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Sample records for chicago pile-2 reactor

  1. G 2 reactor project; Projet de pile a double fin: G 2

    Energy Technology Data Exchange (ETDEWEB)

    Ailleret, [Electricite de France (EDF), Dir. General des Etudes de Recherches, 75 - Paris (France); Taranger, P; Yvon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    The CEA actually constructs the G-2 reactor core working with natural uranium, which will use graphite as moderator, and gas under pressure as cooling fluid. This report presents the specificity of the new reactor: - the different elements of the reactor core, - the control and the security of the reactor, - the renewal of the fuel, - the biologic surrounding wall, - and the cooling circuit. (M.B.) [French] le Commissariat a l'Energie Atomique construit actuellement la pile G-2 a Uranium naturel, qui utilisera le graphite comme moderateur, et le gaz sous pression comme fluide de refroidissement. Ce rapport presente les specificite du nouveau reacteur: - les differents elements de la pile, - le controle et la securite du reacteur, - le renouvellement du combustible, - l'enceinte biologique, - et le circuit de refroidissement. (M.B.)

  2. Seismic analysis of the pile foundation of the reactor building of the NPP ANGRA 2

    International Nuclear Information System (INIS)

    Wolf, J.P.; Arx, G.A. von; Barros, F.C.P. de; Kakubo, M.

    1981-01-01

    A pile foundation subjected to dynamic loads interacts with the surrounding soil. Frequency-dependent stiffness and radiation damping must be properly taken into account in pile-soil-pile interaction. Assuming that the soil consists of horizontal layers of elastic material with hysteretic damping, the dynamic stiffness of a group of (even battered) piles can be determined, accounting rigorously for the cavities where the soil is subsequently replaced by the piles. By way of illustration, this substructure procedure, which works in the frequency domain, is applied to the final design of the pile foundation of the Reactor Building of Angra 2 in Brazil. Below the basemat, a strongly horizontally-layered compressive soil of 36 m thickness rests on bedrock. The reactor building is founded on 202 endbearing piles and 88 floating piles of 15 m length. Every pile is modelled. Along each pile, compatibility between the pile and the soil in all three directions is formulated in seven nodes. The basemat is assumed to be rigid. On the level of bedrock a broad-banded response spectrum specifies the excitation (outcropping). (orig./WL)

  3. The reactor Cabri; La pile cabri

    Energy Technology Data Exchange (ETDEWEB)

    Ailloud, J; Millot, J P [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1964-07-01

    circumstances... - experimental investigations on power excursions linked with precise initial conditions: the aim of this work is to define the basis for theoretical research, and the limits beyond which the risks of explosion cease to be negligible. The research work will be done so as to enable checking with outside reactor experiments and to continue them in the explosion field. - studies of the behaviour of the reactor control-instrumentation. - experimental investigations related with transient operation with initial short life (study of boiling, temperature measurements, vacuum pressure and fraction...) with the aim of defining the hypotheses of a theory on swimming-pool reactor kinetics related to heat transfer phenomena, - investigations of the behaviour of fuels in reactors (these experiments are planned to be carried out in loops) Preliminary experimental results. CABRI went critical on the 21 December 1963. The first transient experiments are expected for March 1964. (authors) [French] II devenait necessaire de construire en France une pile qui permette d'etudier les conditions de fonctionnement des installations futures, de choisir, tester et mettre au point les dispositifs de securite a adopter. On a choisi une pile a eau, type de pile qui correspond aux constructions les plus nouvelles du CEA en matiere de piles laboratoire ou d'universite; il importe en effet de pouvoir evaluer les risques presentes et d'etudier les possibilites d'augmentation de puissance constamment demandees par les utilisateurs: il est particulierement interessant d'eclaircir les phenomenes d'oscillation de puissance et les risques de calefaction (burn out). Les programmes de travaux sur CABRI seront harmonises avec les travaux effectues sur les Spert americains de meme type; lors de sa construction des contacts fructueux ont ete etablis avec les specialistes americains qui ont defini les premiers de ces reacteurs. La communication donne une description sommaire de la pile et decrit le

  4. The reactor Cabri; La pile cabri

    Energy Technology Data Exchange (ETDEWEB)

    Ailloud, J.; Millot, J.P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1964-07-01

    exceptional circumstances... - experimental investigations on power excursions linked with precise initial conditions: the aim of this work is to define the basis for theoretical research, and the limits beyond which the risks of explosion cease to be negligible. The research work will be done so as to enable checking with outside reactor experiments and to continue them in the explosion field. - studies of the behaviour of the reactor control-instrumentation. - experimental investigations related with transient operation with initial short life (study of boiling, temperature measurements, vacuum pressure and fraction...) with the aim of defining the hypotheses of a theory on swimming-pool reactor kinetics related to heat transfer phenomena, - investigations of the behaviour of fuels in reactors (these experiments are planned to be carried out in loops) Preliminary experimental results. CABRI went critical on the 21 December 1963. The first transient experiments are expected for March 1964. (authors) [French] II devenait necessaire de construire en France une pile qui permette d'etudier les conditions de fonctionnement des installations futures, de choisir, tester et mettre au point les dispositifs de securite a adopter. On a choisi une pile a eau, type de pile qui correspond aux constructions les plus nouvelles du CEA en matiere de piles laboratoire ou d'universite; il importe en effet de pouvoir evaluer les risques presentes et d'etudier les possibilites d'augmentation de puissance constamment demandees par les utilisateurs: il est particulierement interessant d'eclaircir les phenomenes d'oscillation de puissance et les risques de calefaction (burn out). Les programmes de travaux sur CABRI seront harmonises avec les travaux effectues sur les Spert americains de meme type; lors de sa construction des contacts fructueux ont ete etablis avec les specialistes americains qui ont defini les premiers de ces reacteurs. La communication donne une

  5. Seismological analysis of group pile foundation for reactor

    International Nuclear Information System (INIS)

    Wang Demin.

    1984-01-01

    In the seismic analysis for reactor foundation of nuclear power plant, the local raise of base mat is of great significance. Base on the study of static and dynamic stability as well as soil-structure interaction of group piles on stratified soil, this paper presents a method of seismic analysis for group piles of reactor foundation at abroad, and a case history is enclosed. (Author)

  6. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-15

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

  8. 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)

  9. 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.

  10. Fuel elements for pressurised-gas reactors; Elements combustibles des piles a gaz sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Stohr, J A; Englander, M; Gauthron, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The design and fabrication of fuel elements for the first CO{sub 2} pressurized reactors have induced to investigate: various cladding materials, natural uranium base fuels, canning processes. The main analogical tests used in connection with the fuel element study are described. These various tests have enabled, among others, the fabrication of the fuel element for the EL2 reactor. Lastly, future solutions for electrical power producing reactors are foreseen. (author)Fren. [French] L'etude et la realisation d'elements combustibles pour les premieres piles a CO{sub 2} sous pression ont conduit a examiner: les divers materiaux de gaine, les combustibles a base d'uranium naturel, les modes de gainage. Les principaux essais analogiques ayant servi au cours de l'etude de la cartouche sont decrits. Ces divers essais ont notamment permis la realisation de la cartouche de la pile EL2. Enfin sont envisagees les solutions futures pour les piles productrices d'energie electrique. (auteur)

  11. 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.

  12. 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

  13. 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

  14. The first reactor [40th anniversary commemorative edition

    Energy Technology Data Exchange (ETDEWEB)

    None

    1982-12-01

    This updated and revised story of the first reactor, or 'pile,' commemorates the 40th anniversary of the first controlled, self-sustaining nuclear chain reaction created by mankind. Enrico Fermi and his team of scientists initiated the reaction on December 2, 1941, underneath the West Stands of Stagg Field at the University of Chicago. Firsthand accounts of the participants as well as postwar recollections by Enrico and Laura Fermi are included.

  15. BR2 Reactor: Irradiation of fuels

    International Nuclear Information System (INIS)

    Verwimp, A.

    2005-01-01

    Safe, reliable and economical operation of reactor fuels, both UO 2 and MOX types, requires in-pile testing and qualification up to high target burn-up levels. In-pile testing of advanced fuels for improved performance is also mandatory. The objectives of research performed at SCK-CEN are to perform Neutron irradiation of LWR (Light Water Reactor) fuels in the BR2 reactor under relevant operating and monitoring conditions, as specified by the experimenter's requirements and to improve the on-line measurements on the fuel rods themselves

  16. The story of fission reactors: from Chicago Pile to advanced energy systems

    International Nuclear Information System (INIS)

    Kannan, Umasankari

    2017-01-01

    Nuclear reactors have been designed which cater to different applications from small research reactors of a few watts to power reactors of several Giga Watts. Based on the neutron energy, there are thermal, intermediate and fast reactors operating are being designed. On the fuel utilization front, there are designs ranging from reactors using natural uranium fuel to enriched uranium to more efficient thorium based reactors. Reactors have also been designed which are neutron eaters, minor actinide burners and breeders. There have been variety of coolant and moderating materials used for different applications from water, gas cooled, liquid sodium cooled to molten salt cooled reactors. Several new reactor designs have been developed using innovative concepts in high temperature reactors, nuclear power packs and compact reactors for special purposes. The design challenges are many from modest designs to complicated hybrid reactors. The GEN-IV forum of IAEA has selected a few of these reactor designs for commercial power production in the coming years based on several quantified indicators. The evolutionary and revolutionary design approaches have been made over the years catering to different need of energy generation. A glimpse of some of the reactors being currently developed and the design modifications done in existing reactors have been given in this paper

  17. 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

  18. 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.

  19. 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)

  20. Shutdown channels and fitted interlocks in atomic reactors; Chaines de securite et verrouillages installes sur les piles atomiques

    Energy Technology Data Exchange (ETDEWEB)

    Furet, J; Landauer, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1968-07-01

    This catalogue consists of tables (one per reactor) giving the following information: number and type of detectors, range of the shutdown channels, nature of the associated electronics, thresholds setting off the alarms, fitted interlocks. These cards have been drawn up with a view to an examination of the reactors safety by the 'Reactor Safety Sub-Commission', they take into account the latest decisions. The reactors involved in this review are: Azur, Cabri, Castor-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, and Ulysse. (authors) [French] Ce catalogue est compose d'un ensemble de tableaux (a raison de un tableau par pile) donnant les renseignements suivants: nombre et nature des detecteurs, dynamique des chaines, nature de l'electronique associee, seuils provoquant des actions de securite, verrouillages installes. Ces fiches ont ete etablies en vue de l'examen de la securite des piles par la 'Sous-Commission de Surete des Piles', et tiennent compte des decisions de celle-ci. Les reacteurs concernes sont: Azur, Cabri, Cator-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, et Ulysse. (auteurs)

  1. Neutronic study of the two french heavy water reactors; Etude neutronique des deux piles francaises a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

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

    1955-07-01

    The two french reactors - the reactor of Chatillon, named Zoe, and the reactor of Saclay - P2 - were the object of detailed neutronic studies which the main ideas are exposed in this report. These studies were mostly done by the Department of the Reactor Studies (D.E.P.). We have thus studied the distribution of neutronic fluxes; the factors influencing reactivity; the link between reactivity and divergence with the formula of Nordheim; the mean time life of neutrons; neutron spectra s of P2; the xenon effect; or the effect of the different adjustments of the plates and controls bar. (M.B.) [French] Les deux reacteurs francais - la pile de Chatillon, appelee ZOE, et la pile de Saclay, designee dans la suite par P2 - ont fait l'objet d'etudes neutroniques detaillees dont les principales sont exposees dans ce rapport. Ces etudes ont ete pour la plupart effectuees dans le cadre du Departement des Etudes de Piles (D.E.P.). Nous avons ainsi entre autre etudie la distribution du flux neutronique; les facteurs influencants la reactivite; le lien entre reactivite et divergence par la formule de Nordheim; le temps de vie moyen des neutrons; les spectres de neutrons de P2; l'effet xenon; ou encore l'effet des differents reglages des plaques et barres de controles. (M.B.)

  2. 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.

  3. 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

  4. 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)

  5. EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1969-02-01

    The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99). Elle est

  6. EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1969-02-01

    The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99

  7. Graphite reactor physics; Physique des piles a graphite

    Energy Technology Data Exchange (ETDEWEB)

    Bacher, P; Cogne, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Noc, B [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    The study of graphite-natural uranium power reactor physics, undertaken ten years ago when the Marcoule piles were built, has continued to keep in step with the development of this type of pile. From 1960 onwards the critical facility Marius has been available for a systematic study of the properties of lattices as a function of their pitch, of fuel geometry and of the diameter of cooling channels. This study has covered a very wide field: lattice pitch varying from 19 to 38 cm. uranium rods and tubes of cross-sections from 6 to 35 cm{sup 2}, channels with diameters between 70 and 140 mm. The lattice calculation methods could thus be checked and where necessary adapted. The running of the Marcoule piles and the experiments carried out on them during the last few years have supplied valuable information on the overall evolution of the neutronic properties of the fuel as a function of irradiation. More detailed experiments have also been performed in Marius with plutonium-containing fuels (irradiated or synthetic fuels), and will be undertaken at the beginning of 1965 at high temperature in the critical facility Cesar, which is just being completed at Cadarache. Spent fuel analyses complement these results and help in their interpretation. The thermalization and spectra theories developed in France can thus be verified over the whole valid temperature range. The efficiency of control rods as a function of their dimensions, the materials of which they are made and the lattices surrounding them has been measured in Marius, and the results compared with calculation on the one hand and with the measurements carried out in EDF 1 on the other. Studies on the control proper of graphite piles were concerned essentially with the risks of spatial instability and the means of detecting and controlling them, and with flux distortions caused by the control rods. (authors) [French] Entreprise il y a dix ans a l'occasion de la construction des piles de Marcoule, l'etude de la

  8. Applicability of out-of-pile fretting wear tests to in-reactor fretting wear-induced failure time prediction

    Science.gov (United States)

    Kim, Kyu-Tae

    2013-02-01

    In order to investigate whether or not the grid-to-rod fretting wear-induced fuel failure will occur for newly developed spacer grid spring designs for the fuel lifetime, out-of-pile fretting wear tests with one or two fuel assemblies are to be performed. In this study, the out-of-pile fretting wear tests were performed in order to compare the potential for wear-induced fuel failure in two newly-developed, Korean PWR spacer grid designs. Lasting 20 days, the tests simulated maximum grid-to-rod gap conditions and the worst flow induced vibration effects that might take place over the fuel life time. The fuel rod perforation times calculated from the out-of-pile tests are greater than 1933 days for 2 μm oxidized fuel rods with a 100 μm grid-to-rod gap, whereas those estimated from in-reactor fretting wear failure database may be about in the range of between 60 and 100 days. This large discrepancy in fuel rod perforation may occur due to irradiation-induced cladding oxide microstructure changes on the one hand and a temperature gradient-induced hydrogen content profile across the cladding metal region on the other hand, which may accelerate brittleness in the grid-contacting cladding oxide and metal regions during the reactor operation. A three-phase grid-to-rod fretting wear model is proposed to simulate in-reactor fretting wear progress into the cladding, considering the microstructure changes of the cladding oxide and the hydrogen content profile across the cladding metal region combined with the temperature gradient. The out-of-pile tests cannot be directly applicable to the prediction of in-reactor fretting wear-induced cladding perforations but they can be used only for evaluating a relative wear resistance of one grid design against the other grid design.

  9. Windscale pile core surveys

    International Nuclear Information System (INIS)

    Curtis, R.F.; Mathews, R.F.

    1996-01-01

    The two Windscale Piles were closed down, defueled as far as possible and mothballed for thirty years following a fire in the core of Pile 1 in 1957 resulting from the spontaneous release of stored Wigner energy in the graphite moderator. Decommissioning of the reactors commenced in 1987 and has reached the stage where the condition of both cores needs to be determined. To this end, non-intrusive and intrusive surveys and sampling of the cores have been planned and partly implemented. The objectives for each Pile differ slightly. The location and quantity of fuel remaining in the damaged core of Pile 1 needed to be established, whereas the removal of all fuel from Pile 2 needed to be confirmed. In Pile 1, the possible existence of a void in the core is to be explored and in Pile 2, the level of Wigner energy remaining required to be quantified. Levels of radioactivity in both cores needed to be measured. The planning of the surveys is described including strategy, design, safety case preparation and the remote handling and viewing equipment required to carry out the inspection, sampling and monitoring work. The results from the completed non-intrusive survey of Pile 2 are summarised. They confirm that the core is empty and the graphite is in good condition. The survey of Pile 1 has just started. (UK)

  10. Can-rupture detection in gas-cooled nuclear reactors; La detection des ruptures de gaine dans les piles nucleaires refroidies par gaz

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

    Can-rupture detection (DRG) is one important aspect of pile safety, more particularly so in the case of gas-cooled reactors. A rapid and sure detection constitutes also an improvement as far as the efficiency of electricity-producing nuclear power stations are concerned. Among the numerous can-rupture detection methods, that based on the measurement of the concentration of short-lived fission gases in the heat-carrying fluid has proved to be the most sensitive and the most rapid. A systematic study of detectors based on the electrostatic collection of the daughter products of fission gases has been undertaken with a view to equip the reactors EL 2, G 3, EDF 1, EDF 2 and EDF 3, the gas loops of PEGASE and EL 4. The different parameters are studied in detail in order to obtain a maximum sensitivity and to make it possible to construct detection devices having the maximum operational reliability and requiring the minimum maintenance. The primary applications of these devices are examined in the case of the above-mentioned reactors. (author) [French] La Detection des Ruptures de Gaines (D. R. G.) est un aspect important de la securite des piles et plus particulierement des piles refroidies par un gaz. Une detection rapide et sure constitue aussi un element d'amelioration du rendement des centrales nucleaires productrices d'energie electrique. Parmi les nombreuses methodes de detection des ruptures de gaines, la mesure de la concentration dans le fluide caloporteur des gaz de fission a vie courte s'est revelee comme la plus sensible et la plus rapide. Une etude systematique des detecteurs a collection electrostatique des descendants des gaz de fission a ete entreprise en vue d'equiper les piles EL 2, G 3, EDF 1, EDF 2 et EDF 3, les boucles a gaz de la pile Pegase et la pile EL 4. Les divers parametres sont etudies en detail pour obtenir une sensibilite maximum et permettre la realisation de dispositifs de detection ayant le maximum de securite de fonctionnement et le

  11. Can-rupture detection in gas-cooled nuclear reactors; La detection des ruptures de gaine dans les piles nucleaires refroidies par gaz

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

    Can-rupture detection (DRG) is one important aspect of pile safety, more particularly so in the case of gas-cooled reactors. A rapid and sure detection constitutes also an improvement as far as the efficiency of electricity-producing nuclear power stations are concerned. Among the numerous can-rupture detection methods, that based on the measurement of the concentration of short-lived fission gases in the heat-carrying fluid has proved to be the most sensitive and the most rapid. A systematic study of detectors based on the electrostatic collection of the daughter products of fission gases has been undertaken with a view to equip the reactors EL 2, G 3, EDF 1, EDF 2 and EDF 3, the gas loops of PEGASE and EL 4. The different parameters are studied in detail in order to obtain a maximum sensitivity and to make it possible to construct detection devices having the maximum operational reliability and requiring the minimum maintenance. The primary applications of these devices are examined in the case of the above-mentioned reactors. (author) [French] La Detection des Ruptures de Gaines (D. R. G.) est un aspect important de la securite des piles et plus particulierement des piles refroidies par un gaz. Une detection rapide et sure constitue aussi un element d'amelioration du rendement des centrales nucleaires productrices d'energie electrique. Parmi les nombreuses methodes de detection des ruptures de gaines, la mesure de la concentration dans le fluide caloporteur des gaz de fission a vie courte s'est revelee comme la plus sensible et la plus rapide. Une etude systematique des detecteurs a collection electrostatique des descendants des gaz de fission a ete entreprise en vue d'equiper les piles EL 2, G 3, EDF 1, EDF 2 et EDF 3, les boucles a gaz de la pile Pegase et la pile EL 4. Les divers parametres sont etudies en detail pour obtenir une sensibilite maximum et permettre la realisation de dispositifs de detection ayant le maximum de securite de

  12. 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

  13. 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.

  14. Study of the dynamic behaviour of the reactor Rapsodie; Etude du comportement dynamique de la pile rapsodie

    Energy Technology Data Exchange (ETDEWEB)

    Abdon, R; Chaigne, M [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1964-07-01

    . The investigation of the control, carried out on analog computer, served to determine the different possible means of starting and changing the conditions of the reactor as well as its automatic control. The calculations were examined in the totality by the construction of a training simulator composed of a board similar to the control board of the reactor, all of whose commands (reactivity and flows) work on an analogue computer which resolves in the real time the dynamic equations of the reactor and which reproduces simultaneously all the parameters representing the state of the installation (power, period, temperatures, etc. ) in the case of various incidents as well as under normal conditions of functioning. (authors) [French] On sait que le developpement des reacteurs surgenerateurs a neutrons rapides pose des problemes nouveaux d'une part dans les domaines mecanique et thermique et d'autre part en ce qui concerne leur comportement dynamique et leur surete. La pile RAPSODIE a ete l'objet de tres nombreuses etudes dynamiques effectuees sur machines analogiques et digitales, pour deux versions du combustible (metal et oxyde). Apres elaboration des modeles mathematiques representatifs de l'ensemble de l'installation (bloc pile et circuit de refroidissement) tant du point de vue neutronique que du point de vue thermodynamique, on a mis au point les schemas analogiques et les codes digitaux utilisables pour mener a bien les simulations d'incidents, de conduite et de stabilite du reacteur. On s'est attache, par rapport aux methodes habituelles a obtenir une precision plus grande, par un decoupage en zones plus fines, par l'emploi de formulations plus representatives du systeme reel, voire solubles analytiquement. Les etudes d'incidents ont ete effectuees par voie analogique pour l'ensemble de l'installation et par voie digitale pour l'etude du bloc pile seul ou de l'installation fonctionnant avec un seul circuit thermique. Un programme complementaire special - qui, a

  15. 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)

  16. Modeling temperature noise in a fast-reactor pile

    International Nuclear Information System (INIS)

    Kebadze, B.V.; Pykhtina, T.V.; Tarasko, M.Z.

    1987-01-01

    To observe partial overlapping of the heat carrier cross section in piles, leading to local temperature rise or boiling of the sodium, provision is made for individual monitoring of the fuel assemblies with respect to the output temperature. Since the deviation of the mean flow rate through the pile and the output temperature is slight with this anomaly, the temperature fluctuations may provide a more informative index. The change in noise characteristics with partial overlapping of the cross sections occurs because of strong distortion of the temperature profile in the overlap region. The turbulent flow in the upper part of the pile transforms this nonuniformity into temperature pulsations which may be recorded by a sensor at the pile output. In this paper the characteristics of temperature noise are studied for various pile conditions and sensor locations by statistical modeling

  17. Radiation hazards in the neighbourhood of uranium reactors; Dangers des rayonnements aupres des piles a uranium

    Energy Technology Data Exchange (ETDEWEB)

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

    1956-07-01

    Radiation hazards near uranium reactors may be divided in two groups. Hazards when the reactor is normally operating: {gamma} radiation from hot uranium or air contamination by fission gases, {gamma} radiation or contamination by the coolant (air, nitrogen, heavy-water), {gamma} radiation from radioisotopes. Hazards in the case of an accident: presence of hot uranium in the atmosphere, soil contamination. (author) [French] Les dangers d'irradiation aupres des piles a uranium sont a classer essentiellement en deux groupes. Les dangers existant aupres d'une pile exploitee normalement: irradiation {gamma} par l'uranium irradie ou contamination de l'air par des gaz de fission, irradiation {gamma} ou contamination par les fluides de refroidissement (air, azote, eau lourde), irradiation {gamma} par les radioelements fabriques. Les dangers en cas d'accident survenant a un reacteur en fonctionnement, ayant pour consequence : la presence dans l'air d'uranium irradie, la contamination du sol. (auteur)

  18. The Windscale piles - past, present and future

    International Nuclear Information System (INIS)

    Jones, J.M.; Adams, A.L.

    1987-01-01

    The paper concerns the Windscale reactor piles, in which a fire occurred in the core of pile 1 thirty years ago. A description is given of the two Windscale piles, along with the events leading up to the accident, and the state of the piles following shutdown. The surveillance and maintenance to ensure that the pile and associated buildings were in a safe condition is outlined. The present state of the core, water ducts and pile chimneys is described. The present and future programme of work to ensure long term safety is discussed. This includes the initial steps in decommissioning of the piles. (U.K.)

  19. Study on the behavior of irradiated light water reactor fuel during out-of-pile annealing

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Kanazawa, Hiroyuki; Uno, Hisao; Sasajima, Hideo

    1988-11-01

    Using the pre-irradiated light water reactor fuel (burnup: 35 MWd/kgU) and the slightly irradiated NSRR fuel (burnup: 5.6 x 10 -6 MWd/kgU), FP gas release rate up to the temperature of 2273 K was measured through out-of-pile annealing test. Results of this experiment were compared with those of ORNL annealing test (SFD/HI-test series) performed in USA. Obtained conclusions are: (1) Maximum release rate of Kr gas in light water reactor fuel was 6.4 % min -1 at temperature of 2273 K. This was in good agreement with ORNL data. FP gas release rate during annealing test was increased greatly with increasing fuel burnup and annealing temperature. (2) No FP was detected in NSRR slightly irradiated fuel up to the temperature of 1913 K. (author)

  20. Generalities on the dynamic behaviour of rapid reactors. Preliminary studies on Rapsodie; Generalites sur le comportement dynamique des piles rapides. Etudes preliminaires de rapsodie

    Energy Technology Data Exchange (ETDEWEB)

    Campan, J L; Chaumont, J P; Clauzon, P P; Ghesquiere, G; Leduc, J; Schmitt, A P; Zaleski, C P [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1963-07-01

    The study of the dynamic behaviour of fast reactors may be divided into three section: 1. Stability studies around equilibrium power only the linear case was examining. S. Transient studies in the case of usual reactor operation (shut down, scram, etc.) with thermal shocks evaluation, for instance. 3. Explosion studies, for the maximum credible accidents. This report presents the status of the studies performed at the 'Physics Research Department' at Cadarache. Methods used are detailed and illustrated with the results obtained on a preliminary metallic core of the Rapsodie Reactor. (authors) [French] Le comportement dynamique des piles rapides, se presente tout naturellement sous trois aspects: 1. Etude de stabilite autour d'un regime d'equilibre (nous nous sommes bornes ici au cas lineaire). 2. Etude de regimes transitoires lors des operations normales de pile (arret, arret d'urgence, etc.) avec evaluation des chocs thermiques par exemple. 3. Etude des regimes transitoires de caractere explosif lors des accidents les plus graves possibles. Ce rapport presente l'etat des etudes a la date du 20 decembre 1961 a la Section d'Etudes de Piles Rapides a CADARACHE. Les methodes employees ont ete detaillees et illustrees a partir des resultats obtenus sur une premiere version 'combustible metallique' de Rapsodie. (auteurs)

  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. 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.

  3. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2002-01-01

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

  4. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2002-04-01

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

  5. Fuel slugs considered for use in the high flux reactor EL3; Elements combustibles envisages pour la pile a haut flux EL 3

    Energy Technology Data Exchange (ETDEWEB)

    Stohr, J A; Caillat, R; Gauthron, M; Montagne, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    EL3 was designed essentially for the study, under irradiation conditions, of materials used in the construction of atomic reactors. The study schedule allocates considerable time and effort to new types of fuel slugs. The present report described the various types of slug being tested or scheduled for tests. After laboratory study, each slug is tested in an experimental cell in the pile. The best are retained and used to charge the reactor (the present charge is purely provisional to permit first criticality and power rise tests)ren. [French] La pile EL3 est essentiellement destinee a l'etude sous irradiation des materiaux utilises dans la construction des reacteurs atomiques. Dans ce programme, une tres large part est reservee a l'etude de nouveaux elements combustibles. Le present rapport decrit les differentes solutions de cartouches dont l'essai est envisage ou en cours. Apres etude en laboratoire, chacune de ces solutions est testee dans une cellule experimentale en pile. Les meilleures seront retenues pour constituer le chargement normal de la pile (le chargement actuel etant essentiellement une solution provisoire qui a permis la divergence de la pile et les premiers essais de montee en puissance). (auteur)

  6. Critical sizes and flux distributions in the shut down pile

    International Nuclear Information System (INIS)

    Banchereau, A.; Berthier, P.; Genthon, J.P.; Gourdon, C.; Lattes, R.; Martelly, J.; Mazancourt, R. de; Portes, L.; Sagot, M.; Schmitt, A.P.; Tanguy, P.; Teste du Bailler, A.; Veyssiere, A.

    1957-01-01

    An important part of the experiments carried out on the reactor G1 during a period of shut-down has consisted in determinations of critical sizes, and measurements of flux distribution by irradiations of detectors. This report deals with the following points: 1- Critical sizes of the flat pile, the long pile and the uranium-thorium pile. 2- Flux charts of the same piles, and study of an exponential experiment. 3- Determination of the slit effect. 4- Calculation of the anisotropy of the lattice. 5- Description of the experimental apparatus of the irradiation measurements. (author) [fr

  7. 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.

  8. 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

  9. Irradiated fuel behavior under accident heating conditions and correlation with fission gas release and swelling model (Chicago)

    International Nuclear Information System (INIS)

    Kryger, B.; Ducamp, F.; Combette, P.

    1981-08-01

    We analyse the mixed oxide fast fuel response to off normal conditions obtained by means of an out-of-pile transient simulation apparatus designed to provide direct observations (temperature, pressure, fuel motion) of fuel fission gas phenomena that might occur during the transients. The results are concerning fast transient tests (0,1 to 1 second) obtained with high gas concentration irradiated fuel (4 to 7 at % burn up, 0,4 cm 3 Xe + Kr /g.UPuO 2 ). The kinetics of fission gas release during the transients have been directly measured and then compared with the calculated results issued of the Chicago model. This model agrees, quite well, with other experiments done in the silene prompt reactor. Other gases than xenon and krypton (such as hydrogen and carbon monoxide) do not play any role in fuel behavior, since they have been completely ruled out

  10. The Windscale piles initial decommissioning programme

    International Nuclear Information System (INIS)

    Boorman, T.; Woodacre, A.

    1992-01-01

    The two Windscale Piles, the first large scale nuclear reactors built in the UK were constructed in the late 1940's and operated until the accident in Pile No 1 caused their permanent shutdown in 1957. Following a period of care and maintenance, a programme of initial decommissioning has begun aimed at establishing a satisfactory long-term safe condition for the Windscale Piles Complex with minimum maintenance commitments. For the chimneys this involves the removal of the top filter sections. The pond will be emptied and cleaned. For the Piles the initial phase includes the consideration of options for long-term decommissioning solutions. (author)

  11. 30 CFR 77.215-2 - Refuse piles; reporting requirements.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Refuse piles; reporting requirements. 77.215-2... COAL MINES Surface Installations § 77.215-2 Refuse piles; reporting requirements. (a) The proposed location of a new refuse pile shall be reported to and acknowledged in writing by the District Manager...

  12. Cimentaciones actuales de los rascacielos de Chicago

    Directory of Open Access Journals (Sweden)

    López Ruiz, Álvaro

    1972-05-01

    Full Text Available The main characteristics of the land and the supporting rock are given, and the pattern of present performance of Chicago's deep foundations is briefly described. This city probably has some of the highest skyscrapers in the world with deep foundations resting on rock, and possibly as well, the highest building with foundations resting on soil. The construction of piles of large diameter, resting on rock and concreted dry, as prescribed by the City Code, is the weak point which sometimes makes it necessary to insert a permanent lining tube deep into the rock in order to obtain the required leaktightness, or to pump out significant quantities of water from the bottom of the excavations. The latter can cause the dragging away of earth from adjacent foundations resting on soil at a higher level, as with the drift of the phreatic level, with the consequent risk of producing appreciable subsidence, and even undermining. To overcome this problem we carried out for the first time in Chicago, a pre-sealing, by means of chemical injections, of the supporting rock of the periphery piles of some deep foundations, by drilling into the centre of each pile, which has also been useful as a means of examining the subsoil before proceeding to the perforation of each pile. This operation was carried out on the large diameter piles of the foundations of the Northern Trust Bank skyscraper, at present under construction, with very satisfactory results. The method of pre-sealing the rock used and the results obtained are described.Se presentan las características principales del terreno y de la roca de apoyo y se describe someramente la forma de ejecución actual de las cimentaciones profundas de Chicago. Esta ciudad tendrá (en breve varios de los rascacielos más altos del mundo apoyados en cimentaciones profundas sobre roca y, posiblemente, también el edificio más alto apoyado sobre suelo. La construcción de pilotes de gran diámetro apoyados sobre roca y

  13. Experience of the remote dismantling of the Windscale advanced gas-cooled reactor and Windscale pile chimneys

    International Nuclear Information System (INIS)

    Wright, E.M.

    1993-01-01

    This paper gives brief descriptions of some of the remote dismantling work and equipment used on two large decommissioning projects: the BNFL Windscale Pile Chimneys Project (remote handling machine, waste packaging machine, remotely controlled excavator, remotely controlled demolition machine) and the AEA Windscale Advanced Gas-cooled Reactor Project (remote dismantling machine, operational waste, bulk removal techniques, semi-remote cutting operations)

  14. Impedance function of a group of vertical piles

    International Nuclear Information System (INIS)

    Wolf, J.P.; Arx, G.A. von

    1978-01-01

    Impedance and transfer functions of a group of vertical piles located in any desired configuration in plan in a horizontally stratified soil layer are derived. Hysteretic and radiation damping are accounted for. The method separates the piles and the soil, introducing unknown interaction forces. The total flexibility matrix of the soil is constructed, superposing the (complex) flexibility coefficients caused by the interaction forces of a single pile only. The dependence of the impedance and transfer functions on the oscllating frequency for foundations with different numbers of piles is investigated. Pile-soil-pile interaction is shown to be very important for all modes of vibration. The procedure is used in the seismic analysis of a reactor building. (Author)

  15. Applications of a lead pile coupled with fast reactor core of Yayoi as an intermediate energy neutron standard field

    International Nuclear Information System (INIS)

    Kosako, Toshiso; Nakazawa, Masaharu; Sekiguchi, Akira; Wakabayashi, Hiroaki.

    1976-10-01

    Intermediate neutron column of YAYOI reactor is here evaluated as an intermediate energy neutron standard field which provides a base of the measurements of various reaction rates in that energy region, including detector calibration and Doppler coefficient determination. The experiments were performed using YAYOI's core as a fast neutron source by coupling with the large lead pile, which is a 160 ton's octagon of 2.5 m high and with a thickness of about 2.5 m face to face distance. Spatial variation of the neutron flux in the lead pile was estimated by gold activation foils, and the neutron spectrum by sandwich foils, a helium-3 proportional counter and a proton recoil counter. The calculated results were obtained using one and two- dimensional discrete ordinate code, ANISN and TWOTRAN II. Through comparison of experiment with calculation, it became clear that the neutron field at the central block has simple energy spectrum and stable spatial distribution of the neutron flux, the absolute of which was 5.0 x 10 4 (n/cm 2 /sec/Watt) at the representative energy of 1 KeV. The energy spectrum of the position and the spatial dependent neutron flux in the lead pile are both represented by the semiempirical formula, which must be useful both for evaluation of experimental data and for future applications. (auth.)

  16. 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.)

  17. 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.

  18. The pile EL3; Pile EL3

    Energy Technology Data Exchange (ETDEWEB)

    Robert, J.; Raievski, V. [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires; Hainzelin, J. [Chantiers de l' Atlantique (Penhoet-Loire), 75 - Paris (France)

    1959-07-01

    The programme of the high flux laboratory pile EL3 was laid down in october 1954. It is a heavy-water moderated and cooled pile. The fuel rods are of uranium metal with 1.6 per cent - 2 per cent of molybdenum, with aluminium canning. The maximum thermal flux in the moderator is 10{sup 14} n/cm{sup 2}/s. Studies began in january 1955, construction in may 1955, and the first divergence took place in July 1957. This report gives a general description of the pile and its adjacent buildings, the physical study of the pile, and certain technological studies carried out for the construction of EL3. (author) [French] Le programme de la pile laboratoire a haut flux EL3, a ete fixe en octobre 1954. C'est une pile moderee et refroidie a l'eau lourde. Les barres de combustible sont en uranium metal a 1,6 pour cent - 2 pour cent de molybdene, gainees a l'aluminium. Le flux thermique maximum dans le moderateur est de 10{sup 14} n/cm{sup 2}/s. Les etudes ont commence en janvier 1955, la construction en mai 1955, la premiere divergence a eu lieu en juillet 1957. On trouvera dans ce rapport, une description generale de la pile et de ses batiments annexes, l'etude physique de la pile et un certain nombre d'etudes technologiques executees pour la construction d'EL3. (auteur)

  19. The pile EL3; Pile EL3

    Energy Technology Data Exchange (ETDEWEB)

    Robert, J; Raievski, V [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires; Hainzelin, J [Chantiers de l' Atlantique (Penhoet-Loire), 75 - Paris (France)

    1959-07-01

    The programme of the high flux laboratory pile EL3 was laid down in october 1954. It is a heavy-water moderated and cooled pile. The fuel rods are of uranium metal with 1.6 per cent - 2 per cent of molybdenum, with aluminium canning. The maximum thermal flux in the moderator is 10{sup 14} n/cm{sup 2}/s. Studies began in january 1955, construction in may 1955, and the first divergence took place in July 1957. This report gives a general description of the pile and its adjacent buildings, the physical study of the pile, and certain technological studies carried out for the construction of EL3. (author) [French] Le programme de la pile laboratoire a haut flux EL3, a ete fixe en octobre 1954. C'est une pile moderee et refroidie a l'eau lourde. Les barres de combustible sont en uranium metal a 1,6 pour cent - 2 pour cent de molybdene, gainees a l'aluminium. Le flux thermique maximum dans le moderateur est de 10{sup 14} n/cm{sup 2}/s. Les etudes ont commence en janvier 1955, la construction en mai 1955, la premiere divergence a eu lieu en juillet 1957. On trouvera dans ce rapport, une description generale de la pile et de ses batiments annexes, l'etude physique de la pile et un certain nombre d'etudes technologiques executees pour la construction d'EL3. (auteur)

  20. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    Dirar, H. M.

    2012-06-01

    The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

  1. Study of the light emitted in the moderation of a heavy-water pile

    International Nuclear Information System (INIS)

    Breton, D.

    1958-06-01

    During the running of a reactor which uses water as a neutron moderator, a bluish light is seen to appear inside the liquid. A detailed study of this radiation, undertaken on the Fontenay-aux-Roses pile, has shown that the spectrum is identical with that which characterises the light produced by the Cerenkov effect. The light intensity as a function of the pile power grows exponentially as a function of time when the pile diverges, with a lifetime equal to that of the rise in power. An examination of the various particles present in the pile has led to the conclusion that only electrons with an energy greater than 260 keV con produce the Cerenkov light. The light source thus produced is about 2.10 6 photons/cm 2 of water, when the pile power equals 1 watt. (author) [fr

  2. 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)

  3. 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.

  4. 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

  5. Welcome to the home page of the Decontamination and Decommissioning Program at Argonne National Laboratory

    International Nuclear Information System (INIS)

    1996-01-01

    This report presents the details of the Argonne National Laboratory Home Page. Topics discussed include decontamination and decommissioning of the following: hot cells; remedial action; Experimental Boiling Water Reactor; glove boxes; the Chicago Pile No. 5 Research Reactor Facility; the Janus Reactor; Building 310 Retention Tanks; Zero Power Reactors 6 and 9; Argonne Thermal Source Reactor; cyclotron facility; and Juggernaut reactor

  6. Analysis of pile foundations under dynamic loads

    International Nuclear Information System (INIS)

    Waas, G.; Hartmann, H.G.

