maximum power output has been found out. • The overall efficiency and power output of hybrid PV–TEG system has been improved. - Abstract: In this study, a thermodynamic model for analysing the performance of a concentrated photovoltaic–thermoelectric generator (CPV–TEG) hybrid system including Thomson effect in conjunction with Seebeck, Joule and Fourier heat conduction effects has been developed and simulated in MATALB environment. The expressions for calculating the temperature of photovoltaic (PV) module, hot and cold sides of thermoelectric (TE) module are derived analytically as well. The effect of concentration ratio, number of thermocouples in TE module, solar irradiance, PV module current and TE module current on power output and efficiency of the PV, TEG and hybrid PV–TEG system have been studied. The optimum concentration ratio corresponding to maximum power output of the hybrid system has been found out. It has been observed that by considering Thomson effect in TEG module, the power output of the PV, TE and hybrid PV–TEG systems decreases and at C = 1 and 5, it reduces the power output of hybrid system by 0.7% and 4.78% respectively. The results of this study may provide basis for performance optimization of a practical irreversible CPV–TEG hybrid system.

  10. Development and research of in-core transducers at IAE (Institute of Atomic Energy)

    International Nuclear Information System (INIS)

    Huang Yucai; Qian Shunfa; Jia Guozhen

    1989-10-01

    The development of in-core transducers at IAE (Institute of Atomic Energy) and their applications in in-pile fuel assembly test are mentioned. These transducers include mainly tubed tungsten-rhenium thermocouple assembly, displacement transducer of linear variable differential transformer, pressure transducer of membrane type, gamma thermometer, turbine flow meter, self-powered neutron detector etc

  11. Development of reactor water level sensor for extreme conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miura, K; Ogasawara, T [Sukegawa Electric Co., Ltd., Hitachi, Ibaraki (Japan); Shibata, Akira; Nakamura, Jinichi; Saito, Takashi; Tsuchiya, Kunihiko [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    In the Fukushima accident, measurement failure of water level was one of the most important factors which caused serious situation. The differential pressure type water level indicators are widely used in various place of nuclear power plant but after the accident of TMI-2, the need of other reliable method has been required. The BICOTH type and the TRICOTH type water level indicator for light water power reactors had been developed for in-pile water level indicator but currently those are not adopted to nuclear power plant. In this study, the development of new type water level indicator composed of thermocouple and heater is described. Demonstration test and characteristic evaluation of the water level indicator were performed and we had obtained satisfactory results. (author)

  12. Isotopic modelling using the ENIGMA-B fuel performance code

    International Nuclear Information System (INIS)

    Rossiter, G.D.; Cook, P.M.A.; Weston, R.

    2001-01-01

    A number of experimental programmes by BNFL and other MOX fabricators have now shown that the in-pile performance of MOX fuel is generally similar to that of conventional UO 2 fuel. Models based on UO 2 fuel experience form a good basis for a description of MOX fuel behaviour. However, an area where the performance of MOX fuel is sufficiently different from that of UO 2 to warrant model changes is in the radial power and burnup profile. The differences in radial power and burnup profile arise from the presence of significant concentrations of plutonium in MOX fuel, at beginning of life, and their subsequent evolution with burnup. Amongst other effects, plutonium has a greater neutron absorption cross-section than uranium. This paper focuses on the development of a new model for the radial power and burnup profile within a UO 2 or MOX fuel rod, in which the underlying fissile isotope concentration distributions are tracked during irradiation. The new model has been incorporated into the ENIGMA-B fuel performance code and has been extended to track the isotopic concentrations of the fission gases, xenon and krypton. The calculated distributions have been validated against results from rod puncture measurements and electron probe micro-analysis (EPMA) linescans, performed during the M501 post irradiation examination (PIE) programme. The predicted gas inventory of the fuel/clad gap is compared with the isotopic composition measured during rod puncture and the measured radial distributions of burnup (from neodymium measurements) and plutonium in the fuel are compared with the calculated distributions. It is shown that there is good agreement between the code predictions and the measurements. (author)

  13. Environmental Performance

    DEFF Research Database (Denmark)

    Svabo, Connie; Lindelof, Anja Mølle

    from the perspective of time and liveness as experienced in art on environmental performance discussing how environmental performances frame the temporality of the world. The paper engages with contemporary examples of environmental performances from various disciplines (sound, video, television...

  14. Real-time measurements of temperature, pressure and moisture profiles in High-Performance Concrete exposed to high temperatures during neutron radiography imaging

    Energy Technology Data Exchange (ETDEWEB)

    Toropovs, N., E-mail: nikolajs.toropovs@rtu.lv [Empa, Swiss Federal Laboratories for Materials Science and Technology, Dübendorf (Switzerland); Riga Technical University, Institute of Materials and Structures, Riga (Latvia); Lo Monte, F. [Politecnico di Milano, Department of Civil and Environmental Engineering, Milan (Italy); Wyrzykowski, M. [Empa, Swiss Federal Laboratories for Materials Science and Technology, Dübendorf (Switzerland); Lodz University of Technology, Department of Building Physics and Building Materials, Lodz (Poland); Weber, B. [Empa, Swiss Federal Laboratories for Materials Science and Technology, Dübendorf (Switzerland); Sahmenko, G. [Riga Technical University, Institute of Materials and Structures, Riga (Latvia); Vontobel, P. [Paul Scherrer Institute, Laboratory for Neutron Scattering and Imaging, Villigen (Switzerland); Felicetti, R. [Politecnico di Milano, Department of Civil and Environmental Engineering, Milan (Italy); Lura, P. [Empa, Swiss Federal Laboratories for Materials Science and Technology, Dübendorf (Switzerland); ETH Zürich, Institute for Building Materials (IfB), Zürich (Switzerland)

    2015-02-15

    High-Performance Concrete (HPC) is particularly prone to explosive spalling when exposed to high temperature. Although the exact causes that lead to spalling are still being debated, moisture transport during heating plays an important role in all proposed mechanisms. In this study, slabs made of high-performance, low water-to-binder ratio mortars with addition of superabsorbent polymers (SAP) and polypropylene fibers (PP) were heated from one side on a temperature-controlled plate up to 550 °C. A combination of measurements was performed simultaneously on the same sample: moisture profiles via neutron radiography, temperature profiles with embedded thermocouples and pore pressure evolution with embedded pressure sensors. Spalling occurred in the sample with SAP, where sharp profiles of moisture and temperature were observed. No spalling occurred when PP-fibers were introduced in addition to SAP. The experimental procedure described here is essential for developing and verifying numerical models and studying measures against fire spalling risk in HPC.