    1981-01-01

    A method is presented for the analysis of pile foundations which are subjected to horizontal dynamic loads from earthquakes, airplane impact, gas explosion or other sources. The motion of the pile cap and the pile forces are computed. - The loads may be applied to the pile cap or directly to the piles (e.g. by earthquake wave motion). The soil may be stratified and is considered to be an elastic or visco-elastic medium. The piles are assumed vertical. The method makes use of an approximate fundamental solution for displacements caused by a dynamic point load in a layered visco-elastic medium. The approximation involves a discretization of the medium in the vertical direction. In horizontal directions the medium is treated by continuum theory. The soil medium supports each pile at about 10 to 20 nodes. A dynamic flexiblity matrix for the soil is derived which relates the elastic, damping and inertial forces of the soil to the displacements at each node. It includes effects of radiation damping. All piles are coupled through the soil flexibility matrix. The piles are modelled by beam elements. Transient response is computed using fast discrete Fourier transforms. The arrangement of the piles is arbitrary. However, simple and double symmetry can be accounted for by the computer program. When the pile arrangement is axisymmetric, the degrees of freedom can be reduced to only those of two piles per ring. The influence of the number of piles and the influence of the pile spacing on group stiffness and on pile forces is presented for two soil profiles. Dynamic effects on pile forces of a foundation for a reactor building are studied. They are significant when soils are soft. (orig.)

  7. Bracing system of the reflecting sheets making up an insulating pile

    International Nuclear Information System (INIS)

    Carr, R.W.

    1976-01-01

    In order to reduce heat and radiation losses, the body of nuclear reactors and the connected pipe work are encased in reflecting and insulating piles of thin spaced sheets of aluminium or stainless steel. These spaced sheets are then encased in thicker and more solid internal and external shells. The piles and shells are generally shaped to follow the contour of the reactor and connected piping. It is therefore necessary to have available a study bracing system to keep the pile intact during the various handling and assembly operations. The fastening system must also exert an effect on the edge of the pile to prevent the sheets making it up from shifting in relation to each other. The description is given of a fastening system that includes an oblong section to be fitted along the edges of the piles up sheets; bracing substantially perpendicular to the oblong section, to space the sheets of the stack in pairs; and a maintaining system, normally perpendicular to the oblong section, to enable the fastener to be clipped to the edge of the sheets by bending it around the edge of each sheet of the pile [fr

  8. Reactor building seismic analysis of a PWR type - NPP

    International Nuclear Information System (INIS)

    Kakubo, Masao

    1983-01-01

    Earthquake engineering studies raised up in Brazil during design licensing and construction phases of Almirante Alvaro Alberto NPP, units 1 and 2. State of art of soil - structure interaction analysis with particular reference to the impedance function calculation analysis with particular reference to the impedance function calculation of a group of pile is presented in this M.Sc. Dissertation, as an example the reactor building dynamic response of a 1325 MWe NPP PWR type is calculated. The reactor building is supported by a pile foundation with 2002 end bearing piles. Upper and lower bound soil parameters are considered in order to observe their influence on dynamic response of structure. Dynamic response distribution on pile heads show pile-soil-pile interaction effects. (author)

  9. Fabrication of the 4. set of fuel elements for the experimental pile EL2; Fabrication du 4. jeu de barreaux de la pile d'essai EL2

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-07-01

    The reactor EL2 is the second atomic reactor built in France. It is a laboratory reactor using heavy water and natural uranium. Its cooling circuit operates with compressed CO{sub 2} gas at 8 kg/cm{sup 2} pressure. The subject of this lecture is the manufacturing of the fourth set of rods. The principle of uranium-can connection is exposed: that is the principle of a pre-pressed bound can. The EL2 reactor has been a prototype with respect to this aspect of the question, and a prototype which has been quite satisfactory. The main steps of the fabrication are exposed: the {gamma} phase extension of uranium, the machining, the three canning (die canning, hydraulic canning, compressed air treatment), the automatic argon arc welding of cups and the different manufacturing controls. (author) [French] Le reacteur EL2 est le deuxieme reacteur construit en France. C'est un reacteur de recherches qui utilise de l'eau lourde et de l'uranium naturel. Il est refroidi par du gaz carbonique sous 8 kg/cm{sup 2} de pression. On etudie dans cet expose la fabrication du quatrieme jeu d'elements combustibles. Le principe de la liaison uranium-gaine est expose: c'est celui d'une gaine precontrainte. La pile EL2 a constitue un prototype a ce point de vue, prototype qui a donne entiere satisfaction. Les principales etapes de la fabrication sont ensuite expliquees: le filage {gamma} de l'uranium, l'usinage des barreaux, les trois operations de gainages (gainage par filiere, gainage hydraulique, gainage a chaud), la soudure automatique des bouchons a l'argon-arc et les differents controles de fabrication. (auteur)

  10. 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

  11. Elementary calculation of the shutdown delay of a pile; Calcul elementaire de la periode d'extinction d'une pile

    Energy Technology Data Exchange (ETDEWEB)

    Yvon, J

    1949-04-01

    This study analyzes theoretically the progress of the shutdown of a nuclear pile (reactor) when a cadmium rod is introduced instantaneously. For simplification reasons, the environment of the pile is considered as homogenous and only thermal neutrons are considered (delayed neutrons are neglected). Calculation is made first for a plane configuration (plane vessel, plane multiplier without reflector, and plane multiplier with reflector), and then for a cylindrical configuration (multiplier without reflector, multiplier with infinitely thick reflector, finite cylindrical piles without reflector and with reflector). The self-sustain conditions are calculated for each case and the multiplication length and the shutdown delay are deduced. (J.S.)

  12. 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.

  13. Radio-active pollution near natural uranium-graphite-gas reactors; La pollution radioactive aupres des piles uranium naturel - graphite - gaz

    Energy Technology Data Exchange (ETDEWEB)

    Chassany, J; Pouthier, J; Delmar, J [Commissariat a l' Energie Atomique, Chusclan (France). Centre de Production de Plutonium de Marcoule

    1967-07-01

    The results of numerous evaluations of the contamination are given: - Reactors in operation during maintenance operations. - Reactors shut-down during typical repair operations (coolants, exchangers, interior of the vessel, etc. ) - Following incidents on the cooling circuit and can-rupture. They show that, except in particular cases, it is the activation products which dominate. Furthermore, after ten years operation, the points at which contamination liable to emit strong doses accumulates are very localized and the individual protective equipment has not had to be reinforced. (authors) [French] Les resultats de nombreuses evaluations de la contamination sont donnes: - Piles en marche pendant les operations d'entretien - Piles a l'arret au cours des chantiers caracteristiques (refrigerants, echangeurs, interieur du caisson, etc.) - A la suite d'incidents sur le circuit de refroidissement et de rupture de gaine. Ils montrent que, sauf cas particulier, ce sont essentiellement les produits d'activation qui dominent. Par ailleurs apres 10 ans de fonctionnement, les points d'accumulation de la contamination susceptibles de delivrer des debits de dose importants restent tres localises et les moyens de protection individuels utilises n'ont pas du etre renforces. (auteurs)

  14. Radio-active pollution near natural uranium-graphite-gas reactors; La pollution radioactive aupres des piles uranium naturel - graphite - gaz

    Energy Technology Data Exchange (ETDEWEB)

    Chassany, J.; Pouthier, J.; Delmar, J. [Commissariat a l' Energie Atomique, Chusclan (France). Centre de Production de Plutonium de Marcoule

    1967-07-01

    The results of numerous evaluations of the contamination are given: - Reactors in operation during maintenance operations. - Reactors shut-down during typical repair operations (coolants, exchangers, interior of the vessel, etc. ) - Following incidents on the cooling circuit and can-rupture. They show that, except in particular cases, it is the activation products which dominate. Furthermore, after ten years operation, the points at which contamination liable to emit strong doses accumulates are very localized and the individual protective equipment has not had to be reinforced. (authors) [French] Les resultats de nombreuses evaluations de la contamination sont donnes: - Piles en marche pendant les operations d'entretien - Piles a l'arret au cours des chantiers caracteristiques (refrigerants, echangeurs, interieur du caisson, etc.) - A la suite d'incidents sur le circuit de refroidissement et de rupture de gaine. Ils montrent que, sauf cas particulier, ce sont essentiellement les produits d'activation qui dominent. Par ailleurs apres 10 ans de fonctionnement, les points d'accumulation de la contamination susceptibles de delivrer des debits de dose importants restent tres localises et les moyens de protection individuels utilises n'ont pas du etre renforces. (auteurs)

  15. Study of the light emitted in the moderation of a heavy-water pile; Etude de la lumiere emise dans le moderateur d'une pile a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-06-15

    During the running of a reactor which uses water as a neutron moderator, a bluish light is seen to appear inside the liquid. A detailed study of this radiation, undertaken on the Fontenay-aux-Roses pile, has shown that the spectrum is identical with that which characterises the light produced by the Cerenkov effect. The light intensity as a function of the pile power grows exponentially as a function of time when the pile diverges, with a lifetime equal to that of the rise in power. An examination of the various particles present in the pile has led to the conclusion that only electrons with an energy greater than 260 keV con produce the Cerenkov light. The light source thus produced is about 2.10{sup 6} photons/cm{sup 2} of water, when the pile power equals 1 watt. (author) [French] Lors du fonctionnement d'un reacteur utilisant l'eau comme moderateur de neutrons, on constate l'apparition d'une lumiere bleutee au sein du liquide. Une etude approfondie de ce rayonnement, entreprise sur la pile Fontenay-aux-Roses a montre que le spectre est identique a celui caracterisant la lumiere produite par effet Cerenkov. L'intensite lumineuse en fonction de Ia puissance de la pile, lors d'une divergence croit exponentiellement en fonction du temps avec une periode egale a celle de la montee en puissance. L'examen des diverses particules presentes dans la pile a permis de conclure que seuls les electrons ayant une energie superieure a 260 keV peuvent produire la lumiere Cerenkov. La source lumineuse ainsi constituee est d'environ 2.10{sup 6} photons/cm{sup 2} d'eau, lorsque la puissance de la pile est egale a 1 watt. (auteur)

  16. 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.

  17. Analysis of transients in the SRP test pile

    International Nuclear Information System (INIS)

    Church, J.P.

    1976-11-01

    Analysis of the hypothetical upper limit accident in the Savannah River Test Pile showed that the offsite thyroid dose from fission product release would be -3 of the 10-CFR-100 guideline dose for 95 percent of measured meteorological conditions. Offsite whole body dose would be negligible. The Test Pile was modified to limit the length of test piece that can be charged to the pile. These modifications reduce the potential offsite dose to -5 of the regulatory guidelines. Assessment of Test Pile safety included calculations of transients initiated by a variety of reactivity additions that were either terminated or not terminated by safety systems. Reactivity addition mechanisms considered were abnormally driving control rods out of the pile and charging abnormal test pieces into the pile. The transients were evaluated in the adiabatic approximation in which three-dimensional calculations of static flux shapes and reactivity were superimposed on point reactor kinetics calculations. Negative reactivity feedback effects appropriate for the pile and the temperature dependence of material properties, such as specific heat and thermal conductivity, were included. The results show that, for the worst initiators, safety systems can prevent the temperature rise from exceeding 1 0 C anywhere in the Test Pile. If the safety systems do not function, the pile temperatures will increase until the transient is ended by the inherent negative reactivity effects, including the melting of some fuel

  18. Study of the light emitted in the moderation of a heavy-water pile; Etude de la lumiere emise dans le moderateur d'une pile a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-06-15

    During the running of a reactor which uses water as a neutron moderator, a bluish light is seen to appear inside the liquid. A detailed study of this radiation, undertaken on the Fontenay-aux-Roses pile, has shown that the spectrum is identical with that which characterises the light produced by the Cerenkov effect. The light intensity as a function of the pile power grows exponentially as a function of time when the pile diverges, with a lifetime equal to that of the rise in power. An examination of the various particles present in the pile has led to the conclusion that only electrons with an energy greater than 260 keV con produce the Cerenkov light. The light source thus produced is about 2.10{sup 6} photons/cm{sup 2} of water, when the pile power equals 1 watt. (author) [French] Lors du fonctionnement d'un reacteur utilisant l'eau comme moderateur de neutrons, on constate l'apparition d'une lumiere bleutee au sein du liquide. Une etude approfondie de ce rayonnement, entreprise sur la pile Fontenay-aux-Roses a montre que le spectre est identique a celui caracterisant la lumiere produite par effet Cerenkov. L'intensite lumineuse en fonction de Ia puissance de la pile, lors d'une divergence croit exponentiellement en fonction du temps avec une periode egale a celle de la montee en puissance. L'examen des diverses particules presentes dans la pile a permis de conclure que seuls les electrons ayant une energie superieure a 260 keV peuvent produire la lumiere Cerenkov. La source lumineuse ainsi constituee est d'environ 2.10{sup 6} photons/cm{sup 2} d'eau, lorsque la puissance de la pile est egale a 1 watt. (auteur)

  19. Taming Windscale's piles

    International Nuclear Information System (INIS)

    Adams, A.L.

    1989-01-01

    The options as to what to do with the Windscale Piles are being assessed before a final decision on decommissioning is made. Both Piles were shutdown in 1957 following the fire in the Pile number 1. Pile 1 still contains 22 tons of natural uranium fuel. The details of graphite moderator content, biological shielding and other components and containment are given. The fuel and isotope channels in Pile 2 have been examined and the air and water ducts have been inspected. The chimneys of both Piles are contaminated and all entrances have been sealed. Before any work starts the air outlet ducts will be sealed from the chimney and a ventilation system installed. A manipulator is being prepared to remove the remaining fuel elements from both Piles. The proposed decommissioning programme for both Piles is outlined. (U.K.)

  20. 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

  1. 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

  2. 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

  3. Piles of dislocation loops in real crystals. 2. Evolution of dislocation piles under irradiation

    International Nuclear Information System (INIS)

    Dubinko, V.I.; Turkin, A.A.; Yanovskij, V.V.

    1985-01-01

    The given paper considers evolution of piles in a real molybdenum crystal under neutron irradiation. Obtained was a stability criterium, when meeting it interstitial piles (one-dimensional periodical structures of interstitial loops) in the crystal tend to stationary state under the irradiation and, when disturbing the criterium, they disintegrate into rapidly growing interstitial isolated loops. It was also shown that the generation of dense vacancy piles results in the formation of an ordering structure of isolated vacancy loops. Theoretical results agree good with experimental data

  4. Neutron spectrum in small iron pile surrounded by lead reflector

    International Nuclear Information System (INIS)

    Kimura, Itsuro; Hayashi, S.A.; Kobayashi, Katsuhei; Matsumura, Tetsuo; Nishihara, Hiroshi.

    1978-01-01

    In order to save the quantity of sample material, a possibility to assess group constants of a reactor material through measurement and analysis of neutron spectrum in a small sample pile surrounded by a reflector of heavy moderator, was investigated. As the sample and the reflector, we chose iron and lead, respectively. Although the time dispersion in moderation of neutrons was considerably prolonged by the lead reflector, this hardly interferes with the assessment of group constants. Theoretical calculation revealed that both the neutron flux spectrum and the sensitivity coefficient of group constants in an iron sphere, 35 cm in diameter surrounded by the lead reflector, 25 cm thick, were close to those of the bare iron sphere, 108 cm in diameter. The neutron spectra in a small iron pile surrounded by a lead reflector were experimentally obtained by the time-of-flight method with an electron linear accelerator and the result was compared with the predicted values. It could be confirmed that a small sample pile surrounded by a reflector, such as lead, was as useful as a much larger bulk pile for the assessment of group constants of a reactor material. (auth.)

  5. Earthquakes - dynamic loads on the foundation piles of a nuclear reactor

    International Nuclear Information System (INIS)

    Henning, F.; Dufour, C.; Fuzier, J.P.; Cheyrezy, M.

    1976-01-01

    The purpose of this paper is to analyse the behaviour of a piled foundation of a nuclear power plant submitted to earthquake loadings. The actual horizontal stiffness of the compound 'soil + structure' is calculated from the stiffness of the piles and the dynamic characteristics of the different layers of soil. The maximum amplitude of the horizontal displacement of the raft foundation is obtained by an iterative calculation from the displacement spectrum applied to the surrounding earth. At this stage, the moment diagrams for the different types of piles and their required reinforcement are calculated. The response of the nuclear plant components is determined from the displacement spectrum previously obtained for the raft. (author)

  6. 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.

  7. 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

  8. Elementary calculation of the shutdown delay of a pile

    International Nuclear Information System (INIS)

    Yvon, J.

    1949-04-01

    This study analyzes theoretically the progress of the shutdown of a nuclear pile (reactor) when a cadmium rod is introduced instantaneously. For simplification reasons, the environment of the pile is considered as homogenous and only thermal neutrons are considered (delayed neutrons are neglected). Calculation is made first for a plane configuration (plane vessel, plane multiplier without reflector, and plane multiplier with reflector), and then for a cylindrical configuration (multiplier without reflector, multiplier with infinitely thick reflector, finite cylindrical piles without reflector and with reflector). The self-sustain conditions are calculated for each case and the multiplication length and the shutdown delay are deduced. (J.S.)

  9. 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.

  10. Behaviour of heavy water in nuclear reactors of the CEA; Comportement de l'eau lourde dans les piles du C.E.A

    Energy Technology Data Exchange (ETDEWEB)

    Chenouard, J; Dirian, G; Roth, E [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    In the two heavy water reactors of the CEA: Zoe and P-2, we do: A) the supervision of the isotopic composition of the heavy water; B) the supervision of gases released by the decomposition of the heavy water under radiation, and to their recombination; C) periodic analyses of impurities. (M.B.) [French] Dans les deux piles a eau lourde du Commissariat a l'Energie Atomique: Zoe et P 2, nous effectuons: A) la surveillance de la composition isotopique de l'eau lourde; B) la surveillance des gaz degages par la decomposition de l'eau lourde sous radiation, et a leur recombinaison; C) des analyses periodiques d'impuretes. (M.B.)

  11. 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)

  12. Decommissioning and dismantling of 305-M test pile at the Savannah River Plant

    International Nuclear Information System (INIS)

    Horton, H.L.

    1985-01-01

    The 305-M Test Pile was started up at the Savannah River Plant in 1952 and operated until 1981. The pile was used to measure the uranium content of reactor fuel. In 1984 work began to decommission and dismantle the pile. Extensive procedures were used that included a detailed description of the radiological controls and safety measures. These controls allowed the job to be completed with radiation doses as low as reasonably achievable

  13. Research reactors; Les piles de recherche

    Energy Technology Data Exchange (ETDEWEB)

    Kowarski, L. [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires]|[Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)

    1955-07-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  14. Research reactors; Les piles de recherche

    Energy Technology Data Exchange (ETDEWEB)

    Kowarski, L [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires; [Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)

    1955-07-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  15. 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)

  16. 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

  17. Big Pile or Small Pile?

    Science.gov (United States)

    Branca, Mario; Quidacciolu, Rossana G.; Soletta, Isabella

    2013-01-01

    The construction of a voltaic pile (battery) is a simple laboratory activity that commemorates the invention of this important device and is of great help in teaching physics. The voltaic pile is often seen as a scientific toy, with the "pile" being constructed from fruit. These toys use some strips of copper and zinc inserted in a piece…

  18. Critical sizes and flux distributions in the shut down pile; Tailles critiques et cartes de flux a froid

    Energy Technology Data Exchange (ETDEWEB)

    Banchereau, A; Berthier, P; Genthon, J P; Gourdon, C; Lattes, R; Martelly, J; Mazancourt, R de; Portes, L; Sagot, M; Schmitt, A P; Tanguy, P; Teste du Bailler, A; Veyssiere, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    An important part of the experiments carried out on the reactor G1 during a period of shut-down has consisted in determinations of critical sizes, and measurements of flux distribution by irradiations of detectors. This report deals with the following points: 1- Critical sizes of the flat pile, the long pile and the uranium-thorium pile. 2- Flux charts of the same piles, and study of an exponential experiment. 3- Determination of the slit effect. 4- Calculation of the anisotropy of the lattice. 5- Description of the experimental apparatus of the irradiation measurements. (author) [French] Une part importante des experiences a froid effectuees sur le reacteur G1 a consiste en des determinations de tailles critiques et des mesures de distributions de flux par irradiations de detecteurs. Le present rapport traite les points suivants: 1- Tailles critiques de la pile plate, de la pile longue, de la pile a uranium-thorium. 2 - Cartes de flux des memes piles et etude d'une experience exponentielle. 3 - Determination de l'effet de fente. 4 - Calcul de l'anisotropie du reseau. 5 - Description de l'appareillage experimental des mesures d'irradiations. (auteur)

  19. Critical sizes and flux distributions in the shut down pile; Tailles critiques et cartes de flux a froid

    Energy Technology Data Exchange (ETDEWEB)

    Banchereau, A.; Berthier, P.; Genthon, J.P.; Gourdon, C.; Lattes, R.; Martelly, J.; Mazancourt, R. de; Portes, L.; Sagot, M.; Schmitt, A.P.; Tanguy, P.; Teste du Bailler, A.; Veyssiere, A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    An important part of the experiments carried out on the reactor G1 during a period of shut-down has consisted in determinations of critical sizes, and measurements of flux distribution by irradiations of detectors. This report deals with the following points: 1- Critical sizes of the flat pile, the long pile and the uranium-thorium pile. 2- Flux charts of the same piles, and study of an exponential experiment. 3- Determination of the slit effect. 4- Calculation of the anisotropy of the lattice. 5- Description of the experimental apparatus of the irradiation measurements. (author) [French] Une part importante des experiences a froid effectuees sur le reacteur G1 a consiste en des determinations de tailles critiques et des mesures de distributions de flux par irradiations de detecteurs. Le present rapport traite les points suivants: 1- Tailles critiques de la pile plate, de la pile longue, de la pile a uranium-thorium. 2 - Cartes de flux des memes piles et etude d'une experience exponentielle. 3 - Determination de l'effet de fente. 4 - Calcul de l'anisotropie du reseau. 5 - Description de l'appareillage experimental des mesures d'irradiations. (auteur)

  20. On-line fast flux measurements in the BR2 reactor

    International Nuclear Information System (INIS)

    Vermeeren, L.

    2009-01-01

    Since 2001, CEA-Cadarache and the Belgian Nuclear Research Centre SCK-CEN are collaborating on the development and in-pile qualification of subminiature fission chambers (diameter of 1.5 mm). Initially, efforts concentrated on fission chambers for the in-pile measurement of thermal fluxes (with 235 U as fissile material). Meanwhile successful long-term tests of the prototypes have been performed in various environments: in low temperature (40-100 degress Celsius) BR2 pool water (up to a thermal neutron fluence of 3 1 0 21 n/cm 2 ) and in the CALLISTO PWR loop (300 degrees Celsius, 155 bars). The long-term qualification of derived industrial detectors (Photonis CFUZ53) in CALLISTO is still ongoing. However, for various types of irradiations in research reactors, the knowledge of the evolution of the fast neutron flux is even of more interest than the thermal flux data. Therefore the collaboration program was extended to the development and the in-pile qualification of subminiature or miniature fission chambers (with 3 mm diameter) for fast neutron detection, for which 242 Pu was selected as the optimal fissile material. In order to achieve the on-line in-pile measurement of fast neutron flux, the fission chambers will be operated in the Campbelling mode (based on the mean square fluctuation of the detector current). In this mode the gamma induced contribution to the signal can be efficiently suppressed. Moreover, a data processing software will take into account the evolution of the fissile deposit in order to assess on-line the fast flux sensitivity and to correct for the low energy neutron contributions. The final objective is to qualify a Fast Neutron Detector System (FNDS) able to provide on-line data for local fast neutron fluxes in Material Testing Reactors. The on-line measurement of the fast neutron flux would contribute significantly to the characterization of the irradiation conditions during test experiments with materials and innovative fuel elements

  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. Out-pile test plan for lifetime extension of shutoff units in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Joe, Y. G.; Lee, J. H.; Jeong, Y. H.; Woo, S. I.; Ryu, J. S.; Kim, Y. G.; Park, Y. C.; Kim, H. G.; Woo, J. S. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    It is estimated that the number of drops of shutoff rods in HANARO will reach the endurance verified numbers before the end of the reactor life. To resolve this situation, we have a plan to prepare of a new spare unit by the performance verification test for the local product, and extend the lifetime of shutoff units installed in the reactor by performing an additional endurance test in the out-pile test facility using an existing spare unit. This paper describes the overall situations and test plan for the out-pile test to extend the lifetime extension of shutoff unit.

  3. 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

  4. Radioprotection problems resulting from the presence of experimental devices around an atomic reactor; Problemes de radioprotection poses par l'implantation de dispositifs experimentaux aupres d'une pile atomique

    Energy Technology Data Exchange (ETDEWEB)

    Fitoussi, L; Lebouleux, Ph; Bricard, Ph; Moreau, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    The setting up of experimental devices around a reactor produces dangers of irradiation and radioactive contamination which can become very great in the case of an accident, especially if the in-pile portion contains fissile matter. This may result in irradiation of personnel, prohibition of access to the experimental zones until the sources of irradiation and contamination have been eliminated, and a prolonged stoppage of the reactor. The plans for an in-pile experiment should take into account radioprotection factors; the aim of these is to reduce to a minimum the radioactive risks normally encountered during the experiment and to eliminate any risks of bad accidents and their consequences. In this report are classified the various types of experiments requiring installations outside the pile itself; for each of these experiments the particular radioprotection factors are given. In order to make possible a study of the radioactive dangers likely to arise during a projected experiment, the authors summarize the physical and technical data required by radioprotection specialists and give the rules and general advice concerning radioprotection which should be useful during the planning of an in-pile experiment and the setting-up of the equipment. (authors) [French] L'implantation de dispositifs experimentaux aupres des reacteurs cree des risques d'irradiation et de contamination radioactive qui peuvent devenir importants en cas d'accident, surtout si la partie en pile comprend des matieres fissiles. Il peut en resulter des irradiations de personne, l'interdiction des aires experimentales jusqu'a elimination des sources d'irradiation et de contamination, un arret prolonge de la pile. L'etude d'un projet d'experience en pile doit donc tenir compte des considerations de radioprotection dont le souci est de reduire aux niveaux tolerables les risques radioactifs inherents au fonctionnement normal de l'experience et d'eliminer les risques d'accidents graves et leurs

  5. 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

  6. The direct conversion of heat into electricity in reactors; Conversion directe de la chaleur en electricite dans les piles

    Energy Technology Data Exchange (ETDEWEB)

    Devin, B; Bliaux, J; Lesueur, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The direct conversion of heat into electricity by thermionic emission in an atomic reactor has been studied with the triple aim of its utilisation: as an energy source for a space device, at the head of a conventional conversion system in power installations, or finally in association with the thermoelectric conversion in very low power installations. The laboratory experiments were mainly orientated towards the electron extraction of metals and compounds and their behaviour at high temperatures. Converters furnishing up to 50 amps at 0. 4 volts with an efficiency close to 10 p. 100 have been constructed in the laboratory; the emitters were heated by electron bombardment and were composed of tungsten covered with an uranium carbide deposit or molybdenum covered with cesium. The main aspects of the coupling between the converter and the reactor have been covered from the point of view of electronics: the influence of the mismatching of the load on the temperature of the emitter and the influence of thermal flux density on the temperature of the emitter and the stability of the converter. Converters using uranium carbide as the electron emitter have been tested in reactors. Tests have been made under dynamic conditions in order to determine the dynamic characteristics. The load matching curves have been constructed and the overall performances of several cells coupled in such a way as to form a reactor rod have been deduced. This information is fundamental to the design of a control system for a thermionic conversion reactor. The problems associated with the reliability of thermionic converters connected in series in the same reactor rod have been examined theoretically. Finally, the absorption isotherms have been drawn at the ambient temperatures for krypton and xenon on activated carbon with the aim of investigating the escape of fission products in a converter. (author) [French] La conversion directe de chaleur en electricite par emission thermionique dans une

  7. Experiments prior to construction of the Rapsodie reactor (1962); Experiences preliminaires a la construction de la pile rapsodie (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Vautrey, L; Zaleski, C P [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1962-07-01

    Before proceeding to the construction of the various reactor components described in the paper 'Fast Breeder Reactor Rapsodie', many experimental studies of a hydraulic, thermal and mechanical character have been carried out, or are under consideration, to test the validity of the principles adopted in the Preliminary Project. This paper deals with the most important of these: 1. Studies of coolant circuit components: sodium pumps (mechanical or electromagnetic), Na-NaK and NaK ir heat exchangers, measuring instruments (flow rates, temperatures), sodium purification circuits, etc. 2. Studies in cooling of fuel and fertile assemblies: a) study of the sodium cooling carried out by means of hydraulic mockups (scale of 1: 1 or over) reproducing the flow of the coolant fluid in the piping, upstream from and inside the fuel and fertile elements. b) study of the cooling by gas and by immersion in lead, employed during handling and storage operations. 3. Studies of special reactor devices: fusible rotating linkage, parts of the control rod mechanisms. 4. Study of the reactor block and coolant circuits as a whole. This study is to begin at the end of the year. The mock-up, now nearing completion, reproduces on a scale of 1: 1 the installation provided in the Preliminary Project and includes: the reactor block, to which is connected a high flow ate sodium circuit, permitting of long-term tests and thermal shocks, and also, a control rod testing circuit; complete installation of the 1 MW and 10 MW coolant circuits, the performances of which it will be possible to check under various operational conditions. 5. A safety study carried out on a 3: 10 scale mock p comprising the whole of the reactor block and shielding, with the object of limiting the effects of any accidental liberation of energy of an explosive character. (authors) [French] Avant d'entreprendre la realisation des divers elements du reacteur decrit dans le rapport 'Reacteur rapide surregenerateur RAPSODIE

  8. 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

  9. 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)

  10. Swimming-pool piles; Piles piscines

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-07-01

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

  11. 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

  12. 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

  13. Design and fabrication report on capsule (11M 19K for out of pile test) for irradiation testing of research reactor materials at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.G.; Yang, S.W.; Park, S.J.; Shim, K.T.; Choo, K.N.; Oh, J.M.; Lee, B.C.; Choi, M.H.; Kim, D.J.; Kim, J.M.; Kang, S.H.; Chun, Y.B.; Kim, T.K.; Jeong, Y.H.

    2012-05-15

    As a part of the research reactor development project with a plate type fuel, the irradiation tests of graphite (Gr), beryllium (Be), and zircaloy 4 materials using the capsule have been investigating to obtain the mechanical characteristics such as an irradiation growth, hardness, swelling and tensile strength at the temperature below 100 .deg. C and the 30 MW reactor power. Then, A capsule to be able to irradiate materials(graphite, Be, zircaloy 4) under 100 .deg. C at the HANARO was designed and fabricated. After performing out of pile testing in single channel test loop by using the capsule, the final design of the capsules to be irradiated in CT and IR2 test hole of HANARO was approved, and 2 sets of capsule were fabricated. These capsules will be loaded in CT and IR2 test hole of HANARO, and be started the irradiation from the end of June, 2012. After performing the irradiation testing of 2 sets of capsule, PIE (Post Irradiation Examination) on irradiated specimens (Gr, Be, and zircaloy 4) will be carry out in IMEF (Irradiated Material Examination Facility). So, the irradiation testing will be contributed to obtain the characteristic data induced neutron irradiation on Gr, Be, and zircaloy 4. And then, it is convinced that these data will be also contributed to obtain the license for JRTR (Jordan Research and Training Reactor) and new research reactor in Korea, and export research reactors.

  14. The pile EL3

    International Nuclear Information System (INIS)

    Robert, J.; Raievski, V.

    1959-01-01

    The programme of the high flux laboratory pile EL3 was laid down in october 1954. It is a heavy-water moderated and cooled pile. The fuel rods are of uranium metal with 1.6 per cent - 2 per cent of molybdenum, with aluminium canning. The maximum thermal flux in the moderator is 10 14 n/cm 2 /s. Studies began in january 1955, construction in may 1955, and the first divergence took place in July 1957. This report gives a general description of the pile and its adjacent buildings, the physical study of the pile, and certain technological studies carried out for the construction of EL3. (author) [fr

  15. Mono pile foundation. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lyngesen, S.; Brendstrup, C.

    1997-02-01

    The use of mono piles as foundations for maritime structures has been developed during the last decades. The installation requirements within the offshore sector have resulted in equipment enabling driving of piles up to 3-4 m to large penetration depths. The availability of this equipment has made the use of large mono piles feasible as foundations for structures like wind turbines. The mono pile foundations consists of three parts; the bare pile, a conical transition and a boat landing. All parts are prefitted at the yard in order to minimise the installation work that has to be carried out offshore. The study of a mono pile foundations for a 1.5 MW wind turbine has been conducted for two locations, Horns Rev and Roedsand. Three different water depths: 5, 8 and 11 m have been investigated in the study. The on-site welding between pile and conical transition is performed by an automatic welding machine. Final testing and eventually repair of the weld are conducted at least 16 hours after welding. This is followed by final installation of J-tube, tie-in to subsea cables and installation of the impressed current system for corrosive protection of the mono pile. The total cost for procurement and installation of the mono pile using the welded connection is estimated. The price does not include procurement and installation of access platform and boat landing. These costs are estimated to 250.000 DKK. Depending on water depth the cost of the pile ranges from 2,2 to 2,7 million DKK. Procurement and fabrication of the pile are approx. 75% of the total costs. The remaining 25% are due to installation. The total costs are very sensitive to the unit price of pile steel. During the project it became obvious that ice load has a very large influence on the dimensions of the mono pile. (EG)

  16. Fast neutron flux in heavy water reactors; Flux de neutrons rapides dans les piles a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

    Brisbois, J; Katz, S [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, 92 (France)

    1966-07-01

    The possibility of calculating the fast neutron flux in a natural uranium-heavy water lattice by superposition of the individual contributions of the different fuel elements was verified using a one-dimension Monte-Carlo code. The results obtained are in good agreement with experimental measurements done in the core and reflector of the reactor AQUILON. (author) [French] La possibilite de calculer le flux de neutrons rapides dans un reseau d'uranium naturel a eau lourde par superposition des apports des divers barreaux, a ete verifiee en utilisant un code Monte-Carlo monodimensionel. Les resultats obtenus concordent avec des mesures experimentales effectuees dans le coeur et reacteur de la pile Aquilon. (auteurs)

  17. 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

  18. Summer 1942 in Chicago: Nuclear power

    International Nuclear Information System (INIS)

    Goldschmidt, B.

    1982-01-01

    On 2 December 1942 the first man-made nuclear reactor went critical. The nuclear age was born. In his recently completed 'political history of nuclear energy' M. Goldschmidt traces the whole story of the nuclear age from the discovery of fission to the present day. In the extract from his book printed below, M. Goldschmidt tells of his personal involvement in the US nuclear research programme and of his contact with the workers at the University of Chicago; he reminds us that Fermi's achievement, historic as it was, was not the first chain reaction to take place on earth

  19. Chicago's urban forest ecosystem: results of the Chicago Urban Forest Climate Project

    Science.gov (United States)

    Gregory E. McPherson; David J. Nowak; Rowan A. Rowntree

    1994-01-01

    Results of the 3-year Chicago Urban Forest Climate Project indicate that there are an estimated 50.8 million trees in the Chicago area of Cook and DuPage Counties; 66 percent of these trees rated in good or excellent condition. During 1991, trees in the Chicago area removed an estimated 6,145 tons of air pollutants, providing air cleansing valued at $9.2 million...

  20. Brighter future predicted at nuclear meetings in Chicago

    International Nuclear Information System (INIS)

    Stein, H.

    1993-01-01

    This article discusses the future of nuclear power in the United States and the rest of the world. It is a summary of a meeting of the American Nuclear Society/European Nuclear Society in Chicago. Some topics discussed include advanced reactor design, public relations, and nuclear safety

  1. A study of some radioprotection apparatuses used in the case of pool reactors; Etude de quelques dispositifs de radioprotection en service aupres des piles piscines

    Energy Technology Data Exchange (ETDEWEB)

    Robien, E de; Choudens, H de; Delpuech, J [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1965-07-01

    Various problems of radioprotection concerning swimming-pool reactors in Grenoble have led us to study adequate solutions: a) The automatic verification of the staff-radioactivity when coming out of Melusine or Siloe has been realized thanks to a {beta}{gamma} gate which is insensitive to the ambient background in the reactor-hall; b) The automatic verification of the contamination of the shoes of the agents working in these reactors has been realized with a dedicated device; c) The necessity to measure precisely {gamma} doses with the help of an autonomous apparatus has led to the making of a plastic-scintillator {gamma} dosimeter; d) The obligation to forbid the opening of doors in some places where there might be a great intensity of radiation, has led us to make doors open according to the intensity of radiation inside the rooms; e) The releases of radioactive iodine have been measured with activated charcoal cartridges that surround a scintillator connected with a unique channel selector; f) Finally the control of reactor safety rod fall in case of a radioactive accident has been secured by a chain whose detector is a chamber immersed in the swimming-pool, which offers, in the particular case of the hot thickness swimming-pool reactor a double advantage: first it enables us to regulate the upper hot water layer, second to get free of transitory radiations which appear in the reactor hall as the experimental apparatuses are taken out from the core. (authors) [French] Differents problemes de radioprotection se posant aupres des piles piscines de Grenoble, ils ont necessite l'etude de solutions particulieres: a) le controle automatique de la radioactivite du personnel sortant de Melusine ou de Siloe a ete realise a l'aide d'un portique {beta}{gamma} insensible au bruit de fond ambiant du hall des piles; b) le controle automatique de la contamination des souliers des agents travaillant dans ces piles a ete realise par une passerelle pieds {beta}{gamma}; c) la

  2. Protection measurements on the Ulysse pile during its start-up

    International Nuclear Information System (INIS)

    Tattegrain, A.; Satge, H.

    1963-01-01

    The authors report measurements which aimed at verifying protection calculations made for the Ulysse atomic pile. They measured thermal, epithermal flows and the fast flow (when high enough to be measured). They also measured the gamma flow in some parts of the reactor. The authors describe the protections present on the different faces of the pile. These are made of baryte concrete and borated concrete. They indicate the detectors used to measure the different flows, and discuss the results obtained in the different channels

  3. Analysis on pile testing results of post-grouting bored pile

    Science.gov (United States)

    Zheng, A. R.

    2017-04-01

    Based on static load test results, the bearing capacity of bored piles with pile-toe and pile-shaft post-grouting has been analyzed. The analysis reveals that: with post-grouting, the interface between pile and surrounding soil are strengthened and the relative sliding displacement in between is reduced; end resistance of pile is enhanced and can be mobilized at earlier stage with smaller sliding displacement. As a result, the performance of bored pile is improved with increased bearing capacity and reduced settlement.

  4. DRIVEN POLYSTRONG REINFORCED CONCRETE PILES AND NEW DESIGN OF PILE CAPS

    Directory of Open Access Journals (Sweden)

    I. I. Bekbasarov

    2015-01-01

    Full Text Available The paper presents constructional and technological features for manufacturing driven piles with variable strength of pile shaft. Economical efficiency of their production has been shown in the paper. The paper provides a pile cap design that ensures perception of hammer impacts with the help of lateral edges of the pile cap. Driven reinforced concrete piles which are manufactured from three shaft sections having various strength have been proposed in the paper. Material strength (concrete grade and diameter of bars and length of shaft sections are given on a case by case basis in accordance with nature and rate of stresses in piles during their driving process. Manufacturing of polystrong piles provides an opportunity to select them for a particular construction site with due account of their preservation during driving process.A pile cap has been developed that as opposed to existing analogous designs makes it possible to transmit impact efforts from a hammer to the pile through lateral surface of its head part. The pile cap provides the possibility to increase an area for perception of hammer impact efforts by the pile and in doing so it is possible significantly to reduce a damage risk and destruction of pile concrete during its driving. Application of polystrong piles and their driving with the help of new pile cap are considered as a basis for defect-free and resource-saving technology for pile foundations in the construction.

  5. 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

  6. 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

  7. 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)

  8. Remote dismantlement tasks for the CP5 reactor: Implementation, operations, and lessons learned

    International Nuclear Information System (INIS)

    Noakes, M.W.

    1998-01-01

    This paper presents a developer's perspective on lessons learned from one example of the integration of new prototype technology into a traditional operations environment. The dual arm work module was developed by the Robotics Technology Development Program as a research and development activity to examine manipulator controller modes and deployment options. It was later reconfigured for the dismantlement of the Argonne National Laboratory Chicago Pile number-sign 5 reactor vessel as the crane-deployed dual arm work platform. Development staff worked along side operations staff during a significant part of the deployment to provide training, maintenance, and tooling support. Operations staff completed all actual remote dismantlement tasks. At the end of available development support funding, the Dual Arm Work Platform was turned over to the operations staff, who is still using it to complete their dismantlement tasks

  9. Thermal and chemical analysis of carbon dioxide reforming of methane using the out-of-pile test facility

    International Nuclear Information System (INIS)

    Huang Ziyong; Ohashi, Hirofumi; Inagaki, Yoshiyuki

    2000-03-01

    In the Japan Atomic Energy Research Institute, a hydrogen production system is being designed to produce hydrogen by means of steam reforming of natural gas (its main composition is methane(CH 4 )) using nuclear heat (10 MW, 1178 K) supplied by the High Temperature Engineering Test Reactor (HTTR). Prior to coupling of the steam reforming system with the HTTR, an out-of-pile demonstration test was planned to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions of the prototype. The out-of-pile test facility simulates key components downstream to an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 : 30 and has a hydrogen production capacity of 110 Nm 3 /h using an electric heater as a reactor substitute. The test facility is presently under construction. Reforming of natural gas with carbon dioxide CO 2 (CO 2 reforming) using the out-of-pile test facility is also being considered. In recent years, catalytic reforming of natural gas with CO 2 to synthesis gas (CO and H 2 ) has been proposed as one of the most promising technologies for utilization of those two greenhouse gases. Numerical analysis on heat and mass balance has practical significance in CO 2 reforming when the steam reforming process is adopted in the out-of-pile test. Numerical analysis of CO 2 reforming and reforming of natural gas with CO 2 and steam (CO 2 +H 2 O reforming) have been carried out using the mathematical model. Results such as the methane conversion rate, product gas composition, and the components temperature distribution considering the effects of helium gas temperature, reforming pressure, molar ratio of process gases and so on have been obtained in the numerical analysis. Heat and mass balance of the out-of-pile test facility considering chemical reactions are evaluated well. The methane conversation rates are about 0.36 and 0.35 which correspond to the equilibrium at 1085 and 1100 K for

  10. Thermal and chemical analysis of carbon dioxide reforming of methane using the out-of-pile test facility

    Energy Technology Data Exchange (ETDEWEB)

    Huang Ziyong [Institute of Nuclear Energy Technology, Tsinghua University (China); Ohashi, Hirofumi; Inagaki, Yoshiyuki [Department of Advanced Nuclear Heat Technology, Oarai Research Establishment, Japan Atomic Energy Research Institute, Oarai, Ibaraki (Japan)

    2000-03-01

    In the Japan Atomic Energy Research Institute, a hydrogen production system is being designed to produce hydrogen by means of steam reforming of natural gas (its main composition is methane(CH{sub 4})) using nuclear heat (10 MW, 1178 K) supplied by the High Temperature Engineering Test Reactor (HTTR). Prior to coupling of the steam reforming system with the HTTR, an out-of-pile demonstration test was planned to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions of the prototype. The out-of-pile test facility simulates key components downstream to an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 : 30 and has a hydrogen production capacity of 110 Nm{sup 3}/h using an electric heater as a reactor substitute. The test facility is presently under construction. Reforming of natural gas with carbon dioxide CO{sub 2} (CO{sub 2} reforming) using the out-of-pile test facility is also being considered. In recent years, catalytic reforming of natural gas with CO{sub 2} to synthesis gas (CO and H{sub 2}) has been proposed as one of the most promising technologies for utilization of those two greenhouse gases. Numerical analysis on heat and mass balance has practical significance in CO{sub 2} reforming when the steam reforming process is adopted in the out-of-pile test. Numerical analysis of CO{sub 2} reforming and reforming of natural gas with CO{sub 2} and steam (CO{sub 2}+H{sub 2}O reforming) have been carried out using the mathematical model. Results such as the methane conversion rate, product gas composition, and the components temperature distribution considering the effects of helium gas temperature, reforming pressure, molar ratio of process gases and so on have been obtained in the numerical analysis. Heat and mass balance of the out-of-pile test facility considering chemical reactions are evaluated well. The methane conversation rates are about 0.36 and 0.35 which

  11. Tests for validation of fast neutron reactors safety

    International Nuclear Information System (INIS)

    Nagata, T.; Yamashita, H.

    2001-01-01

    Japanese scientific research and design enterprises in cooperation with industrial and power generating corporations implement a project on creating a fast neutron reactor of the ultimate safety. One of the basic expected results from such a development is creation of a reactor core structure that is able to eliminate recriticality occurrence in the course of reactor accident involving fuel melting. One of the possible ways to solve this problem is to include pipes (meant for specifying directed (controlled) molten fuel relocation) into fuel assembly structure. In the course of conduction and subsequent implementation of such a design the basic issue is to experimentally confirm the operating capacity of FA having such a structure and that is called FAIDUS. Within EAGLE Project on experimental basis of IAE NNC RK an activity has been started on preparation and conduction of out-of-pile and in-pile tests. During tests a sodium coolant will be used. Studies are conducted by NNC RK in cooperation with the Japanese corporations JAPC and JNC. Basic objective of out-of-pile tests was to obtain preliminary information on fuel relocation behavior under conditions simulating accident involving melting of core consisting of FAIDUS FA, which will help to clarify simulation criteria and to develop the most optimum structure of the experimental channel for reactor experiments conduction. The basic objective of in-pile tests was the experimental confirmation of operating capacity of FAIDUS FA model under reactor conditions. According to the program two tests are planned to be performed at IGR reactor: tests for validation of fast neutron reactor safety, and out-of-pile tests at EAGLE experimental facility without sodium coolant

  12. Measurement and analysis of angular neutron spectra in a manganese pile

    International Nuclear Information System (INIS)

    Selvi, S.; Hayashi, S.A.; Kimura, I.; Kobayashi, K.; Yamamoto, S.; Mori, T.; Nishihara, H.; Kanazawa, S.; Nakagawa, M.

    1984-01-01

    The energy and angular distribution of neutrons in a Mn pile were measured by the linac time-of-flight method. A cylindrical Pb target for the production of photoneutrons was placed at the center of the pile. The experimental results were compared with the theoretical calculations using the group constants from the nuclear data files, JENDL-2 and ENDF/B-IV. Good agreement can be seen in the general shapes between calculated and measured angular spectra in three decades of energy range form a few keV to a few MeV. As far as can be concluded from the intercomparison, the neutron cross section data for Mn in ENDF/B-IV may be applicable to reactor design: however, several improvements for its resonance parameters can be recommended. A little more improvements are recommended for that in JENDL-2 from this intercomparison. (orig.) [de

  13. Preliminary work for stage 2 decommissioning of B16 pile chimney

    International Nuclear Information System (INIS)

    Wright, E.M.; Mathews, R.F.

    1991-01-01

    Planning of the second stage of decommissioning of the two pile chimneys at Sellafield started while work was underway on the first stage, which involved removal of the sections above the filters. The second stage requires the removal of all radio-active parts and the dismantling of the filter and diffuser sections, and has to be completed by 1997. The planning involved studying the many possible options and their effects on both radiological and industrial safety. This decommissioning project employs a high proportion of civil engineering and construction techniques, which are then developed to eliminate the hazards from radioactive dusts, and to minimise the effect of radiation on operatives working on the project. Much of this equipment is modified forms of standard construction equipment and includes cutting equipment and remotely operated vehicles. The initial phases of the work involve: provision of a waste packaging and access building; provision of temporary ventilation systems to control the dust generated by the work, cutting of 3 m square access doorway through the 1.5 m thick reinforced concrete wall of the chimney; provision of Remotely Operated Vehicle (ROV) to act as a tool carrier for lining stripping work; removal of the thermal lining from the floor and lower walls of the chimney, and installation of precast concrete walls which separate the pile reactor core from the chimney flue. (author)

  14. Improved soil characterization for pipe piles in sand in API RP-2A

    International Nuclear Information System (INIS)

    Hossain, M.K.; Briaud, J.L.

    1993-01-01

    In the offshore, most foundations are steel pipe piles and most of them are designed using the API RP 2A guidelines. For axial capacity of piles in sand the current guidelines in many cases show definite discrepancies when compared against actual load capacities of piles. An updated data base analysis shows that there are three major weaknesses in the current guidelines with respect to soil characterization: (a) the consideration of the lateral earth pressure coefficient, K, as a constant (1.0 or 0.8); (b) the consideration of the unit point bearing resistance, q, as a linear function of depth; and (c) the absence of an unambiguous soil parameter determination process based on reliable in-situ test results. In this paper, specific modifications to the current API RP 2A guidelines are proposed on the basis of a data base analysis to account for the discrepancies arising from (a), (b), and (c) above. These modifications will reduce the discrepancies in the current API RP 2A method and increase the accuracy of the prediction of axial capacity of pipe piles in sand. Furthermore this will make the method fundamentally more consistent with soil behavior in deep foundations

  15. 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)

  16. Starting up a programme of atomic piles using compressed gas; Le demarrage d'un programme de piles atomiques a gaz comprime

    Energy Technology Data Exchange (ETDEWEB)

    Horowitz, J; Yvon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1959-07-01

    1) An examination of the intellectual and material resources which have directed the French programme towards: a) the natural uranium and plutonium system, b) the use of compressed gas as heat transfer fluid (primary fluid). 2) The parts played in exploring the field by the pile EL2 and G1, EL2 a natural uranium, heavy water and compressed gas pile, G1 a natural uranium, graphite and atmospheric air pile. 3) Development of the neutronics of graphite piles: physical study of G1. 4) The examination of certain problem posed by centres equipped with natural uranium, graphite and compressed carbon dioxide piles: structure, special materials, fluid circuits, maximum efficiency. Economic aspects. 5) Aids to progress: a) piles for testing materials and for tests on canned fuel elements, b) laboratory and calculation facilities. 6) Possible new orientations of compressed gas piles: a) raising of the pressure, b) enriched fuel, c) higher temperatures, d) use of heavy water. (author) [French] 1) Examen des ressources - intellectuelles et materielles - qui ont oriente le programme fran is vers: a) la voie de l'uranium naturel et du plutonium; b) l'emploi comme fluide pour le transfert de la chaleur (fluide primaire) d'un gaz comprime. 2) Le role d'exploration des piles EL2 et G1, EL2 pile a uranium naturel, eau lourde et gaz comprime, G1 pile a uranium naturel, graphite et air atmospherique. 3) Developpement de la neutronique des piles a graphite: l'etude physique de G1. 4) Examen de certains problemes poses par les centrales equipees de piles a uranium naturel, graphite et gaz carbonique comprime: structure, materiaux speciaux, circuits de fluides, optimisation. Aspects economiques. 5) Les auxiliaires du progres: a) piles pour essai de materiaux et pour essais de cartouches, b) moyens de laboratoire et moyens de calcul. 6) Orientations nouvelles possibles des piles a gaz comprime: a) elevation de la pression, b) combustible enrichi, c) temperatures elevees, d) emploi de l

  17. A pilot application of the RELAP file to the steady state and transient analysis of a test section inside the BR2 reactor

    International Nuclear Information System (INIS)

    Ferri, M. G.; D'Auria, F.; Forasassi, G.; Giot, M.

    2000-01-01

    BR2 is a material test reactor sited in the Belgian Nuclear Research Centre in Mol. The main research programs carried out in BR2 are related to the safety of nuclear reactor structural materials and fuels, in normal and accidental conditions, plant lifetime evaluation and ageing of components. In this framework, a computer program that allows the performance of detailed, steady state analysis of several kinds of in-pile sections with an axisymmetrical geometry has been developed. Furthermore, comparing its results with those of the well known, extensively used, Relap5/Mod 3.2 code on a test problem has validated this program. This was performed in three steps: 1. modalisation development of a subsystem of a typical in-pile section. 2. steady state analysis and comparison with the above-mentioned program. 3. transient simulation of the same system; the considered transient consists of a loss of coolant flow. (author)

  18. Pile Driving

    Science.gov (United States)

    1987-01-01

    Machine-oriented structural engineering firm TERA, Inc. is engaged in a project to evaluate the reliability of offshore pile driving prediction methods to eventually predict the best pile driving technique for each new offshore oil platform. Phase I Pile driving records of 48 offshore platforms including such information as blow counts, soil composition and pertinent construction details were digitized. In Phase II, pile driving records were statistically compared with current methods of prediction. Result was development of modular software, the CRIPS80 Software Design Analyzer System, that companies can use to evaluate other prediction procedures or other data bases.

  19. 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.

  20. Seismic response of pile foundations and pile forces caused by kinematic and inertial interaction

    International Nuclear Information System (INIS)

    Hartmann, H.G.; Waas, G.

    1985-01-01

    The horizontal motion and pile forces of pile groups subjected to earthquake excitation are analysed. The piles are modelled as linear elastic beam elements embedded in a layered linear visco-elastic soil medium. Pile-soil-pile interaction is included. The earthquake excitation results from vertically propagating shear waves. Kinematic and inertial interaction effects on foundation motion and pile forces are studied for a single pile, a small pile group and a large pile group. Soft and stiff soil conditions are considered, and the effect of a flexible vs. a rigid halfspace below the soil layers is shown. (orig.)

  1. Pile load test on large diameter steel pipe piles in Timan-Pechora, Russia

    Energy Technology Data Exchange (ETDEWEB)

    McKeown, S. [Golder Associates Inc., Houston, TX (United States); Tart, B. [Golder Associates Inc., Anchorage, AK (United States); Swartz, R. [Paragon Engineering Services Inc., Houston, TX (United States)

    1994-12-31

    Pile load testing conducted in May and June of 1993 at the Polar Lights Ardalin project in Arkangelsk province, Russia, was documented. Pile load testing was conducted to determine the ultimate and allowable pile loads for varying pile lengths and ground temperature conditions and to provide creep test data for deformation under constant load. The piles consisted of 20 inch diameter steel pipe piles driven open ended through prebored holes into the permafrost soils. Ultimate pile capacities, adfreeze bond, and creep properties observed were discussed. 10 figs., 4 tabs.

  2. Some particular problems put by operating experimental reactors; Quelques problemes particuliers poses par le fonctionnement des piles laboratoires

    Energy Technology Data Exchange (ETDEWEB)

    Candiotti, C; Mabeix, R; Uguen, R [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    On basis of a six years experience in operating research reactors, the authors explain, first, the difference in their utilization between these piles and another similar ones and, after, in consequence, they set off corresponding servitudes. These servitudes put very particular problems in operating itself, maintenance, modifications or additions on these apparatus. (author) [French] Les redacteurs se basant sur six annees d'experience dans l'exploitation de reacteurs de recherche, exposent tout d'abord les differences d'utilisation entre ces engins et d'autres appareils fonctionnellement similaires et font ressortir, par voie de consequence, les servitudes correspondantes. Ces servitudes posent des problemes tres particuliers dans les domaines de l'exploitation proprement dite, de l'entretien, des modifications ou adjonctions apportees a l'ensemble. (auteur)

  3. 3D CAD model of the subcritical nuclear reactor of IPN; Modelo CAD 3D del reactor nuclear subcritico del IPN

    Energy Technology Data Exchange (ETDEWEB)

    Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ibarra R, G.; Del Valle G, E.; Sanchez R, A., E-mail: narehc@hotmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN, Edif. 9, Unidad Profesional Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico)

    2016-09-15

    The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

  4. Piles of objects

    KAUST Repository

    Hsu, Shu-Wei

    2010-01-01

    We present a method for directly modeling piles of objects in multi-body simulations. Piles of objects represent some of the more interesting, but also most time-consuming portion of simulation. We propose a method for reducing computation in many of these situations by explicitly modeling the piles that the objects may form into. By modeling pile behavior rather than the behavior of all individual objects, we can achieve realistic results in less time, and without directly modeling the frictional component that leads to desired pile shapes. Our method is simple to implement and can be easily integrated with existing rigid body simulations. We observe notable speedups in several rigid body examples, and generate a wider variety of piled structures than possible with strict impulse-based simulation. © 2010 ACM.

  5. 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.

  6. 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

  7. Validity and Utilization of the Out-Pile Testing Facilities at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Kee-Nam; Cho, Man-Soon; Yang, Sung-Woo; Shin, Yoon-Taek; Park, Seng-Jae; Jun, Byung-Hyuk; Kim, Myong-Seop [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Various neutron irradiation facilities such as rabbit irradiation facilities, loop facilities and the capsule irradiation facilities for irradiation tests of nuclear materials, fuels and radioisotope products have been developed at HANARO. Among these irradiation facilities, the capsule is the most useful device for coping with the various test requirements at HANARO. To support the national research and development programs on nuclear reactors and the nuclear fuel cycle technology in Korea, new irradiation capsules have been developed and actively utilized for the irradiation tests requested by numerous users. The environmental conditions for these reactors are generally beyond present day reactor technology, especially regarding the higher neutron fluence and higher operating temperature. To effectively support the national R and Ds relevant to the future nuclear systems, the development of advanced irradiation technologies concerning higher neutron fluence and irradiation temperature are being preferentially developed at HANARO. The utilization of the out-pile testing facilities to satisfy the criteria of safety evaluation for a new device installed in the core of HANARO was summarized. In addition, the validity of the out-pile testing facilities was evaluated and proved to be effective for verifying the integrity of irradiation capsule.

  8. Validity and Utilization of the Out-Pile Testing Facilities at HANARO

    International Nuclear Information System (INIS)

    Choo, Kee-Nam; Cho, Man-Soon; Yang, Sung-Woo; Shin, Yoon-Taek; Park, Seng-Jae; Jun, Byung-Hyuk; Kim, Myong-Seop

    2016-01-01

    Various neutron irradiation facilities such as rabbit irradiation facilities, loop facilities and the capsule irradiation facilities for irradiation tests of nuclear materials, fuels and radioisotope products have been developed at HANARO. Among these irradiation facilities, the capsule is the most useful device for coping with the various test requirements at HANARO. To support the national research and development programs on nuclear reactors and the nuclear fuel cycle technology in Korea, new irradiation capsules have been developed and actively utilized for the irradiation tests requested by numerous users. The environmental conditions for these reactors are generally beyond present day reactor technology, especially regarding the higher neutron fluence and higher operating temperature. To effectively support the national R and Ds relevant to the future nuclear systems, the development of advanced irradiation technologies concerning higher neutron fluence and irradiation temperature are being preferentially developed at HANARO. The utilization of the out-pile testing facilities to satisfy the criteria of safety evaluation for a new device installed in the core of HANARO was summarized. In addition, the validity of the out-pile testing facilities was evaluated and proved to be effective for verifying the integrity of irradiation capsule

  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. 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)

  11. A history of ZED-2

    International Nuclear Information System (INIS)

    Jones, R.

    2010-01-01

    The ZED-2 Reactor at Chalk River Laboratories was 50 years old this fall. First criticality occurred in September 1960. ZED-2 is perhaps not very well known in the Canadian Nuclear Industry, certainly not as well known as the various CANDU power reactors or the research reactors NRU and NRX. Part of the reason for this I suspect is that when casually judging the importance of reactors the first parameters that spring to mind are power generated (for power reactors) or neutron flux (for research reactors), bigger being 'better in both cases. By these standards ZED-2 does indeed appear puny: the maximum allowed power is 200W and the corresponding flux about 10 8 to 10 9 neutrons cm -2 s -1 , both numbers being about a factor of 500,000 smaller than the corresponding values for NRU. So, what is it all about? How is it that such an apparently insignificant reactor has operated for 50 years?, longer than any other Canadian reactor except NRU and the McMaster Reactor. What is it used for? What contributions has it made to the Canadian industry? Maybe one might also ask for how long is it going to continue? Well, that's what this talk is about, although I think I will leave the final question to wiser heads than mine. ZED-2 is a descendant of famous progenitors: starting with Enrico Farm's first critical pile of graphite and uranium (created at the University of Chicago in 1942) through Canada's ZEEP (first reactor to go critical outside the USA) that went critical in 1945. These early critical facilities were first about proof of principle that a self sustaining nuclear chain reaction could be established and controlled in a reasonable sized facility and second, in the longer term, developing understanding of the underlying reactor physics and the development of theories and methods to accurately predict the important properties of critical assemblies generally. (author)

  12. 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

  13. Some particular aspects of control in nuclear power reactors; Quelques aspects particuliers du controle dans les piles atomiques de puissance

    Energy Technology Data Exchange (ETDEWEB)

    Furet, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Pupponi, J [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    There are still many problems in the field of measurement and control of neutron flux. The present studies in connexion with high flux reactors contribute to the solution of these problems which concern specialists in reactor control. The present state of this investigation and the results of different studies carried out in France by the C A and the EDF are pointed out: A - In the nuclear instrumentation field, work is at present devoted to the technologies used to develop detectors and cables, which have to work at high temperature and in a high {gamma} background; fast electronic techniques are applied to fission counters to measure low neutron fluxes in a high {gamma} background (10 Rh). B - In the control and safety field, there is a real need for studies on the behaviour of reactors in the subcritical state. This increases the margin of security during restarts when poison effects must be overcome The perturbations due to control rod movements necessitate a new organisation of power level safety and control assemblies, in connexion with thermal or activation measurements. Two methods of fast start-up are described. They are related to the fission rate measurement as a function of time. This is done either continuously by a constant and high reactivity change, or step by step. The application of automatic techniques to detector motion seems to give the answer to control and safety in normal start-up. C - The scope of these studies covers the methods used for the control of E.D.F. 3, which are described. (authors) [French] La mesure et le controle du flux neutronique dans les piles de puissance posent encore de nombreux problemes. Les etudes actuellement entreprises dans le domaine des piles a haut flux, doivent apporter une contribution importante a la solution de ces problemes qui interessent les specialistes du controle des piles de puissance. On analyse l'etat actuel de ces etudes et on donne les resultats des differents travaux effectues en France, dans

  14. 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)

  15. Underwater noise reduction of marine pile driving using a double pile.

    Science.gov (United States)

    2015-12-01

    Impact pile driving of steel piles in marine environments produces extremely high sound levels in the water. : It has been shown that current pile driving noise attenuation techniques, such as bubble curtains and : cofferdams, provide limited noise r...

  16. 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.

  17. Enrico Fermi: The First Chain Reactor (with Film) and Pion-Proton Scattering

    International Nuclear Information System (INIS)

    Martin, Ron

    2003-01-01

    A twenty minute film will be shown depicting the first nuclear chain reactor at the University of Chicago on December 2, 1942. The film was made of a re-enactment in 1952 and is narrated by Arthur Compton and Enrico Fermi. After the film, Ronald Martin will talk about his experiences on pion-proton scattering with Enrico Fermi at the Chicago synchrocyclotron in the fifties.

  18. Piloting procedure for a pile running below critical level

    International Nuclear Information System (INIS)

    Lacour, J.; Raievski, V.

    1957-01-01

    The knowledge of the subcritical state of a reactor in the course of starting up makes it possible to avoid passing too quickly through the critical state. The problem arises every time the pile is put into action again following, for example, an appreciable modification in the fuel charge, or an accidental fall of the security rods during a run at high flux or at high temperature. The method described provides a mean of knowing at each moment the anti-reactivity value of the pile by means of a direct-reading instrument mounted on the control board. This result is obtained by superimposing a fixed frequency oscillation on the normal movement of a control rod, and reading on a phase-meter the dephasing of the neutron density. Theory shows, and experiments confirm, that for a given frequency the dephasing depends only on the lifetime of the fast neutrons, the characteristics of the slow neutrons and the anti-reactivity of the pile. The minimum time necessary for an anti-reactivity determination is equal to a modulation period (from 1 to 4 seconds). (authors) [fr

  19. Can rupture detector for water cooled piles; Detecteur de rupture de gaine pour piles refroidies a l'eau

    Energy Technology Data Exchange (ETDEWEB)

    Choudens, H de; Guitton, P

    1962-07-01

    The object of this study was to develop a simple, easy to-build, apparatus for showing the appearance of a defect on a fuel element can of a swimming pool reactor. The apparatus used consists of a coil of activated carbon around a NaI(Tl) crystal. Through this coil pass the gases obtained by degassing a sample of water from the reactor; the fission gases which appear when a can leaks are trapped in the carbon; the NaI(Tl) crystal is coupled with a photomultiplier followed by a single-channel selector fixed on a photo-electric peak characteristic of the {gamma} spectrum of fission gases. Preliminary experiments were carried out in laboratory; a more complete device was then built and is now working in the reactor Melusine. (author) [French] Le but de cette etude a ete la realisation d'un appareil simple et facile a realiser destine a indiquer l'apparition d'un defaut sur une gaine des elements combustibles d'une pile piscine. A cet effet, l'appareillage utilise est compose d'un serpentin de charbon active entourant un cristal de NaI (Tl). Ce serpentin est parcouru par les gaz provenant du degazage d'un prelevement d'eau de la piscine du reacteur; les gaz de fission apparaissant lors d'une rupture de gaine sont retenus dans le charbon; le cristal INa (Tl) est couple avec un PM suivi d'un selecteur monocanal cale sur un pic photoelectrique caracteristique du spectre {gamma} des gaz de fission. Des manipulations preliminaires ont ete faites en laboratoire, un dispositif plus complet a ete ensuite monte et fonctionne a la Pile Melusine. (auteurs)

  20. 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

  1. Static pile load tests on driven piles into Intermediate-Geo Materials.

    Science.gov (United States)

    2016-09-01

    The Wisconsin Department of Transportation (WisDOT) has concerns with both predicting pile lengths and pile capacities for H-piles driven into Intermediate-Geo Materials (IGM). The goal of the research was to perform 7 static axial load tests at 7 lo...

  2. 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.

  3. 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.

  4. The Tensile Capacity Of Bored Piles In Frictional Soils

    DEFF Research Database (Denmark)

    Krabbenhøft, Sven; Andersen, Allan; Damkilde, Lars

    2008-01-01

    Three series of 10 piles each were installed in two different locations. The length of the piles varied from 2 to 6 m and the diameters were 14 and 25 cm. The piles were constructed above the groundwater table using continuous flight augers and the concrete was placed by gravity free fall. The pi....... The piles were tested to failure in axial uplift and the load-displacement relations were recorded.......Three series of 10 piles each were installed in two different locations. The length of the piles varied from 2 to 6 m and the diameters were 14 and 25 cm. The piles were constructed above the groundwater table using continuous flight augers and the concrete was placed by gravity free fall...

  5. 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.

  6. Some particular problems put by operating experimental reactors; Quelques problemes particuliers poses par le fonctionnement des piles laboratoires

    Energy Technology Data Exchange (ETDEWEB)

    Candiotti, C.; Mabeix, R.; Uguen, R. [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    On basis of a six years experience in operating research reactors, the authors explain, first, the difference in their utilization between these piles and another similar ones and, after, in consequence, they set off corresponding servitudes. These servitudes put very particular problems in operating itself, maintenance, modifications or additions on these apparatus. (author) [French] Les redacteurs se basant sur six annees d'experience dans l'exploitation de reacteurs de recherche, exposent tout d'abord les differences d'utilisation entre ces engins et d'autres appareils fonctionnellement similaires et font ressortir, par voie de consequence, les servitudes correspondantes. Ces servitudes posent des problemes tres particuliers dans les domaines de l'exploitation proprement dite, de l'entretien, des modifications ou adjonctions apportees a l'ensemble. (auteur)

  7. Some particular aspects of control in nuclear power reactors

    International Nuclear Information System (INIS)

    Vathaire, F. de; Vernier, Ph.; Pascouet, A.

    1964-01-01

    This paper reviews the experience acquired in France on the question, of reactor safety. Since a special paper is being presented on reactors of the graphite gas type, the safety of the other types studied in France is discussed here: - heavy water-gas reactors, - fast neutron reactors, - water research reactors of the swimming-pool and tank types. The safety rules peculiar to the different types are explained, with emphasis on their influence on the reactor designs and on the power limits they impose. The corresponding safety studies are presented, particular stress being placed on the original work developed in these fields. Special mention is made of the experimental systems constructed for these studies: the reactor CABRI, pile loop for depressurization tests, loops outside the pile, mock-ups etc. (authors) [fr

  8. General outline of the operation and utilization of the BR2 reactor

    International Nuclear Information System (INIS)

    Baugnet, J.M.; Leonard, F.; Gandolfo, J.M.; Lenders, H.

    1978-01-01

    The BR2 reactor is a high-flux material testing reactor of the thermal heterogeneous type. The fuel is 93% 235 U enriched uranium in the form of plates clad in aluminium. The moderator consists of beryllium and light water, the water being pressurized (12.5kg/cm 2 )and acting also as coolant. The pressure vessel is of aluminium, and is placed in a pool of demineralized water. One should stress the following main features of the design: the experimental channels are skew, the tube bundle presenting the form of a hyperboloid of revolution (see figure 1)-this gives easy access at the top and bottom reactor covers allowing complex instrumented devices, while maintaining a very high neutron flux at the core; great flexibilty of utilization, due to the fact that it is possible to adapt the core configuration to the experimental loading as the fissile charge can be centred on different experimental channels; although BR2 is a thermal reactor, it is possible to achieve neutron spectra very similar to those obtained in a fast reactor, either by the use of absorbing screens or by the use of fissile material within the experimental device; five 200mm diameter channels are available for loading large experimental irradiation devices, as in-pile sodium, gas or water loops. (author)

  9. Introduction to Single Piles under Lateral Loading

    DEFF Research Database (Denmark)

    Augustesen, Anders; Ibsen, Lars Bo

    .2). The description is based on results of laboratory tests, full-scale field tests as well as numerical investigations presented in literature. Second, general methods that attempt to model lateral pile response are discussed in section 1.4. Third, focus is paid to a widely used method for prediction of the response......The purpose of this chapter is to give a short introduction to single piles subjected to lateral loading. First, the observed behaviour of laterally loaded piles is described, i.e. the effects of loading conditions, installation procedure, pile type etc. on pile behaviour are presented (section 1...... of a lateral loaded pile, namely the Winkler approach in which the pile is modelled as an elastic beam on an elastic foundation (section 1.5). The soil response and thereby the elastic foundation is represented by springs with nonlinear behaviour (p-y curves). In section 1.6 different types and formulations...

  10. 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

  11. CP-5 reactor remote dismantlement activities: Lessons learned in the integration of new technology in an operations environment

    International Nuclear Information System (INIS)

    Noakes, M.W.

    1998-01-01

    This paper presents the developer's perspective on lessons learned from one example of the integration of new prototype technology into a traditional operations environment. The dual arm work module was developed by the Robotics Technology Development Program as a research and development activity to examine manipulator controller modes and deployment options. It was later reconfigured for the dismantlement of the Argonne National Laboratory Chicago Pile No. 5 reactor vessel as the crane-deployed dual arm work platform. Development staff worked along side operations staff during a significant part of the deployment to provide training, maintenance, and tooling support. Operations staff completed all actual remote dismantlement tasks. At the end of available development support funding, the Dual Arm Work Platform was turned over to the operations staff, who are still using it to complete their dismantlement tasks

  12. 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)

  13. 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

  14. 33 CFR 110.83 - Chicago Harbor, Ill.

    Science.gov (United States)

    2010-07-01

    .... Beginning at a point 2,120 feet South of the intersection of the North line of the Chicago Yacht Club... the first described line, passing 100 feet East of the Chicago Yacht Club bulkhead, 440 feet; thence.... Beginning at a point 145 feet North of the North line of the Chicago Yacht Club bulkhead, as constructed in...

  15. An MCNP parametric study of George C. Laurence's subcritical pile experiment

    International Nuclear Information System (INIS)

    Dranga, R.; Blomeley, L.; Carrington, R.

    2014-01-01

    In the early 1940s at the National Research Council (NRC) Laboratories in Ottawa, Canada, Dr. George Laurence conducted several experiments to determine if a sustained nuclear fission chain reaction in a carbon-uranium arrangement (or 'pile') was possible. Although Dr. Laurence did not achieve criticality, these pioneering experiments marked a significant historical event in nuclear science, and they provided a valuable reference for subsequent experiments that led to the design of Canada's first heavy-water reactors at the Chalk River Nuclear Laboratories. This paper summarizes the results of a recent collaborative project between Atomic Energy of Canada Limited and the Deep River Science Academy undertaken to numerically explore the experiments carried out at the NRC Laboratories by Dr. Laurence, while teaching high school students about nuclear science and technology. In this study, a modern Monte Carlo reactor physics code, MCNP6, was utilized to identify and study the key parameters impacting the subcritical pile's neutron multiplication factor (e.g., moderation, geometry, material impurities) and quantify their effect on the extent of subcriticality. The findings presented constitute the first endeavour to model, using a current computational reactor physics tool, the seminal experiment that provided the foundation of Canada's nuclear science and technology program. (author)

  16. Trends in PM2.5 emissions, concentrations and apportionments in Detroit and Chicago

    Science.gov (United States)

    Milando, Chad; Huang, Lei; Batterman, Stuart

    2016-03-01

    PM2.5 concentrations throughout much of the U.S. have decreased over the last 15 years, but emissions and concentration trends can vary by location and source type. Such trends should be understood to inform air quality management and policies. This work examines trends in emissions, concentrations and source apportionments in two large Midwest U.S. cities, Detroit, Michigan, and Chicago, Illinois. Annual and seasonal trends were investigated using National Emission Inventory (NEI) data for 2002 to 2011, speciated ambient PM2.5 data from 2001 to 2014, apportionments from positive matrix factorization (PMF) receptor modeling, and quantile regression. Over the study period, county-wide data suggest emissions from point sources decreased (Detroit) or held constant (Chicago), while emissions from on-road mobile sources were constant (Detroit) or increased (Chicago), however changes in methodology limit the interpretation of inventory trends. Ambient concentration data also suggest source and apportionment trends, e.g., annual median concentrations of PM2.5 in the two cities declined by 3.2-3.6%/yr (faster than national trends), and sulfate concentrations (due to coal-fired facilities and other point source emissions) declined even faster; in contrast, organic and elemental carbon (tracers of gasoline and diesel vehicle exhaust) declined more slowly or held constant. The PMF models identified nine sources in Detroit and eight in Chicago, the most important being secondary sulfate, secondary nitrate and vehicle emissions. A minor crustal dust source, metals sources, and a biomass source also were present in both cities. These apportionments showed that the median relative contributions from secondary sulfate sources decreased by 4.2-5.5% per year in Detroit and Chicago, while contributions from metals sources, biomass sources, and vehicles increased from 1.3 to 9.2% per year. This first application of quantile regression to trend analyses of speciated PM2.5 data reveals

  17. Complementary and alternative medicine use for arthritis pain in 2 Chicago community areas.

    Science.gov (United States)

    Feinglass, Joe; Lee, Chin; Rogers, Michelle; Temple, Leslie Mendoza; Nelson, Cynthia; Chang, Rowland W

    2007-01-01

    To compare the use of complementary and alternative medicine (CAM) for arthritis between 2 ethnically distinct metropolitan Chicago community areas. A telephone interview survey of adults age 45 years or above living in North (88.9% white) or South (79.7% African American) areas. Of 763 respondents, 405 reported arthritis or chronic joint symptoms and were asked about use and satisfaction with 7 CAM therapies. Differences between areas were compared with population-weighted tests; multiple logistic regression was used to analyze the likelihood of CAM use controlled for demographics, behavioral risk factors, and arthritis severity. South Chicago respondents had a higher prevalence and more severe arthritis symptoms such as mean joint pain and more functional limitations. Use of CAM therapy by South Chicago respondents, most commonly massage and relaxation techniques, was 10% greater than North Chicago respondents (61.5% to 51%) but this was not significantly different. Among CAM users, South Chicago respondents reported higher satisfaction with 6 of the 7 CAM therapies and greater future interest in CAM therapies. Poor overall health status (P=0.03), arthritis pain (P=0.005), and concomitant use of prescription medications (P=0.03) were the only significant predictors of CAM use. Although there were only small differences in overall CAM use by area, older residents of largely African American communities were enthusiastic users of relaxation, massage, and nutritional and dietary techniques. CAM modalities could be important adjuncts to traditional medical treatment of arthritis pain for minority communities.

  18. Pile Driving Analysis for Pile Design and Quality Assurance

    Science.gov (United States)

    2017-08-01

    Driven piles are commonly used in foundation engineering. The most accurate measurement of pile capacity is achieved from measurements made during static load tests. Static load tests, however, may be too expensive for certain projects. In these case...

  19. Out-of-pile demonstration test of hydrogen production system coupling with HTTR

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Nishihara, Tetsuo; Takeda, Tetsuaki; Hada, Kazuhiko; Hayashi, Koji

    1999-01-01

    In Japan Atomic Energy Research Institute, a hydrogen production system is being designed to produce hydrogen by means of a steam reforming process of natural gas using nuclear heat (10 MW, 905degC) supplied by the High Temperature Engineering Test Reactor (HTTR). The safety principle and criteria are also being investigated in the HTTR hydrogen production system. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile demonstration test was planned to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions of the HTTR hydrogen production system. The out-of-pile test facility simulates key components downstream an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 to 30 has a hydrogen production capacity of 110 Nm 3 /h using an electric heater as a reactor substitute. The test facility is under manufacturing aiming at completion in 2000 and followed by the test till 2004. In parallel to this, a hydrogen permeation test and a corrosion test of a catalyst tube of a steam reformer are being carried out to obtain data necessary for the design of the system. This report describes outline of the out-of-pile hydrogen production facility and demonstration test program for the HTTR hydrogen production system at present status. (author)

  20. Out-of-pile demonstration test of hydrogen production system coupling with HTTR

    Energy Technology Data Exchange (ETDEWEB)

    Inagaki, Yoshiyuki; Nishihara, Tetsuo; Takeda, Tetsuaki; Hada, Kazuhiko; Hayashi, Koji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1999-07-01

    In Japan Atomic Energy Research Institute, a hydrogen production system is being designed to produce hydrogen by means of a steam reforming process of natural gas using nuclear heat (10 MW, 905degC) supplied by the High Temperature Engineering Test Reactor (HTTR). The safety principle and criteria are also being investigated in the HTTR hydrogen production system. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile demonstration test was planned to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions of the HTTR hydrogen production system. The out-of-pile test facility simulates key components downstream an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 to 30 has a hydrogen production capacity of 110 Nm{sup 3}/h using an electric heater as a reactor substitute. The test facility is under manufacturing aiming at completion in 2000 and followed by the test till 2004. In parallel to this, a hydrogen permeation test and a corrosion test of a catalyst tube of a steam reformer are being carried out to obtain data necessary for the design of the system. This report describes outline of the out-of-pile hydrogen production facility and demonstration test program for the HTTR hydrogen production system at present status. (author)

  1. Apparatus for examination of irradiated fuel elements of industrial reactors at Marcoule; Appareillage d'examen des elements combustibles des piles industrielles de Marcoule

    Energy Technology Data Exchange (ETDEWEB)

    Pesenti, P; Wallet, Ph [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    The authors describe a viewing and measurement cell for the slugs of Marcoule industrial reactors. This cell allows visual inspection, and photography of slugs. Length measurements are also made possible by horizontal motion of the slug both in translation and rotation. (author) [French] Les auteurs decrivent une cellule d'observation et de mesure des elements combustibles des piles industrielles de Marcoule. La cellule permet l'examen a vue, la photographie, la radioscopie et la radiographie des elements combustibles. Elle permet en outre la mesure de longueurs sur ces elements, ces derniers pouvant etre deplaces horizontalement en translation, et en rotation. (auteur)

  2. 78 FR 46810 - Safety Zone; Motion Picture Filming; Chicago River; Chicago, IL

    Science.gov (United States)

    2013-08-02

    ... portion of the Chicago River due to the filming of a motion picture. These temporary safety zones are...-AA00 Safety Zone; Motion Picture Filming; Chicago River; Chicago, IL AGENCY: Coast Guard, DHS. ACTION: Temporary final rule. SUMMARY: The Coast Guard is establishing three temporary safety [[Page 46811

  3. Three dimensional modeling of laterally loaded pile groups resting in sand

    Directory of Open Access Journals (Sweden)

    Amr Farouk Elhakim

    2016-04-01

    Full Text Available Many structures often carry lateral loads due to earth pressure, wind, earthquakes, wave action and ship impact. The accurate predictions of the load–displacement response of the pile group as well as the straining actions are needed for a safe and economic design. Most research focused on the behavior of laterally loaded single piles though piles are most frequently used in groups. Soil is modeled as an elastic-perfectly plastic model using the Mohr–Coulomb constitutive model. The three-dimensional Plaxis model is validated using load–displacement results from centrifuge tests of laterally loaded piles embedded in sand. This study utilizes three dimensional finite element modeling to better understand the main parameters that affect the response of laterally loaded pile groups (2 × 2 and 3 × 3 pile configurations including sand relative density, pile spacing (s = 2.5 D, 5 D and 8 D and pile location within the group. The fixity of the pile head affects its load–displacement under lateral loading. Typically, the pile head may be unrestrained (free head as the pile head is allowed to rotate, or restrained (fixed head condition where no pile head rotation is permitted. The analyses were performed for both free and fixed head conditions.