  15. Environmental Performance

    DEFF Research Database (Denmark)

    Lindelof, Anja Mølle; Schmidt, Ulrik; Svabo, Connie

    2017-01-01

    Do ants and grasshoppers perform? Do clouds, plants and melting ice? Do skyscrapers, traffic jams and computer vira? And what happens to our understanding of liveness if that is the case? This chapter takes ongoing theoretical disputes about the nature of live performance in performance studies...... as its starting point to investigate liveness within a specific kind of contemporary performance: ‘environmental performances’. Environmental performances are arts practices that take environmental processes as their focus by framing activities of non-human performers such as clouds, wind and weeds - key...

  16. Performing Identities

    DEFF Research Database (Denmark)

    von Wallpach, Sylvia; Hemetsberger, Andrea; Espersen, Peter

    2017-01-01

    performative approaches to branding, this study applies a performativity theory perspective. Brand performances—encompassing playing and liking, basement building and showcasing, creating and innovating, community building and facilitating, storytelling, missionizing, and marketplace developing—exhibit generic...

  17. Dj Performance

    DEFF Research Database (Denmark)

    Dj Performance at a late concert at The Hub, Plymouth, in support of Sileni, Superconductor and others.......Dj Performance at a late concert at The Hub, Plymouth, in support of Sileni, Superconductor and others....

  18. Combustion performance evaluation of air staging of palm oil blends.

    Science.gov (United States)

    Mohd Jaafar, Mohammad Nazri; Eldrainy, Yehia A; Mat Ali, Muhammad Faiser; Wan Omar, W Z; Mohd Hizam, Mohd Faizi Arif

    2012-02-21

    The problems of global warming and the unstable price of petroleum oils have led to a race to develop environmentally friendly biofuels, such as palm oil or ethanol derived from corn and sugar cane. Biofuels are a potential replacement for fossil fuel, since they are renewable and environmentally friendly. This paper evaluates the combustion performance and emission characteristics of Refined, Bleached, and Deodorized Palm Oil (RBDPO)/diesel blends B5, B10, B15, B20, and B25 by volume, using an industrial oil burner with and without secondary air. Wall temperature profiles along the combustion chamber axis were measured using a series of thermocouples fitted axially on the combustion chamber wall, and emissions released were measured using a gas analyzer. The results show that RBDPO blend B25 produced the maximum emission reduction of 56.9% of CO, 74.7% of NOx, 68.5% of SO(2), and 77.5% of UHC compared to petroleum diesel, while air staging (secondary air) in most cases reduces the emissions further. However, increasing concentrations of RBDPO in the blends also reduced the energy released from the combustion. The maximum wall temperature reduction was 62.7% for B25 at the exit of the combustion chamber.

  19. Diagnostics for Evaluating Performance of NSTX Liquid Lihium Divertor

    Science.gov (United States)

    Kaita, R.; Kugel, H.; Kallman, J.; Leblanc, B.; Paul, S.; Roquemore, A. L.; Skinner, C.; Soukhanovskii, V.; Maingi, R.; Ahn, J.-W.; Wilgen, J.; Allain, J.-P.; Taylor, C.

    2009-11-01

    A Liquid Lithium Divertor (LLD) is being installed on NSTX to investigate particle control and power handling with liquid lithium as plasma-facing component (PFC). The LLD is expected to provide a low-recycling plasma-facing component (PFC). To study the effects of such a PFC on plasma performance, a variety of edge measurements are required. Since its surface is highly reflective at visible wavelengths, a Lyman-alpha detector array will be used to monitor the recycling. To understand changes in edge transport, electron temperature and density measurements will be made with Langmuir probes mounted in PFC's near the LLD, and the edge sightlines of a multipoint Thomson scattering system. A frequency-scanning reflectometer will also provide scrapeoff layer electron density profiles. The LLD response to heat loads will be examined with infrared cameras and thermocouples. Diagnostics are also needed to measure the erosion and codeposition of lithium. They include quartz deposition monitors and a retractable probe for exposing samples to the plasma.

  20. Performance analysis

    International Nuclear Information System (INIS)

    2008-05-01

    This book introduces energy and resource technology development business with performance analysis, which has business division and definition, analysis of current situation of support, substance of basic plan of national energy, resource technique development, selection of analysis index, result of performance analysis by index, performance result of investigation, analysis and appraisal of energy and resource technology development business in 2007.

  1. Performative Work

    DEFF Research Database (Denmark)

    Beunza, Daniel; Ferraro, Fabrizio

    2018-01-01

    by attending to the normative and regulative associations of the device. We theorize this route to performativity by proposing the concept of performative work, which designates the necessary institutional work to enable translation and the subsequent adoption of the device. We conclude by considering...... the implications of performative work for the performativity and the institutional work literatures.......Callon’s performativity thesis has illuminated how economic theories and calculative devices shape markets, but has been challenged for its neglect of the organizational, institutional and political context. Our seven-year qualitative study of a large financial data company found that the company...

  2. A study on KMRR utilization for fuel development

    International Nuclear Information System (INIS)

    Kang, Young Hwan; Ryu, Woo Seog; Park, Ji Yeon; Joo, Kee Nam; Park, Jong Man; Park, Se Jin

    1991-01-01

    The most effective utilization scheme of the KMRR was studied in the field of nuclear fuel development through reviewing literatural documents on irradiation facilities and in-pile test. It is suggested that the KMRR should be used for verification tests of advanced fuels and for power ramping / cycling tests of fuel rods. In addition, the characterization tests for fuel development and the basic material research should be also performed. In-pile loops for fuel verification and/or power ramping / cycling tests are proposed to be installed in advance, and capsules are necessary for power ramping / cycling tests, fuel characterization tests and / or material tests. Instrumentation technologies for thermocouple, SPND (Self-Powered Neutron Detector) and pressure transducer, and the in-situ dimensional measuring systems have to be developed to obtain the useful and various results from irradiation tests in the KMRR. A mock-up test rod for characterizing fuel thermal response was manufactured and the related technologies as well as the design specification were set up. An equipment for microdrilling and grooving of fuel pellets and an apparatus for diffusion-bonding between zircaloy-4 and stainless steel were made. A study to verify the integrity of test rod weldments is presented using out-of pile corrosion test. (Author)

  3. The data requirements for the verification and validation of a fuel performance code - the transuranus perspective

    International Nuclear Information System (INIS)

    Schubert, A.; Di Marcello, V.; Rondinella, V.; Van De Laar, J.; Van Uffelen, P.

    2013-01-01

    In general, the verification and validation (V and V) of a fuel performance code like TRANSURANUS consists of three basic steps: a) verifying the correctness and numerical stability of the sub-models; b) comparing the sub-models with experimental data; c) comparing the results of the integral fuel performance code with experimental data Only the second and third steps of the V and V rely on experimental information. This scheme can be further detailed according to the physical origin of the data: on one hand, in-reactor ('in-pile') experimental data are generated in the course of the irradiation; on the other hand ex-reactor ('out-of-pile') experimental data are obtained for instance from various postirradiation examinations (PIE) or dedicated experiments with fresh samples. For both categories, we will first discuss the V and V of sub-models of TRANSURANUS related to separate aspects of the fuel behaviour: this includes the radial variation of the composition and fissile isotopes, the thermal properties of the fuel (e.g. thermal conductivity, melting temperature, etc.), the mechanical properties of fuel and cladding (e.g. elastic constants, creep properties), as well as the models for the fission product behaviour. Secondly, the integral code verification will be addressed as it treats various aspects of the fuel behaviour, including the geometrical changes in the fuel and the gas pressure and composition of the free volume in the rod. (authors)

  4. Gender & performance

    NARCIS (Netherlands)

    Röttger, K.; Buchheim, E.; Groot, M.; Jonker, E.; Müller-Schirmer, A.; de Vos, M.; Walhout, E.; van der Zande, H.

    2012-01-01

    This Yearbook for Women’s History (Jaarboek voor Vrouwengeschiedenis) examines the theme of gender and performance. It is supervised by guest editor Kati Röttger, professor in Theatre Studies at the University of Amsterdam. The term performance - a temporary and active presentation, expression, or

  5. Performing compliance

    DEFF Research Database (Denmark)

    Wimmelmann, Camilla Lawaetz

    2017-01-01

    the local policy workers front-staged some practices in the implementation process and back-staged others. The local policy workers deliberately performed ‘guideline compliance’ by using information control and impression management techniques. The findings suggest that local guideline compliance should...... be regarded as a staged performance in which deliberate techniques are used to produce and manage certain impressions of compliance....

  6. School Performance

    Science.gov (United States)

    Lamas, Héctor A.

    2015-01-01

    The school performance study of students is, due to its relevance and complexity, one of the issues of major controversy in the educational research, and it has been given special attention in the last decades. This study is intended to show a conceptual approach to the school performance construct, contextualizing the reality in the regular basic…

  7. Aesthetic Performance

    DEFF Research Database (Denmark)

    Landgrebe, Jeanette

    2013-01-01

    -verbal actions, gaze orientation, active and static interactional strategies and props. From the data investigated, it seems that the performance act is divided into different stages which each calls for different strategies: the group's initiation of the entire performance act reveals that the group stand out......This article deals with how an aesthetic performance is enacted and coordinated by a performance group attracting attention and engaging commuters in a public space. Multimodal interactional resources and the way they are coordinated by interactants are investigated, and include verbal and non...... as uncoordinated and it may have a significance for whether the 'street' performers manage to stay in character or not. Once attention from commuters is obtained, a continued gaze from these commuters opens up for subsequent interaction, which then ultimately may result in the successful handing over of a card...

  8. Organizational Performance

    Directory of Open Access Journals (Sweden)

    Renata Peregrino de Brito

    2016-01-01

    Full Text Available This paper presents a theoretical and empirical analysis of the relationship between human resource management (HRM and organizational performance. Theoretically, we discuss the importance of HRM for the development of resources and its impact on business performance. Empirically, we evaluated articles published on Brazilian academic journals that addressed such relationships. The results showed a lack of studies conducted at this intersection. From the universe of 2,469 articles, only 16 (0.6% sought to relate HRM and organizational performance. We observed a dominance of isolated HR practices, which does not consider HRM as a system, and of operational performance measures, relative to financial and efficiency variables. Most studies show a positive relationship between HRM practices and performance, in line with the literature. However, we point out some methodological issues, such as the difficulty of isolating the HR practices from its context, the failure to consider the temporality of this relationship, and the comparison between companies from different industries.

  9. School Performance

    Directory of Open Access Journals (Sweden)

    Héctor A. Lamas

    2015-03-01

    Full Text Available The school performance study of students is, due to its relevance and complexity, one of the issues of major controversy in the educational research, and it has been given special attention in the last decades. This study is intended to show a conceptual approach to the school performance construct, contextualizing the reality in the regular basic education classrooms. The construct of learning approaches is presented as one of the factors that influences the school performance of students. Besides, an outlook of the empirical research works related to variables that are presented as relevant when explaining the reason for a specific performance in students is shown. Finally, some models and techniques allowing an appropriate study of school performance are presented.

  10. Performances of new reconstruction algorithms for CT-TDLAS (computer tomography-tunable diode laser absorption spectroscopy)

    International Nuclear Information System (INIS)

    Jeon, Min-Gyu; Deguchi, Yoshihiro; Kamimoto, Takahiro; Doh, Deog-Hee; Cho, Gyeong-Rae

    2017-01-01

    Highlights: • The measured data were successfully used for generating absorption spectra. • Four different reconstruction algorithms, ART, MART, SART and SMART were evaluated. • The calculation speed of convergence by the SMART algorithm was the fastest. • SMART was the most reliable algorithm for reconstructing the multiple signals. - Abstract: Recent advent of the tunable lasers made to measure simultaneous temperature and concentration fields of the gases. CT-TDLAS (computed tomography-tunable diode laser absorption spectroscopy) is one the leading techniques for the measurements of temperature and concentration fields of the gases. In CT-TDLAS, the accuracies of the measurement results are strongly dependent upon the reconstruction algorithms. In this study, four different reconstruction algorithms have been tested numerically using experimental data sets measured by thermocouples for combustion fields. Three reconstruction algorithms, MART (multiplicative algebraic reconstruction technique) algorithm, SART (simultaneous algebraic reconstruction technique) algorithm and SMART (simultaneous multiplicative algebraic reconstruction technique) algorithm, are newly proposed for CT-TDLAS in this study. The calculation results obtained by the three algorithms have been compared with previous algorithm, ART (algebraic reconstruction technique) algorithm. Phantom data sets have been generated by the use of thermocouples data obtained in an actual experiment. The data of the Harvard HITRAN table in which the thermo-dynamical properties and the light spectrum of the H_2O are listed were used for the numerical test. The reconstructed temperature and concentration fields were compared with the original HITRAN data, through which the constructed methods are validated. The performances of the four reconstruction algorithms were demonstrated. This method is expected to enhance the practicality of CT-TDLAS.