  4. 3D CAD model of the subcritical nuclear reactor of IPN

    International Nuclear Information System (INIS)

    Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A.; Ibarra R, G.; Del Valle G, E.; Sanchez R, A.

    2016-09-01

    The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

  5. 76 FR 11334 - Safety Zone; Soil Sampling; Chicago River, Chicago, IL

    Science.gov (United States)

    2011-03-02

    ...The Coast Guard is establishing a temporary safety zone on the North Branch of the Chicago River near Chicago, Illinois. This zone is intended to restrict vessels from a portion of the North Branch of the Chicago River due to soil sampling in this area. This temporary safety zone is necessary to protect the surrounding public and vessels from the hazards associated with the soil sampling efforts.

  6. Soil heating during burning of forest slash piles and wood piles

    Science.gov (United States)

    Matt D. Busse; Carol J. Shestak; Ken R. Hubbert

    2013-01-01

    Pile burning of conifer slash is a common fuel reduction practice in forests of the western United States that has a direct, yet poorly quantified effect on soil heating. To address this knowledge gap, we measured the heat pulse beneath hand-built piles ranging widely in fuel composition and pile size in sandy-textured soils of the Lake Tahoe Basin. The soil heat pulse...

  7. Starting up a programme of atomic piles using compressed gas

    International Nuclear Information System (INIS)

    Horowitz, J.; Yvon, J.

    1959-01-01

    1) An examination of the intellectual and material resources which have directed the French programme towards: a) the natural uranium and plutonium system, b) the use of compressed gas as heat transfer fluid (primary fluid). 2) The parts played in exploring the field by the pile EL2 and G1, EL2 a natural uranium, heavy water and compressed gas pile, G1 a natural uranium, graphite and atmospheric air pile. 3) Development of the neutronics of graphite piles: physical study of G1. 4) The examination of certain problem posed by centres equipped with natural uranium, graphite and compressed carbon dioxide piles: structure, special materials, fluid circuits, maximum efficiency. Economic aspects. 5) Aids to progress: a) piles for testing materials and for tests on canned fuel elements, b) laboratory and calculation facilities. 6) Possible new orientations of compressed gas piles: a) raising of the pressure, b) enriched fuel, c) higher temperatures, d) use of heavy water. (author) [fr

  8. Use of geothermal piles combined with pile foundations

    Directory of Open Access Journals (Sweden)

    Ivan Kuzytskyi

    2016-07-01

    Full Text Available The possibility of use of geothermal piles in conditions of cold climate is considered. Full-scale experiment is conducted for using this technology in Kiev. Obtained results testify about a possibility for using the system in conditions of Ukraine, but this technology requires more detailed study and simulation of multiannual cycle of use of geothermal piles 

  9. Pile noise experiment in MINERVE reactor to estimate kinetic parameters using various data processing methods

    Energy Technology Data Exchange (ETDEWEB)

    Geslot, Benoit; Gruel, Adrien; Pepino, Alexandra; Di Salvo, Jacques; Izarra, Gregoire de; Jammes, Christian; Destouches, Christophe; Blaise, Patrick [CEA, DEN, DER/SPEx, Cadarache, F-13108 St Paul Lez Durance (France)

    2015-07-01

    MINERVE is a two-zone pool type zero power reactor operated by CEA (Cadarache, France). Kinetic parameters of the core (prompt neutron decay constant, delayed neutron fraction, generation time) have been recently measured using various pile noise experimental techniques, namely Feynman-α, Rossi-α and Cohn-α. Results are discussed and compared to each other's. The measurement campaign has been conducted in the framework of a tri-partite collaboration between CEA, SCK.CEN and PSI. Results presented in this paper were obtained thanks to a time-stamping acquisition system developed by CEA. PSI performed simultaneous measurements which are presented in a companion paper. Signals come from two high efficiency fission chambers located in the graphite reflector next to the core driver zone. Experiments were conducted at critical state with a reactor power of 0.2 W. The core integral fission rate is obtained from a calibrated miniature fission chamber located at the center of the core. Other results obtained in two sub-critical configurations will be presented elsewhere. Best estimate delayed neutron fraction comes from the Cohn-α method: 747 ± 15 pcm (1σ). In this case, the prompt decay constant is 79 ± 0.5 s{sup -1} and the generation time is 94.5 ± 0.7 μs. Other methods give consistent results within the confidence intervals. Experimental results are compared to calculated values obtained from a full 3D core modeling with the CEA-developed Monte Carlo code TRIPOLI4.9 associated with its continuous energy JEFF3.1.1-based library. A very good agreement is observed for the calculated delayed neutron fraction (748.7 ± 0.4 pcm at 1σ), that is a difference of -0.3% with the experiment. On the contrary, a 10% discrepancy is observed for the calculated generation time (104.4 ± 0.1 μs at 1σ). (authors)

  10. 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.

  11. 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.

  12. Description of the Triton reactor; Pile Triton, rapport descriptif

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1967-09-01

    The Triton reactor is an enriched uranium pool type reactor. It began operation in 1959, after a divergence made on the June 30 the same year. Devoted to studies of radiation protection, its core can be displaced in the longitudinal direction. The pool can be separated in two unequal compartments by a wall. The Triton core is placed in a small compartment, the Nereide core in the big compartment. A third compartment without water is called Naiade II, is separated by a concrete wall in which is made a window closed by an aluminium plate (2.50 m x 2.70 m). The Naiade II hole is useful for protection experiments using the Nereide core. After a complete refitting, the power of the triton reactor that reached progressively from 1.2 MW to 2 MW, then 3 MW has reached in August 1965 6.5 MW. The reactor has been specialized in irradiations in fix position, the core become fix, the nereide core has been hung mobile. Since it has been used for structure materials irradiation, for radioelements fabrication and fundamental research. The following descriptions are valid for the period after August 1965. [French] Le reacteur Triton est un reacteur piscine, a uranium enrichi. Il est entre en fonctionnement en 1959, apres une divergence effectuee le 30 juin de cette meme annee. Destine a des etudes de protection contre les rayonnements, son coeur pouvait se deplacer dans le sens longitudinal. La piscine peut etre separee en deux compartiments inegaux par un batardeau. Le coeur triton est place dans le petit compartiment, le coeur Nereide dans le grand compartiment. Un troisieme compartiment sans eau, appele Naiade II, est separe par une paroi en beton dans laquelle est amenagee une fenetre obturee par une plaque d'aluminium (2,50 m x 2,70 m). La fosse Naiade II sert a des experiences de protection utilisant le coeur nereide. Apres une refonte complete, la puissance du reacteur triton qui etait passee progressivement de 1,2 MW a 2 MW, puis 3 MW, a atteint en aout 1965 6, 5 MW. Le

  13. FORECASTING PILE SETTLEMENT ON CLAYSTONE USING NUMERICAL AND ANALYTICAL METHODS

    Directory of Open Access Journals (Sweden)

    Ponomarev Andrey Budimirovich

    2016-06-01

    Full Text Available In the article the problem of designing pile foundations on claystones is reviewed. The purpose of this paper is comparative analysis of the analytical and numerical methods for forecasting the settlement of piles on claystones. The following tasks were solved during the study: 1 The existing researches of pile settlement are analyzed; 2 The characteristics of experimental studies and the parameters for numerical modeling are presented, methods of field research of single piles’ operation are described; 3 Calculation of single pile settlement is performed using numerical methods in the software package Plaxis 2D and analytical method according to the requirements SP 24.13330.2011; 4 Experimental data is compared with the results of analytical and numerical calculations; 5 Basing on these results recommendations for forecasting pile settlement on claystone are presented. Much attention is paid to the calculation of pile settlement considering the impacted areas in ground space beside pile and the comparison with the results of field experiments. Basing on the obtained results, for the prediction of settlement of single pile on claystone the authors recommend using the analytical method considered in SP 24.13330.2011 with account for the impacted areas in ground space beside driven pile. In the case of forecasting the settlement of single pile on claystone by numerical methods in Plaxis 2D the authors recommend using the Hardening Soil model considering the impacted areas in ground space beside the driven pile. The analyses of the results and calculations are presented for examination and verification; therefore it is necessary to continue the research work of deep foundation at another experimental sites to improve the reliability of the calculation of pile foundation settlement. The work is of great interest for geotechnical engineers engaged in research, design and construction of pile foundations.

  14. Chemical states of piled-up phosphorus and arsenic atoms at the SiO2/Si interface

    International Nuclear Information System (INIS)

    Yoshimura, Yusuke; Ono, Kanta; Fujioka, Hiroshi; Hayakawa, Shinjiro; Sato, Yoshiyuki; Uematsu, Masashi; Baba, Yuji; Hirose, Kazuyuki; Oshima, Masaharu

    1999-01-01

    We have investigated the positions of the piled-up phosphorus atoms at the SiO 2 /Si interface using the extended X-ray absorption fine structure (EXAFS) and X-ray photoelectron spectroscopy (XPS). The EXAFS and XPS data can be well explained on the assumption that the piled-up arsenic atoms exist at the tetrahedral sites. On the contrary, phosphorus atoms exist not at the tetrahedral sites but at the denser sites. The depth profile measurements of XPS have revealed that the piled-up arsenic and phosphorus atoms exist within 20A from the interface. (author)

  15. Dynamic stiffness of pile groups in a multilayered soil. Part 1

    International Nuclear Information System (INIS)

    Ohta, Y.; Hijikata, K.; Kobayashi, Y.

    1989-01-01

    For evaluating the dynamic stiffness of the pile group foundations, forced vibration tests are executed on pile group foundation models. Two types of test models are used, one is a single pile model and the other a four-pile model. Dividing the tests into 4 steps, the forced vibration tests are performed. Step 1 is for the single pile model, and steps 2 to 4 are for the four-pile model. In step 2 and step 3, the gap effects between the foundation bottom and the ground surface are examined. In step 4, the backfill effects are obtained. Based on the test results, the pile group effects, the gap effects and the backfill effects on the dynamic characteristics of the pile group foundations are described in this paper

  16. New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

    International Nuclear Information System (INIS)

    Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

    1996-01-01

    The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

  17. Adiabatic calorimeter with static vacuum for the measurement of the heating of in- pile materials; Calorimetre adiabatique a vide statique pour la mesure d'echauffements de materiaux en pile

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    After having reviewed the various interaction processes occurring between radiations present in nuclear reactors and matter, the author describes the different calorimetric methods which may be used for measuring the energy absorbed in the materials. He then gives a detailed description of the adiabatic calorimeter, the associated measurement device and the calibration methods which have been chosen. He finally gives values for the heating produced at various experimental positions in the reactors EL-2 and EL-3 for several materials currently used in reactor construction. (author) [French] Apres avoir passe en revue les differents processus d'interaction des rayonnements, existant dans les reacteurs nucleaires, avec la matiere, l'auteur decrit les differentes methodes calorimetriques qui peuvent etre utilisees pour mesurer l'energie absorbee dans les materiaux. II presente ensuite en detail le calorimetre adiabatique, le dispositif de mesure associe et les methodes d'etalonnage qui ont ete retenus. Enfin il donne des valeurs d'echauffement dans divers emplacements experimentaux des piles EL-2 et EL-3 pour differents materiaux d'utilisation courante dans les reacteurs. (auteur)

  18. Control Rods in high-Flux Swimming-Pool Reactors; Les Barres de Controle dans les Piles Piscines a Haut Flux; Reguliruyushchie sterzhni dlya reaktorov bassejnovogo tipa s vysokoj plotnost'yu nejtronnogo potoka; Las Barras de Control en los Reactores Tipo Piscina de Flujo Elevado

    Energy Technology Data Exchange (ETDEWEB)

    Ageroni, P.; Blum, P.; Denielou, G.; Denis, P.; Meunier, C. [Centre d' Etudes Nucleaires de Grenoble (France)

    1964-06-15

    Control-rod problems in open swimming-pool high-flux and high specific power research reactors are examined in the light of the calibrations and experiments made during the construction of the SILOE reactor. Control-rod operating experience for this reactor at 13 MW is also described. 2. The following are considered in turn: (a) Reactivity balances and reactivity values for the different types of rod tested (cadmium, B4C , rare earths and combinations of these different elements). (b) Flux peaks set up in the core by the presence of the control rods, their incidence on the specific power, the fast fluxes that can be obtained and means of increasing them. (c ) The technological problems involved in constructing the rods. (d) In-pile cooling, vibration, deformation and scram-time problems. 3. In conclusion, current studies on control rods in open swimming-pool reactors operating in the 10 - 30 1W range are briefly summarized. (author) [French] 1. Les problemes poses par les barres de controle dans les reacteurs de recherche de type piscine ouverte a haute puissance specifique et haut flux sont examines a la lumiere des calculs et des experiences effectues pendant la construction du reacteur SILOE. Les resultats de l'experience de fonctionnement a 13 MW de ce reacteur sont egalement presentes en ce qui concerne les barres de controle. 2. On examine successivement: a) les bilans de reactivite et les valeurs en reactivite des differents types de barres qui ont ete essayes (Cadmium, B 4C , terres rares et combinaisons de ces differents elements). b) Les pics de flux crees dans le coeur par la presence de barres de controle, leur incidence sur la puissance specifique, et les flux rapides que l'on peut obtenir ainsi que les moyens correspondants d'accroitre ces flux. c) Les problemes technologiques poses par la construction des barres. d) Les problemes de refrigeration, de vibration, de deformation, de temps de chute en pile. 3. En conclusion on decrit sommairement les

  19. Innovative hybrid pile oscillator technique in the Minerve reactor: open loop vs. closed loop

    Science.gov (United States)

    Geslot, Benoit; Gruel, Adrien; Bréaud, Stéphane; Leconte, Pierre; Blaise, Patrick

    2018-01-01

    Pile oscillator techniques are powerful methods to measure small reactivity worth of isotopes of interest for nuclear data improvement. This kind of experiments has long been implemented in the Mineve experimental reactor, operated by CEA Cadarache. A hybrid technique, mixing reactivity worth estimation and measurement of small changes around test samples is presented here. It was made possible after the development of high sensitivity miniature fission chambers introduced next to the irradiation channel. A test campaign, called MAESTRO-SL, took place in 2015. Its objective was to assess the feasibility of the hybrid method and investigate the possibility to separate mixed neutron effects, such as fission/capture or scattering/capture. Experimental results are presented and discussed in this paper, which focus on comparing two measurements setups, one using a power control system (closed loop) and another one where the power is free to drift (open loop). First, it is demonstrated that open loop is equivalent to closed loop. Uncertainty management and methods reproducibility are discussed. Second, results show that measuring the flux depression around oscillated samples provides valuable information regarding partial neutron cross sections. The technique is found to be very sensitive to the capture cross section at the expense of scattering, making it very useful to measure small capture effects of highly scattering samples.

  20. Results of oscillation experiments on the Cesar and Marius piles - Uranium-Plutonium fuels

    International Nuclear Information System (INIS)

    Laponche, Bernard; Brunet, Max; Menessier, Denise; Morier, Francis; Basiuk, Marie-Jose; Tonolli, Jacky; Vanuxeem, Jacqueline

    1969-05-01

    The authors present, comment and discuss results obtained during three measurement campaigns performed on the Cesar and Marius atomic piles between 1965 and 1967 for the determination of some physical quantities (like the Plutonium η or its cross sections) from measurements of two signals which characterize the pile response to a central disturbance caused by the fuel to be studied. They more particularly address mass-corrected signals, the Uranium-235 and Boron calibration of the reactor, the local signal of the equivalent sample to a measured UPu sample. They indicate the different steps of interpretation of these results, present and discuss the measured results

  1. 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)

  2. 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

  3. Safety precautions in atomic pile control (1962); Securite dans le controle des piles atomiques (1962)

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-07-01

    We have been led to study the problem of safety in atomic pile control as a result of our participation on the one hand in the planning of C.E.A. atomic piles, and on the other hand in the pile safety sub omission considering atomic pile safety of operational or planned C.E.A. piles. We have thus had to consider the wishes occurring in piles during their operation and also their behaviour in the dynamic state The present work deals mainly with the importance of intrinsic safety devices, with the influence of reactivity variations on the power fluctuations during accidental operation, and with the development of robust and reliable safety appliances. The starting p accident has been especially studied both for low-flux piles where a compromise is necessary between the response time of the safety appliances and the statistical fluctuations and for high lux piles where xenon poisoning has an effect on the lower limit of the velocity of reactivity liberation. The desirability has been stressed of automation as a safety factor in atomic pile control. The details required for an understanding of the diagrams of the apparatus are given. (author) [French] Nous avons aborde le probleme de la securite dans le controle des piles atomiques a la suite de notre participation d'une part aux avant rojets de piles atomiques du CE.A. et d'autre part a l'examen au sein de la sous ommission de surete des piles, de la securite des piles du CE.A. en fonctionnement ou en projet. Nous avons ete amenes a nous interesser alors aux risques encourus par les piles pendant leur fonctionnement et par la meme a leur comportement en regime dynamique. Ce travail traite principalement de l'importance des securites intrinseques, de l'influence des variations de reactivite sur les evolutions de puissance en regime d'accident et du developpement d'appareillages de securite robustes et de fonctionnement tres sur. L'accident de demarrage a ete particulierement developpe aussi bien pour les piles a bas

  4. Safety precautions in atomic pile control (1962); Securite dans le controle des piles atomiques (1962)

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-07-01

    We have been led to study the problem of safety in atomic pile control as a result of our participation on the one hand in the planning of C.E.A. atomic piles, and on the other hand in the pile safety sub omission considering atomic pile safety of operational or planned C.E.A. piles. We have thus had to consider the wishes occurring in piles during their operation and also their behaviour in the dynamic state The present work deals mainly with the importance of intrinsic safety devices, with the influence of reactivity variations on the power fluctuations during accidental operation, and with the development of robust and reliable safety appliances. The starting p accident has been especially studied both for low-flux piles where a compromise is necessary between the response time of the safety appliances and the statistical fluctuations and for high lux piles where xenon poisoning has an effect on the lower limit of the velocity of reactivity liberation. The desirability has been stressed of automation as a safety factor in atomic pile control. The details required for an understanding of the diagrams of the apparatus are given. (author) [French] Nous avons aborde le probleme de la securite dans le controle des piles atomiques a la suite de notre participation d'une part aux avant rojets de piles atomiques du CE.A. et d'autre part a l'examen au sein de la sous ommission de surete des piles, de la securite des piles du CE.A. en fonctionnement ou en projet. Nous avons ete amenes a nous interesser alors aux risques encourus par les piles pendant leur fonctionnement et par la meme a leur comportement en regime dynamique. Ce travail traite principalement de l'importance des securites intrinseques, de l'influence des variations de reactivite sur les evolutions de puissance en regime d'accident et du developpement d'appareillages de securite robustes et de fonctionnement tres sur. L'accident de demarrage a ete particulierement

  5. An experimental study on pile spacing effects under lateral loading in sand.

    Science.gov (United States)

    Khari, Mahdy; Kassim, Khairul Anuar; Adnan, Azlan

    2013-01-01

    Grouped and single pile behavior differs owing to the impacts of the pile-to-pile interaction. Ultimate lateral resistance and lateral subgrade modulus within a pile group are known as the key parameters in the soil-pile interaction phenomenon. In this study, a series of experimental investigation was carried out on single and group pile subjected to monotonic lateral loadings. Experimental investigations were conducted on twelve model pile groups of configurations 1 × 2, 1 × 3, 2 × 2, 3 × 3, and 3 × 2 for embedded length-to-diameter ratio l/d = 32 into loose and dense sand, spacing from 3 to 6 pile diameter, in parallel and series arrangement. The tests were performed in dry sand from Johor Bahru, Malaysia. To reconstruct the sand samples, the new designed apparatus, Mobile Pluviator, was adopted. The ultimate lateral load is increased 53% in increasing of s/d from 3 to 6 owing to effects of sand relative density. An increasing of the number of piles in-group decreases the group efficiency owing to the increasing of overlapped stress zones and active wedges. A ratio of s/d more than 6d is large enough to eliminate the pile-to-pile interaction and the group effects. It may be more in the loose sand.

  6. 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

  7. 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

  8. 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

    fabricated for both room temperature proof-of-concept evaluations and high temperature testing. Evaluations have been performed jointly by the INL and the French Alternative Energies and Atomic Energy Commission (CEA), both in Idaho Falls (USA) and in Cadarache (France), in the framework of a collaborative program for instrumentation of Material Testing Reactors. Initial tests were conducted on samples with a large range of thermal conductivities and temperatures ranging from 20 deg. C to 600 deg. C. Particularly, tests were recently performed on a sample having thermal conductivity and dimensions similar to UO 2 and MOX nuclear fuels, in order to validate the ability of this sensor to operate for in-pile characterization of Light Water Reactors fuels. The results of the tests already completed at INL and CEA indicate that the Transient Hot Wire Needle Probe offers an enhanced method for in-pile detection of thermal conductivity. (authors)

  9. 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

    fabricated for both room temperature proof-of-concept evaluations and high temperature testing. Evaluations have been performed jointly by the INL and the French Alternative Energies and Atomic Energy Commission (CEA), both in Idaho Falls (USA) and in Cadarache (France), in the framework of a collaborative program for instrumentation of Material Testing Reactors. Initial tests were conducted on samples with a large range of thermal conductivities and temperatures ranging from 20 deg. C to 600 deg. C. Particularly, tests were recently performed on a sample having thermal conductivity and dimensions similar to UO{sub 2} and MOX nuclear fuels, in order to validate the ability of this sensor to operate for in-pile characterization of Light Water Reactors fuels. The results of the tests already completed at INL and CEA indicate that the Transient Hot Wire Needle Probe offers an enhanced method for in-pile detection of thermal conductivity. (authors)

  10. 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

  11. 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

  12. Study of fast reactor safety test facilities. Preliminary report

    International Nuclear Information System (INIS)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods

  13. Protection for work in the pressure tank of the pile G2; Protection des travaux dans le caisson de la pile G2

    Energy Technology Data Exchange (ETDEWEB)

    Chassany, J Ph; Rodier, J [Commissariat a l' Energie Atomique, Service de Protection contre les Radiations, Marcoule (France). Centre d' Etudes Nucleaires

    1961-07-01

    While the pile was shut down after a three-month run at full power, the secondary circuit was cleaned and some alterations were carried out. The pile contained 100 tons of uranium, half of which made up the periphery and was irradiated uranium. The possibility of carrying on work inside the pressure tank was not considered at the time of construction. Because of the heat and the irradiation it was only possible to remain in the pressure tank for a limited period of time, and several operators received doses of the order of 1.5 rem. Cotton clothing gave satisfactory protection against contamination and was more comfortable than the vinyl equipment. The work lasted for 17 days and involved 881 incursions into the pressure tank. (author) [French] Un nettoyage et des modifications ont ete realises dans le circuit secondaire pendant l'arret de la pile apres un fonctionnement de trois mois a pleine puissance. La pile contenait 100 tonnes d'uranium dont la moitie, composant la peripherie, etait de l'uranium irradie. La possibilite d'entreprendre un travail a l'interieur du caisson n'avait pas ete envisage lors de la construction. Le temps de sejour dans le caisson etait a la fois limite par l'irradiation et la chaleur, plusieurs operateurs, ont integre une dose de l'ordre de 1,5 rem. Les vetements de coton ont apporte une protection contre la contamination satisfaisante et un confort relatif par rapport aux equipements de vinyle. L'intervention a dure 17 jours, a comporte 881 entrees dans le caisson. (auteur)

  14. Experimental Verification of Integrity of Low-Pressure Injection Piles Structure - Pile Internal Capacity

    Science.gov (United States)

    Pachla, Henryk

    2017-12-01

    The idea of strengthening the foundation using injection piles lies in transferring loads from the foundation to the piles anchorage in existing structure and formed in the soil. Such a system has to be able to transfer loads from the foundation to the pile and from the pile onto the soil. Pile structure often reinforced with steel element has to also be able to transfer such a loading. According to the rules of continuum mechanics, the bearing capacity of such a system and a deformation of its individual elements can be determined by way of an analysis of the contact problem of three interfaces. Each of these surfaces is determined by different couples of materials. Those surfaces create: pile-foundation anchorage, bonding between reinforcement and material from which the pile is formed and pilesoil interface. What is essential is that on the contact surfaces the deformation of materials which adhere to each other can vary and depends on the mechanical properties and geometry of these surfaces. Engineering practice and experimental research point out that the failure in such structures occurs at interfaces. The paper is concentrating on presenting the experiments on interaction between cement grout and various types of steel reinforcement. The tests were conducted on the special low pressure injection piles widely used to strengthen foundations of already existing structures of historical buildings due to the technology of formation and injection pressure.

  15. 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)

  16. Characterizing hand-piled fuels

    Science.gov (United States)

    Clinton S. Wright; Paige C. Eagle; Cameron S. Balog

    2010-01-01

    Land managers throughout the West pile and burn surface fuels to mitigate fire hazard in dry forests. Whereas piling was historically conducted with heavy machinery following commercial harvesting operations, land managers are increasingly prescribing the use of hand piling and burning to treat surface fuels created by thinning and brush cutting. An estimate of the...

  17. Analysis of effect of different construction methods of piles on the end effect on skin friction of piles

    Institute of Scientific and Technical Information of China (English)

    ZHOU Hongbo; CHEN Zhuchang

    2007-01-01

    Based on the comparative analysis of end effect on skin friction of displacement-pile (driven pile),the end effect on skin friction of bored pile is studied.The end effect on skin friction between driven pile and bored pile is different and the end effect on skin friction of bored pile is reduce of skin friction in the soil layer adjacent to the pile end.The degradation degree of skin friction is deduced with the increase of the distance from pile end.The concept of additional mud cake formed by the effect of cushion at the bottom of borehole during pouring concrete is introduced to explain the mechanism of end effect on skin friction of the bored pile.The test results of post-grouting piles indicate that the post-grouting technique is an effective way to improve the end effect on skin friction of bored pile.

  18. Screw piles for cold climate foundations

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, R.; Sakr, M. [Almita Manufacturing Ltd., Edmonton, AB (Canada)

    2008-07-01

    Almita Manufacturing is an Alberta-based company that designs and builds screw piles with its own installation teams. It also engineers and supplies piles to numerous other companies and independent installers. The company services industries such as oil and gas; power transmission and distribution; and commercial construction. This presentation discussed the design and technical aspects of screw piles. A screw pile was defined as a steel pipe shaft with a 45 degree cut at the bottom and a formed helical plate welded to the outside of the pipe near the base and at a selected point on the shaft. The pile is screwed into the ground with a planetary drive head of suitable torque rating. The helical plate or helix helps facilitate the installation of the pile and gives the screw pile increased bearing capacity and pull-out resistance over a traditional straight-shaft pile. Screw piles were compared against cast in place concrete piles and steel driven piles. Screw piles were reported to have no tailings; no concrete curing time; no rebar, anchor belts, and no liners; and no dewatering. Screw piles can also be installed in all types of weather. Rhe Cree Burn Camp case study near Fort McMurray, Alberta was also presented. This residential camp and recreation complex consists of pre-fabricated units that make up three storey housing buildings and a single floor multi-use building. The case study provided information on soil; design parameter inputs; load testing program and pile configuration; geotechnical and structural design results; compression load test arrangement; pile test setup; and test results. The presentation also discussed fabrication as well as installation equipment. Various applications were also presented through a series of project pictures. Last, the presentation provided a simple cost analysis. tabs., figs.

  19. PAHR experiments in the MELUSINE reactor

    International Nuclear Information System (INIS)

    Rousseau, D.; Dereymez, P.; Guyon, H.; Junod, E.; Ploujoux, M.; Tournebize, F.; Backs, H.

    1983-01-01

    After a hypothetical accident in a fast neutron reactor core, the nuclear fuel and construction materials melt partially. In several out-of-pile devices, the melting materials and the sodium coolant come to interact thermodynamically. In short, a few seconds after the accident a bed of debris immersed in sodium is formed on a plane of steel. The PAHR program has as principal objective to study the thermodynamic behaviour of this bed in the MELUSINE reactor, taking into account the most crucial parameters that rule the phenomena. More particularly, the aim is to draw attention to the bed behaviour beyond the fusion point of the steel up to the partial fusion of the fuel. The authors describe the CELIA capsule and its instrumentation; the operation conditions of the reactor and the coupling factor; the out-of-pile materials and their operation conditions. (Auth.)

  20. 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

  1. Description of the Triton reactor; Pile Triton, rapport descriptif

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1967-09-01

    The Triton reactor is an enriched uranium pool type reactor. It began operation in 1959, after a divergence made on the June 30 the same year. Devoted to studies of radiation protection, its core can be displaced in the longitudinal direction. The pool can be separated in two unequal compartments by a wall. The Triton core is placed in a small compartment, the Nereide core in the big compartment. A third compartment without water is called Naiade II, is separated by a concrete wall in which is made a window closed by an aluminium plate (2.50 m x 2.70 m). The Naiade II hole is useful for protection experiments using the Nereide core. After a complete refitting, the power of the triton reactor that reached progressively from 1.2 MW to 2 MW, then 3 MW has reached in August 1965 6.5 MW. The reactor has been specialized in irradiations in fix position, the core become fix, the nereide core has been hung mobile. Since it has been used for structure materials irradiation, for radioelements fabrication and fundamental research. The following descriptions are valid for the period after August 1965. [French] Le reacteur Triton est un reacteur piscine, a uranium enrichi. Il est entre en fonctionnement en 1959, apres une divergence effectuee le 30 juin de cette meme annee. Destine a des etudes de protection contre les rayonnements, son coeur pouvait se deplacer dans le sens longitudinal. La piscine peut etre separee en deux compartiments inegaux par un batardeau. Le coeur triton est place dans le petit compartiment, le coeur Nereide dans le grand compartiment. Un troisieme compartiment sans eau, appele Naiade II, est separe par une paroi en beton dans laquelle est amenagee une fenetre obturee par une plaque d'aluminium (2,50 m x 2,70 m). La fosse Naiade II sert a des experiences de protection utilisant le coeur nereide. Apres une refonte complete, la puissance du reacteur triton qui etait passee progressivement de 1,2 MW a 2 MW, puis 3 MW, a atteint en aout 1965 6, 5 MW

  2. Development and testing of the EDF-2 reactor fuel element

    International Nuclear Information System (INIS)

    Delpeyroux, P.

    1964-01-01

    This technical report reviews the work which has been necessary for defining the EDF-2 fuel element. After giving briefly the EDF-2 reactor characteristics and the preliminary choice of parameters which made it possible to draw up a draft plan for the fuel element, the authors consider the research proper: - Uranium studies: tests on the passage into the β phase of an internal crown of a tube, bending of the tube under the effect of a localized force, welding of the end-pellets and testing for leaks. The resistance of the tube to crushing and of the pellets to yielding under the external pressure have been studied in detail in another CEA report. - Can studies: conditions of production and leak proof testing of the can, resistance of the fins to creep due to the effect of the gas flow. - Studies of the extremities of the element: creep under compression and welding of the plugs to the can. - Cartridge studies: determination of the characteristics of the can fuel fixing grooves and of the canning conditions, verification of the resistance of the fuel element to thermal cycling, determination of the temperature drop at the can-fuel interface dealt with in more detail in another CEA report. - Studies of the whole assembly: this work which concerns the graphite jacket, the support and the cartridge vibrations has been carried out by the Mechanical and Thermal Study Service (Mechanics Section). In this field the Fuel Element Study Section has investigated the behaviour of the centering devices in a gas current. The outcome of this research is the defining of the plan of the element the production process and the production specifications. The validity of ail these out-of-pile tests will be confirmed by the in-pile tests already under way and by irradiation of the elements in the EDF-2 reactor itself. In conclusion the programme is given for improving the fuel element and for defining the fuel element for the second charge. (authors) [fr

  3. Static and dynamic pile testing of reinforced concrete piles with structure integrated fibre optic strain sensors

    Science.gov (United States)

    Schilder, Constanze; Kohlhoff, Harald; Hofmann, Detlef; Basedau, Frank; Habel, Wolfgang R.; Baeßler, Matthias; Niederleithinger, Ernst; Georgi, Steven; Herten, Markus

    2013-05-01

    Static and dynamic pile tests are carried out to determine the load bearing capacity and the quality of reinforced concrete piles. As part of a round robin test to evaluate dynamic load tests, structure integrated fibre optic strain sensors were used to receive more detailed information about the strains along the pile length compared to conventional measurements at the pile head. This paper shows the instrumentation of the pile with extrinsic Fabry-Perot interferometers sensors and fibre Bragg gratings sensors together with the results of the conducted static load test as well as the dynamic load tests and pile integrity tests.

  4. 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

  5. An Experimental Study on Pile Spacing Effects under Lateral Loading in Sand

    Science.gov (United States)

    Khari, Mahdy; Kassim, Khairul Anuar; Adnan, Azlan

    2013-01-01

    Grouped and single pile behavior differs owing to the impacts of the pile-to-pile interaction. Ultimate lateral resistance and lateral subgrade modulus within a pile group are known as the key parameters in the soil-pile interaction phenomenon. In this study, a series of experimental investigation was carried out on single and group pile subjected to monotonic lateral loadings. Experimental investigations were conducted on twelve model pile groups of configurations 1 × 2, 1 × 3, 2 × 2, 3 × 3, and 3 × 2 for embedded length-to-diameter ratio l/d = 32 into loose and dense sand, spacing from 3 to 6 pile diameter, in parallel and series arrangement. The tests were performed in dry sand from Johor Bahru, Malaysia. To reconstruct the sand samples, the new designed apparatus, Mobile Pluviator, was adopted. The ultimate lateral load is increased 53% in increasing of s/d from 3 to 6 owing to effects of sand relative density. An increasing of the number of piles in-group decreases the group efficiency owing to the increasing of overlapped stress zones and active wedges. A ratio of s/d more than 6d is large enough to eliminate the pile-to-pile interaction and the group effects. It may be more in the loose sand. PMID:24453900

  6. Global and local scour at pile groups

    DEFF Research Database (Denmark)

    Sumer, B. Mutlu; Bundgaard, Klavs; Fredsøe, Jørgen

    2005-01-01

    This paper presents the results of an experimental investigation on scour around pile groups with different configurations exposed to steady current. Two kinds of tests were carried out: (1) Rigid-bed tests, and (2) Actual scour tests. In the former tests, the mean and turbulence properties...... of the flow were measured across the pile groups. The pile group configurations were such that the global scour was distinguished from the local scour. The results show that the global scour can be quite substantial....

  7. An MCNP parametric study of George C. Laurence's subcritical pile experiment

    Energy Technology Data Exchange (ETDEWEB)

    Dranga, R.; Blomeley, L., E-mail: ruxandra.dranga@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carrington, R. [McGill Univ., Dept. of Mathematics and Statistics, Montreal, Quebec (Canada)

    2014-12-01

    In the early 1940s at the National Research Council (NRC) Laboratories in Ottawa, Canada, Dr. George Laurence conducted several experiments to determine if a sustained nuclear fission chain reaction in a carbon-uranium arrangement (or 'pile') was possible. Although Dr. Laurence did not achieve criticality, these pioneering experiments marked a significant historical event in nuclear science, and they provided a valuable reference for subsequent experiments that led to the design of Canada's first heavy-water reactors at the Chalk River Nuclear Laboratories. This paper summarizes the results of a recent collaborative project between Atomic Energy of Canada Limited and the Deep River Science Academy undertaken to numerically explore the experiments carried out at the NRC Laboratories by Dr. Laurence, while teaching high school students about nuclear science and technology. In this study, a modern Monte Carlo reactor physics code, MCNP6, was utilized to identify and study the key parameters impacting the subcritical pile's neutron multiplication factor (e.g., moderation, geometry, material impurities) and quantify their effect on the extent of subcriticality. The findings presented constitute the first endeavour to model, using a current computational reactor physics tool, the seminal experiment that provided the foundation of Canada's nuclear science and technology program. (author)

  8. Stability of Slopes Reinforced with Truncated Piles

    Directory of Open Access Journals (Sweden)

    Shu-Wei Sun

    2016-01-01

    Full Text Available Piles are extensively used as a means of slope stabilization. A novel engineering technique of truncated piles that are unlike traditional piles is introduced in this paper. A simplified numerical method is proposed to analyze the stability of slopes stabilized with truncated piles based on the shear strength reduction method. The influential factors, which include pile diameter, pile spacing, depth of truncation, and existence of a weak layer, are systematically investigated from a practical point of view. The results show that an optimum ratio exists between the depth of truncation and the pile length above a slip surface, below which truncating behavior has no influence on the piled slope stability. This optimum ratio is bigger for slopes stabilized with more flexible piles and piles with larger spacing. Besides, truncated piles are more suitable for slopes with a thin weak layer than homogenous slopes. In practical engineering, the piles could be truncated reasonably while ensuring the reinforcement effect. The truncated part of piles can be filled with the surrounding soil and compacted to reduce costs by using fewer materials.

  9. Piles of objects

    KAUST Repository

    Hsu, Shu-Wei; Keyser, John

    2010-01-01

    We present a method for directly modeling piles of objects in multi-body simulations. Piles of objects represent some of the more interesting, but also most time-consuming portion of simulation. We propose a method for reducing computation in many

  10. 3D FEM Analysis of a Pile-Supported Riverine Platform under Environmental Loads Incorporating Soil-Pile Interaction

    Directory of Open Access Journals (Sweden)

    Denise-Penelope N. Kontoni

    2018-01-01

    Full Text Available An existing riverine platform in Egypt, together with its pile group foundation, is analyzed under environmental loads using 3D FEM structural analysis software incorporating soil-pile interaction. The interaction between the transfer plate and the piles supporting the platform is investigated. Two connection conditions were studied assuming fixed or hinged connection between the piles and the reinforced concrete platform for the purpose of comparison of the structural behavior. The analysis showed that the fixed or hinged connection condition between the piles and the platform altered the values and distribution of displacements, normal force, bending moments, and shear forces along the length of each pile. The distribution of piles in the pile group affects the stress distribution on both the soil and platform. The piles were found to suffer from displacement failure rather than force failure. Moreover, the resulting bending stresses on the reinforced concrete plate in the case of a fixed connection between the piles and the platform were almost doubled and much higher than the allowable reinforced concrete stress and even exceeded the ultimate design strength and thus the environmental loads acting on a pile-supported riverine offshore platform may cause collapse if they are not properly considered in the structural analysis and design.

  11. 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)

  12. Measurement and Analysis of Horizontal Vibration Response of Pile Foundations

    Directory of Open Access Journals (Sweden)

    A. Boominathan

    2007-01-01

    Full Text Available Pile foundations are frequently used in very loose and weak deposits, in particular soft marine clays deposits to support various industrial structures, power plants, petrochemical complexes, compressor stations and residential multi-storeyed buildings. Under these circumstances, piles are predominantly subjected to horizontal dynamic loads and the pile response to horizontal vibration is very critical due to its low stiffness. Though many analytical methods have been developed to estimate the horizontal vibration response, but they are not well validated with the experimental studies. This paper presents the results of horizontal vibration tests carried out on model aluminium single piles embedded in a simulated Elastic Half Space filled with clay. The influence of various soil and pile parameters such as pile length, modulus of clay, magnitude of dynamic load and frequency of excitation on the horizontal vibration response of single piles was examined. Measurement of various response quantities, such as the load transferred to the pile, pile head displacement and the strain variation along the pile length were done using a Data Acquisition System. It is found that the pile length, modulus of clay and dynamic load, significantly influences the natural frequency and peak amplitude of the soil-pile system. The maximum bending moment occurs at the fundamental frequency of the soil-pile system. The maximum bending moment of long piles is about 2 to 4 times higher than that of short piles and it increases drastically with the increase in the shear modulus of clay for both short and long piles. The active or effective pile length is found to be increasing under dynamic load and empirical equations are proposed to estimate the active pile length under dynamic loads.

  13. The use and evolution of the CEA research reactors; Utilisation et evolution des reacteurs de recherche du C.E.A

    Energy Technology Data Exchange (ETDEWEB)

    Rossillon, F; Chauvez, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The authors successively examine the different research reactors in use in the French C.E.A. Nuclear Centres. They trace briefly their histories, describing how they have been used up to the present, and how they have been adapted to changes in programme by means of certain modifications. They also describe the reasons which have led to the elaboration of the project for the new reactor Osiris. Zoe, the oldest reactor in the CEA, has been in service in the Centre de Fontenay-aux-Roses since 1948. It is used mainly for measurements of absorption cross-sections in graphite, and for various short irradiations which do not require high fluxes. The reactor EL 2, in service since 1952, was used for the first studies on gas cooling. It has also been widely used for the production of radioisotopes and for a large number of experiments in the fields of physics, metallurgy and physical chemistry. The ageing of certain elements of the reactor has led to the decision to close it down in the near future The reactor EL 3 has been widely used for experiments in physics and in the investigation of fuels. The possibilities of the reactor in fast neutron irradiations will be considerably improved by the adoption of a new type of core (the 'snow crystal' structure). Triton-I, a 2 MW swimming-pool reactor, is used for the most part for fast neutron and gamma irradiations. The modifications being carried out on it at present should result in an increase in the power of the reactor up to 4 or 5 MW. In a neighbouring compartment is housed Triton-II which is of the same general structure, as Triton-I, but whose maximum power is 100 kW. Triton-II is used solely for studies on shielding. Melusine, a 2 MW swimming-pool reactor, has been in use in the Centre d'Etudes Nucleaires de Grenoble since 1959. It has supported a very high programme concerned mainly with solid state physics, fundamental research into refractory fissile materials and special graphites, and the study of the behaviour of

  14. 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

  15. Swimming-pool piles

    International Nuclear Information System (INIS)

    Trioulaire, M.

    1959-01-01

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

  16. Laboratory Test Setup for Cyclic Axially Loaded Piles in Sand

    DEFF Research Database (Denmark)

    Thomassen, Kristina; Ibsen, Lars Bo; Andersen, Lars Vabbersgaard

    2017-01-01

    This paper presents a comprehensive description and the considerations regarding the design of a new laboratory test setup for testing cyclic axially loaded piles in sand. The test setup aims at analysing the effect of axial one-way cyclic loading on pile capacity and accumulated displacements....... Another aim was to test a large diameter pile segment with dimensions resembling full-scale piles to model the interface properties between pile and sand correctly. The pile segment was an open-ended steel pipe pile with a diameter of 0.5 m and a length of 1 m. The sand conditions resembled the dense sand...... determined from the API RP 2GEO standard and from the test results indicated over consolidation of the sand. Two initial one-way cyclic loading tests provided results of effects on pile capacity and accumulated displacements in agreement with other researchers’ test results....

  17. Decontamination and decommissioning activities photobriefing book FY 1999

    International Nuclear Information System (INIS)

    2000-01-01

    The Chicago Pile 5 (CP-5) Reactor, the first reactor built on the Argonne National Laboratory-East site, followed a rich history that had begun in 1942 with Enrico Fermi's original pile built under the west stands at the Stagg Field Stadium of The University of Chicago. CP-5 was a 5-megawatt, heavy water-moderated, enriched uranium-fueled reactor used to produce neutrons for scientific research from 1954--79. The reactor was shut down and defueled in 1979, and placed into a lay-up condition pending funding for decontamination and decommissioning (D and D). In 1990, work was initiated on the D and D of the facility in order to alleviate safety and environmental concerns associated with the site due to the deterioration of the building and its associated support systems. A decision was made in early Fiscal Year (FY) 1999 to direct focus and resources to the completion of the CP-5 Reactor D and D Project. An award of contract was made in December 1998 to Duke Engineering and Services (Marlborough, MA), and a D and D crew was on site in March 1999 to begin work, The project is scheduled to be completed in July 2000. The Laboratory has determined that the building housing the CP-5 facility is surplus to the Laboratory's needs and will be a candidate for demolition. In addition to a photographic chronology of FY 1999 activities at the CP-5 Reactor D and D Project, brief descriptions of other FY 1999 activities and of projects planned for the future are provided in this photobriefing book

  18. Experimental Verification of Integrity of Low-Pressure Injection Piles Structure – Pile Internal Capacity

    Directory of Open Access Journals (Sweden)

    Pachla Henryk

    2017-12-01

    Full Text Available The idea of strengthening the foundation using injection piles lies in transferring loads from the foundation to the piles anchorage in existing structure and formed in the soil. Such a system has to be able to transfer loads from the foundation to the pile and from the pile onto the soil. Pile structure often reinforced with steel element has to also be able to transfer such a loading. According to the rules of continuum mechanics, the bearing capacity of such a system and a deformation of its individual elements can be determined by way of an analysis of the contact problem of three interfaces. Each of these surfaces is determined by different couples of materials. Those surfaces create: pile-foundation anchorage, bonding between reinforcement and material from which the pile is formed and pilesoil interface. What is essential is that on the contact surfaces the deformation of materials which adhere to each other can vary and depends on the mechanical properties and geometry of these surfaces. Engineering practice and experimental research point out that the failure in such structures occurs at interfaces. The paper is concentrating on presenting the experiments on interaction between cement grout and various types of steel reinforcement. The tests were conducted on the special low pressure injection piles widely used to strengthen foundations of already existing structures of historical buildings due to the technology of formation and injection pressure.

  19. Qualification of an out-of-pile Thermohydraulic test Bench (BETHY) developed to calibrate calorimetric cells under specific JHR experimental conditions

    International Nuclear Information System (INIS)

    De Vita, C.; Brun, J.; Carette, M.; Reynard-Carette, C.; Lyoussi, A.; Fourmentel, D.; Villard, J.F.; Guimbal, P.; Malo, J.Y.

    2013-06-01

    Online in-pile measurement methods are crucial during irradiations in material testing reactors to better understand the behavior of materials under accelerated ageing conditions and of nuclear fuels under high irradiation levels. Thus, the construction of a new Material Testing Reactor such as the Jules Horowitz Reactor (JHR) leads to new research and development programs devoted to innovative instrumentation and measurement methods. The presented works are performed in the framework of the IN-CORE program, 'Instrumentation for Nuclear radiations and Calorimetry Online in Reactor', between CEA and Aix-Marseille University. The program aim is to develop experimental devices and test bench to quantify more precisely the nuclear heating parameter in the JHR experimental channels. This in-pile parameter is usually measured by means of calorimeter or gamma thermometer. This paper focuses on a new out-of-pile test bench called BETHY. This bench was developed to study the response of a differential calorimeter during its preliminary calibration step according to specific thermal and hydraulic conditions occurring inside one type of JHR core channel. The first section of this paper is dedicated to a detailed description of the bench. The second part presents the study of the thermal characteristics established in the bench for two main thermal running modes. The last one concerns the calibration curve of the reference cell of the differential calorimeter in the case of homogenous temperature. (authors)

  20. Vibration tests on pile-group foundations using large-scale blast excitation

    International Nuclear Information System (INIS)

    Tanaka, Hideo; Hijikata, Katsuichirou; Hashimoto, Takayuki; Fujiwara, Kazushige; Kontani, Osamu; Miyamoto, Yuji; Suzuki, Atsushi

    2005-01-01

    Extensive vibration tests have been performed on pile-supported structures at a large-scale mining site. Ground motions induced by large-scale blasting operations were used as excitation forces for vibration tests. The main objective of this research is to investigate the dynamic behavior of pile-supported structures, in particular, pile-group effects. Two test structures were constructed in an excavated 4 m deep pit. One structure had 25 steel tubular piles and the other had 4 piles. The super-structures were exactly the same. The test pit was backfilled with sand of appropriate grain size distributions in order to obtain good compaction, especially between the 25 piles. Accelerations were measured at the structures, in the test pit and in the adjacent free field, and pile strains were measured. The vibration tests were performed six times with different levels of input motions. The maximum horizontal acceleration recorded at the adjacent ground surface varied from 57 cm/s 2 to 1683 cm/s 2 according to the distances between the test site and the blast areas. Maximum strains were 13,400 micro-strains were recorded at the pile top of the 4-pile structure, which means that these piles were subjected to yielding

  1. Characterizing Axial Stiffness of Individual Batter Piles with Emphasis on Elevated, Laterally Loaded, Clustered Pile Groups

    Science.gov (United States)

    2016-11-01

    using the appropriate stiffness based on the direction of the calculated pile load. 1...load cases. CPGA utilizes the stiffness method (Saul 1968) of three-dimensional pile group analysis for user-specified static loadings. The pile...CPGA analysis and coordinate systems (global and pile) As discussed in Chapter 1, the CPGA software utilizes the stiffness method (Saul 1968) of

  2. 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.

  3. Modelling the pile load test

    OpenAIRE

    Prekop Ľubomír

    2017-01-01

    This paper deals with the modelling of the load test of horizontal resistance of reinforced concrete piles. The pile belongs to group of piles with reinforced concrete heads. The head is pressed with steel arches of a bridge on motorway D1 Jablonov - Studenec. Pile model was created in ANSYS with several models of foundation having properties found out from geotechnical survey. Finally some crucial results obtained from computer models are presented and compared with these obtained from exper...

  4. Use of pile driving analysis for assessment of axial load capacity of piles : [technical summary].

    Science.gov (United States)

    2012-01-01

    The dynamic response of a pile during driving is very : complex, involving the interactions of the hammer, cushion, : pile and soil during application of an impact load. : The first analysis aimed at simulating a hammer blow on : a pile was published...

  5. Teaching about City Life in Chicago.

    Science.gov (United States)

    LaBianco, Claudine R.

    These materials discuss the history of Chicago, Illinois, including prominent persons in the city's past, and landmarks for which Chicago is well known. A number of activities are suggested, some of which concern Chicago's industries, historical sites, architecture, newspapers, ethnic groups, and history. A list of books about Chicago for…

  6. The development of fast neutron reactors in France

    International Nuclear Information System (INIS)

    Petit, J.

    1983-01-01

    The French strategy is based on a coherent and carefully defined development program launched in 1950, each new stage of which is decided upon after the results of the preceding stage are analyzed. Rapsodie, the forst experimental reactor began operating in 1967 after more than 10 years of full scale test of its components. The in-pile fuel and component behaviour experience gained was put to immediate use for the design and out-of-pile tests of components for Phenix. Phenix is a prototype power generating demonstration reactor which has been operating since 1974. Rapsodie was developed for rigorous and statistical test of fuel behaviour. Super Phenix, the 1200 MWe reactor of the Creys Malville plant was ordered in 1977 and benefited from the 3 years operating experience gained with Phenix and the 10 years of operating experience acquired with Rapsodie. In 1986, after only one year of experience with Super Phenix, it is expected that all the parties involved in the financial and technical aspects of Super Phenix will be in the position to suggest the next stage in the development of large commercial plants to the government. The next reactor in the series, Super Phenix 2, is currently being studied

  7. Innovative hybrid pile oscillator technique in the Minerve reactor: open loop vs. closed loop

    Directory of Open Access Journals (Sweden)

    Geslot Benoit

    2018-01-01

    Full Text Available Pile oscillator techniques are powerful methods to measure small reactivity worth of isotopes of interest for nuclear data improvement. This kind of experiments has long been implemented in the Mineve experimental reactor, operated by CEA Cadarache. A hybrid technique, mixing reactivity worth estimation and measurement of small changes around test samples is presented here. It was made possible after the development of high sensitivity miniature fission chambers introduced next to the irradiation channel. A test campaign, called MAESTRO-SL, took place in 2015. Its objective was to assess the feasibility of the hybrid method and investigate the possibility to separate mixed neutron effects, such as fission/capture or scattering/capture. Experimental results are presented and discussed in this paper, which focus on comparing two measurements setups, one using a power control system (closed loop and another one where the power is free to drift (open loop. First, it is demonstrated that open loop is equivalent to closed loop. Uncertainty management and methods reproducibility are discussed. Second, results show that measuring the flux depression around oscillated samples provides valuable information regarding partial neutron cross sections. The technique is found to be very sensitive to the capture cross section at the expense of scattering, making it very useful to measure small capture effects of highly scattering samples.

  8. Modelling the pile load test

    Directory of Open Access Journals (Sweden)

    Prekop Ľubomír

    2017-01-01

    Full Text Available This paper deals with the modelling of the load test of horizontal resistance of reinforced concrete piles. The pile belongs to group of piles with reinforced concrete heads. The head is pressed with steel arches of a bridge on motorway D1 Jablonov - Studenec. Pile model was created in ANSYS with several models of foundation having properties found out from geotechnical survey. Finally some crucial results obtained from computer models are presented and compared with these obtained from experiment.

  9. Laterally Loaded Piles in Clay

    DEFF Research Database (Denmark)

    Christensen, Helle; Niewald, Gitte

    1992-01-01

    The ultimate lateral resistance of a pile element moved horizontally can be analyzed by the theory of plasticity. At a certain depth the movements around the pile are purely horizontal and upper bound solutions can be estimated theoretically under undrained circumstances. Model tests...... in the laboratory show ultimate resistances close to the estimated limits and p - y curves close to curves based on test results from full-scale piles. Rough and smooth piles with circular and square cross sections are investigated....

  10. Aspects of nuclear reactor safety

    International Nuclear Information System (INIS)

    Hardt, P. von der; Rottger, H.

    1980-01-01

    The Colloquium on 'Irradiation Tests for Reactor Safety Programmes' has been organised by JRC Petten in order to determine the present state of technology in the field. The role of research and test reactors for studies of structural material and fuel elements under transient and off-normal conditions was to be explained. The Colloquium has been attended by 110 participants from outside and inside Europe. 27 papers were presented covering the major ongoing projects in Japan, the United States, and in Europe, and elaborating in particular: - design rationale and layout of safety irradiation experiments; - design, manufacture, and performance of irradiation equipment with particular attention to generation and control of transient conditions, fast response in-pile instrumentation and its out-of-pile data retrieval; - post-irradiation evaluation; - results and analytical support

  11. 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.

  12. Kinetic parameters of the GUINEVERE reference configuration in VENUS-F reactor obtained from a pile noise experiment using Rossi and Feynman methods

    Energy Technology Data Exchange (ETDEWEB)

    Geslot, Benoit; Pepino, Alexandra; Blaise, Patrick; Mellier, Frederic [CEA, DEN, DER/SPEx, Cadarache, F-13108 St Paul Lez Durance (France); Lecouey, Jean-Luc [LPC Caen, ENSICAEN, Universite de Caen, CNRS/IN2P3, 6 Bd. Marechal Juin 14050 Caen cedex (France); Carta, Mario [ENEA, UTFISST-REANUC, C.R. Casaccia, S.P.040 via Anguillarese 301, 00123 S. Maria Di Galeria, Roma (Italy); Kochetkov, Anatoly; Vittiglio, Guido [SCK.CEN, Belgian Nuclear Research Centre, Boeretang 200, BE-2400, Mol (Belgium); Billebaud, Annick [LPSC, CNRS, IN2P3/UJF/INPG, 53 Avenue des Martyrs, 38026 Grenoble cedex (France)

    2015-07-01

    A pile noise measurement campaign has been conducted by the CEA in the VENUS-F reactor (SCK-CEN, Mol Belgium) in April 2011 in the reference critical configuration of the GUINEVERE experimental program. The experimental setup made it possible to estimate the core kinetic parameters: the prompt neutron decay constant, the delayed neutron fraction and the generation time. A precise assessment of these constants is of prime importance. In particular, the effective delayed neutron fraction is used to normalize and compare calculated reactivities of different subcritical configurations, obtained by modifying either the core layout or the control rods position, with experimental ones deduced from the analysis of measurements. This paper presents results obtained with a CEA-developed time stamping acquisition system. Data were analyzed using Rossi-α and Feynman-α methods. Results were normalized to reactor power using a calibrated fission chamber with a deposit of Np-237. Calculated factors were necessary to the analysis: the Diven factor was computed by the ENEA (Italy) and the power calibration factor by the CNRS/IN2P3/LPC Caen. Results deduced with both methods are consistent with respect to calculated quantities. Recommended values are given by the Rossi-α estimator, that was found to be the most robust. The neutron generation time was found equal to 0.438 ± 0.009 μs and the effective delayed neutron fraction is 765 ± 8 pcm. Discrepancies with the calculated value (722 pcm, calculation from ENEA) are satisfactory: -5.6% for the Rossi-α estimate and -2.7% for the Feynman-α estimate. (authors)

  13. Reactor G1: high power experiments

    International Nuclear Information System (INIS)

    Laage, F. de; Teste du Baillet, A.; Veyssiere, A.; Wanner, G.

    1957-01-01

    The experiments carried out in the starting-up programme of the reactor G1 comprised a series of tests at high power, which allowed the following points to be studied: 1- Effect of poisoning by Xenon (absolute value, evolution). 2- Temperature coefficients of the uranium and graphite for a temperature distribution corresponding to heating by fission. 3- Effect of the pressure (due to the coiling system) on the reactivity. 4- Calibration of the security rods as a function of their position in the pile (1). 5- Temperature distribution of the graphite, the sheathing, the uranium and the air leaving the canals, in a pile running normally at high power. 6- Neutron flux distribution in a pile running normally at high power. 7- Determination of the power by nuclear and thermodynamic methods. These experiments have been carried out under two very different pile conditions. From the 1. to the 15. of August 1956, a series of power increases, followed by periods of stabilisation, were induced in a pile containing uranium only, in 457 canals, amounting to about 34 tons of fuel. A knowledge of the efficiency of the control rods in such a pile has made it possible to measure with good accuracy the principal effects at high temperatures, that is, to deal with points 1, 2, 3, 5. Flux charts giving information on the variations of the material Laplacian and extrapolation lengths in the reflector have been drawn up. Finally the thermodynamic power has been measured under good conditions, in spite of some installation difficulties. On September 16, the pile had its final charge of 100 tons. All the canals were loaded, 1,234 with uranium and 53 (i.e. exactly 4 per cent of the total number) with thorium uniformly distributed in a square lattice of 100 cm side. Since technical difficulties prevented the calibration of the control rods, the measurements were limited to the determination of the thermodynamic power and the temperature distributions (points 5 and 7). This report will

  14. 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

  15. Technical committee meeting on evaluation of radioactive materials release and sodium fires in fast reactors

    International Nuclear Information System (INIS)

    1996-01-01

    The objectives of the Technical Committee Meeting was to review the activities of research on radioactive materials release and sodium fires in fast reactors in each of the participating countries. It covered: out-of-pile experiments and analysis codes on source term; in-pile experiments on source term; core disruptive accidents; sodium leak experience in liquid metal fast reactors; evaluation of sodium fire; and aerosol behaviour

  16. Technical committee meeting on evaluation of radioactive materials release and sodium fires in fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    The objectives of the Technical Committee Meeting was to review the activities of research on radioactive materials release and sodium fires in fast reactors in each of the participating countries. It covered: out-of-pile experiments and analysis codes on source term; in-pile experiments on source term; core disruptive accidents; sodium leak experience in liquid metal fast reactors; evaluation of sodium fire; and aerosol behaviour.

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

    International Nuclear Information System (INIS)

    Smith, D.L.

    1987-10-01

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

  18. The First Reactor, 40th Anniversary (rev.)

    Energy Technology Data Exchange (ETDEWEB)

    Allardice, Corbin; Trapnell, Edward R; Fermi, Enrico; Fermi, Laura; Williams, Robert C

    1982-12-01

    This booklet, an updated version of the original booklet describing the first nuclear reactor, was written in honor of the 40th anniversary of the first reactor or "pile". It is based on firsthand accounts told to Corbin Allardice and Edward R. Trapnell, and includes recollections of Enrico and Laura Fermi.

  19. 40 CFR 264.554 - Staging piles.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 25 2010-07-01 2010-07-01 false Staging piles. 264.554 Section 264.554... for Cleanup § 264.554 Staging piles. This section is written in a special format to make it easier to... staging pile? A staging pile is an accumulation of solid, non-flowing remediation waste (as defined in...

  20. Some equipment for graphite research in swimming pool reactors; Quelques dispositifs d'etude du graphite dans les piles piscines

    Energy Technology Data Exchange (ETDEWEB)

    Seguin, M; Arragon, Ph; Dupont, G; Gentil, J; Tanis, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-07-01

    The irradiation devices described are used for research concerning reactors of the natural uranium type, moderated by graphite and cooled by carbon dioxide. The devices are generally designed for use in swimming pool reactors. The following points have been particularly studied: - maximum use of the irradiation volume, - use of the simplest technological solutions, - standardization of certain constituent parts. This standardization calls for precision machining and careful assembling; these requirements are also true when a relatively low irradiation temperature is required and the nuclear heating is pronounced. Finally, the design of these devices is suitable for the irradiation of other fissile or non-fissile materials. (authors) [French] Les dispositifs d'irradiation decrits servent aux etudes relatives a la filiere des reacteurs a uranium naturel, moderes au graphite et refroidis par le gaz carbonique. Ils sont generalement concus pour etre utilises dans des piles piscines. L'accent a ete mis sur: - l'utilisation au maximum du volume d'irradiation, - le recours aux solutions technologiques les plus simples, - la standardisation de certaines parties constitutives. Cette standardisation impose un usinage precis et un montage soigne, lesquels sont egalement necessaires lorsqu'on doit obtenir une temperature d'irradiation relativement basse alors que l'echauffement nucleaire est important. Enfin, la conception de ces dispositifs est valable pour irradier d'autres materiaux non fissiles ou fissiles. (auteurs)

  1. 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)

  2. 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).

  3. 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).

  4. Pullout capacity of batter pile in sand.

    Science.gov (United States)

    Nazir, Ashraf; Nasr, Ahmed

    2013-03-01

    Many offshore structures are subjected to overturning moments due to wind load, wave pressure, and ship impacts. Also most of retaining walls are subjected to horizontal forces and bending moments, these forces are due to earth pressure. For foundations in such structures, usually a combination of vertical and batter piles is used. Little information is available in the literature about estimating the capacity of piles under uplift. In cases where these supporting piles are not vertical, the behavior under axial pullout is not well established. In order to delineate the significant variables affecting the ultimate uplift shaft resistance of batter pile in dry sand, a testing program comprising 62 pullout tests was conducted. The tests are conducted on model steel pile installed in loose, medium, and dense sand to an embedded depth ratio, L/d, vary from 7.5 to 30 and with various batter angles of 0°, 10°, 20°, and 30°. Results indicate that the pullout capacity of a batter pile constructed in dense and/or medium density sand increases with the increase of batter angle attains maximum value and then decreases, the maximum value of Pα occurs at batter angle approximately equal to 20°, and it is about 21-31% more than the vertical pile capacity, while the pullout capacity for batter pile that constructed in loose sand decreases with the increase of pile inclination. The results also indicated that the circular pile is more resistant to pullout forces than the square and rectangular pile shape. The rough model piles tested is experienced 18-75% increase in capacity compared with the smooth model piles. The suggested relations for the pullout capacity of batter pile regarding the vertical pile capacity are well predicted.

  5. Pullout capacity of batter pile in sand

    Directory of Open Access Journals (Sweden)

    Ashraf Nazir

    2013-03-01

    Full Text Available Many offshore structures are subjected to overturning moments due to wind load, wave pressure, and ship impacts. Also most of retaining walls are subjected to horizontal forces and bending moments, these forces are due to earth pressure. For foundations in such structures, usually a combination of vertical and batter piles is used. Little information is available in the literature about estimating the capacity of piles under uplift. In cases where these supporting piles are not vertical, the behavior under axial pullout is not well established. In order to delineate the significant variables affecting the ultimate uplift shaft resistance of batter pile in dry sand, a testing program comprising 62 pullout tests was conducted. The tests are conducted on model steel pile installed in loose, medium, and dense sand to an embedded depth ratio, L/d, vary from 7.5 to 30 and with various batter angles of 0°, 10°, 20°, and 30°. Results indicate that the pullout capacity of a batter pile constructed in dense and/or medium density sand increases with the increase of batter angle attains maximum value and then decreases, the maximum value of Pα occurs at batter angle approximately equal to 20°, and it is about 21–31% more than the vertical pile capacity, while the pullout capacity for batter pile that constructed in loose sand decreases with the increase of pile inclination. The results also indicated that the circular pile is more resistant to pullout forces than the square and rectangular pile shape. The rough model piles tested is experienced 18–75% increase in capacity compared with the smooth model piles. The suggested relations for the pullout capacity of batter pile regarding the vertical pile capacity are well predicted.

  6. Numerical Analysis of Helical Pile-Soil Interaction under Compressive Loads

    Science.gov (United States)

    Polishchuk, A. I.; Maksimov, F. A.

    2017-11-01

    The results of the field tests of full-scale steel helical piles in clay soils intended for prefabricated temporary buildings foundations are presented in this article. The finite element modeling was used for the evaluation of stress distribution of the clay soil around helical piles. An approach of modeling of the screw-pile geometry has been proposed through the Finite Element Analysis. Steel helical piles with a length of 2.0 m, shaft diameter of 0.108 m and a blade diameter of 0.3 m were used in the experiments. The experiments have shown the efficiency of double-bladed helical piles in the clay soils compared to single-bladed piles. It has been experimentally established that the introduction of the second blade into the pile shaft provides an increase of the bearing capacity in clay soil up to 30% compared to a single-bladed helical pile with similar geometrical dimensions. The numerical results are compared with the measurements obtained by a large scale test and the bearing capacity has been estimated. It has been found that the model results fit the field results. For a double-bladed helical pile it was revealed that shear stresses upon pile loading are formed along the lateral surface forming a cylindrical failure surface.

  7. The out-of-pile test for internal pressure measurement of nuclear fuel rod using LVDT

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Kim, D. S.; Joo, K. N.; Park, S. J.; Kang, Y. H.; Kim, Y. K.; Yeum, K. I. [KAERI, Taejon (Korea, Republic of)

    2002-05-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT(Linear Variable Differential Transformer). The objectives of this test were to understand the LVDT's characteristics and to study its application techniques for fuel irradiation technology. It will be required to analyze the acquired internal pressure of fuel rod during fuel irradiation test in HANARO. Therefore, the out of pile test system for pressure measurement was developed, and the test with the LVDT at room temperature were performed. This test were implemented in 1 kg/cm{sup 2} increment from 1 kg/cm{sup 2} to 30 kg/cm{sup 2}, and repeated 6 times at same condition. The LVDT's sensitivities were obtained by following two ways, the one by test and the other by calculation from characteristics data. These two sensitivities were compared and analyzed. The calculation method for internal pressure of nuclear fuel rod at specified temperature was also established. The results of the out-of-pile test will be used to predict accurately the internal pressure of fuel rod during irradiation test. And, the well qualified out-of-pile tests are needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation capsule.

  8. Evaluation of axial pile bearing capacity based on pile driving analyzer (PDA) test using Neural Network

    Science.gov (United States)

    Maizir, H.; Suryanita, R.

    2018-01-01

    A few decades, many methods have been developed to predict and evaluate the bearing capacity of driven piles. The problem of the predicting and assessing the bearing capacity of the pile is very complicated and not yet established, different soil testing and evaluation produce a widely different solution. However, the most important thing is to determine methods used to predict and evaluate the bearing capacity of the pile to the required degree of accuracy and consistency value. Accurate prediction and evaluation of axial bearing capacity depend on some variables, such as the type of soil, diameter, and length of pile, etc. The aims of the study of Artificial Neural Networks (ANNs) are utilized to obtain more accurate and consistent axial bearing capacity of a driven pile. ANNs can be described as mapping an input to the target output data. The method using the ANN model developed to predict and evaluate the axial bearing capacity of the pile based on the pile driving analyzer (PDA) test data for more than 200 selected data. The results of the predictions obtained by the ANN model and the PDA test were then compared. This research as the neural network models give a right prediction and evaluation of the axial bearing capacity of piles using neural networks.

  9. 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

  10. 76 FR 58108 - Safety Zone; Ryder Cup Captain's Duel Golf Shot, Chicago River, Chicago, IL

    Science.gov (United States)

    2011-09-20

    ...-AA00 Safety Zone; Ryder Cup Captain's Duel Golf Shot, Chicago River, Chicago, IL AGENCY: Coast Guard... the Chicago River during a golfing event that will involve hitting golf balls from land onto a... vessels from the hazards associated with golf balls being hit from land onto a stationary barge in the...

  11. Simplified analysis of laterally loaded pile groups

    Directory of Open Access Journals (Sweden)

    F.M. Abdrabbo

    2012-06-01

    Full Text Available The response of laterally loaded pile groups is a complicated soil–structure interaction problem. Although fairly reliable methods are developed to predicate the lateral behavior of single piles, the lateral response of pile groups has attracted less attention due to the required high cost and complication implication. This study presents a simplified method to analyze laterally loaded pile groups. The proposed method implements p-multiplier factors in combination with the horizontal modulus of subgrade reaction. Shadowing effects in closely spaced piles in a group were taken into consideration. It is proven that laterally loaded piles embedded in sand can be analyzed within the working load range assuming a linear relationship between lateral load and lateral displacement. The proposed method estimates the distribution of lateral loads among piles in a pile group and predicts the safe design lateral load of a pile group. The benefit of the proposed method is in its simplicity for the preliminary design stage with a little computational effort.

  12. The first nuclear chain reaction

    International Nuclear Information System (INIS)

    Zinn, W.H.

    1989-01-01

    The author offers his recollections of the experimental efforts beginning in 1939 which culminated in the Chain Reaction in the squash court on December 2, 1942. Recalled are Columbia University experiments which did much to establish the feasibility of the chain in natural uranium and which stimulated the creation of the Manhattan District. The Columbia group moved to the University of Chicago, where, in early summer of 1942, construction and analysis of a number of subcritical reactors (piles) gave assurance with a high probability that only a reasonable amount of uranium and moderator would be required

  13. Research reactors for power reactor fuel and materials testing - Studsvik's experience

    International Nuclear Information System (INIS)

    Grounes, M.

    1998-01-01

    Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

  14. Developing a Data Visualization System for the Bank of America Chicago Marathon (Chicago, Illinois USA).

    Science.gov (United States)

    Hanken, Taylor; Young, Sam; Smilowitz, Karen; Chiampas, George; Waskowski, David

    2016-10-01

    As one of the largest marathons worldwide, the Bank of America Chicago Marathon (BACCM; Chicago, Illinois USA) accumulates high volumes of data. Race organizers and engaged agencies need the ability to access specific data in real-time. This report details a data visualization system designed for the Chicago Marathon and establishes key principles for event management data visualization. The data visualization system allows for efficient data communication among the organizing agencies of Chicago endurance events. Agencies can observe the progress of the race throughout the day and obtain needed information, such as the number and location of runners on the course and current weather conditions. Implementation of the system can reduce time-consuming, face-to-face interactions between involved agencies by having key data streams in one location, streamlining communications with the purpose of improving race logistics, as well as medical preparedness and response. Hanken T , Young S , Smilowitz K , Chiampas G , Waskowski D . Developing a data visualization system for the Bank of America Chicago Marathon (Chicago, Illinois USA). Prehosp Disaster Med. 2016;31(5):572-577.

  15. Pile Load Capacity – Calculation Methods

    Directory of Open Access Journals (Sweden)

    Wrana Bogumił

    2015-12-01

    Full Text Available The article is a review of the current problems of the foundation pile capacity calculations. The article considers the main principles of pile capacity calculations presented in Eurocode 7 and other methods with adequate explanations. Two main methods are presented: α – method used to calculate the short-term load capacity of piles in cohesive soils and β – method used to calculate the long-term load capacity of piles in both cohesive and cohesionless soils. Moreover, methods based on cone CPTu result are presented as well as the pile capacity problem based on static tests.

  16. Static pile load tests on driven piles in Intermediate-Geo Materials : research brief.

    Science.gov (United States)

    2017-02-01

    Research Objectives: : Investigate the use of modified standard penetration tests (MSPT) : Compare field results with predictions made by the WisDOT driving formula, PDA and CAPWAP : Improve prediction of pile lengths and pile capacities ...

  17. Test Exponential Pile

    Science.gov (United States)

    Fermi, Enrico

    The Patent contains an extremely detailed description of an atomic pile employing natural uranium as fissile material and graphite as moderator. It starts with the discussion of the theory of the intervening phenomena, in particular the evaluation of the reproduction or multiplication factor, K, that is the ratio of the number of fast neutrons produced in one generation by the fissions to the original number of fast neutrons, in a system of infinite size. The possibility of having a self-maintaining chain reaction in a system of finite size depends both on the facts that K is greater than unity and the overall size of the system is sufficiently large to minimize the percentage of neutrons escaping from the system. After the description of a possible realization of such a pile (with many detailed drawings), the various kinds of neutron losses in a pile are depicted. Particularly relevant is the reported "invention" of the exponential experiment: since theoretical calculations can determine whether or not a chain reaction will occur in a give system, but can be invalidated by uncertainties in the parameters of the problem, an experimental test of the pile is proposed, aimed at ascertaining if the pile under construction would be divergent (i.e. with a neutron multiplication factor K greater than 1) by making measurements on a smaller pile. The idea is to measure, by a detector containing an indium foil, the exponential decrease of the neutron density along the length of a column of uranium-graphite lattice, where a neutron source is placed near its base. Such an exponential decrease is greater or less than that expected due to leakage, according to whether the K factor is less or greater than 1, so that this experiment is able to test the criticality of the pile, its accuracy increasing with the size of the column. In order to perform this measure a mathematical description of the effect of neutron production, diffusion, and absorption on the neutron density in the

  18. Stress transfer from pile group in saturated and unsaturated soil using theoretical and experimental approaches

    Directory of Open Access Journals (Sweden)

    al-Omari Raid R.

    2017-01-01

    Full Text Available Piles are often used in groups, and the behavior of pile groups under the applied loads is generally different from that of single pile due to the interaction of neighboring piles, therefore, one of the main objectives of this paper is to investigate the influence of pile group (bearing capacity, load transfer sharing for pile shaft and tip in comparison to that of single piles. Determination of the influence of load transfer from the pile group to the surrounding soil and the mechanism of this transfer with increasing the load increment on the tip and pile shaft for the soil in saturated and unsaturated state (when there is a negative pore water pressure. Different basic properties are used that is (S = 90%, γd = 15 kN / m3, S = 90%, γd = 17 kN / m3 and S = 60%, γd =15 kN / m3. Seven model piles were tested, these was: single pile (compression and pull out test, 2×1, 3×1, 2×2, 3×2 and 3×3 group. The stress was measured with 5 cm diameter soil pressure transducer positioned at a depth of 5 cm below the pile tip for all pile groups. The measured stresses below the pile tip using a soil pressure transducer positioned at a depth of 0.25L (where L is the pile length below the pile tip are compared with those calculated using theoretical and conventional approaches. These methods are: the conventional 2V:1H method and the method used the theory of elasticity. The results showed that the method of measuring the soil stresses with soil pressure transducer adopted in this study, gives in general, good results of stress transfer compared with the results obtained from the theoretical and conventional approaches.

  19. Test Pile Reactivity Loss Due to Trichloroethylene

    International Nuclear Information System (INIS)

    Plumlee, K.E.

    2001-01-01

    The presence of trichloroethylene in the test pile caused a continual decrease in pile reactivity. A system which removed, purified, and returned 12,000 cfh helium to the pile has held contamination to a negligible level and has permitted normal pile operation

  20. Fugitive emission rates assessment of PM2.5 and PM10 from open storage piles in China

    Science.gov (United States)

    Cao, Yiqi; Liu, Tao; He, Jiao

    2018-03-01

    An assessment of the fugitive emission rates of PM2.5 and PM10 from an open static coal and mine storage piles. The experiment was conducted at a large union steel enterprises in the East China region to effectively control the fugitive particulate emissions pollution on daily work and extreme weather conditions. Wind tunnel experiments conducted on the surface of static storage piles, and it generated specific fugitive emission rates (SERs) at ground level of between ca.10-1 and ca.102 (mg/m2·s) for PM2.5 and between ca.101 and ca.103 (mg/m2·s) for PM10 under the u*(wind velocity) between ca.3.0 (m/s) and 10.0 (m/s). Research results show that SERs of different materials differ a lot. Material particulate that has lower surface moisture content generate higher SER and coal material generate higher SER than mine material. For material storage piles with good water infiltrating properties, aspersion is a very effective measure for control fugitive particulate emission.

  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. Corrosion of magnesium and some magnesium alloys in gas cooled reactors; Corrosion du magnesium et de certains de ses alliages dans les piles refroidies par gaz

    Energy Technology Data Exchange (ETDEWEB)

    Caillat, R; Darras, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The results of corrosion tests on magnesium and some magnesium alloys (Mg-Zr and Mg-Zr-Zn) in moist air (like G1 reactor) and in CO{sub 2}: (like G2, G3, EDF1 reactors) are reported. The maximum temperature for exposure of magnesium to moist air without any risk of corrosion is 350 deg. C. Indeed, the oxidation rate follows a linear law above 350 deg. C although it reaches a constant level and keeps on very low under 350 deg. C. However, as far as corrosion is concerned this temperature limit can be raised up to 500 deg. C if moist air is very slightly charged with fluorinated compounds. Under pressure of CO{sub 2}, these three materials oxidate much more slowly even if 500 deg. C is reached. The higher is the temperature, the higher is the constant level of the weight increase and the quicker is reached this one. However, Mg-Zr alloy behaves quite better than pure magnesium and especially than Mg-Zr-Zn alloy. (author)Fren. [French] On expose essentiellement les resultats d'etudes sur la corrosion du magnesium et de certains de ses alliages (Mg-Zr et Mg-Zr-Zn) dans l'air humide (cas de la pile G1) et dans le gaz carbonique (cas des piles G2, G3, EDF1, etc...). La temperature limite d'exposition du magnesium dans l'air humide sans risque de corrosion se situe a 350 deg. C; en effet l'oxydation a un caractere lineaire au-dessus de cette temperature, alors qu'elle atteint un palier et reste tres limitee au-dessous de 350 deg. C. Du point de vue de la corrosion, cette temperature limite d'emploi peut cependant etre elevee jusqu'a 500 deg. C si l'on introduit dans l'air humide de tres faibles teneurs de composes fluores. Dans le gaz carbonique sous pression, l'oxydation est beaucoup plus faible, meme jusqu'a 50g. C pour les trois materiaux: l'augmentation de poids atteint un palier d'autant plus eleve et ceci d'autant plus rapidement que la temperature est elle-meme plus elevee. Cependant, l'alliage Mg-Zr se comporte nettement mieux que le magnesium pur et surtout que l

  4. Corrosion of magnesium and some magnesium alloys in gas cooled reactors; Corrosion du magnesium et de certains de ses alliages dans les piles refroidies par gaz

    Energy Technology Data Exchange (ETDEWEB)

    Caillat, R.; Darras, R. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The results of corrosion tests on magnesium and some magnesium alloys (Mg-Zr and Mg-Zr-Zn) in moist air (like G1 reactor) and in CO{sub 2}: (like G2, G3, EDF1 reactors) are reported. The maximum temperature for exposure of magnesium to moist air without any risk of corrosion is 350 deg. C. Indeed, the oxidation rate follows a linear law above 350 deg. C although it reaches a constant level and keeps on very low under 350 deg. C. However, as far as corrosion is concerned this temperature limit can be raised up to 500 deg. C if moist air is very slightly charged with fluorinated compounds. Under pressure of CO{sub 2}, these three materials oxidate much more slowly even if 500 deg. C is reached. The higher is the temperature, the higher is the constant level of the weight increase and the quicker is reached this one. However, Mg-Zr alloy behaves quite better than pure magnesium and especially than Mg-Zr-Zn alloy. (author)Fren. [French] On expose essentiellement les resultats d'etudes sur la corrosion du magnesium et de certains de ses alliages (Mg-Zr et Mg-Zr-Zn) dans l'air humide (cas de la pile G1) et dans le gaz carbonique (cas des piles G2, G3, EDF1, etc...). La temperature limite d'exposition du magnesium dans l'air humide sans risque de corrosion se situe a 350 deg. C; en effet l'oxydation a un caractere lineaire au-dessus de cette temperature, alors qu'elle atteint un palier et reste tres limitee au-dessous de 350 deg. C. Du point de vue de la corrosion, cette temperature limite d'emploi peut cependant etre elevee jusqu'a 500 deg. C si l'on introduit dans l'air humide de tres faibles teneurs de composes fluores. Dans le gaz carbonique sous pression, l'oxydation est beaucoup plus faible, meme jusqu'a 50g. C pour les trois materiaux: l'augmentation de poids atteint un palier d'autant plus eleve et ceci d'autant plus rapidement que la temperature est elle-meme plus elevee. Cependant, l

  5. 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

  6. Numerical Study on Dynamic Response of Pile Group Foundation of Geotechnical Centrifuge

    Directory of Open Access Journals (Sweden)

    Mao Quansheng

    2015-01-01

    Full Text Available Based on National Engineering Laboratory for Harbor Engineering Structure-Geotechnical Centrifuge Laboratory construction project, the dynamical response of piles foundation under horizontal-rocking vibration was analyzed by using finite element software Abaqus, and the displacement and stress characteristics of piles were discussed with soil between the piles reinforced by high pressure jet piles. The result indicates that in the operation of the centrifuge, foundation changes of vertical load of center pile are very small; the vertical displacement of the pile head is increasing, the vertical displacement of the pile head is no longer changed until the vibration time reaches 3 times period,; the horizontal load of piles varies with sinusoidal, the horizontal displacement amplitude is increasing, , and the vibration amplitude reaches to fixed value at 2 times vibration period.