  11. Key differences in the fabrication of US and German TRISO-coated particle fuel, and their implications on fuel performance

    International Nuclear Information System (INIS)

    Petti, D.A.; Buongiorno, J.; Maki, J.T.; Miller, G.K.; Hobbins, R.R.

    2002-01-01

    Historically, the irradiation performance of TRISO-coated gas reactor particle fuel in Germany has been superior to that in the US. German fuel generally displayed in-pile gas release values that were three orders of magnitude lower than US fuel. Thus, we have critically examined the TRISO-coated fuel fabrication processes in the US and German and the associated irradiation database with a goal of understanding why the German fuel behaves acceptably, why the US fuel has not faired as well, and what process/production parameters impart the reliable performance to this fuel form. The postirradiation examination results are also reviewed to identify failure mechanisms that may be the cause of the poorer US irradiation performance. This comparison will help determine the roles that particle fuel process/product attributes and irradiation conditions (burnup, fast neutron fluence, temperature, degree of acceleration, power per particle) have on the behavior of the fuel during irradiation and provide a more quantitative linkage between acceptable processing parameters, as-fabricated fuel properties and subsequent in-reactor performance. (author)

  12. Performance managenemt

    DEFF Research Database (Denmark)

    Jacobi, Claus Brygger

    This paper attempts to identify barriers that prevent performance management from being genuinely result-based. By observing what happened when a Danish workfare reform was implemented by applying performance management, it becomes apparent that there exists internal decouplings on and between two...... levels; a decoupling between the monitoring/evaluation of established performance indicators and the revising of these for policy-making on future interventions, and a decoupling between the strategic political/administrative level and operational street-level, inhibiting its adaption to local...

  13. Performative Environments

    DEFF Research Database (Denmark)

    Thomsen, Bo Stjerne

    2008-01-01

    The paper explores how performative architecture can act as a collective environment localizing urban flows and establishing public domains through the integration of pervasive computing and animation techniques. The NoRA project introduces the concept of ‘performative environments,' focusing on ...... of local interactions and network behaviour, building becomes social infrastructure and prompts an understanding of architectural structures as quasiobjects, which can retain both variation and recognisability in changing social constellations.......The paper explores how performative architecture can act as a collective environment localizing urban flows and establishing public domains through the integration of pervasive computing and animation techniques. The NoRA project introduces the concept of ‘performative environments,' focusing...

  14. High Speed Surface Thermocouples Interface to Wireless Transmitters

    Science.gov (United States)

    2017-03-15

    above engines , or virtually anywhere manufacturers. Electric power plants Inside the mechanisms producing the power Kiln manufacturers, artists...ORGANIZATION REPORT NUMBER 9. SPONSORING/MONITORING AGENCY NAME(S) AND ADDRESS(ES) 10. SPONSOR/MONITOR’S ACRONYM(S) US Army Tank- automotive Research...Development & Engineering Center TARDEC Warren , Michigan 48397-5000 • 11 . SPONSOR/MONITOR’S REPORT NUMBER(S) 12. DISTRIBUTION/AVAILABILITY

  15. Performance assessment

    International Nuclear Information System (INIS)

    Doe, T.

    1985-01-01

    The purpose of performance assessment is to show that the repository is expected to serve its stated function - disposing of radioactive waste safely both during operation and for the postclosure period. Performance assessment is a straightforward concept, but its application may be very complicated. The concept of performance assessment has been clarified by the Nuclear Regulatory Commission (NRC) in their Draft Generic Technical Position on Licensing Assessment Methodology for High-Level Waste Geologic Repositories (NRC, 1984). This document has gone a long way toward defining the criteria that the NRC will use to determine whether or not information from site characterization is adequate to meet the regulations of the Nuclear Regulatory Commission and the Environmental Protection Agency (EPA). A favorable determination is required for issuance of a construction authorization, which is the first major regulatory requirement for developing a working repository. It is, therefore, essential that a research program be developed that not only resolves the outstanding technical issues, but also does it in such a way that the results are clearly applicable to the formal performance assessment and licensing procedures. The definitions of performance assessment are reviewed and the current NRC thinking is summarized

  16. Computer modelling of water reactor fuel element performance and life time

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Golovnin, I.S.; Elesin, V.F.

    1983-01-01

    Well calibrated models and methods of calculation permit the confident prediction of fuel element behaviour under most different operational conditions; based on the prediction of this kind one can improve designs and fuel element behaviour. Therefore, in the Soviet Union in the development of reactor cores for NPP one of the leading parts is given to design problems associated with computer modelling of fuel element performance and reliability. Special attention is paid to methods of calculation that permit the prediction of fuel element behaviour under conditions which either make experimental studies very complicated (practically impossible) or require laborious and expensive in-pile tests. Primarily it concerns accidents of different types, off-normal conditions, transients, fuel element behaviour at high burn-up, when an accumulation of a great amount of fission fragments is accompanied by changes in physical and mechanical properties as induced by irradiation damage, mechanical fatigue, physical and chemical reactions with a coolant, fission products etc. Some major computer modelling programs for the prediction of water reactor fuel behaviour are briefly described below and tendencies in the further development of work in this area are summarized

  17. Performative Silences

    DEFF Research Database (Denmark)

    Dupret, Katia

    2018-01-01

    static nor neutral. It has performative effects. Silencing as an act, rather than a noun, is conceptualised as a central ‘configurating actor’ of change. Through the description of minute details from a videotaped supervision session in the mental healthcare sector, it is shown how different performative...... configurations of silence makes people relate to each other in new ways and influence new work practices. In spite of its somewhat immaterial connotations, using an Actor-Network Theory approach to organization studies, silencing is conceptualised as both a means and an effect of change efforts, which are socio...