  7. 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.)

  8. 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

  9. Simulation of bearing capacity of bored piles

    Directory of Open Access Journals (Sweden)

    Majeed Ahmed

    2018-01-01

    Full Text Available This study focuses on how one can possibly predict the ultimate load for the piles that did not reach failure. This challenge was acquired through Chin- Konder method by which, the estimated settlement that correspond to failure load is well defined. Hence, this research aims to make a comparative study between the results of pile load tests carried out in Al-Basrah sewage treatment plant project, and those results induced from the numerical analysis in term of ultimate pile capacity. Consequently, it may give a clear idea on the ability of numerical simulation in getting close to the actual behavior of piles. In the current study, a numerical study using Plaxis 3D Foundation program has been performed on bored piles by the assistance of site investigations of soil. Mohr- Coulomb and linear elastic models were adopted in the simulation for soil and pile respectively. Ten bored piles were used in this analysis under different values of loading. The diameter and length of pile are 0.6m and 24m respectively. The test results indicate that, an excellent agreement has been found as a response of pile capacity between the field and numerical studies. Also, ideal load- settlement curves were created using Chin- Konder method to predict the failure load of bored piles. Also, the results have demonstrated that, the pile capacity obtained from the simulation process is larger about 51% than that design load estimated before the design of piles. This may present a priority to use the finite element method to be accounted as an effective approach in the primary analysis.

  10. Pile Model Tests Using Strain Gauge Technology

    Science.gov (United States)

    Krasiński, Adam; Kusio, Tomasz

    2015-09-01

    Ordinary pile bearing capacity tests are usually carried out to determine the relationship between load and displacement of pile head. The measurement system required in such tests consists of force transducer and three or four displacement gauges. The whole system is installed at the pile head above the ground level. This approach, however, does not give us complete information about the pile-soil interaction. We can only determine the total bearing capacity of the pile, without the knowledge of its distribution into the shaft and base resistances. Much more information can be obtained by carrying out a test of instrumented pile equipped with a system for measuring the distribution of axial force along its core. In the case of pile model tests the use of such measurement is difficult due to small scale of the model. To find a suitable solution for axial force measurement, which could be applied to small scale model piles, we had to take into account the following requirements: - a linear and stable relationship between measured and physical values, - the force measurement accuracy of about 0.1 kN, - the range of measured forces up to 30 kN, - resistance of measuring gauges against aggressive counteraction of concrete mortar and against moisture, - insensitivity to pile bending, - economical factor. These requirements can be fulfilled by strain gauge sensors if an appropriate methodology is used for test preparation (Hoffmann [1]). In this paper, we focus on some aspects of the application of strain gauge sensors for model pile tests. The efficiency of the method is proved on the examples of static load tests carried out on SDP model piles acting as single piles and in a group.

  11. Safety precautions in atomic pile control (1962)

    International Nuclear Information System (INIS)

    Furet, J.

    1962-01-01

    We have been led to study the problem of safety in atomic pile control as a result of our participation on the one hand in the planning of C.E.A. atomic piles, and on the other hand in the pile safety sub omission considering atomic pile safety of operational or planned C.E.A. piles. We have thus had to consider the wishes occurring in piles during their operation and also their behaviour in the dynamic state The present work deals mainly with the importance of intrinsic safety devices, with the influence of reactivity variations on the power fluctuations during accidental operation, and with the development of robust and reliable safety appliances. The starting p accident has been especially studied both for low-flux piles where a compromise is necessary between the response time of the safety appliances and the statistical fluctuations and for high lux piles where xenon poisoning has an effect on the lower limit of the velocity of reactivity liberation. The desirability has been stressed of automation as a safety factor in atomic pile control. The details required for an understanding of the diagrams of the apparatus are given. (author) [fr

  12. Piles of dislocation loops in real crystals

    International Nuclear Information System (INIS)

    Dubinko, V.I.; Turkin, A.A.; Yanovskij, V.V.

    1985-01-01

    Behaviour of piles of dislocation loops in crystals was studied in order to define metal swelling under irradiation. Energy of pile interaction with point defects and intrinsic pile energy are studied in the framework of the linear elasticity theory. Preference of dislocation pile calculated in the paper decreases with radiation dose hence, material swelling rate also decreases. Creation of conditions, which assume an existence of piles of dislocation loops being stable under irradiation, is of particular interest

  13. Monitoring moisture content, temperature, and humidity in whole-tree pine chip piles

    Science.gov (United States)

    John Klepac; Dana Mitchell; Jason Thompson

    2015-01-01

    Two whole-tree chip piles were monitored for moisture content, temperature, and relative humidity from October 8th, 2010 to March 16th, 2011 at a location in south Alabama. Initial moisture content samples were collected immediately after chips were delivered to the study location on October 8th for Pile 1 and October 22nd for Pile 2. During pile construction, Lascar...

  14. Physical Modelling of Large Diameter Piles in Coarse-Grained Soil

    DEFF Research Database (Denmark)

    Brødbæk, K. T.; Augustesen, Anders Hust; Møller, M.

    2011-01-01

    of increasing the effective stresses. The test setup is thoroughly described in the paper. Two non-slender aluminium pipe piles subjected to lateral loads have been tested in the laboratory. The piles are heavily instrumented with strain gauges in order to obtain p-y curves, displacement and bending moment......Monopiles are an often-used foundation concept for offshore wind turbine converters. These piles are highly subjected to lateral loads and overturning bending moments due to wind and wave forces. To ensure enough stiffness of the foundation and an acceptable pile-head deflection, monopiles...... with diameters of 4 to 6 m are typically employed. In current practice these piles are traditionally designed by means of the p-y curve method although the method is developed and verified for slender piles in sand with diameters up to approximately 2 m. One of the limitations of the p-y curves used in current...

  15. 3D Centrifuge Modeling of the Effect of Twin Tunneling to an Existing Pile Group

    Directory of Open Access Journals (Sweden)

    M. A. Soomr

    2017-10-01

    Full Text Available In densely built urban areas, it is inevitable that tunnels will be constructed near existing pile groups. The bearing capacity of a pile group depends on shear stress along the soil-pile interface and normal stress underneath the pile toe while the two would be adversely affected by the unloading process of tunneling. Although extensive studies have been conducted to investigate the effects of tunnel construction on existing single piles, the influence of twin tunnel advancement on an existing pile group is merely reported in the literature. In this study, a series of three-dimensional centrifuge tests were carried out to investigate the response of an existing pile group under working load subjected to twin tunneling at various locations in dry Toyoura sand. In each twin tunneling test, the first tunnel is constructed near the mid-depth of the pile shaft, while the second tunnel is subsequently constructed either next to, below or right underneath the pile toe (Tests G_ST, G_SB and G_SU, respectively. Among the three tests, the 2nd tunnel excavated near the pile toe (Test G_ST results in the smallest settlement but the largest transverse tilting (0.2% of pile group. Significant bending moment was induced at the pile head (1.4 times of its bending moment capacity due to the 2nd tunnel T. On the contrary, tunneling right underneath the toe of pile (i.e., Test G_SU results in the smallest tilting but largest settlement of the pile group (4.6% of pile diameter and incremental mobilisation of shaft resistance (13%. Due to stress release by the twin tunneling, the axial force taken by the front piles close to tunnels was reduced and partially transferred to the rear piles. This load transfer can increase the axial force in rear piles by 24%.

  16. 30 CFR 77.214 - Refuse piles; general.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Refuse piles; general. 77.214 Section 77.214... Installations § 77.214 Refuse piles; general. (a) Refuse piles constructed on or after July 1, 1971, shall be..., tipples, or other surface installations and such piles shall not be located over abandoned openings or...

  17. 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

  18. 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)

  19. A computational model of pile vertical vibration in saturated soil based on the radial disturbed zone of pile driving

    International Nuclear Information System (INIS)

    Li Qiang; Shi Qian; Wang Kuihua

    2010-01-01

    In this study, a simplified computational model of pile vertical vibration was developed. The model was based on the inhomogeneous radial disturbed zone of soil in the vicinity of a pile disturbed by pile driving. The model contained two regions: the disturbed zone, which was located in the immediate vicinity of the pile, and the undisturbed region, external to the disturbed zone. In the model, excess pore pressure in the disturbed zone caused by pile driving was assumed to follow a logarithmic distribution. The relationships of stress and strain in the disturbed zone were based on the principle of effective stress under plain strain conditions. The external zone was governed by the poroelastic theory proposed by Biot. With the use of a variable separation method, an analytical solution in the frequency domain was obtained. Furthermore, a semi-analytical solution was attained by employing a numerical convolution method. Numerical results from the frequency and time domain indicated that the equivalent radius of the disturbed zone and the ratio of excess pore pressure had a significant effect on pile dynamic response. However, actual interactions between pile and soil will be weaker due to the presence of the radial disturbed zone, which is caused by pile driving. Consequently, the ideal undisturbed model overestimates the interaction between pile and soil; however, the proposed model reflects the interaction of pile and soil better than the perfect contact model. Numerical results indicate that the model can account for the time effect of pile dynamic tests.

  20. Some Remarks on Foundation Pile Testing Procedures

    Science.gov (United States)

    Rybak, Jarosław

    2017-10-01

    This work presents the review of pile capacity testing techniques. In an overview, the key points in pile designing are: determination of the appropriate computational schemes, reliable data on loads and the properties of structural materials (in particular, of the soil mass, which is marked by the greatest variability). The procedure of constructing a pile foundation should include: carrying out soil tests in the scope that ensures safe designing, selecting a piling technology that is relevant both to geotechnical conditions and expected loads, drafting a piling design together with the design of load tests, setting up a testing station for further load tests, static and/or dynamic tests of pile load capacity, preceded by supplementary soil tests when the conditions of test pile installation fail to comply with the design assumptions or when the pile length exceeds the depth of the previously investigated soil, making documentation of load capacity tests (with an additional correction of the piling design), the actual piling (ongoing analysis of pile driving logs and, if necessary, testing the piles’ integrity), drawing up the as-built documentation. Unfortunately, the design is corrected after the load test have been conducted only if the piles fail to show the designed bearing capacity. The designer is then obliged to revise the design assumptions on the basis of tests results. If the test results account for the a greater bearing capacity than necessary and it would be recommendable to limit the extent of the planned (i.e. set out in the contract) piling works, usually neither the contractor nor the designer, nor even the Construction Site Supervisor, acting for the benefit of the Investor, are willing to take on the responsibility for reducing the scope of the piling works. The necessity of conducting additional control tests before and during the implementation of the construction project is often treated by the investors as an attempt at extorting extra

  1. 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

  2. Literature review Quasi-static and Dynamic pile load tests : Primarily report on non-static pile load tests

    NARCIS (Netherlands)

    Huy, N.Q.

    2010-01-01

    Pile testing, which plays an importance role in the field of deep foundation design, is performed by static and non-static methods to provide information about the following issues: (Poulos, 1998) - The ultimate capacity of a single pile. - The load-displacement behavior of a pile. - The performance

  3. 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

  4. Multisignal detecting system of pile integrity testing

    Science.gov (United States)

    Liu, Zuting; Luo, Ying; Yu, Shihai

    2002-05-01

    The low strain reflection wave method plays a principal rule in the integrating detection of base piles. However, there are some deficiencies with this method. For example, there is a blind area of detection on top of the tested pile; it is difficult to recognize the defects at deep-seated parts of the pile; there is still the planar of 3D domino effect, etc. It is very difficult to solve these problems only with the single-transducer pile integrity testing system. A new multi-signal piles integrity testing system is proposed in this paper, which is able to impulse and collect signals on multiple points on top of the pile. By using the multiple superposition data processing method, the detecting system can effectively restrain the interference and elevate the precision and SNR of pile integrity testing. The system can also be applied to the evaluation of engineering structure health.

  5. Test Setup for Axially Loaded Piles in Sand

    DEFF Research Database (Denmark)

    Thomassen, Kristina

    The test setup for testing axially static and cyclic loaded piles in sand is described in the following. The purpose for the tests is to examine the tensile capacity of axially loaded piles in dense fully saturated sand. The pile dimensions are chosen to resemble full scale dimension of piles used...... in offshore pile foundations today....

  6. The controlled of the materials by the method of oscillation at the reactor core of Chatillon; Le controle des materiaux par la methode d'oscillation a la pile de Chatillon

    Energy Technology Data Exchange (ETDEWEB)

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

    1955-07-01

    Nuclear controls has for aim to determine the validity of materials intended to be used for the construction of the reactor core. The cross-section of capture of these materials has to be measured while comparing them either to a standard of the same material, either to an element of cross-section supposed known. We studied the disruption of the working of the reactor generated by the periodic introduction of a sample of the studied material. This method is based on the measure of the phase angle of the signal provided by the ionization chamber. This signal results from the composition of a local signal and an aggregate signal due to the effects of diffusion and capture. This method permits the comparison of the capture of 2 samples very dispersive and few capturing as the graphite, the beryllium, the beryllium oxide, with a good precision. It permits to determine the cross-section of capture of elements as magnesium or aluminum. (M.B.) [French] Le controle nucleaire a pour but de determiner la valeur des materiaux destines a etre utilises pour la construction des piles. II s'agit de mesurer la section efficace de capture de ces materiaux en les comparant soit a un echantillon etalon du meme materiau, soit a un element de section efficace supposee connue. On etudie la perturbation du fonctionnement de la pile engendree par l'introduction periodique d'un echantillon du materiau a etudier. Cette methode est basee sur la mesure de l'angle de phase du signal fourni par la chambre d'ionisation. Ce signal resulte de la composition d'un signal local et d'un signal global dus aux effets de diffusion et de capture. Cette methode permet la comparaison de la capture de 2 echantillons de corps tres diffusants et peu capturants comme le graphite,le beryllium, l'oxyde de beryllium, avec une bonne precision. Elle permet par ailleurs de determiner la section efficace de capture de corps tels que le magnesium ou l'aluminium. (M.B.)

  7. Energy piles. A fundamental energy pile; Energiepfaehle. Eine fundamentale Energiequelle

    Energy Technology Data Exchange (ETDEWEB)

    Kaiser, Holger; Beldermann, Nico [GF-Tec GmbH, Roedermark (Germany)

    2013-03-01

    The Maintower, the new airport in Berlin/Brandenburg, a lot of Ikea buildings, and also small office buildings or residential buildings may exchange energy with the underground by means of pile fundaments. At the correct planning and execution, energy piles are low-cost geothermal power plants which sustainable generate heating and cooling for the buildings standing on them. Even more energy can be generated safely under compliance with the groundwater protection by means of a new development of the material and the transfer.

  8. Pile Design Based on Cone Penetration Test Results

    OpenAIRE

    Salgado, Rodrigo; Lee, Junhwan

    1999-01-01

    The bearing capacity of piles consists of both base resistance and side resistance. The side resistance of piles is in most cases fully mobilized well before the maximum base resistance is reached. As the side resistance is mobilized early in the loading process, the determination of pile base resistance is a key element of pile design. Static cone penetration is well related to the pile loading process, since it is performed quasi-statically and resembles a scaled-down pile load test. In ord...

  9. 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.

  10. Tackling Absenteeism in Chicago

    Science.gov (United States)

    Allensworth, Elaine; Evans, Shayne

    2016-01-01

    Incisive research from the University of Chicago Urban Education Institute on the prevalence and consequences of absenteeism in Chicago schools has highlighted the dramatic effects of even moderate amounts of absences on grades, graduation rates, and student success in college. These insights spurred not only an ambitious 98% attendance goal on…

  11. THE NEW STRUCTURE OF A PLATE-PILE FOUNDATION

    Directory of Open Access Journals (Sweden)

    SAMORODOV О. V.

    2016-01-01

    Full Text Available Raising of problem. In the construction of high-rise buildings with significant loads on foundations and bedding at the base of not rocky soils tend to use the most common pile foundation to reduce the deformation and correspond to regulations [1] on the maximum permissible deformation. Monitoring of the stress-strain state (SSS pile foundations constructed buildings shows the existing reserves of bearing capacity on the one hand - soil bases of the second group of limit states, on the other - the foundation of the first group of limit states by regulating the SSS foundation during construction and exploitation. Therefore, are increasingly using more progressive structure of foundations consisting of piles and of plate, as well as methods for their construction [2 - 10], however, in their design there are a number of disadvantages associated with the ambiguity of the application of methods for the calculation of such structures, which allows to do only partial conclusions and recommendations. Purpose. Is proposing the new structure of a plate-pile foundation, which overcomes the drawbacks of the existing design solutions and methods of consruction their, as well as is proposing an engineering method of a determination of basic parameters. Conclusion. Is proposes the new structure of a plate-pile foundation and the method of a determination of basic parameters his in the design of a soil base to the maximum permissible deformation Su buildings. Efficiency of application this type plate-pile foundation obtained by rational distribution resistance between a plate and piles, when under load from the building to the first work fully incorporated plate that allows maximum deformed for plate, and then the piles - due to of the hinge connection with the plate. Thus, depending on the maximum permissible deformation for buildings resistance of plate part of a full load of more than 50%, that significantly reduces the consumption of concrete.

  12. Test Procedure for Axially Loaded Piles in Sand

    DEFF Research Database (Denmark)

    Thomassen, Kristina

    The test procedure described in the following is used when examining the effects of static or cyclic loading on the skin friction of an axially loaded pile in dense sand. The pile specimen is only loaded in tension to avoid any contribution from the base resistance. The pile dimensions are chosen...... to resemble full scale dimension of piles used in offshore pile foundations today. In this report is given a detailed description of the soil preparation and pile installation procedures as well data acquisition methods....

  13. 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

  14. 78 FR 27304 - Safety Zone; Melrose Pyrotechnics Fireworks Display; Chicago Harbor, Chicago, IL

    Science.gov (United States)

    2013-05-10

    ... environmental risk to health or risk to safety that may disproportionately affect children. 11. Indian Tribal... Pier in Chicago Harbor, Chicago, IL. The Captain of the Port, Lake Michigan, has determined that these fireworks displays will pose a significant risk to public safety and property. Such hazards include falling...

  15. 75 FR 45478 - Safety Zone; Transformers 3 Movie Filming, Chicago River, Chicago, IL

    Science.gov (United States)

    2010-08-03

    ...-AA00 Safety Zone; Transformers 3 Movie Filming, Chicago River, Chicago, IL AGENCY: Coast Guard, DHS... the different types of stunts that will be performed during the filming of this movie. DATES... our analyses based on 13 of these statutes or executive orders. Regulatory Planning and Review This...

  16. Failed fuel diagnosis during WWER reactor operation using the RTOP-CA code

    International Nuclear Information System (INIS)

    Likhanskii, V.; Afanasieva, E.; Sorokin, A.; Evdokimov, I.; Kanukova, V.; Khromov, A.

    2006-01-01

    The mechanistic code RTOP-CA is developed for objectives of failed fuel diagnosis during WWER reactor operation. The RTOP-CA code enables to solve a direct problem: modelling the failed fuel behavior and prediction of primary coolant activity if characteristics of failures in the reactor core are known. Results of verification of the RTOP-CA code are presented. Separate physical models were verified on small-scale in-pile and out-of-pile experiments. Integral verification cases included data obtained at research reactors and at nuclear power plants. The RTOP-CA code is used for development of a neural-network approach to the inverse problem: detection of failure characteristics on the base of data on primary coolant activity during reactor operation. Preliminary results of application of the neural-network approach for evaluation of fuel failure characteristics are presented. (authors)

  17. 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

  18. 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

  19. Development of out-of-pile version of instrumented irradiation capsule for determination of online creep deformation

    International Nuclear Information System (INIS)

    Venkatesu, Sadu; Saxena, Rajesh; Chaurasia, P.K.; Muthuganesh, M.; Murugan, S.; Venugopal, S.

    2016-01-01

    Materials used for fuel cladding and structural components in fast reactors can undergo significant dimensional and physical changes due to exposure to high energy neutrons. At high temperatures in nuclear environment, material undergoes considerable deformation due to thermal and irradiation creep. Diametral increase of fuel pin due to thermal and irradiation creep, apart from irradiation swelling, reduces the coolant flow area around the fuel pins affecting the effective removal of heat generated in the fuel pins. The changes due to creep can be determined by two types of material irradiation tests in reactor. The first type includes non-instrumented irradiation tests with specimen dimensional evaluations carried out in post-irradiation examinations. The second type includes instrumented irradiation tests with online monitoring and/or controlling of test conditions and real time measurement of changes in dimensions of the specimen. During instrumented irradiation tests, parameters such as specimen temperature, the load exerted on the specimen, specimen elongation, etc. can be monitored and/or controlled using suitable components such as linear variable differential transformers (LVDTs), bellows, thermocouples, etc. Instrumented irradiation experiments in reactors are relatively complex in design but can provide full information on the experimental parameters. Such benefits provide motivation for development of instrumented irradiation capsule to measure creep behavior online during in-pile instrumented irradiation tests. Out-of-pile version of the instrumented irradiation capsule for determination of online creep deformation has been developed and tested in the furnace by raising the temperature gradually up to 330 °C. This paper discusses the details of the design, assembly of experimental set up and experimental results of the out-of-pile version of instrumented capsule developed in our laboratory for determination of online creep deformation. (author)

  20. Response of shallow geothermal energy pile from laboratory model tests

    Science.gov (United States)

    Marto, A.; Amaludin, A.

    2015-09-01

    occurred was greater than the limiting value when the pile was loaded with thermo-axial loads of 40°C and 200 N. It is therefore recommended that the global factor of safety to be applied for energy pile installed in firm soil should be more than 2.3 to prevent any hazard to occur in the future, should the pile also be subjected to thermal load of 40°C or greater.

  1. Preliminary design of steam reformer in out-pile demonstration test facility for HTTR heat utilization system

    Energy Technology Data Exchange (ETDEWEB)

    Haga, Katsuhiro; Hino, Ryutaro; Inagaki, Yosiyuki; Hata, Kazuhiko; Aita, Hideki; Sekita, Kenji; Nishihara, Tetsuo; Sudo, Yukio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Yamada, Seiya

    1996-11-01

    One of the key objectives of HTTR is to demonstrate effectiveness of high-temperature nuclear heat utilization system. Prior to connecting a heat utilization system to HTTR, an out-pile demonstration test is indispensable for the development of experimental apparatuses, operational control and safety technology, and verification of the analysis code of safety assessment. For the first heat utilization system of HTTR, design of the hydrogen production system by steam reforming is going on. We have proposed the out-pile demonstration test plan of the heat utilization system and conducted preliminary design of the test facility. In this report, design of the steam reformer, which is the principal component of the test facility, is described. In the course of the design, two types of reformers are considered. The one reformer contains three reactor tubes and the other contains one reactor tube to reduce the construction cost of the test facility. We have selected the steam reformer operational conditions and structural specifications by analyzing the steam reforming characteristics and component structural strength for each type of reformer. (author)

  2. 75 FR 41760 - Safety Zone; Transformers 3 Movie Filming, Chicago River, Chicago, IL

    Science.gov (United States)

    2010-07-19

    ...-AA00 Safety Zone; Transformers 3 Movie Filming, Chicago River, Chicago, IL AGENCY: Coast Guard, DHS... that will be performed during the filming of this movie. DATES: Effective Date: this rule is effective.... Regulatory Planning and Review This rule is not a significant regulatory action under section 3(f) of...

  3. Efficiency of the Shut-Down and Safety Equipment and the Kinetic Characteristics of the G2 and G3 Reactors; Efficacite des dispositifs de secours et de securite et caracteristiques cinetiques des piles G2 et G3; Ehffektivnost' sistem avarijnoj zashchity reaktorov G.2 i G.3 i kineticheskie kharakteristiki ehtikh sistem; Caracteristicas cineticas y eficacia de los dispositivos de auxilio y de seguridad de los reactores G2 y G3

    Energy Technology Data Exchange (ETDEWEB)

    Henri, C.; Plisson, J.; Teste duBailler, A. [Centre d' Etudes Nucleaires de Saclay (France)

    1963-10-15

    The experience gained in several years of operating the G2 and G3 reactors confirms that natural uranium-graphite-gas reactors are extremely safe. The built-in shut-down and safety mechanisms which minimize operational incidents such as lack of power from the mains, blower failure, lack of water etc., together with accidents such as cladding bursts, local overheating, loss of coolant etc. are described and their operation explained by means of diagrams. The main points examined are as follows: (a) power distribution and controlability during accident conditions; (b) distribution of emergency water; and (c) the safety chain. The performance of the installations and the successive improvements incorporated in them are mentioned. The built-in safety characteristics of the reactors are shown by means of an experimental study of their behaviour in transient operation. These studies make it possible to check the validity of the calculation model. The machine calculation programmes can subsequently be used to study the consequences of possible accidents. Special attention is given to the depressurization accident, taking into account the performance of the safety device installed. (author) [French] L'experience acquise'au cours de plusieurs annees d'exploitation des piles G2 et G3 permet de confirmer le haut degre de securite du fonctionnement des piles de la filiere uranium naturel-graphitegaz. Les installations fixes de secours et de securite permettant de pallier, d'une part aux incidents d'exploitation tels que manque d'alimentation du reseau de distribution, arret de soufflage, manque d'alimentation en eau, etc., d'autre part, a des accidents tels que rupture de gaine, echauffements locaux, perte de fluide caloporteur, etc., sont decrites et leur fonctionnement explicite au moyen de schemas de principe. On examine principalement (a) la distribution ''puissance'' et ''controle'' des installations secourues, (b) la distribution d'eau secourue, et (c) la chaine de

  4. Tension Tests On Bored Piles In Sand

    DEFF Research Database (Denmark)

    Krabbenhøft, Sven; Clausen, Johan; Damkilde, Lars

    2006-01-01

    The lengths of the bored piles varied from 2 m to 6 m and all were of a diameter of 140 mm. The piles were tested to failure in tension and the load-displacement relations were recorded. The investigation has shown pronounced differences between the load bearing capacities obtained by different...... design methods. The methods proposed by Fleming et al. and Reese & O’Neill seem to produce the best match with the test results....

  5. Release of fission products and post-pile creep behaviour of irradiated fuel rods stored under dry conditions

    International Nuclear Information System (INIS)

    Kaspar, G.; Peehs, M.; Bokelmann, R.; Jorde, D.; Schoenfeld, H.; Haas, W.; Bleier, A.; Rutsch, F.

    1985-06-01

    The release of moisture and fission products (Kr-85, H-3 and I-129) under dry storage conditions has been examined on six fuel rods which have become defective in the reactor. During the examinations, inert conditions prevailed and limited air inlet was allowed temporarily. The storage temperature was 400 0 C. The residual moisture content of the fuel rods was approx. 5 g. At the beginning of the test, the total moisture content and 0,05% (max.) of the fission gas inventory were released. Under inert conditions, fission gas was not released during a prolonged period of time. Under oxidizing conditions, however, fission gas was released in the course of UO 2 oxidation. Post-pile creep of Zircaloy cladding tubes was measured at temperatures between 350 and 395 0 C and interval gauge pressures between 69 and 110 bar. The creep curves indicate that the irradiated cladding tube specimens still bear internal residual stresses which contribute through their relaxation to the post-pile creep. (orig.) [de

  6. ANALYSIS OF EXISTING SCHEMES AND THE OPTIMIZING SETTLEMENT CHOIS OF PILES WORK SCHEMES IN CLAY SOILS

    Directory of Open Access Journals (Sweden)

    BOLSHAKOV V. I.

    2016-09-01

    Full Text Available Summary. It were considered and analyzed the existing schemes of piles work in clay soils. 1. Leningrad scientific school, where the formation of pile bearing capacity use as the basis of the thixotropic clay soils hardening and radial soil pressing around the pile shaft during the piles driving with pile-driving equipment for the exploitation period. 2. Odessa scientific school, in which the uplift soil formation from the edge pile use as the basis of the pile bearing capacity during the piles driving, the formation of the pressed zones (platform in the piles edge plane, the gap formation around the pile shaft during its diving by ground pushed moving with the pile edge. 3. Preconditions of the pile bearing capacity formation of the pile by the thixotropic soil hardening in time and the radial soil pressing around the pile shaft can not give an answer to the following questions: 1 Why during the pile driving is formed the gap around the trunk of dived piles, when by condition there is a radial soil hardening around the trunk? 2 Why in the interpiled space is formed the lune (deflection, not the soil mass swelling (due to the radial hardening? 3 By what is formed the calculated soil resistance under the lower end (edge of the pile? which is about 10 times higher than the calculated soil resistance in the edge plane, according to the Building Code V.2.1-10. 2009? The justified answers on all these and other technical and technological matters give perquisites of the Odessa scientific school with additions and authors developments

  7. Large scale vibration tests on pile-group effects using blast-induced ground motion

    International Nuclear Information System (INIS)

    Katsuichirou Hijikata; Hideo Tanaka; Takayuki Hashimoto; Kazushige Fujiwara; Yuji Miyamoto; Osamu Kontani

    2005-01-01

    Extensive vibration tests have been performed on pile-supported structures at a large-scale mining site. Ground motions induced by large-scale blasting operations were used as excitation forces for vibration tests. The main objective of this research is to investigate the dynamic behavior of pile-supported structures, in particular, pile-group effects. Two test structures were constructed in an excavated 4 m deep pit. Their test-structures were exactly the same. One structure had 25 steel piles and the other had 4 piles. The test pit was backfilled with sand of appropriate grain size distributions to obtain good compaction, especially between the 25 piles. Accelerations were measured at the structures, in the test pit and in the adjacent free field, and pile strains were measured. Dynamic modal tests of the pile-supported structures and PS measurements of the test pit were performed before and after the vibration tests to detect changes in the natural frequencies of the soil-pile-structure systems and the soil stiffness. The vibration tests were performed six times with different levels of input motions. The maximum horizontal acceleration recorded at the adjacent ground surface varied from 57 cm/s 2 to 1,683 cm/s 2 according to the distances between the test site and the blast areas. (authors)

  8. Full Scale Model Test of Consolidation Acceleration on Soft Soil deposition with Combination of Timber Pile and PVD (Hybrid Pile)

    OpenAIRE

    Sandyutama, Y.; Samang, L.; Imran, A. M.; Harianto4, T.

    2015-01-01

    This research aims to analyze the effect of composite pile-PVD (hybrid pile) as the reinforcement in embankment on soft soil by the means of numerical simulation and Full-Scale Trial Embankment. The first phase cunducted by numerical analysis and obtained 6-8 meters hybrid pile length effective. Full-Scale trial embankment. was installed hybrid pile of 6 m and preloading of 4,50 height. Full-scale tests were performed to investigate the performances of Hybrid pile reinforcement. This research...

  9. Ultimate capacity of piles penetrating in weak soil layers

    Directory of Open Access Journals (Sweden)

    Al-Obaidi Ahmed

    2018-01-01

    Full Text Available A pile foundation is one of the most popular forms of deep foundations. They are routinely employed to transfer axial structure loads through the soft soil to stronger bearing strata. Piles generally used to increase the load carrying capacity of the foundation and reduce the settlement of the foundation. On the other hand, many cases in practice where piles pass through different layers of soil that contain weak layers located at different depths and extension, also some time cavities with a different shape, size, and depth are found. In this study, a total of 96 cases is considered and simulated in PLAXIS 2D program aiming to understand the influence of weak soil on the ultimate pile capacity. The piles embedded in the dense sand with a layer of weak soil at different extension and location. The cross section of the geometry used in this study was designed as an axisymmetric model with the 15-node element; the boundary condition recommended at least 5D in the horizontal direction, and (L+5D in the vertical direction where D and L are the diameter and length of pile, respectively. The soil is modeled as Mohr-Coulomb, with five input parameters and the behavior of pile material represented by the linear elastic model. The results of the above cases are compared with the results found in a pile embedded in dense soil without weak layers or cavities. The results indicated that the existence of weak soil layer within the surrounding soil around the pile decreases the ultimate capacity. Furthermore, it has been found that increase in the weak soil width (extension leads to reduction in the ultimate capacity of the pile. This phenomenon is applicable to all depth of weak soil. The influence of weak layer extension on the ultimate capacity is less when it is presentin the upper soil layers.

  10. Pulse pile-up IV

    International Nuclear Information System (INIS)

    Wilkinson, D.H.

    1991-05-01

    The study of pulse pile-up is extended from the case of unipolar pulses, for which ruin theory is an excellent approximation, to the case of bipolar pulses for which ruin theory is not applicable to the effect of the back-kicks in reducing the pile-up: an appropriate solution is presented. (Author) 3 refs., 11 figs

  11. 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

  12. The Chicago Landscape of Career and Technical Education. Feature on Research and Leadership. Vol. 2, No. 1

    Science.gov (United States)

    Owens, Devean R.; Welton, Anjalé D.

    2016-01-01

    In this brief, Owens and Welton provide an introductory overview of career and technical education (CTE) programs in both Chicago Public Schools (CPS) and the City Colleges of Chicago (CCC). They report that the State of Illinois, and the City of Chicago in particular, have some of the highest unemployment rates in the country for youth ages 16 to…

  13. 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

  14. Introduction of effective piles in a base structure

    Directory of Open Access Journals (Sweden)

    В.Б. Кашка

    2005-03-01

    Full Text Available  Design features of effective piles such as СВ and their advantages in use are considered at the device of the pile bases in comparison with widely widespread types of piles. From results of comparative tests of piles under static pressing loading in different earth conditions the tendency of redistribution of bearing (carrying ability between a trunk and expansions an effective pile such as СВ was determined on earth conditions.

  15. 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)

  16. Nuclear data needs for the analysis of generation and burn-up of actinide isotopes in nuclear reactors

    International Nuclear Information System (INIS)

    Kuesters, H.

    1980-04-01

    A reliable prediction of the in-pile and out-of-pile physics characteristics of nuclear fuel is one of the objectives of present-day reactor physics. The paper describes the main production paths of important actinides for light water and fast breeder reactors. The accuracy of recent nuclear data is examined by comparisons of theoretical predictions with the results from post-irradiation analysis of nuclear fuel from power reactors, and partly with results obtained in zero-power facilities. A world-wide comparison of nuclear data to be used in large fast power reactor burn-up and long term considerations is presented. The needs for further improvement of nuclear data are discussed. (orig.) [de

  17. Pile foundation of nuclear power plant structures

    International Nuclear Information System (INIS)

    Jurkiewicz, W.J.; Thomaz, E.; Rideg, P.; Girao, M.

    1978-01-01

    The subject of pile foundation used for nuclear power plant structures, considering the experience gained by the designers of the Angra Nuclear Power Plant, Units 2 and 3 in Brazil is dealt with. The general concept of the pile foundations, including types and execution of the piles, is described briefly. Then the two basic models, i.e. the static model and the dynamic one, used in the design are shown, and the pertinent design assumptions as related to the Angra project are mentioned. The criteria which established the loading capacity of the piles are discussed and the geological conditions of the Angra site are also explained briefly, justifying the reasons why pile foundations are necessary in this project. After that, the design procedures and particularly the tools - i.e. the computer programs - are described. It is noted that the relatively simple but always time consuming job of loading determination calculations can be computerized too, as it was done on this project through the computer program SEASA. The interesting aspects of soil/structure interaction, applicable to static models, are covered in detail, showing the theoretical base wich was used in the program PILMAT. Then the advantage resulting from computerizing of the job of pile reinforcement design are mentioned, describing briefly the jobs done by the two special programs PILDES and PILTAB. The point is stressed that the effort computerizing the structural design of this project was not so much due to the required accuracy of the calculations, but mainly due to the need to save on the design time, as to allow to perform the design task within the relatively tight time schedule. A conclusion can be drawn that design of pile foundations for nuclear power plant structures is a more complex task than the design of bearing type of foundation for the same structures, but that the task can be always made easier when the design process can be computerized. (Author)

  18. In-situ grouting of uranium-mill-tailings piles: an assessment

    International Nuclear Information System (INIS)

    Tamura, T.; Boegly, W.J. Jr.

    1983-05-01

    Passage in 1978 of the Uranium Mill Tailings Radiation Control Act (UMTRCA) initiated a program of remedial action for 22 existing mill tailings piles generated in the period 1940 to 1970 as part of the nation's defense and nuclear power programs. The presence of these piles poses potential health and environmental contamination concerns. Possible remedial actions proposed include multilayer covers over the piles to reduce water infiltration, reduce radon gas releases, and reduce airborne transport of tailings fines. In addition, suggested remedial actions include (1) the use of liners to prevent groundwater contamination by leachates from the piles and (2) chemical stabilization of the tailings to retain the radioactive and nonradioactive sources of contamination. Lining of the piles would normally be applicable only to piles that are to be moved from their present location such that the liner could be placed between the tailings and the groundwater. However, by using civil engineering techniques developed for grouting rocks and soils for strength and water control, it may be possible to produce an in situ liner for piles that are not to be relocated. The Department of Energy (DOE) Uranium Mill Tailings Remedial Action Project Office requested that ORNL assess the potential application of grouting as a remedial action. This report examines the types of grouts, the equipment available, and the costs, and assesses the possibility of applying grouting technology as a remedial action alternative for uranium mill tailings piles

  19. In-situ grouting of uranium-mill-tailings piles: an assessment

    Energy Technology Data Exchange (ETDEWEB)

    Tamura, T.; Boegly, W.J. Jr.

    1983-05-01

    Passage in 1978 of the Uranium Mill Tailings Radiation Control Act (UMTRCA) initiated a program of remedial action for 22 existing mill tailings piles generated in the period 1940 to 1970 as part of the nation's defense and nuclear power programs. The presence of these piles poses potential health and environmental contamination concerns. Possible remedial actions proposed include multilayer covers over the piles to reduce water infiltration, reduce radon gas releases, and reduce airborne transport of tailings fines. In addition, suggested remedial actions include (1) the use of liners to prevent groundwater contamination by leachates from the piles and (2) chemical stabilization of the tailings to retain the radioactive and nonradioactive sources of contamination. Lining of the piles would normally be applicable only to piles that are to be moved from their present location such that the liner could be placed between the tailings and the groundwater. However, by using civil engineering techniques developed for grouting rocks and soils for strength and water control, it may be possible to produce an in situ liner for piles that are not to be relocated. The Department of Energy (DOE) Uranium Mill Tailings Remedial Action Project Office requested that ORNL assess the potential application of grouting as a remedial action. This report examines the types of grouts, the equipment available, and the costs, and assesses the possibility of applying grouting technology as a remedial action alternative for uranium mill tailings piles.

  20. 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.

  1. Reliability of Estimation Pile Load Capacity Methods

    Directory of Open Access Journals (Sweden)

    Yudhi Lastiasih

    2014-04-01

    Full Text Available None of numerous previous methods for predicting pile capacity is known how accurate any of them are when compared with the actual ultimate capacity of piles tested to failure. The author’s of the present paper have conducted such an analysis, based on 130 data sets of field loading tests. Out of these 130 data sets, only 44 could be analysed, of which 15 were conducted until the piles actually reached failure. The pile prediction methods used were: Brinch Hansen’s method (1963, Chin’s method (1970, Decourt’s Extrapolation Method (1999, Mazurkiewicz’s method (1972, Van der Veen’s method (1953, and the Quadratic Hyperbolic Method proposed by Lastiasih et al. (2012. It was obtained that all the above methods were sufficiently reliable when applied to data from pile loading tests that loaded to reach failure. However, when applied to data from pile loading tests that loaded without reaching failure, the methods that yielded lower values for correction factor N are more recommended. Finally, the empirical method of Reese and O’Neill (1988 was found to be reliable enough to be used to estimate the Qult of a pile foundation based on soil data only.

  2. 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)

  3. 30 CFR 77.215-4 - Refuse piles; abandonment.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Refuse piles; abandonment. 77.215-4 Section 77... MINES Surface Installations § 77.215-4 Refuse piles; abandonment. When a refuse pile is to be abandoned... refuse pile shall be abandoned in accordance with a plan submitted by the operator and approved by the...

  4. Seismic behavior analysis of piled drums

    International Nuclear Information System (INIS)

    Aoki, H.; Kosaka, T.; Mizushina, T.; Shimizu, M.; Uji, S.; Tsuchiya, H.

    1987-01-01

    In general, low level radioactive waste is packed in drums and stored in a warehouse being piled vertically, or laid horizontally. To observe the behavior of piled drums during an earthquake, an experimental study was reported. The experimental study is limited by the vibrating platform capacity. To carry out these tests up to the supporting limit is not recommended, in view of the vibrating platform curing as well as the operators' security. It is very useful to develop the analytical method for simulating the behavior of the drums. In this report, a computer program of piled drum's dynamic motion is shown, and the analytical result is referred to the experimental result. From the result of experiment on piled drums, the sliding effect has been found to be very important for the stability of drum, and the rocking motion observed, showing a little acceleration is less than the static estimated value. Behavior of piled drums is a complex phenomena comprising of sliding, rocking and jumping

  5. Attenuation of pressure dips underneath piles of spherocylinders.

    Science.gov (United States)

    Zhao, Haiyang; An, Xizhong; Gou, Dazhao; Zhao, Bo; Yang, Runyu

    2018-05-30

    The discrete element method (DEM) was used to simulate the piling of rod-like (elongated sphero-cylindrical) particles, mainly focusing on the effect of particle shape on the structural and force properties of the piles. In this work, rod-like particles of different aspect ratios were discharged on a flat surface to form wedge-shaped piles. The surface properties of the piles were characterized in terms of angle of repose and stress at the bottom of the piles. The results showed that the rise of the angle of repose became slower with the increase of particle aspect ratio. The pressure dip underneath the piles reached the maximum when the particle aspect ratio was around 1.6, beyond which the pressure dip phenomenon became attenuated. Both the pressure dip and the shear stress dip were quantitatively examined. The structure and forces inside the piles were further analyzed to understand the change in pressure dip, indicating that "bridging" or "arching" structures within the piles were the cause of the pressure dip.

  6. Studies in Phebus reactor of fuel behaviour upon LOCA conditions

    International Nuclear Information System (INIS)

    Manin, A.; Del Negro, R.; Reocreux, M.

    1980-09-01

    The fuel behaviour upon LOCA conditions is studied in an in-pile loop, in Phebus reactor. This paper presents: a short description of Phebus reactor; the current program (adjusting the thermohydraulic conditions in order to get cladding failure); the program developments (consequences involved by cladding failure); the fuel test conditions determination [fr

  7. 30 CFR 817.83 - Coal mine waste: Refuse piles.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 3 2010-07-01 2010-07-01 false Coal mine waste: Refuse piles. 817.83 Section... ACTIVITIES § 817.83 Coal mine waste: Refuse piles. Refuse piles shall meet the requirements of § 817.81, the... drainage may not be diverted over the outslope of the refuse pile. Runoff from areas above the refuse pile...

  8. Nanoindentation-induced pile-up in hydrogenated amorphous silicon

    International Nuclear Information System (INIS)

    Pantchev, B; Danesh, P; Wiezorek, J; Schmidt, B

    2010-01-01

    Nanoindentation-induced material extrusion around the nanoindent (pile-up) leads to an overestimation of elastic modulus, E, and nanohardness, H, when the test results are evaluated using the Oliver and Pharr method. Factors affecting the pile-up during testing are residual stresses in film and ratio of film and substrate mechanical properties. Nanoindentation of hydrogenated amorphous silicon (a-Si:H) films has been carried out with the aim to study the effect of residual compressive stress on the pile-up in this material. To distinguish the contribution of compressive stress to the appearance of pile-up ion implantation has been used as a tool, which reduces the compressive stress in a-Si:H. Scanning probe microscope has been used for the imaging of the indent and evaluation of the pile-up. The values of E and H have been obtained from the experimental load-displacement curves using depth profiling with Berkovich tip, which has created negligible pile-up. A sharper cube corner tip has been used to study the pile-up. It has been established that pile-up is determined by the material plasticity, when the compressive stress is below 200 MPa. The contribution of mechanical stress to the pile-up is essential for the stress as high, as about 500 MPa.

  9. 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

  10. Operational reactor physics analysis codes (ORPAC)

    International Nuclear Information System (INIS)

    Kumar, Jainendra; Singh, K.P.; Singh, Kanchhi

    2007-07-01

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

  11. Grouting for Pile Foundation Improvement

    NARCIS (Netherlands)

    Van der Stoel, A.E.C.

    2001-01-01

    The aim of this research was to examine the use of grouting methods for pile foundation improvement, a generic term that is used here to define both foundation renovation (increasing the bearing capacity of a pile foundation that has insufficient bearing capacity) and foundation protection

  12. Pile foundation response in liquefiable soil deposit during strong earthquakes. ; Centrifugal test for pile foundation model and correlation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Y.; Miura, K. (Kajima Corp., Tokyo (Japan)); Scott, R.; Hushmand, B. (California Inst. of Technology, California, CA (United States))

    1992-09-30

    For the purpose of studying the pile foundation response in liquefiable soil deposit during earthquakes, a centrifugal loading system is employed which can reproduce the stress conditions of the soil in the actual ground, and earthquake wave vibration tests are performed in dry and saturated sand layers using a pile foundation model equipped with 4 piles. In addition, the result of the tests is analyzed by simulation using an analytic method for which effective stress is taken into consideration to investigate the effectiveness of this analytical model. It is clarified from the result of the experiments that the bending moment of the pile and the response characteristics of the foundation in the pile foundation response in saturated sand are greatly affected by the longer period of acceleration wave form of the ground and the increase in the ground displacement due to excess pore water pressure buildup. It is shown that the analytical model of the pile foundation/ground system is appropriate, and that this analytical method is effective in evaluating the seismic response of the pile foundation in nonlinear liquefiable soil. 23 refs., 21 figs., 3 tabs.

  13. Atomic pile Directorate, Department of Metallurgy, Departments of Technology, Department of Fuel Elements and Structures, Division of Study of Fuel Elements - Semi annual report on the 1968-10-1

    International Nuclear Information System (INIS)

    Arnaud, M.; Tortel, J.; Viallet, H.; Marinot, R.; Rulleau, A.; Lestiboudois, G.; Rousseau, G.; Faussat, A.; Ollier, H.; Truffert, J.; Ferrier, C.; Courcon, P.; Rendu, M.; Dieumegard, M.; Bret, A.

    1968-01-01

    This document gathers a set of reports of studies performed on nuclear fuel elements. The addressed topics are: creep behaviour of UMo and UMoAl tubes and pellets under the action of an external pressure (creep strength of tubes under external pressure, creep strength of pellets under external pressure, uncertainties on irradiation parameters in Pegase), problems related to centring devices (measurements and tests), irradiations of ring elements in power reactors, uranium/sheath metallurgical relationship for Bugey and influence of irradiation (cartridge behaviour in Pegase, long duration irradiation in power reactors, extrapolation in Bugey of results obtained in G2), theoretical study of kinetic oxidation phenomena in metal fuels, tests of leaking cartridges in EdF2, evolution of pressure in EL4 type irradiated fuel rods with ZrCu liners with respect to the conductivity integral, a focus on irradiations of Z0 type fuel elements in Pegase, cluster safety tests with uranium carbide in pile and out of pile, a review of studies performed on fuel elements with blowhole, and application of neutrography to fuel elements

  14. Analysis of Dynamic Stiffness of Bridge Cap-Pile System

    Directory of Open Access Journals (Sweden)

    Jinhui Chu

    2018-01-01

    Full Text Available In order to investigate the applicability of dynamic stiffness for bridge cap-pile system, a laboratory test was performed. A numerical model was also built for this type of system. The impact load was applied on the cap top and the dynamic stiffness was analysed. Then, the effect of the effective friction area between pile and soil was also considered. Finally, the dynamic stiffness relationship between the single pile and the cap-pile system was also compared. The results show that the dynamic stiffness is a sensitive index and can well reflect the static characteristics of the pile at the elastic stage. There is a significant positive correlation between the vertical dynamic stiffness index and bearing capacity of the cap-pile system in the similar formation environment. For the cap-pile system with four piles, the dynamic stiffness is about four times as large as the single pile between 10 and 20 Hz.

  15. Some conclusions about the concrete strength of the bored piles of Angra 2 nuclear power plant

    International Nuclear Information System (INIS)

    Thomaz, E.C.S.; Joia, L.A.

    1984-01-01

    The concrete of the bored piles of Angra 2 was submitted to a deep control, so more than five thousand core samples were analyzed to verify the quality of the concrete. Based on these samples and using statistics regression theory some conclusions could be done. It was analyzed the dependence of the concrete strength upon the depth of the pile. Also based on these samples some probability distribution functions that could simulate the concrete strength were studied applying the Kolmogorov - Smirnov fitness test. Finally, a method for evaluating a confidence interval of one of the probability function (Weibull distribution) was developed adopting the Monte Carlo simulation technique. (Author) [pt

  16. Integrity and As-built capacity of bored pile group

    International Nuclear Information System (INIS)

    Shaw, D.E.; Kissenpfennig, J.F.; Huemmer, M.R.

    1983-01-01

    This paper discusses the application of statistical methods to the reliability evaluation of cast-in-place concrete piles. The difficulties associated with pile construction can lead to larger uncertainties than would be associated with normal reinforced concrete structures both due to uncertainty in concrete quality and end bearing capacity. These uncertainties can be dealt with through the use of statistical methods. A statistical model of an individual pile is formulated along with a methodology for determining necessary statistical parameters from results of concrete batch tests, core strength tests and visual logs, sonic geophysical testing methods, and proof tests. Strength models for both static vertical and seismic horizontal loadings are discussed. The overall safety of a pile foundation is dependent upon the distribution of individual pile strength as well as the additional reliability due to the use of a large number of parallel load paths provided by a pile group foundation. The paper presents a mechanical model of global pile behavior which accounts for individual pile ductility along with the possibility of redistribution of loads from weaker to stronger piles. The use of the Monte Carlo method to determine the overall reliability of the pile foundation is discussed. Numerical results for both individual pile behavior as well as overall foundation behavior are presented. (orig.)

  17. Thermomechanical Behavior of Energy Pile Embedded in Sandy Soil

    Directory of Open Access Journals (Sweden)

    Xu Huang

    2018-01-01

    Full Text Available The traditional energy pile (solid energy pile has been implemented for decades. However, the design of different kinds of energy piles is still not well understood. In this study, a series of model tests were performed on an aluminum pipe energy pile (PEP in dry sandy soil to investigate the thermal effects on the mechanical behaviors of pipe energy pile. The thermal responses of the PEP were also analyzed. Steady temperatures of the PEP under different working conditions were also compared with that of the solid energy pile. Different loading tests were carried out on four pipe energy piles under three different temperatures of 5, 35, and 50°C, respectively. The bearing capacity change can be interpreted through the load-displacement curves. Experiment results were also compared with the solid energy pile to evaluate bearing capacities of the PEP and the solid energy pile under different temperature conditions. The mobilized shaft resistance was also calculated and compared with the solid energy pile data and the results show that the PEP has a similar load transfer mechanism with the solid energy pile. It could also be found that, for PEPs under working load, plastic displacement would appear after a whole heating cycle.

  18. 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.

  19. Characteristics of severely damaged fuel from PBF tests and the TMI-2 accident

    International Nuclear Information System (INIS)

    Osetek, D.J.; Cook, B.A.; Dallman, R.J.; Broughton, J.M.

    1986-01-01

    As a result of the TMI-2 reactor accident, the US Nuclear Regulatory Commission initiated a research program to investigate phenomena associated with severe fuel damage accidents. This program is sponsored by several countries and includes in-pile and out-of-pile experiments, separate effects studies, and computer code development. The principal in-pile testing portion of the program includes four integral severe fuel damage (SFD) tests in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The INEL is also responsible for examining the damaged core in the Three Mile Island-Unit 2 (TMI-2) reactor, which offers the unique opportunity to directly compare the findings of an experimental program to those of an actual reactor accident. The principal core damage phenomena which can occur during a severe accident are discussed, and examples from the INEL research programs are used to illustrate the characteristics of these phenomena. The preliminary results of the programs are presented, and their impact on plant operability during severe accidents is discussed

  20. Early history of cosmic rays at Chicago

    Science.gov (United States)

    Yodh, Gaurang B.

    2013-02-01

    Cosmic ray studies at the University of Chicago were started by Arthur Compton during the late 1920s. The high points of cosmic ray studies at Chicago under Compton and Marcel Schein are the focus of this report, which summarizes the research done at Chicago up to the end of World War II.

  1. Load-bearing Characters Analysis of Large Diameter Rock-Socketed Filling Piles Based on Self-Balanced Method

    Science.gov (United States)

    tongqing, Wu; liang, Li; xinjian, Liu; Xu, nianchun; Tian, Mao

    2018-03-01

    Self-balanced method is carried out on the large diameter rock-socketed filling piles of high-pile wharf at Inland River, to explore the distribution laws of load-displacement curve, pile internal force, pile tip friction resistance and pile side friction resistance under load force. The results showed that: the tip resistance of S1 and S2 test piles accounted for 53.4% and 53.6% of the pile bearing capacity, respectively, while the total side friction resistance accounted for 46.6% and 46.4% of the pile bearing capacity, respectively; both the pile tip friction resistance and pile side friction resistance can be fully played, and reach to the design requirements. The reasonability of large diameter rock-socketed filling design is verified through test analysis, which can provide basis for the optimization of high-pile wharf structural type, thus reducing the wharf project cost, and also providing reference for the similar large diameter rock-socketed filling piles of high-pile wharf at Inland River.

  2. 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%.

  3. Underwater Sound Propagation from Marine Pile Driving.

    Science.gov (United States)

    Reyff, James A

    2016-01-01

    Pile driving occurs in a variety of nearshore environments that typically have very shallow-water depths. The propagation of pile-driving sound in water is complex, where sound is directly radiated from the pile as well as through the ground substrate. Piles driven in the ground near water bodies can produce considerable underwater sound energy. This paper presents examples of sound propagation through shallow-water environments. Some of these examples illustrate the substantial variation in sound amplitude over time that can be critical to understand when computing an acoustic-based safety zone for aquatic species.

  4. 29 CFR 1926.603 - Pile driving equipment.

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 8 2010-07-01 2010-07-01 false Pile driving equipment. 1926.603 Section 1926.603 Labor... Operations § 1926.603 Pile driving equipment. (a) General requirements. (1) Boilers and piping systems which are a part of, or used with, pile driving equipment shall meet the applicable requirements of the...

  5. 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

  6. Settlement during vibratory sheet piling

    NARCIS (Netherlands)

    Meijers, P.

    2007-01-01

    During vibratory sheet piling quite often the soil near the sheet pile wall will settle. In many cases this is not a problem. For situations with houses, pipelines, roads or railroads at relative short distance these settlements may not be acceptable. The purpose of the research described in this

  7. 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)

  8. 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)

  9. Experimental apparatus for in-pile studies of: creep of nuclear fuels, and Young's-modulus of structural materials; Dispositifs experimentaux pour etudes en pile du fluage des materiaux combustibles, et du module d'elasticite des materiaux de structure

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, A; Le Bret, P; Alfille, L; Pesenti, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    Creep test under compression: the apparatus described allows to study, in an horizontal beam hole of a research reactor such as EL2, the creep behaviour of nuclear fuel samples under neutron flux. The maximum stress applied on the specimens is a constant compression chooses between 0.200 and 0.400 kg/mm{sup 2} (285 psi and 570 psi). - Young's Modulus measurement: in another horizontal beam hole of such a reactor, an apparatus allows to study the irradiation effect on Young's Modulus of a structural material specimen. (author)Fren. [French] Essai de fluage en compression: l'appareillage decrit permet d'etudier, dans un canal horizontal d'une pile experimentale type EL2, le fluage en compression d'eprouvettes de materiaux fissiles sous le flux de neutrons, sous une contrainte maximum de 500 g/mm{sup 2}. - Mesure du module d'Young: dans un canal identique au precedent, un appareillage permet de suivre l'influence du rayonnement sur le module d'elasticite d'une eprouvette d'un materiau de structure. (auteur)

  10. 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

  11. BDI behavior evaluation of an upgraded Monju core and a demonstration core. (1) Plans for the out of pile bundle compressive tests for large diameter pins

    International Nuclear Information System (INIS)

    Ichikawa, Shoichi; Haga, Hiroyuki; Katsuyama, Kozo; Uwaba, Tomoyuki; Maeda, Koji; Nishinoiri, Kenji

    2012-07-01

    The life of FBR (Fast Breeder Reactor) fuel assembly is restricted by BDI (Bundle-Duct Interaction). Therefore, it is very important to carry out the out pile bundle compressive tests which can imitate BDI, in order to evaluate BDI behavior. The target of the conventional BDI behavior was small diameter pins (φ6.5mm) for fuel pellets which were used with the assembly of Monju (the Monju prototype fast breeder reactor) etc. Furthermore by an upgraded Monju core and a demonstration core, adoption of large diameter pins for the holler annular pellets is planned. Therefore, it was necessary to carry out BDI evaluation of a large diameter pin. Then, the plans for out of pile bundle compressive test for large diameter pins were are reported. (author)

  12. Chicago's urban forest ecosystem: Results of the Chicago Urban Forest Climate Project. (Includes executive summary). Forest Service general technical report (Final)

    International Nuclear Information System (INIS)

    McPherson, E.G.; Nowak, D.J.; Rowntree, R.A.

    1994-06-01

    Results of the 3-year Chicago Urban Forest Climate Project indicate that there are an estimated 50.8 million trees in the Chicago area of Cook and DuPage Counties; 66 percent of these trees rated in good or excellent condition. During 1991, trees in the Chicago area removed an estimated 6,145 tons of air pollutants, providing air cleansing valued at $9.2 million dollars. These trees also sequester approximately 155,000 tons of carbon per year, and provide residential heating and cooling energy savings that, in turn, reduce carbon emissions from power plants by about 12,600 tons annually. Shade, lower summer air temperatures, and a reduction in windspeed associated with increasing tree cover by 10 percent can lower total heating and cooling energy use by 5 to 10 percent annually ($50 to $90 per dwelling unit). The projected net present value of investment in planting and care of 95,000 trees in Chicago is $38 million ($402 per planted tree), indicating that the long-term benefits of trees are more than twice their costs

  13. Grouting of uranium mill tailings piles

    International Nuclear Information System (INIS)

    Boegly, W.J. Jr.; Tamura, T.; Williams, J.D.

    1984-03-01

    A program of remedial action was initiated for a number of inactive uranium mill tailings piles. These piles result from mining and processing of uranium ores to meet the nation's defense and nuclear power needs and represent a potential hazard to health and the environment. Possible remedial actions include the application of covers to reduce radon emissions and airborne transport of the tailings, liners to prevent groundwater contamination by leachates from the piles, physical or chemical stabilization of the tailings, or moving the piles to remote locations. Conventional installation of liners would require excavation of the piles to emplace the liner; however, utilization of grouting techniques, such as those used in civil engineering to stabilize soils, might be a potential method of producing a liner without excavation. Laboratory studies on groutability of uranium mill tailings were conducted using samples from three abandoned piles and employing a number of particulate and chemical grouts. These studies indicate that it is possible to alter the permeability of the tailings from ambient values of 10 -3 cm/s to values approaching 10 -7 cm/s using silicate grouts and to 10 -8 cm/s using acrylamide and acrylate grouts. An evaluation of grouting techniques, equipment required, and costs associated with grouting were also conducted and are presented. 10 references, 1 table

  14. 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

  15. Technical and safe development features of modern research reactor

    International Nuclear Information System (INIS)

    Wang Jiaying; Dong Duo

    1998-01-01

    The development trend of research reactor in the world, and development situation in China are introduced. Up to now, some research reactors have serviced for long time and equipment have aged, not to be satisfied for requirement of science and technology development. New research reactors must been developed. The technical features and safe features of new type research reactor in China, for example: multi-pile utilization, compact core of high flux, high automation level of control, reactor two independent shutdown systems, great coefficient of negative temperature, passive safety systems, reliable residual heat removal system are studied

  16. Temperature response functions (G-functions) for single pile heat exchangers

    International Nuclear Information System (INIS)

    Loveridge, Fleur; Powrie, William

    2013-01-01

    Foundation piles used as heat exchangers as part of a ground energy system have the potential to reduce energy use and carbon dioxide emissions from new buildings. However, current design approaches for pile heat exchangers are based on methods developed for boreholes which have a different geometry, with a much larger aspect (length to diameter) ratio. Current methods also neglect the transient behaviour of the pile concrete, instead assuming a steady state resistance for design purposes. As piles have a much larger volume of concrete than boreholes, this neglects the significant potential for heat storage within the pile. To overcome these shortcomings this paper presents new pile temperature response functions (G-functions) which are designed to reflect typical geometries of pile heat exchangers and include the transient response of the pile concrete. Owing to the larger number of pile sizes and pipe configurations which are possible with pile heat exchangers it is not feasible to developed a single unified G-function and instead upper and lower bound solutions are provided for different aspects ratios. - Highlights: • We present new temperature response functions for pile heat exchangers. • The functions include transient heat transfer within the pile concrete. • Application of the functions reduces the resulting calculated temperature ranges. • Greater energy efficiency is possible by accounting for heat storage in the pile

  17. Prediction of pile set-up for Ohio soils.

    Science.gov (United States)

    2011-02-01

    ODOT typically uses small diameter driven pipe piles for bridge foundations. When a pile is driven into the subsurface, it disturbs and displaces the soil. As the soil surrounding the pile recovers from the installation disturbance, a time dependant ...

  18. Control panel for radiation around reactors (1963); Tableaux de controle des radiations aupres des piles (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Candes, P; Barthoux, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    This report outlines the general philosophy of radiation control in French reactors and their annexes. The supervision is carried out continuously from a central control panel on which appear all the measurements made and the alarm signals. The equipment is described; one item makes it possible to measure simultaneously the radioactive dusts and gases. The specifications of the alarm system, which is considered to be the most important are given. Finally a new measuring technique is proposed which makes it possible to reduce considerably the cost of radiation control while at the same time providing the results in a form in which they can be easily treated, in particular in the case of the calculation of total doses. (authors) [French] Ce rapport definit la philosophie generale du controle des radiations dans les piles francaises et dans leurs annexes. La surveillance se fait d'une maniere continue a partir d'un tableau de controle centralise ou sont reportees toutes les mesures et les signalisations d'alarme. On decrit les appareils utilises, dont un permet la mesure simultanee des poussieres et gaz radioactifs, et on definit les specifications de la fonction alarme qui est consideree comme la plus importante. Enfin on propose une nouvelle technique de mesure qui permettrait de reduire considerablement le cout du controle des radiations tout en fournissant des resultats plus facilement exploitables, en particulier pour le calcul des doses integrees. (auteurs)

  19. Heat dissipating nuclear reactor with metal liner

    Science.gov (United States)

    Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

  20. Power cycling experiments in INR-TRIGA-SSR Reactor

    International Nuclear Information System (INIS)

    Dumitru, M.

    2008-01-01

    The in-reactor experimental program started this summer with some power cycling experiments to provide date on fuel behaviour under abnormal reactor operating conditions. The paper describes the irradiation device, its operational features and an original 'under-flux' movement system. Also, there are presented main data of irradiation device (pressure, flow, temperature, construction), in-pile section, location, sample, instrumentation, experimental sequences and operating data of Interest for the experimenters. (author)

  1. High resolution esophageal manometry--the switch from "intuitive" visual interpretation to Chicago classification.

    Science.gov (United States)

    Srinivas, M; Balakumaran, T A; Palaniappan, S; Srinivasan, Vijaya; Batcha, M; Venkataraman, Jayanthi

    2014-03-01

    High resolution esophageal manometry (HREM) has been interpreted all along by visual interpretation of color plots until the recent introduction of Chicago classification which categorises HREM using objective measurements. It compares HREM diagnosis of esophageal motor disorders by visual interpretation and Chicago classification. Using software Trace 1.2v, 77 consecutive tracings diagnosed by visual interpretation were re-analyzed by Chicago classification and findings compared for concordance between the two systems of interpretation. Kappa agreement rate between the two observations was determined. There were 57 males (74 %) and cohort median age was 41 years (range: 14-83 years). Majority of the referrals were for gastroesophageal reflux disease, dysphagia and achalasia. By "intuitive" visual interpretation, the tracing were reported as normal in 45 (58.4 %), achalasia 14 (18.2 %), ineffective esophageal motility 3 (3.9 %), nutcracker esophagus 11 (14.3 %) and nonspecific motility changes 4 (5.2 %). By Chicago classification, there was 100 % agreement (Kappa 1) for achalasia (type 1: 9; type 2: 5) and ineffective esophageal motility ("failed peristalsis" on visual interpretation). Normal esophageal motility, nutcracker esophagus and nonspecific motility disorder on visual interpretation were reclassified as rapid contraction and esophagogastric junction (EGJ) outflow obstruction by Chicago classification. Chicago classification identified distinct clinical phenotypes including EGJ outflow obstruction not identified by visual interpretation. A significant number of unclassified HREM by visual interpretation were also classified by it.

  2. Numerical Simulation by using Soldiers Pile of the Embankment on Semarang-Solo Highway

    Science.gov (United States)

    Tumanduk, M. S. S. S.; Maki, T. S.; Pangkey, T. U. Y.; Pandeiroth, Y. C.

    2018-02-01

    Semarang-Solo highway works section II Gedawang-Penggaron constitutes a labile area. It is thought to be effect of the existence of coat clay shale which have moulded. For the purpose of anticipating the embankment mass movement it is placed line bored pile and stringed up (soldiers pile). The objective of this research is to know the efficient use of soldier’s pile of the embankment on Semarang-Solo highway section II Gedawang-Penggaron pursuant based upon numerical simulation. The result of analysis depicts that original slope in a stabil state with horizontal displacement which equal to 0.06 m and safety factor (SF) which equal to 1.31. The strengthened embankment with bored pile is not effective to give am SF improvement at slope so that, at this phase, the slope cannot be slid to be safe enough from landslide namely with horizontal displacement which equal to 0.20 m and SF which equal to 1.09. The effect of traffic load horizontal displacement is which equal to 0.21 m with SF which equal to 1.00. The earthquake simulation results horizontal displacement which equal to 0.75 m with SF which equal to 1.00. Long variation of bored pile of phase II by neglecting bored pile phase III at the depth 35 m yields horizontal displacement which equal to 0.03 m and SF optimum which equal to 2.17. The variation of pile location by placing bored pile under embankment slope foot with distance from the location of bored pile of phase II which equal to 20 m without changing the profile of the existing bored pile creates the horizontal displacement which equals to 0.02 m with SF which equal to 2.29. The result of the horizontal displacement and SF of the two alternative is safer compared to the existing condition (SF>1.5).

  3. Installation effects of auger cast-in-place piles

    Directory of Open Access Journals (Sweden)

    Fathi M. Abdrabbo

    2012-12-01

    Full Text Available Since their introduction in Europe and North America some 50 years ago, auger cast-in-place piles (ACIP have become increasingly popular all over the world. These piles offer considerable environmental advantages during construction including minimal vibration, and low noise beside their high productivity. The most severe limitation of the ACIP is its sensitivity to operator performance, which can lead to a pile of poor integrity or inconsistent quality. Thus the improper use of ACIP equipment can result in piles containing defects or can cause instability of nearby structures. Three case studies are presented and discussed in an effort to illustrate learned lessons. First case study highlights the misuse of ACIP equipment leading to unreliable defective pile foundations. Second and third case studies show the adverse effects of installing ACIP on the stability of nearby structures. The study revealed that it is essential to employ a clever pile crew during the installation of ACIP to observe, interpret, and take corrective actions for unusual situations. The authorities worldwide should oblige pile contractors to employ only experienced and qualified workers in charge of geotechnical engineering works. Tender documents should include precise clauses related to the technological factors affecting the quality of ACIP. Unfavorable side effects of installing ACIP in saturated loose and medium sand can cause tilt of adjacent existing structures; even they are on either shallow or deep foundations. A row of micro-piles and/or soil grouting adjacent to the existing buildings were successfully used to reduce the adverse effects of ACIP. Implementation of different codes on the results of pile loading tests produced different pile working loads. Therefore tender documents should specify the code upon which interpreting the pile test results. At the meantime the geotechnical engineer should implement his experience and judgment during application of the

  4. Reactor BR2

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2000-07-01

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

  5. Reactor BR2

    International Nuclear Information System (INIS)

    Gubel, P.

    2000-01-01

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

  6. John Dewey and early Chicago functionalism.

    Science.gov (United States)

    Backe, A

    2001-11-01

    John Dewey and James Angell are regarded respectively as the founder and systematizer of the Chicago school of functional psychology. The early Chicago school traditionally has been portrayed as a unified theoretical approach based primarily on William James's naturalist theory of mental processes. It is argued in this article that although the psychology systematized by Angell bore a close affinity to James's naturalism, Dewey's own psychology was based primarily on the neo-Hegelian philosophy of Thomas Hill Green. Through a review of a number of Dewey's major writings, Green's neo-Hegelian philosophy is shown to have influenced Dewey's views on psychological concepts such as reaction, emotion, and perception during the formative period of the Chicago school. The interpretation of Dewey's psychology developed in this article leads to the conclusion that early Chicago functionalism should not be regarded as a unified theoretical approach.

  7. The Effects of Time on Soil Behaviour and Pile Capacity

    DEFF Research Database (Denmark)

    Augustesen, Anders

    When designing pile foundations, static design equations, pile driving formulae, static loading tests or stress wave analyses can be employed to estimate the axial capacity of single piles. Both laboratory and field tests show that soil exhibits time-dependent behaviour. An important result...... based on a set of static loading tests. In the literature it is suggested that the pile capacity increases with the logarithm to time after installation which is confirmed in this thesis. In continuation of this, it is analysed whether the magnitude of the set-up is related to the properties of the clay...... circumstances (e.g. load specifications, length of pile, pile material). In order to evaluate the design methods for piles in clay, it is necessary to correct for time between pile driving and pile testing. Results of testing the calculation procedures against the available data by employing different time...

  8. The past, present, and future of test and research reactor physics

    International Nuclear Information System (INIS)

    Ryskamp, J.M.

    1992-01-01

    Reactor physics calculations have been performed on research reactors since the first one was built 50 yr ago under the University of Chicago stadium. Since then, reactor physics calculations have evolved from Fermi-age theory calculations performed with slide rules to three-dimensional, continuous-energy, coupled neutron-photon Monte Carlo computations performed with supercomputers and workstations. Such enormous progress in reactor physics leads us to believe that the next 50 year will be just as exciting. This paper reviews this transition from the past to the future

  9. Development of p-y curves of laterally loaded piles in cohesionless soil.

    Science.gov (United States)

    Khari, Mahdy; Kassim, Khairul Anuar; Adnan, Azlan

    2014-01-01

    The research on damages of structures that are supported by deep foundations has been quite intensive in the past decade. Kinematic interaction in soil-pile interaction is evaluated based on the p-y curve approach. Existing p-y curves have considered the effects of relative density on soil-pile interaction in sandy soil. The roughness influence of the surface wall pile on p-y curves has not been emphasized sufficiently. The presented study was performed to develop a series of p-y curves for single piles through comprehensive experimental investigations. Modification factors were studied, namely, the effects of relative density and roughness of the wall surface of pile. The model tests were subjected to lateral load in Johor Bahru sand. The new p-y curves were evaluated based on the experimental data and were compared to the existing p-y curves. The soil-pile reaction for various relative density (from 30% to 75%) was increased in the range of 40-95% for a smooth pile at a small displacement and 90% at a large displacement. For rough pile, the ratio of dense to loose relative density soil-pile reaction was from 2.0 to 3.0 at a small to large displacement. Direct comparison of the developed p-y curve shows significant differences in the magnitude and shapes with the existing load-transfer curves. Good comparison with the experimental and design studies demonstrates the multidisciplinary applications of the present method.

  10. Development of p-y Curves of Laterally Loaded Piles in Cohesionless Soil

    Science.gov (United States)

    Khari, Mahdy; Kassim, Khairul Anuar; Adnan, Azlan

    2014-01-01

    The research on damages of structures that are supported by deep foundations has been quite intensive in the past decade. Kinematic interaction in soil-pile interaction is evaluated based on the p-y curve approach. Existing p-y curves have considered the effects of relative density on soil-pile interaction in sandy soil. The roughness influence of the surface wall pile on p-y curves has not been emphasized sufficiently. The presented study was performed to develop a series of p-y curves for single piles through comprehensive experimental investigations. Modification factors were studied, namely, the effects of relative density and roughness of the wall surface of pile. The model tests were subjected to lateral load in Johor Bahru sand. The new p-y curves were evaluated based on the experimental data and were compared to the existing p-y curves. The soil-pile reaction for various relative density (from 30% to 75%) was increased in the range of 40–95% for a smooth pile at a small displacement and 90% at a large displacement. For rough pile, the ratio of dense to loose relative density soil-pile reaction was from 2.0 to 3.0 at a small to large displacement. Direct comparison of the developed p-y curve shows significant differences in the magnitude and shapes with the existing load-transfer curves. Good comparison with the experimental and design studies demonstrates the multidisciplinary applications of the present method. PMID:24574932

  11. Development of p-y Curves of Laterally Loaded Piles in Cohesionless Soil

    Directory of Open Access Journals (Sweden)

    Mahdy Khari

    2014-01-01

    Full Text Available The research on damages of structures that are supported by deep foundations has been quite intensive in the past decade. Kinematic interaction in soil-pile interaction is evaluated based on the p-y curve approach. Existing p-y curves have considered the effects of relative density on soil-pile interaction in sandy soil. The roughness influence of the surface wall pile on p-y curves has not been emphasized sufficiently. The presented study was performed to develop a series of p-y curves for single piles through comprehensive experimental investigations. Modification factors were studied, namely, the effects of relative density and roughness of the wall surface of pile. The model tests were subjected to lateral load in Johor Bahru sand. The new p-y curves were evaluated based on the experimental data and were compared to the existing p-y curves. The soil-pile reaction for various relative density (from 30% to 75% was increased in the range of 40–95% for a smooth pile at a small displacement and 90% at a large displacement. For rough pile, the ratio of dense to loose relative density soil-pile reaction was from 2.0 to 3.0 at a small to large displacement. Direct comparison of the developed p-y curve shows significant differences in the magnitude and shapes with the existing load-transfer curves. Good comparison with the experimental and design studies demonstrates the multidisciplinary applications of the present method.

  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. Studies on solid-state physics carried out with the Saclay reactor (1962); Etudes de physique du solide realisees a la pile de Saclay (1962)

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-07-01

    This paper deals only with solid-state physics experiments carried out on outgoing beams: rather than giving a general review of the work performed, if refers to only a few of the most important studies or those nearest completion. These are being made with the experimental beams of the two Saclay reactors EL-2, with a central flux of 10{sup 13} n/cm{sup 2}, and - since 1958 - EL-3, whose central flux is equal ta 10{sup 14} n/cm{sup 2}. The experiments are being carried out by two separate groups of physicists, employing different techniques, namely neutron diffraction using a crystal spectrometer, and inelastic scattering using a time-of-flight spectrometer. (author) [French] Cet expose ne relate que des experiences de physique du solide faites sur des faisceaux sortis; plutot que de donner une revue de l'ensemble des travaux effectues, on ne cite que quelques etudes que l'on peut considerer comme plus essentielles ou mieux achevees. On utilise les faisceaux experimentaux des deux piles de Saclay, EL-2 dont le flux au centre est de 10{sup 13}n/cm{sup 2} et, depuis 1958, EL-3 pour laquelle il est egal a 10{sup 14} n/cm{sup 2}. Les experiences sont realisees par deux groupes de physiciens distincts, employant des techniques differentes, la diffraction des neutrons qui utilise un spectrometre a cristal, et la diffusion inelastique avec un spectrometre a temps de vol. (auteur)

  14. Heat dissipating nuclear reactor

    Science.gov (United States)

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  15. Statistical fluctuations in reactors (1960); Fluctuations statistiques dans les piles (1960)

    Energy Technology Data Exchange (ETDEWEB)

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

    1960-07-01

    The theory of space dependent fluctuations is developed, taking into account the effect of delayed neutrons. The 'diffusion equation' or Fokker-Planck equation is worked out in the case of age and two group theory, but the first one because of in physical significance is used in this report. The theory is applied to the study of the spectral density of fluctuations and fluctuations of counting rate and current flowing through the charge resistor of an ionisation chamber, the effect of the entrance capacity is discussed. The space dependent theory shows that the fluctuations in the core and reflector of a near critical pile obey to the same law. The spectral densities in the core and reflector are similar, there is no sensible attenuation of high frequency fluctuations in the reflector. Compared to the space independent theory, this theory give better agreement with experience, one can use the simple space independent theory but in checking with experiment it is necessary to introduce numerical factors given by the space dependent theory. (author) [French] La theorie des fluctuations statistiques est developpee dans le cas spatial en tenant compte des neutrons retardes, et dans le cadre de la theorie de l'age vitesse. L'equation d'evolution de la probabilite est egalement etablie dans le cadre de la theorie a deux groupes. Ces considerations sont appliquees a l'etude de la densite spectrale des fluctuations et aux fluctuations des taux de comptage et du courant circulant dans la resistance de charge du detecteur. On etudie en particulier l'effet de la constante de temps introduite par la capacite d'entree. Cette theorie etablit que les fluctuations dans le coeur et le reflecteur suivent la meme loi pour une pile critique, il en est de meme pour la densite spectrale meme a frequence elevee. Par rapport a la theorie d'ensemble, la theorie spatiale donne des coefficients numeriques ou facteurs de forme, qui permettent d'obtenir un bon accord entre la theorie et l

  16. Multi-scale sensitivity analysis of pile installation using DEM

    Science.gov (United States)

    Esposito, Ricardo Gurevitz; Velloso, Raquel Quadros; , Eurípedes do Amaral Vargas, Jr.; Danziger, Bernadete Ragoni

    2017-12-01

    The disturbances experienced by the soil due to the pile installation and dynamic soil-structure interaction still present major challenges to foundation engineers. These phenomena exhibit complex behaviors, difficult to measure in physical tests and to reproduce in numerical models. Due to the simplified approach used by the discrete element method (DEM) to simulate large deformations and nonlinear stress-dilatancy behavior of granular soils, the DEM consists of an excellent tool to investigate these processes. This study presents a sensitivity analysis of the effects of introducing a single pile using the PFC2D software developed by Itasca Co. The different scales investigated in these simulations include point and shaft resistance, alterations in porosity and stress fields and particles displacement. Several simulations were conducted in order to investigate the effects of different numerical approaches showing indications that the method of installation and particle rotation could influence greatly in the conditions around the numerical pile. Minor effects were also noted due to change in penetration velocity and pile-soil friction. The difference in behavior of a moving and a stationary pile shows good qualitative agreement with previous experimental results indicating the necessity of realizing a force equilibrium process prior to any load-test to be simulated.

  17. 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

  18. 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

  19. 30 CFR 816.83 - Coal mine waste: Refuse piles.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 3 2010-07-01 2010-07-01 false Coal mine waste: Refuse piles. 816.83 Section... ACTIVITIES § 816.83 Coal mine waste: Refuse piles. Refuse piles shall meet the requirements of § 816.81, the... drainage may not be diverted over the outslope of the refuse piles. Runoff from the areas above the refuse...

  20. Windscale advanced gas-cooled reactor (WAGR) decommissioning project overview

    International Nuclear Information System (INIS)

    Pattinson, A.