  18. Precision Pulse Capsulotomy: Preclinical Safety and Performance of a New Capsulotomy Technology.

    Science.gov (United States)

    Chang, David F; Mamalis, Nick; Werner, Liliana

    2016-02-01

    To assess the preclinical safety and performance of a new precision pulse capsulotomy (PPC) method. Human cadaver eye studies and surgical, slit-lamp, and histopathologic evaluation in a consecutive series of 20 live rabbits. Human cadaver eyes and New Zealand white rabbits. Precision pulse capsulotomy uses a highly focused, fast, multipulse, low-energy discharge to produce a perfectly round anterior capsulotomy instantaneously and simultaneously along all 360°. Capsulotomies are performed using a disposable handpiece with a soft collapsible tip and circular nitinol cutting element. Miyake-Apple imaging and scanning electron microscopy (SEM) of PPC were conducted in human cadaver eyes. Surgical, postoperative slit-lamp, and histopathologic assessments of PPC were performed in 20 live rabbits and were compared with manual continuous curvilinear capsulorrhexis (CCC) in the fellow eye. Anterior chamber (AC) thermocouple temperature measurements were evaluated in a subset of rabbit eyes. Capsulotomy edge circularity, SEM morphologic features and zonular movement with PPC in human cadaver eyes. Anterior chamber temperature during PPC and grading of ocular inflammation, corneal endothelial damage, anterior capsular opacification (ACO), and posterior capsular opacification (PCO). Miyake-Apple imaging showed minimal zonular stress, and thermocouple measurements demonstrated negligible AC temperature changes during PPC. Precision pulse capsulotomy produced round, complete capsulotomies in all 20 rabbit eyes, leading to successful in-the-bag intraocular lens (IOL) implantation. Slit-lamp examinations at 3 days and 1, 2, and 4 weeks after surgery showed no significant differences between PPC and CCC in corneal edema, AC inflammatory reaction, capsular fibrosis, ACO, and PCO. Postmortem studies showed no difference in the corneal endothelium between PPC and CCC eyes. All IOLs were well centered in PPC eyes, and histopathologic analysis showed no greater inflammatory

  19. Performing Brexit

    DEFF Research Database (Denmark)

    Adler-Nissen, Rebecca; Galpin, Charlotte; Rosamond, Ben

    2017-01-01

    constructed from the outside. Brexit signifies more than the technical complexities of the UK withdrawing from the European Union. It works both as a promise of a different future and performatively to establish a particular past. Brexit works as a frame with potential to shape perceptions in three domains...

  20. Performance Design

    DEFF Research Database (Denmark)

    Svabo, Connie

    Contribution to conference: Art and Presence The emerging field of Performance Design is unfolded as a bastard form of research/art/design/practice, with shapeshifting, monstruous, hybrid and transformational qualities. The potential for presencing, which emerges out of momentarily transgressing...

  1. Urban performances

    DEFF Research Database (Denmark)

    Samson, Kristine

    2012-01-01

    Through three different urban performances, the paper investigates how, when and under which circumstances urban space is transformed and distorted from its every day use and power relations. Distortion is an annual street festival in Copenhagen with the objective to distort the functional city...... creates an intensive space for the empowerment and liberation of the body. Occupy Wall street and its action in the autumn 2001 is the ultimate example of how urban political performances intensifies and transform every day spaces. Through examples of how OWS tactically appropriates and transforms urban...... space, I seek to show how representational space, for instance the public square, is transformed and distorted by heterogeneous and unforeseen modes of operating. Despite differing in their goal and outset, I wish to unfold an alternative to urban transformation practices in planning and architecture...

  2. Performative securitization

    DEFF Research Database (Denmark)

    Philipsen, Lise

    2018-01-01

    This piece develops a performative take on securitization theory. It argues that rather than seeing authority as a prerequisite for speaking security, we need to zoom in on how speaking security can be used to claim authority. Such acts of claiming authority are crucial to understand the current...... challenged and changed. Two, following Butler, we must open up who can speak security, seeing how speaking security can be used to take authority, rather than seeing authority as a precondition for speaking security....

  3. ORELA performance

    International Nuclear Information System (INIS)

    Lewis, T.A.

    1976-04-01

    The most recent information concerning the performance of ORELA that would be of interest to experimenters is presented. Included are characteristics of the beam in terms of both time and intensity and descriptions of systems routinely used to monitor these beam characteristics. For example, with klystron power and maximum electron gun output current at nominal values and for pulse repetition rates in the range above 800 pps, output beam energies per pulse vary from 5 J for 2.5 nsec-wide pulses to approximately 32 J for 10 nsec pulses and 65 J for 40 nsec pulses

  4. Performance of

    Directory of Open Access Journals (Sweden)

    Naema A. Ali

    2015-09-01

    Full Text Available Soil collapse occurs when increased moisture causes chemical or physical bonds between the soil particles to weaken, which allows the structure of the soil to collapse. Collapsible soils are generally low-density, fine-grained combinations of clay and sand left by mudflows that have dried, leaving tiny air pockets. When the soil is dry, the cemented materials are strong enough to bond the sand particles together. When natural soil becomes wet, moisture alters the cementation structure and the soil’s strength is compromised, causing collapse or subsidence. Based on soil type and density, the potential for encountering collapsible soils throughout most of the project alignment is low. Conditions in arid and semi-arid climates like Borg El Arab, near Alexandria Egypt favor the formation of the most problematic collapsible soils. The behavior and performance of compacted sand replacement over treated collapsible soil by pre-wetting and compaction are investigated in the current study. Field investigation was performed in the form of plate loading tests conducted on compacted sand replacement over improved collapsible soil. Field plate load tests program was developed to explore the effect of compacted sand replacement thickness on collapsibility potential. Treated collapsible soil was replaced with imported cohesionless soil with variable thickness up to footing width. Results proved that the improvement of collapsible soils by sand/crushed stone replacement is possible to control/mitigate their risk potentials against sudden settlement when exposed to water. Replacement soil increases the rate and reduces the amount of footing settlement. For compacted collapsible soils, partial replacement by compacted sand/crushed stone layers decreases collapsibility potential risk. Results also, introduce the development of practical, economical and environmentally safe geochemical methods for collapsible soil stabilization and collapsible risk mitigation.

  5. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Laug, David V.; Scates, Dawn M.; Reber, Edward L.; Roybal, Lyle G.; Walter, John B.; Harp, Jason M. [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Morris, Robert N. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer A system has been developed for safety testing of irradiated coated particle fuel. Black-Right-Pointing-Pointer FACS system is designed to facilitate remote operation in a shielded hot cell. Black-Right-Pointing-Pointer System will measure release of fission gases and condensable fission products. Black-Right-Pointing-Pointer Fuel performance can be evaluated at temperatures as high as 2000 Degree-Sign C in flowing helium. - Abstract: The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 Degree-Sign C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated

  6. The Jules Horowitz reactor, a new high performance European material testing reactor open to international users: present status and objectives

    International Nuclear Information System (INIS)

    Iracane, D.; Bignan, G.

    2010-01-01

    The development of nuclear power as a sustainable and competitive energy source will continue to require research and development of fuel and material behaviour under irradiation. This necessitates a high performance material testing reactor (MTR). Facing the obsolescence of most of the existing MTR in Europe, France decided a few years ago the construction of the RJH (Jules Horowitz reactor). RJH is designed, built and will be operated as an international user facility. A first set of experimental hosting devices is being designed. For instance, there are the in-core CALIPSO Nak integrated loop for material studies and other loops for fuel studies under nominal or off-normal or accidental conditions. The RJH international program will focus on the following subjects: -) fuel reliability, assessed through power ramps tests and post-irradiation examination; -) Loss of coolant tests done out-of-pile in a first phase and in-pile in a possible second phase; and -) source term tests addressing fission products release. The paper reports also the point of view of VATTENFALL (a Swedish power utility), as a potential European RJH user. (A.C.)

  7. Irradiation performance of HTGR fuel rods in HFIR experiments HRB-7 and -8

    International Nuclear Information System (INIS)

    Valentine, K.H.; Homan, F.J.; Long, E.L. Jr.; Tiegs, T.N.; Montgomery, B.H.; Hamner, R.L.; Beatty, R.L.

    1977-05-01

    The HRB-7 and -8 experiments were designed as a comprehensive test of mixed thorium-uranium oxide fissile particles with Th:U ratios from 0 to 8 for HTGR recycle application. In addition, fissile particles derived from Weak-Acid Resin (WAR) were tested as a potential backup type of fissile particle for HTGR recycle. These experiments were conducted at two temperatures (1250 and 1500 0 C) to determine the influence of operating temperature on the performance parameters studied. The minor objectives were comparison of advanced coating designs where ZrC replaced SiC in the Triso design, testing of fuel coated in laboratory-scale equipment with fuel coated in production-scale coaters, comparison of the performance of 233 U-bearing particles with that of 235 U-bearing particles, comparison of the performance of Biso coatings with Triso coatings for particles containing the same type of kernel, and testing of multijunction tungsten-rhenium thermocouples. All objectives were accomplished. As a result of these experiments the mixed thorium-uranium oxide fissile kernel was replaced by a WAR-derived particle in the reference recycle design. A tentative decision to make this change had been reached before the HRB-7 and -8 capsules were examined, and the results of the examination confirmed the accuracy of the previous decision. Even maximum dilution (Th/U approximately equal to 8) of the mixed thorium-uranium oxide kernel was insufficient to prevent amoeba of the kernels at rates that are unacceptable in a large HTGR. Other results showed the performance of 233 U-bearing particles to be identical to that of 235 U-bearing particles, the performance of fuel coated in production-scale equipment to be at least as good as that of fuel coated in laboratory-scale coaters, the performance of ZrC coatings to be very promising, and Biso coatings to be inferior to Triso coatings relative to fission product retention

  8. The performance of solar heat pump with non-freon refrigerant CF{sub 3}CH{sub 2}F(R-134a) for school classroom heating [II

    Energy Technology Data Exchange (ETDEWEB)

    Sun, K.H.; Jung, H.C. [Kyung Hee University, Seoul (Korea, Republic of); Kim, K.S. [Daebul University (Korea, Republic of)

    1997-03-01

    The goal of this paper is to measure and compare the performance of solar heat pump for school classroom heating. To accomplish the goal, solar heat pump with aluminum roll bond type evaporator and indoor heat exchanger(condenser) was built and fully instrumented with thermocouples and pressure transducers etc. The test results showed that the COP and capacity of R-134a(CF{sub 3}CH{sub 2}F) were higher than those of R-12(CF{sub 2}Cl{sub 2}). The solar heat pump system for room heating was designed to show the best efficiency that the room temperature make 18{approx}20{sup o} C and 23{approx}25{sup o} C in Seoul during November, December, and January. (author) 12 refs., 4 figs., 2 tabs.

  9. Cooling performance and evaluation of automotive refrigeration system for a passenger car

    Science.gov (United States)

    Prajitno, Deendarlianto, Majid, Akmal Irfan; Mardani, Mahardeka Dhias; Wicaksono, Wendi; Kamal, Samsul; Purwanto, Teguh Pudji; Fauzun

    2016-06-01

    A new design of automotive refrigeration system for a passenger car was proposed. To ensure less energy consumption and optimal thermal comfort, the performance of the system were evaluated. This current research was aimed to evaluate the refrigeration characteristics of the system for several types of cooling load. In this present study, a four-passenger wagon car with 1500 cc gasoline engine that equipped by a belt driven compressor (BDC) was used as the tested vehicle. To represent the tropical condition, a set of lamps and wind sources are installed around the vehicle. The blower capacity inside a car is varied from 0.015 m/s to 0.027 m/s and the compressor speed is varied at variable 820, 1400, and 2100 rpm at a set temperature of 22°C. A set of thermocouples that combined by data logger were used to measure the temperature distribution. The system uses R-134a as the refrigerant. In order to determine the cooling capacity of the vehicle, two conditions were presented: without passengers and full load conditions. As the results, cooling capacity from any possible heating sources and transient characteristics of temperature in both systems for the cabin, engine, compressor, and condenser are presented in this work. As the load increases, the outlet temperature of evaporator also increases due to the increase of condensed air. This phenomenon also causes the increase of compressor work and compression ratio which associated to the addition of specific volume in compressor inlet.

  10. Thermal Performance of Precast Concrete Sandwich Panel (PCSP) Design for Sustainable Built Environment

    Science.gov (United States)

    Ern, Peniel Ang Soon; Ling, Lim Mei; Kasim, Narimah; Hamid, Zuhairi Abd; Masrom, Md Asrul Nasid Bin

    2017-10-01

    Malaysia’s awareness of performance criteria in construction industry towards a sustainable built environment with the use of precast concrete sandwich panel (PCSP) system is applied in the building’s wall to study the structural behaviour. However, very limited studies are conducted on the thermal insulation of exterior and interior panels in PCSP design. In hot countries such as Malaysia, proper designs of panel are important to obtain better thermal insulation for building. This study is based on thermal performance of precast concrete sandwich panel design for sustainable built environment in Malaysia. In this research, three full specimens, which are control specimen (C), foamed concrete (FC) panels and concrete panels with added palm oil fuel ash (FC+ POFA), where FC and FC+POFA sandwiched with gypsum board (G) were produced to investigate their thermal performance. Temperature difference of exterior and interior surface of specimen was used as indicators of thermal-insulating performance of PCSP design. Heat transfer test by halogen lamp was carried out on three specimens where the exterior surface of specimens was exposed to the halogen lamp. The temperature reading of exterior and interior surface for three specimens were recorded with the help of thermocouple. Other factors also studied the workability, compressive strength and axial compressive strength of the specimens. This study has shown that FC + POFA specimen has the strength nearer to normal specimen (C + FC specimen). Meanwhile, the heat transfer results show that the FC+POFA has better thermal insulation performance compared to C and FC specimens with the highest temperature difference, 3.4°C compared to other specimens. The results from this research are useful to be implemented in construction due to its benefits such as reduction of energy consumption in air-conditioning, reduction of construction periods and eco-friendly materials.

  11. Thermal performance measurements of a 100 percent polyester MLI [multilayer insulation] system for the Superconducting Super Collider

    International Nuclear Information System (INIS)

    Boroski, W.N.; Gonczy, J.D.; Niemann, R.C.

    1989-09-01

    Thermal performance measurements of a 100 percent polyester multilayer insulation (MLI) system for the Superconducting Super Collider (SSC) were conducted in a Heat Leak Test Facility (HLTF) under three experimental test arrangements. Each experiment measured the thermal performance of a 32-layer MLI blanket instrumented with twenty foil sensors to measure interstitial layer temperatures. Heat leak values and sensor temperatures were monitored during transient and steady state conditions under both design and degraded insulating vacuums. Heat leak values were measured using a heatmeter. MLI interstitial layer temperatures were measured using Cryogenic Linear Temperature Sensors (CLTS). Platinum resistors monitored system temperatures. High vacuum was measured using ion gauges; degraded vacuum employed thermocouple gauges. A four-wire system monitored instrumentation sensors and calibration heaters. An on-line computerized data acquisition system recorded and processes data. This paper reports on the instrumentation and experimental preparation used in carrying out these measurements. In complement with this paper is an associate paper bearing the same title head, but with the title extension 'Part 2: Laboratory results (300K--80K). 13 refs., 7 figs

  12. Performance Management or Performance Based Management?

    OpenAIRE

    Cristina PROTOPOPESCU

    2013-01-01

    In this paper we present some considerations about performance and performance management. Starting with the challenge of defining the performance concept, we intend to establish if „performance management” can be a new management system or it is just a sophisticated term for a HR strategy in order to improve the performance of teams and individuals. We also try to discuss the conection between performance management and management by objectives. Whether or not it is exageratted to talk about...

  13. Analysis of TIMS performance subjected to simulated wind blast

    Science.gov (United States)

    Jaggi, S.; Kuo, S.

    1992-01-01

    The results of the performance of the Thermal Infrared Multispectral Scanner (TIMS) when it is subjected to various wind conditions in the laboratory are described. Various wind conditions were simulated using a 24 inch fan or combinations of air jet streams blowing toward either or both of the blackbody surfaces. The fan was used to simulate a large volume of air flow at moderate speeds (up to 30 mph). The small diameter air jets were used to probe TIMS system response in reaction to localized wind perturbations. The maximum nozzle speed of the air jet was 60 mph. A range of wind directions and speeds were set up in the laboratory during the test. The majority of the wind tests were conducted under ambient conditions with the room temperature fluctuating no more than 2 C. The temperature of the high speed air jet was determined to be within 1 C of the room temperature. TIMS response was recorded on analog tape. Additional thermistor readouts of the blackbody temperatures and thermocouple readout of the ambient temperature were recorded manually to be compared with the housekeeping data recorded on the tape. Additional tests were conducted under conditions of elevated and cooled room temperatures. The room temperature was varied between 19.5 to 25.5 C in these tests. The calibration parameters needed for quantitative analysis of TIMS data were first plotted on a scanline-by-scanline basis. These parameters are the low and high blackbody temperature readings as recorded by the TIMS and their corresponding digitized count values. Using these values, the system transfer equations were calculated. This equation allows us to compute the flux for any video count by computing the slope and intercept of the straight line that relates the flux to the digital count. The actual video of the target (the lab floor in this case) was then compared with a simulated target. This simulated target was assumed to be a blackbody at emissivity of .95 degrees and the temperature was

  14. Instrument for continuous supervision of the radioactivity of CO{sub 2} coolant in piles - DCCA -CO{sub 2} (1960); Dispositif de controle continu de la radioactivite du CO{sub 2} de refroidissement des piles - DCCA - CO{sub 2} (1960)

    Energy Technology Data Exchange (ETDEWEB)

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

    1960-07-01

    This paper describes an apparatus for continuous measurement of CO{sub 2} activity, which can be used on piles cooled by circulation of gas. The first part is devoted mainly to describing the apparatus used and the character of the radioactivity and thermodynamic measurements carried out, and giving the general characteristics of the gas circuit required if the instrument is to be suitably gas-tight. In the second part theoretical calculations are given, particularly on the determination of the ionisation current in an ionisation chamber with circulating gas. Several parameters enter into this determination, such as the mean path of {beta} particles in the ionisation chamber, the linear number of ion pairs formed in the gas by these {beta} particles as a function of their energy, the temperature and pressure of the gas in the ionisation chamber. This part also evaluates the sensitivity areas of the apparatus for measuring the concentrations of radioactive gases such as argon-41 and fission gases from uranium-235 in the CO{sub 2} coolant. In the last part are described the results of measurements performed with such an apparatus on the pile EL2, the special investigations carried out on the CO{sub 2} coolant of this pile, and the information gained during normal operation and during accidents. The DCCA - CO{sub 2} which has just been put in operation at G2 is briefly presented. In the conclusion the possibilities offered by this apparatus are underlined. (author) [French] Ce rapport a pour but de presenter le Dispositif de Controle continu de l'Activite du CO{sub 2} pouvant etre utilise aupres des piles refroidies par une circulation de gaz. La premiere partie du rapport consiste essentiellement a decrire l'ensemble de l'appareillage mis en oeuvre, a preciser la nature des mesures de radioactivite et de thermodynamique effectuees et a citer les caracteristiques generales du circuit de gaz pour avoir un dispositif presentant une etancheite efficace. Dans la seconde

  15. Instrument for continuous supervision of the radioactivity of CO{sub 2} coolant in piles - DCCA -CO{sub 2} (1960); Dispositif de controle continu de la radioactivite du CO{sub 2} de refroidissement des piles - DCCA - CO{sub 2} (1960)

    Energy Technology Data Exchange (ETDEWEB)

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

    1960-07-01

    This paper describes an apparatus for continuous measurement of CO{sub 2} activity, which can be used on piles cooled by circulation of gas. The first part is devoted mainly to describing the apparatus used and the character of the radioactivity and thermodynamic measurements carried out, and giving the general characteristics of the gas circuit required if the instrument is to be suitably gas-tight. In the second part theoretical calculations are given, particularly on the determination of the ionisation current in an ionisation chamber with circulating gas. Several parameters enter into this determination, such as the mean path of {beta} particles in the ionisation chamber, the linear number of ion pairs formed in the gas by these {beta} particles as a function of their energy, the temperature and pressure of the gas in the ionisation chamber. This part also evaluates the sensitivity areas of the apparatus for measuring the concentrations of radioactive gases such as argon-41 and fission gases from uranium-235 in the CO{sub 2} coolant. In the last part are described the results of measurements performed with such an apparatus on the pile EL2, the special investigations carried out on the CO{sub 2} coolant of this pile, and the information gained during normal operation and during accidents. The DCCA - CO{sub 2} which has just been put in operation at G2 is briefly presented. In the conclusion the possibilities offered by this apparatus are underlined. (author) [French] Ce rapport a pour but de presenter le Dispositif de Controle continu de l'Activite du CO{sub 2} pouvant etre utilise aupres des piles refroidies par une circulation de gaz. La premiere partie du rapport consiste essentiellement a decrire l'ensemble de l'appareillage mis en oeuvre, a preciser la nature des mesures de radioactivite et de thermodynamique effectuees et a citer les caracteristiques generales du circuit de gaz pour avoir un dispositif presentant une etancheite efficace

  16. Understanding protocol performance: impact of test performance.

    Science.gov (United States)

    Turner, Robert G

    2013-01-01

    This is the second of two articles that examine the factors that determine protocol performance. The objective of these articles is to provide a general understanding of protocol performance that can be used to estimate performance, establish limits on performance, decide if a protocol is justified, and ultimately select a protocol. The first article was concerned with protocol criterion and test correlation. It demonstrated the advantages and disadvantages of different criterion when all tests had the same performance. It also examined the impact of increasing test correlation on protocol performance and the characteristics of the different criteria. To examine the impact on protocol performance when individual tests in a protocol have different performance. This is evaluated for different criteria and test correlations. The results of the two articles are combined and summarized. A mathematical model is used to calculate protocol performance for different protocol criteria and test correlations when there are small to large variations in the performance of individual tests in the protocol. The performance of the individual tests that make up a protocol has a significant impact on the performance of the protocol. As expected, the better the performance of the individual tests, the better the performance of the protocol. Many of the characteristics of the different criteria are relatively independent of the variation in the performance of the individual tests. However, increasing test variation degrades some criteria advantages and causes a new disadvantage to appear. This negative impact increases as test variation increases and as more tests are added to the protocol. Best protocol performance is obtained when individual tests are uncorrelated and have the same performance. In general, the greater the variation in the performance of tests in the protocol, the more detrimental this variation is to protocol performance. Since this negative impact is increased as

  17. Application of fiber optic grating strain sensor for measurement of strain under irradiation environment

    International Nuclear Information System (INIS)

    Kaji, Y.; Matsui, Y.; Kita, S.; Ide, H.; Tsukada, T.; Tsuji, H.

    2001-01-01

    In Japan Atomic Energy Research Institute (JAERI), in-pile strain measurement techniques have been developed using Japan Materials Testing Reactor (JMTR). In order to evaluate the performance of fiber optic grating sensor under irradiation environment, heat-up and performance tests at elevated temperature before irradiation and in-pile tests were performed in JMTR. (author)

  18. Textiles Performance Testing Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — The Textiles Performance Testing Facilities has the capabilities to perform all physical wet and dry performance testing, and visual and instrumental color analysis...

  19. Sipping machine control system new design to perform integrity of nuclear fuel test in Cofrentes power plant

    Energy Technology Data Exchange (ETDEWEB)

    Palomo, M., E-mail: mpalomo@iqn.upv.es [Departamento de Ingenieria Quimica y Nuclear. Universidad Politecnica de Valencia (Spain); Urrea, M., E-mail: Matias.urrea@iberdrola.es [C.N.Cofrentes - Iberdrola Generacion S.A., Cofrentes, Valencia (Spain); Curiel, M., E-mail: m.curiel@lainsa.com [LAINSA Grupo Dominguis, Valencia (Spain); Arnaldos, A., E-mail: a.arnaldos@titaniast.com [TITANIA Servicios Tecnologicos SL, Grupo Dominguis, Valencia (Spain)

    2011-07-01

    This paper we present is related to SIPPING machine control system new design to perform integrity of nuclear fuel test. This test is a non destructive technique used for evaluating the radiated nuclear fuel coating structural integrity. It is based on the radioactive emission detection of fission elements in the reactor cooling system, using the fuel inspection equipment (SIPPING). SIPPING equipment consists of one simultaneous test bell-shaped vessel of eight fuel elements, and another one for individual element test, a control workstation and some accessories (cables, thermocouples, hoses). SIPPING inspection is carried out by means of fuel element vessel. Through air injection, water flows around the element and heat evacuation is reduced, so fuel elements temperature increases. Those elements with faults shall expelled fission components dissolved in water and/or as a gas component. The project aim is the SIPPING system control design and software based on LabVIEW, for control, monitoring and documentation of the SIPPING Test. This project shall give a major functionality to the system and, at the same time, shall facilitate the user a friendlier and interactive environment allowing: to substitute the present work platform with a real-time electronic system based on cRIO and a control software ad-hoc designed for SIPPING system; to equip new system of a major redundancy for data storage, minimising loss probability of the same. (author)

  20. Evaluation of cleaning and disinfection performance of automatic washer disinfectors machines in programs presenting different cycle times and temperatures.

    Science.gov (United States)

    Bergo, Maria do Carmo Noronha Cominato

    2006-01-01

    Thermal washer-disinfectors represent a technology that brought about great advantages such as, establishment of protocols, standard operating procedures, reduction in occupational risk of a biological and environmental nature. The efficacy of the cleaning and disinfection obtained by automatic washer disinfectors machines in running programs with different times and temperatures determined by the different official agencies was validated according to recommendations from ISO Standards 15883-1/1999 and HTM2030 (NHS Estates, 1997) for the determining of the Minimum Lethality and DAL both theoretically and through the use with thermocouples. In order to determine the cleaning efficacy, the Soil Test, Biotrace Pro-tect and the Protein Test Kit were used. The procedure to verify the CFU count of viable microorganisms was performed before and after the thermal disinfection. This article shows that the results are in compliance with the ISO and HTM Standards. The validation steps confirmed the high efficacy level of the Medical Washer-Disinfectors. This protocol enabled the evaluation of the procedure based on evidence supported by scientific research, aiming at the support of the Supply Center multi-professional personnel with information and the possibility of developing further research.