    2003-01-01

    The current BNFL reactor decommissioning projects are presented. The projects concern power reactor sites at Berkely, Trawsfynydd, Hunterstone, Bradwell, Hinkley Point; UKAEA Windscale Pile 1; Research reactors within UK Scottish Universities at East Kilbride and ICI (both complete); WAGR. The BNFL environmental role include contract management; effective dismantling strategy development; implementation and operation; sentencing, encapsulation and transportation of waste. In addition for the own sites it includes strategy development; baseline decommissioning planning; site management and regulator interface. The project objectives for the Windscale Advanced Gas-Cooled Reactor (WAGR) are 1) Safe and efficient decommissioning; 2) Building of good relationships with customer; 3) Completion of reactor decommissioning in 2005. The completed WAGR decommissioning campaigns are: Operational Waste; Hot Box; Loop Tubes; Neutron Shield; Graphite Core and Restrain System; Thermal Shield. The current campaign is Lower Structures and the remaining are: Pressure vessel and Insulation; Thermal Columns and Outer Vault Membrane. An overview of each campaign is presented

  1. 1938-1942: the fateful 4 years of discovery

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The chain of discovery which led from an experiment in 1938 which demonstrated nuclear fission to the first controlled chain reaction in the Chicago Pile 1 (CP1) on 2nd December 1942 is outlined. The initial experiment by Hahn and Strassman detected barium as a product of the bombardment of uranium by neutrons. An explanation of the presence of the barium was offered by Meitner and Frisch in terms of the fission of the uranium atom into two elements. This involved a net reduction in mass which it was suggested was converted into energy. This was subsequently confirmed experimentally. A further prediction that the process produced secondary neutrons was also confirmed leading to the possibility of a chain reaction. From then on, efforts were directed in Germany, the USA and the United Kingdom towards the building of a structure in which there was a self-propagating, controlled, chain reaction and also into devising a bomb based on an explosive nuclear chain reaction. The elements of an ''atomic pile'' were gradually conceived including neutron moderation, critical size, the neutron multiplication factor and a reactor lattice. Under the direction of Fermi, this culminated in the construction of the uranium, graphite moderated pile CP1. The plutonium produced in such a reactor was one route to a bomb. Of other routes explored in the USA over the same period, the separation of U235 by gaseous diffusion was selected for development. (UK)

  2. Numerical Simulation for the Soil-Pile-Structure Interaction under Seismic Loading

    Directory of Open Access Journals (Sweden)

    Lifeng Luan

    2015-01-01

    Full Text Available Piles are widely used as reinforcement structures in geotechnical engineering designs. If the settlement of the soil is greater than the pile, the pile is pulled down by the soil, and negative friction force is produced. Previous studies have mainly focused on the interaction of pile-soil under static condition. However, many pile projects are located in earthquake-prone areas, which indicate the importance of determining the response of the pile-soil structure under seismic load. In this paper, the nonlinear, explicit, and finite difference program FLAC3D, which considers the mechanical behavior of soil-pile interaction, is used to establish an underconsolidated soil-pile mode. The response processes of the pile side friction force, the pile axial force, and the soil response under seismic load are also analyzed.

  3. Emission of toxic explosive and fire hazardous gases in coal piles stored under atmospheric conditions. Part I

    International Nuclear Information System (INIS)

    Grossman, S.L.; Cohen, H.

    1998-01-01

    Bituminous coal stockpiles stored in open air undergo weathering processes due to low temperature oxidation (40-100 degree C) resulting in quality deterioration. The process is accompanied by emission of hazardous explosive gases such as molecular hydrogen and low molecular weight organic gases. The article describes the process of low temperature oxidation of coal and goes on to report on simulation experiments carried out to assess the oxidation resistance of various coals stored in Israel, performed in small glass batch reactors and on the monitoring of temperatures and gas evolved in large coal piles stored in open air (performed using a portable unit which can penetrate up to 7 meters inside a coal pile). Molecular hydrogen emissions were found in small concentrations, in all types of coal studied. The amount of hydrogen formed in the batch reactors is linearly dependent on the amount of oxygen consumed in the coal oxidation process and also on the temperature. It was only slightly dependent on the coal mass and independent of particle size. Previous published work has only mentioned hydrogen emission at higher temperatures (240 degree C)

  4. Interacting with piles of artifacts on digital tables

    NARCIS (Netherlands)

    Aliakseyeu, D.; Lucero Vera, A.A.; Subramanian, S.

    2007-01-01

    Designers and architects regularly use piles to organise visual artifacts. Recent efforts have now made it possible for users to create piles in digital systems as well. However, there is still little understanding of how users shouldinteract with digital piles. In this paper we investigate this

  5. Interacting with piles of artifacts on digital tables

    NARCIS (Netherlands)

    Aliakseyeu, D.; Subramanian, S.; Lucero Vera, A.A.; Gutwin, C.

    2006-01-01

    Designers and architects regularly use piles to organize visual artifacts. Recent efforts have now made it possible for users to create piles in digital systems as well. However, there is still little understanding of how users should interact with digital piles. In this paper we investigate this

  6. FIELD INVESTIGATIONS OF PILED-RAFT FOUNDATIONS WITH SHORT-LENGTH CONIC PILES IN BUILDING AREAS OF MINSK

    Directory of Open Access Journals (Sweden)

    V. A. Sernov

    2015-01-01

    Full Text Available In recent time piled foundations are extensively applied due to an increase of storeys in buildings constructed in Minsk and load increment on the soil. Preference is given to this approach even in the case when relatively firm soil occurs in the top part of the foundation bed. In this case maximum usage of the foundation bed bearing capacity and reduction of foundation cost are considered as top-priority tasks for designers. One of the ways to increase the bearing capacity of piled foundations is the necessity to take into account resistance of foundation bed soil located under raft bottom. The raft as well as a shallow foundation is capable to transfer a significant part of building load into the soil. Such approach makes it possible to reduce a number of piles in the foundation or shorten their length. Then it results in shortening of the construction period and significant reduction in zero cycle. However up to the present moment reliable calculation methods that permit to take into account soil resistance in the raft base. An analysis of previous investigations on the matter executed by various researchers and a number of field investigations have been carried out with the purpose to develop the proposed methods.The paper presents results of field investigations on foundations consisting of short stamped tapered piles which are joined together with the help of the raft fragment. Strength and deformation characteristics of the bases are increasing while making such foundations in the fill-up soil. In this case the filled-up ground layer becomes a bearing layer both for piles and rafts as well. Improvement of high-plastic clay-bearing soil properties is ensured by ramming dry concrete mix under pile foot. The paper describes an experience on application of the piled-raft foundation in complicated engineering and geological conditions while constructing the Orthodox Church in Minsk.

  7. Pile Structure Program, Projected Start Date : January 1, 2010 (Implementation).

    Energy Technology Data Exchange (ETDEWEB)

    Collins, Chris; Corbett, Catherine [Lower Columbia River Estuary Partnership; Ebberts, Blaine [U.S. Army Corps of Engineers

    2009-07-27

    The 2008 Federal Columbia River Power System Biological Opinion includes Reasonable and Prudent Alternative 38-Piling and Piling Dike Removal Program. This RPA directs the Action Agencies to work with the Estuary Partnership to develop and implement a piling and pile dike removal program. The program has since evolved to include modifying pile structures to enhance their habitat value and complexity by adding large woody debris. The geographic extent of the Pile Structure Program (PSP) includes all tidally-influenced portions of the lower Columbia River below Bonneville Dam; however, it will focus on the mainstem. The overarching goal of the PSP is to enhance and restore ecosystem structure and function for the recovery of federally listed salmonids through the active management of pile structures. To attain this goal, the program team developed the following objectives: (1) Develop a plan to remove or modify pile structures that have lower value to navigation channel maintenance, and in which removal or modification will present low-risk to adjacent land use, is cost-effective, and would result in increased ecosystem function. (2) Determine program benefits for juvenile salmonids and the ecosystem through a series of intensively monitored pilot projects. (3) Incorporate best available science and pilot project results into an adaptive management framework that will guide future management by prioritizing projects with the highest benefits. The PSP's hypotheses, which form the basis of the pilot project experiments, are organized into five categories: Sediment and Habitat-forming Processes, Habitat Conditions and Food Web, Piscivorous Fish, Piscivorous Birds, and Toxic Contaminant Reduction. These hypotheses are based on the effects listed in the Estuary Module (NOAA Fisheries in press) and others that emerged during literature reviews, discussions with scientists, and field visits. Using pilot project findings, future implementation will be adaptively managed

  8. Investigation of a North Sea oil platform drill cuttings pile

    International Nuclear Information System (INIS)

    Hartley, J.P.; Watson, T.N.

    1993-01-01

    A comprehensive study of the drill cuttings pile at North West Hutton was undertaken in August, 1992. Fifty one wells have been drilled in the field, mainly using mineral oil based drill fluids, with the cuttings discharged to sea. The cuttings pile was mapped using a 3D side scan sonar system and the periphery was defined by towed side scan sonar and gamma ray spectrometer surveys. The pile was cored by vibrocorer to a maximum depth of 2.35m. The cores were assessed geotechnically and subsampled for physical and chemical analyses. Environmental impact was investigated by grab sampling at 12 stations out to 7,500m, selected on the basis of cuttings distribution. The results are relevant to the corrosion and long-term environmental effects of oily cuttings piles, the remove/leave alone debate, and abandonment planning. The cores were subsampled for hydrocarbon, trace metals and sulphide content and grain size analysis. Metals analyses included identification of metal species to estimate bioavailability and implications of pile disturbance. Estimates of oil migration within the pile are made from correlation of the chemical analyses results with the drilling history, in particular the change from diesel to low toxicity base oil in 1984. Strong gradients were found in the faunal data which correlate well with the physical and chemical results. Dense populations of opportunists species were present adjacent to the platform, including a novel molluscan opportunist. This is the first comprehensive study of an oily cuttings pile and is a contribution to the debate on their long term impact and fate on abandonment

  9. 30 CFR 77.215-1 - Refuse piles; identification.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Refuse piles; identification. 77.215-1 Section... COAL MINES Surface Installations § 77.215-1 Refuse piles; identification. A permanent identification marker, at least six feet high and showing the refuse pile identification number as assigned by the...

  10. Rational use of anchor pile material of the thin retaining walls

    Directory of Open Access Journals (Sweden)

    Yushkov Boris Semenovich

    2014-12-01

    Full Text Available The article considers the urgency of application of the reinforced concrete anchor piles in the constructions of retaining structures associated with the possibility of establishing rigid joint of element interface and more durable pile constructions in the soil. The features of the inclined anchor piles work as a part of sheet-pile retaining walls are noted. There was performed a study of the stress-strain state of the inclined reinforced concrete anchor piles of the thin sheet-pile wall with the reinforced concrete face members of T-section, combined with piles by a longitudinal beam. The authors consider a constructive scheme of retaining structure and list the applied loads. The efforts in the anchor piles were determined. The bending-moment curves show the character of the force distribution along the pile. A form of the pile ensuring the rational distribution of material along the pile is presented. The distribution of efforts along the length and effect of filling on its operation in the soil were accepted as the criteria of construction solution for a pile. The substantiation of the proposed design of pile is presented in terms of its stress-strain state and the rational use of material. The authors made conclusions on the reasonability of adopted design solutions associated with an increase in the flexural strength of pile, increment of the ultimate pullout capacity, stability improvement, effective use of backfill and exception of the «out of operation» areas of the pile.

  11. Concentration processes under tubesheet sludge piles in nuclear steam generators

    International Nuclear Information System (INIS)

    Gonzalez, F.; Spekkens, P.

    1987-01-01

    The process by which bulk water solutes are concentrated under tubesheet sludge piles in nuclear steam generators was investigated in the laboratory under simulated CANDU operating conditions. Concentration rates were found to depend on the tube heat flux and pile depth, although beyond a critical depth the concentration efficiency decreased. This efficiency could be expressed by a concentration coefficient, and was found to depend also on the sludge pile porosity. Solute concentration profiles in the sludge pile suggested that the concentration mechanism in a high-porosity/permeability pile is characterized by boiling mainly near or at the tube surface, while in low-porosity piles, the change of phase may also become important in the body of the sludge pile. In all cases, the full depth of the pile was active to some extent in the concentration process. As long as the heat transfer under the pile was continued, the solute remained under the pile and slowly migrated toward the bottom. When the heat transfer was stopped, the solute diffused back into the bulk solution at a rate slower than that of the concentration process

  12. Review of vibration effect during piling installation to adjacent structure

    Science.gov (United States)

    Rahman, Nurul Aishah Abd; Musir, Adhilla Ainun; Dahalan, Nurol Huda; Ghani, Abdul Naser Abdul; Khalil, Muhamad Kasimi Abd

    2017-12-01

    Basically, many major structures across the world such as towers, high rise building, houses and bridges utilize pile as a support material. The use of pile is important to strengthen the structures. However, this has led to another problem to the nearest surrounding structures resulted from pile driving. As part of a construction work, unavoidable pile driving activity generates a vibration towards the surrounding structures if uncontrolled may cause damage to the adjacent structure. As the current construction works are frequently located in urban areas where the distance between the nearest building structures is not far, vibration may cause damage to nearby structures. Knowing which part of the building that is mostly affected by various vibration patterns from the impact of pile driving is crucial. Thus, it is very important to predict the impact of vibration during piling installation work. This paper reviews the vibrations generated by piling activity toward surrounding structures in terms sources of vibration, impact of piling installation, pile-soil interaction, and factors affecting the vibration impact of building as well as to study the parameters involved in vibration generation during piling works.

  13. Observation of Spectral Signatures of 1/f Dynamics in Avalanches on Granular Piles

    Science.gov (United States)

    Kim, Yong W.; Nishino, Thomas K.

    1997-03-01

    Granular piles of monodisperse glass spheres, 0.46+0.03 mm in diameter, have been studied. The base diameter of the pile has been varied from 3/8" to 2" in 1/8" increments. A single-grain dispenser with greater than 95consisting of a stepping motor-actuated reciprocating arm with a single-grain scoop. Each grain is dropped on the apex of the pile with lowest possible landing velocity at intervals at least 30longer than the duration of largest avalanches for each given pile. Each grain being added and being lost in avalanches from the pile is optically detected and recorded. The power spectrum of the net addition of grains to the pile as a function of time is found to be robustly 1/f for all base sizes. A wide variety of dynamical properties of 1/f systems, as obtained from the high precision data, will be presented.

  14. Chicago-St. Louis high speed rail plan

    International Nuclear Information System (INIS)

    Stead, M.E.

    1994-01-01

    The Illinois Department of Transportation (IDOT), in cooperation with Amtrak, undertook the Chicago-St. Louis High Speed Rail Financial and Implementation Plan study in order to develop a realistic and achievable blueprint for implementation of high speed rail in the Chicago-St. Louis corridor. This report presents a summary of the Price Waterhouse Project Team's analysis and the Financial and Implementation Plan for implementing high speed rail service in the Chicago-St. Louis corridor

  15. Chicago-St. Louis high speed rail plan

    Energy Technology Data Exchange (ETDEWEB)

    Stead, M.E.

    1994-12-31

    The Illinois Department of Transportation (IDOT), in cooperation with Amtrak, undertook the Chicago-St. Louis High Speed Rail Financial and Implementation Plan study in order to develop a realistic and achievable blueprint for implementation of high speed rail in the Chicago-St. Louis corridor. This report presents a summary of the Price Waterhouse Project Team`s analysis and the Financial and Implementation Plan for implementing high speed rail service in the Chicago-St. Louis corridor.

  16. Heterogeneous dipolar theory of the exponential pile

    International Nuclear Information System (INIS)

    Mastrangelo, P.V.

    1981-01-01

    We present a heterogeneous theory of the exponential pile, closely related to NORDHEIM-SCALETTAR's. It is well adapted to lattice whose pitch is relatively large (D-2O, grahpite) and the dimensions of whose channels are not negligible. The anisotropy of neutron diffusion is taken into account by the introduction of dipolar parameters. We express the contribution of each channel to the total flux in the moderator by means of multipolar coefficients. In order to be able to apply conditions of continuity between the flux and their derivatives, on the side of the moderator, we develop in a Fourier series the fluxes found at the periphery of each channel. Using Wronski's relations of Bessel's functions, we express the multipolar coefficients of the surfaces of each channel, on the side of the moderator, by means of the harmonics of each flux and their derivatives. We retain only monopolar (A 0 sub(g)) and dipolar (A 1 sub(g)) coefficients; those of a higher order are ignored. We deduce from these coefficients the systems of homogeneous equations of the exponential pile with monopoles on their own and monopoles plus dipoles. It should be noted that the systems of homogeneous equations of the critical pile are contained in those of the exponential pile. In another article, we develop the calculation of monopolar and dipolar heterogeneous parameters. (orig.)

  17. 30 CFR 77.215-3 - Refuse piles: certification.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Refuse piles: certification. 77.215-3 Section... COAL MINES Surface Installations § 77.215-3 Refuse piles: certification. (a) Within 180 days following written notification by the District Manager that a refuse pile can present a hazard, the person owning...

  18. Material Science Activities for Fusion Reactors in Kazakhstan

    International Nuclear Information System (INIS)

    Tazhibayeva, I.; Kenzhin, E.; Kulsartov, T.; Shestakov, V.; Chikhray, Y.; Azizov, E.; Filatov, O.; Chernov, V.M.

    2007-01-01

    Full text of publication follows: Paper contains results of fusion material testing national program and results of activities on creation of material testing spherical tokamak. Hydrogen isotope behavior (diffusion, permeation, and accumulation) in the components of the first wall and divertor was studied taking into account temperature, pressure, and reactor irradiation. There were carried out out-of-pile and in-pile (reactors IVG-IM, WWRK, RA) studies of beryllium of various grades (TV-56, TShG-56, DV-56, TGP-56, TIP-56), graphites (RG-T, MPG-8, FP 479, R 4340), molybdenum, tungsten, steels (Cr18Ni10Ti, Cr16Ni15, MANET, F82H), alloys V-(4-6)Cr-( 4-5)Ti, Cu+1%Cr+0.1%Zr, and double Be/Cu and triple Be/Cu/steel structures. Tritium permeability from eutectic Pb+17%Li through steels Cr18Ni10Ti, Cr16Ni15, MANET, and F82H were studied taking into account protective coating effects. The tritium production rate was experimentally assessed during in-pile and post-reactor experiments. There were carried out radiation tests of ceramic Li 2 TiO 3 (96% enrichment by Li-6) with in-situ registration of released tritium and following post-irradiation material tests of irradiated samples. Verification of computer codes for simulation of accidents related to LOCA in ITER reactor was carried out. Codes' verification was carried out for a mockup of first wall in a form of three-layer cylinder of beryllium, bronze (Cu-Cr-Zr) and stainless steel. At present Kazakhstan Tokamak for Material testing (tokamak KTM) is created in National Nuclear Center of Republic of Kazakhstan in cooperation with Russian Federation organizations (start-up is scheduled on 2008). Tokamak KTM allows for expansion and specification of the studies and tests of materials, protection options of first wall, receiving divertor tiles and divertor components, methods for load reduction at divertor, and various options of heat/power removal, fast evacuation of divertor volume and development of the techniques for

  19. Bore pile foundation tall buildings closed in the heritage building

    Science.gov (United States)

    Triastuti, Nusa Setiani

    2017-11-01

    Bore pile foundation for high building surroundings heritage building should be not damage. Construction proses must good, no necking, no mixed deep water, no sliding soil, nonporous concrete. Objective the execution of bore pile so that heritage buildings and neighboring buildings that are old do not experience cracks, damage and tilting. The survey methodology was observe the process of the implementation of the dominant silt, clay soil, in addition a limited space and to analyze the results of loading tests, investigations of soil and daily reports. Construction process determines the success of the structure bore pile in high building structure bearing, without damaging a heritage building. Attainment the hard soil depth, density concrete, observable clean reinforcement in the implementation. Monitoring the implementation of, among others, the face of the ground water little reduce in the area and outside the footprint of the building, no impact of vibration drilling equipment, watching the mud content on the water coming out at the time of drilling, concrete volume was monitored each 2 m bore depth of pile, The result researched heritage building was not damage. The test results bore pile axial, lateral analyzed the results have the appropriate force design required.

  20. Interesting Developments in Testing Methods Applied to Foundation Piles

    Science.gov (United States)

    Sobala, Dariusz; Tkaczyński, Grzegorz

    2017-10-01

    Both: piling technologies and pile testing methods are a subject of current development. New technologies, providing larger diameters or using in-situ materials, are very demanding in terms of providing proper quality of execution of works. That concerns the material quality and continuity which define the integral strength of pile. On the other side we have the capacity of the ground around the pile and its ability to carry the loads transferred by shaft and pile base. Inhomogeneous nature of soils and a relatively small amount of tested piles imposes very good understanding of small amount of results. In some special cases the capacity test itself form an important cost in the piling contract. This work presents a brief description of selected testing methods and authors remarks based on cooperation with Universities constantly developing new ideas. Paper presents some experience based remarks on integrity testing by means of low energy impact (low strain) and introduces selected (Polish) developments in the field of closed-end pipe piles testing based on bi-directional loading, similar to Osterberg idea, but without sacrificial hydraulic jack. Such test is suitable especially when steel piles are used for temporary support in the rivers, where constructing of conventional testing appliance with anchor piles or kentledge meets technical problems. According to the author’s experience, such tests were not yet used on the building site but they bring a real potential especially, when the displacement control can be provided from the river bank using surveying techniques.

  1. 30 CFR 77.215 - Refuse piles; construction requirements.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Refuse piles; construction requirements. 77.215... COAL MINES Surface Installations § 77.215 Refuse piles; construction requirements. (a) Refuse deposited on a pile shall be spread in layers and compacted in such a manner so as to minimize the flow of air...

  2. The Berlin tradition in Chicago: Franz Alexander and the Chicago Institute for Psychoanalysis.

    Science.gov (United States)

    Schmidt, Erika S

    2010-01-01

    Freud considered Franz Alexander, the first graduate of the Berlin Psychoanalytic Institute and an assistant in the Berlin Polyclinic, to be "one of our strongest hopes for the future." Alexander went on to become the first director of the Chicago Institute for Psychoanalysis in 1932 and modeled some of the Chicago Institute's mission on his Berlin experiences. He was also a researcher in psychosomatic medicine, a prolific writer about psychoanalysis and prominent in psychoanalytic organizations. As he proposed modifications in psychoanalytic technique, he became a controversial figure, especially in the elaboration of his ideas about brief therapy and the corrective emotional experience. This paper puts Alexander's achievements in historical context, draws connections between the Berlin and Chicago Institutes and suggests that, despite his quarrels with traditional psychoanalysis, Alexander's legacy may be in his attitude towards psychoanalysis, characterized by a commitment to scientific study, a willingness to experiment, and a conviction about the role of psychoanalysis within the larger culture.

  3. Pressure vessel for a BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

  4. Friction effects on lateral loading behavior of rigid piles

    DEFF Research Database (Denmark)

    Zania, Varvara; Hededal, Ole

    2012-01-01

    taking into account the shear frictional resistance along the pile. For this purpose efficient three dimensional finite element models of different diameter have been developed. The increase of the side friction and of the diameter of the pile is shown to alter the failure pattern and increase...... the lateral capacity of the pile. The obtained p - y curves demonstrate the importance of the aforementioned parameters in the design of rigid piles, as the reduction of friction along the interface reduces not only the ultimate load but also the stiffness of the soil-pile response. Read More: http...

  5. In-pile loop OWL-2 and irradiation tests done with it

    International Nuclear Information System (INIS)

    Suzuki, Shinobu; Ikeshima, Yoshiaki; Kawano, Masakatsu; Watanabe, Hiroyuki; Sato, Hitoshi; Tanaka, Isao

    1990-11-01

    The OWL-2 which was built in the JMTR as the biggest water loop in Japan has been operating for irradiation service since February 1972. The desired objective of the OWL-2, contributing to the development of various nuclear fuels and materials for the light water power reactor and to reactor engineering, has been so fully achieved that the OWL-2 is planned to be dismantled. After the dismantling, a loop, needed for the research and development of the breeding blanket for the fusion reactor, is going to be installed in place of the OWL-2 as a part of the JMTR Modification Program. This paper deals with the history of the OWL-2 with an emphasis on the technical affairs taken into consideration when designing the OWL-2, the irradiation tests, development of the turbine flowmeter, results of the surveillance test of the material of the in-reactor tube, the knowledge gained in the course of the investigation into the cause of transgranular stress corrosion cracking (TGSCC) which developed in the wall of the in-reactor tube, and countermeasures taken to prevent TGSCC from recurring. (author)

  6. 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

  7. 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

  8. Global and local scour at pile groups

    DEFF Research Database (Denmark)

    Sumer, B. Mutlu; Bundgaard, Klavs; Fredsøe, Jørgen

    2005-01-01

    This paper presents the results of an experimental investigation on scour around pile groups with different configurations exposed to steady current. Two kinds of tests were carried out: rigid-bed tests and actual scour tests. In these, the mean and turbulence properties of the flow were measured...... across the pile groups. The pile-group configurations were such that the global scour was distinguished from the local scour. The results show that the global scour can be quite substantial....

  9. Protection for work in the pressure tank of the pile G2

    International Nuclear Information System (INIS)

    Chassany, J.Ph.; Rodier, J.

    1961-01-01

    While the pile was shut down after a three-month run at full power, the secondary circuit was cleaned and some alterations were carried out. The pile contained 100 tons of uranium, half of which made up the periphery and was irradiated uranium. The possibility of carrying on work inside the pressure tank was not considered at the time of construction. Because of the heat and the irradiation it was only possible to remain in the pressure tank for a limited period of time, and several operators received doses of the order of 1.5 rem. Cotton clothing gave satisfactory protection against contamination and was more comfortable than the vinyl equipment. The work lasted for 17 days and involved 881 incursions into the pressure tank. (author) [fr

  10. Uranium absorption study pile

    International Nuclear Information System (INIS)

    Raievski, V.; Sautiez, B.

    1959-01-01

    The report describes a pile designed to measure the absorption of fuel slugs. The pile is of graphite and comprises a central section composed of uranium rods in a regular lattice. RaBe sources and BF 3 counters are situated on either side of the center. A given uranium charge is compared with a specimen charge of about 560 kg, and the difference in absorption between the two noted. The sensitivity of the equipment will detect absorption variations of about a few ppm boron (10 -6 boron per gr. of uranium) or better. (author) [fr

  11. Three dimensional analysis of laterally loaded piles

    International Nuclear Information System (INIS)

    Yilmaz, C.

    1987-01-01

    In this study static analysis of laterally loaded pile is studied by the three models. The first model is the beam on discrete elastic springs. This model is analyzed using a flexibility method. The second model is the beam on a two-parameter elastic foundation. This model is analyzed using the linear finite element method. The third model is the finite element model, using the three-dimensional iso-parametric parabolic brick element. Three-dimensional pile group analysis is also performed using elastic constants of single pile obtained by any one of the above analyses. The main objective is to develop computer programs for each model related to single piles and to group analysis. Then, the deflections, rotations, moments, shears, stresses and strains of the single pile are obtained at any arbitrary point. Comparison is made between each model and with other studies such as Poulos 1971, Desai and Appel 1976. In addition, to provide a benchmark of three-dimensional finite element analysis, the Boussinesq problem is analyzed. (orig.)

  12. OSIRIS reactor radioprotection, radioprotection measurements performed during the power rise and the first 50 megawatt operation; Radioprotection de la pile OSIRIS, mesures de radioprotection effectuees au cours de la montee en puissance et des premiers fonctionnements a 50 megawatts

    Energy Technology Data Exchange (ETDEWEB)

    Fanton, B; Lebouleux, P

    1967-12-01

    The authors supply the results of the measurements that have been made near the Osiris reactor during the power increase and during the first functioning at 50 megawatts. The measurements relate to the absorbed dose rates in the premises, the water activation and the atmospheric contamination. The influence of the heat layer of water movements and the water rate in the core chimney on the absorbed dose rate at the footbridge level overhanging the pile core has been studied. The modifications to the protection devices that have been proposed after the measurements and the effect of these modifications on the results of the measures are given then. The regeneration process of a water purification chain has been examined from the radiation protection point of view. It has been possible to make some twenty radionuclides obvious in the produced effluents and to determine the volume activity of these effluents for each radionuclide. The whole of results show that in a general way, the irradiation levels are low during the usual reactor functioning. [French] Les auteurs fournissent les resultats des mesures de radioprotection oui ont ete effectuees aupres de la pile Osiris pendant la montee en puissance et au cours des premiers fonctionnements a 50 megawatts. Les mesures portent sur les debits de dose absorbee dans les locaux, l'activation de l'eau et la contamination atmospherique. L'influence de la couche chaude des mouvements d'eau et du debit d'eau dans la cheminee du coeur sur le debit de dose absorbee au niveau de la passerelle surplombant le coeur de la pile, a ete etudiee. Les modifications aux dispositifs de protection, qui ont ete proposees a la suite des mesures, et l'effet de ces modifications sur les resultats des mesures sont indiques ensuite. Le processus de regeneration d'une chaine d'epuration de l'eau a ete examine sous l'angle de la radioprotection. Il a ete possible de mettre en evidence une vingtaine de radionucleides dans les effluents produits et de

  13. Foundation heat transfer analysis for buildings with thermal piles

    International Nuclear Information System (INIS)

    Almanza Huerta, Luis Enrique; Krarti, Moncef

    2015-01-01

    Highlights: • A numerical transient thermal model for thermo-active foundations is developed. • Thermal interactions between thermal piles and building foundations are evaluated. • A simplified analysis method of thermal interactions between thermal piles and building foundations is developed. - Abstract: Thermal piles or thermo-active foundations utilize heat exchangers embedded within foundation footings to heat and/or cool buildings. In this paper, the impact of thermal piles on building foundation heat transfer is investigated. In particular, a simplified analysis method is developed to estimate the annual ground-coupled foundation heat transfer when buildings are equipped with thermal piles. First, a numerical analysis of the thermal performance of thermo-active building foundations is developed and used to assess the interactions between thermal piles and slab-on-grade building foundations. The impact of various design parameters and operating conditions is evaluated including foundation pile depth, building slab width, foundation insulation configuration, and soil thermal properties. Based on the results of a series of parametric analyses, a simplified analysis method is presented to assess the impact of the thermal piles on the annual heat fluxes toward or from the building foundations. A comparative evaluation of the predictions of the simplified analysis method and those obtained from the detailed numerical analysis indicated good agreement with prediction accuracy lower than 5%. Moreover, it is found that thermal piles can affect annual building foundation heat loss/gain by up to 30% depending on foundation size and insulation level

  14. Experimental Study on Post Grouting Bearing Capacity of Large Diameter Bored Piles

    Directory of Open Access Journals (Sweden)

    Wang Duanduan

    2015-01-01

    Full Text Available Post grouting can improve the inherent defects such as the formation of the mud cake at pile side and the sediment at pile end in the process of bored pile construction. Thus post grouting has been widely used in Engineering. The purpose of this paper is to research the influences of post grouting to pile bearing capacity more systematically and intuitively. Combined with the static load test of four test piles in Weihe River Bridge test area of new airport highway in Xi’an, the bearing capacity and settlement of routine piles and post grouting piles are comparatively analyzed. The test results show that under the same geological condition, post grouting can improve the properties of pile tip and pile shaft soil of bored piles significantly, enhance the ultimate resistance, improve the ultimate bearing capacity and reduce the pile tip settlement. Then post grouting can aim to optimize pile foundation.

  15. Experience with reactor assembly of FBTR

    International Nuclear Information System (INIS)

    Srinivasan, G.; Ravishankar, K.; Babu, A.; Varadarajan, S.; Arumugam, P.; Sekhar, P.

    2006-01-01

    Reactor Assembly, also called Block Pile, is the heart of FBTR and houses the core, top and lateral shields, control rod drive mechanisms (CRDM), sodium inlet pipe and outlet pipes etc. Two major problems which arose during commissioning were reactor vessel tilt due to convection in cover gas space and failure of inflatable seals. The reactor vessel tilt was solved by Helium injection. Reactor was operated without pressurising the inflatable seals till 2005, when the seals were replaced. Other major problems in the course of twenty years of reactor operation were failure of three CRDM lower parts, Core Cover plate which houses the core thermocouples getting stuck in the fuel handling position, water leaks from the Biological Shield Cooling (BSC) coils around the reactor, failure of core wires in the trailing cables during fuel handling etc. This paper addresses the major problems faced and modifications carried out. (author)

  16. Gamma spectrum measurement in a swimming-pool-type reactor; Mesure du spectre {gamma} d'une pile piscine

    Energy Technology Data Exchange (ETDEWEB)

    Pla, E [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1969-07-01

    After recalling the various modes of interaction of gamma rays with matter, the authors describe the design of a spectrometer for gamma energies of between 0.3 and 10 MeV. This spectrometer makes use of the Compton and pair-production effects without eliminating them. The collimator, the crystals and the electronics have been studied in detail and are described in their final form. The problem of calibrating the apparatus is then considered ; numerous graphs are given. The sensitivity of the spectrometer for different energies is determined mainly for the 'Compton effect' group. Finally, in the last part of the report, are given results of an experimental measurement of the gamma spectrum of a swimming-pool type reactor with new elements. (author) [French] Apres un rappel des differents modes d'interaction des rayons gamma avec la matiere, nous decrivons la conception d'un spectrometre pour les energies gamma s'etendant de 0,3 a 10 MeV. Ce spectrometre utilise les effets Compton et creation de paires sans les eliminer. Le collimateur, les cristaux et l'electronique sont entierement etudies et decrits dans leur realisation definitive. Ensuite, le probleme de l'etalonnage de l'appareil est envisage; de nombreuses courbes sont donnees. La sensibilite du spectrometre pour les differentes energies est determinee principalement pour le groupe ''effet Compton''. Enfin, les resultats d'une experience de mesure du spectre gamma d'une pile piscine avec elements neufs sont donnes dans la derniere partie. (auteur)

  17. The history of ZED-2

    International Nuclear Information System (INIS)

    Jones, R.

    2010-01-01

    This presentation gives the history of ZED-2 reactor at the Chalk River Laboratories. It traces the genealogy from Fermi Pile (1942) and ZEEP (1945) to the present ZED-2 reactor. ZED-2 is larger than ZEEP to allow larger critical lattices of larger CANDU type channels. There are two basic types of measurements made in ZED-2: critical size (buckling) of lattice and detailed reaction rate distribution in a cell. The presentation goes on to discuss research activities on ZED-2 from 1960 to the present.

  18. The history of ZED-2

    Energy Technology Data Exchange (ETDEWEB)

    Jones, R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2010-07-01

    This presentation gives the history of ZED-2 reactor at the Chalk River Laboratories. It traces the genealogy from Fermi Pile (1942) and ZEEP (1945) to the present ZED-2 reactor. ZED-2 is larger than ZEEP to allow larger critical lattices of larger CANDU type channels. There are two basic types of measurements made in ZED-2: critical size (buckling) of lattice and detailed reaction rate distribution in a cell. The presentation goes on to discuss research activities on ZED-2 from 1960 to the present.

  19. Modelling the behaviour of intergranular fission gas during out-of-pile annealing

    International Nuclear Information System (INIS)

    Valin, S.; Mocellin, A.; Eminet, G.; Ravel, S.

    2002-01-01

    Standard UO 2 fuel irradiated in a commercial PWR (burn-up: 14 GWd/t u ) and hyper-stoichiometric large-grained fuel irradiated in an experimental reactor (burn-up: 9 GWd/t u ) were submitted to out-of-pile annealing treatments under sweeping helium. The 85 Kr release was continuously monitored during the annealing experiments, with temperatures ranging from 1 400 to 1 760 deg C. The burst release was not systematically observed during the heating ramp but did sometimes appear during the isothermal period following the ramp. A model based on microstructural observations was developed to explain the experimental results. The release was shown to depend on gas diffusion towards grain boundaries and on temporary gas trapping in grain boundary bubbles. The gas saturation on grain boundaries, leading to release, depends on their crystallographic orientation. (authors)

  20. Structural mechanics and reactor safety

    International Nuclear Information System (INIS)

    Brandes, K.

    1983-01-01

    Operational safety and reliability of nuclear power plants widely depend on the mechanical behaviour of their structural components and their resistance to the various and complex influences. Durability and consistency of structural components are determined by the kind of strain - during the life - and by environmental conditions. The Conferences on Structural Mechanics in Reactor Technology (SMiRT) are dedicated to the discussion of such questions. The 7th of these Conferences taking place in 2-year increments was held in Chicago in August 1983. The number of contributions again increased, the number of participants slightly decreased. There are some trends in this field worth mentioning, in particular the fact that experience from design and operation of nuclear power plants now available is more and more made use of, and that more and more attention is given the problems of fusion reactors. (orig./HP) [de

  1. The high temperature out-of-pile test of LVDT for internal pressure measurement of nuclear fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Kim, D. S.; Yoon, K. B.; Sin, Y. T.; Park, S. J.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT(Linear Variable Differential Transformer). As the results of out-of-pile test at room temperature, it was concluded that the well qualified out-of-pile tests were needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation capsule, because LVDT is very sensitive to variation of temperature. Therefore, the high temperature out-of-pile test system for pressure measurement was developed, and this test was performed under the temperature condition between room temperature and 300 .deg. C increasing the pressure from 0 bar to 30 bar. The LVDT's high temperature characteristics and temperature sensitivity of LVDT were analyzed through this experiment. Based on the result of this test, the method for the application of LVDT at high temperature was introduced. It is known that the results will be used to predict accurately the internal pressure of fuel rod during irradiation test.

  2. Heave induced reduction of friction capacity of pile embedded in clays

    OpenAIRE

    Setyo Budi Gogot; Wibowo Tantri Gondo

    2017-01-01

    Installation of new piles may cause heave which influence friction capacity of existing piles. The heave can be observed from the difference in the elevation of existing pile heads recorded before and after the installation of new piles or through load-settlement diagram from Static Load Test data. This paper presents the study of bearing capacity of hollow cylindrical concrete piles with diameter of 800 mm from two projects. The piles at Project I and Project II were hydraulically jacked int...

  3. Estimating volume, biomass, and potential emissions of hand-piled fuels

    Science.gov (United States)

    Clinton S. Wright; Cameron S. Balog; Jeffrey W. Kelly

    2009-01-01

    Dimensions, volume, and biomass were measured for 121 hand-constructed piles composed primarily of coniferous (n = 63) and shrub/hardwood (n = 58) material at sites in Washington and California. Equations using pile dimensions, shape, and type allow users to accurately estimate the biomass of hand piles. Equations for estimating true pile volume from simple geometric...

  4. Influences on the thermal efficiency of energy piles

    International Nuclear Information System (INIS)

    Cecinato, Francesco; Loveridge, Fleur A.

    2015-01-01

    Energy piles have recently emerged as a viable alternative to borehole heat exchangers, but their energy efficiency has so far seen little research. In this work, a finite element numerical model is developed for the accurate 3D analysis of transient diffusive and convective heat exchange phenomena taking place in geothermal structures. The model is validated by reproducing both the outcome of a thermal response test carried out on a test pile, and the average response of the linear heat source analytical solution. Then, the model is employed to carry out a parametric analysis to identify the key factors in maximising the pile energy efficiency. It is shown that the most influential design parameter is the number of pipes, which can be more conveniently increased, within a reasonable range, compared to increasing the pile dimensions. The influence of changing pile length, concrete conductivity, pile diameter and concrete cover are also discussed in light of their energetic implications. Counter to engineering intuition, the fluid flowrate does not emerge as important in energy efficiency, provided it is sufficient to ensure turbulent flow. The model presented in this paper can be easily adapted to the detailed study of other types of geothermal structures. - Highlights: • A numerical model for 3D thermal transient analysis of energy piles is developed. • The model is validated against both field data and an analytical solution. • Key parameters are then identified for efficient thermal design of energy piles. • Energy efficiency is maximised by large pipe number and concrete conductivity. • Large exchanger fluid velocity does not have a major impact on efficiency

  5. Field Test of Driven Pile Group under Lateral Loading

    Science.gov (United States)

    Gorska, Karolina; Rybak, Jaroslaw; Wyjadlowski, Marek

    2017-12-01

    All the geotechnical works need to be tested because the diversity of soil parameters is much higher than in other fields of construction. Horizontal load tests are necessary to determine the lateral capacity of driven piles subject to lateral load. Various load tests were carried out altogether on the test field in Kutno (Poland). While selecting the piles for load tests, different load combinations were taken into account. The piles with diverse length were chosen, on the basis of the previous tests of their length and integrity. The subsoil around the piles consisted of mineral soils: clays and medium compacted sands with the density index ID>0.50. The pile heads were free. The points of support of the “base” to which the dial gauges (displacement sensors) were fastened were located at the distance of 0.7 m from the side surface of the pile loaded laterally. In order to assure the independence of measurement, additional control (verifying) geodetic survey of the displacement of the piles subject to the load tests was carried out (by means of the alignment method). The trial load was imposed in stages by means of a hydraulic jack. The oil pressure in the actuator was corrected by means of a manual pump in order to ensure the constant value of the load in the on-going process of the displacement of the pile under test. On the basis of the obtained results it is possible to verify the numerical simulations of the behaviour of piles loaded by a lateral force.

  6. Rotary peening with captive shot

    International Nuclear Information System (INIS)

    1998-02-01

    Roto Peen with captive shot removes coatings and surface contamination from concrete floors. The objective of treating radioactively contaminated concrete floors during the Deactivation and Decommissioning (D and D) process is to reduce the surface contamination levels to meet regulatory criteria for unrestricted use. The US Department of Energy (DOE) Chicago Operations office and DOE's Federal Energy Technology Center (FETC) jointly sponsored a Large-Scale Demonstration Project (LSDP) at the Chicago Pile-5 Research Reactor (CP-5) at Argonne National Laboratory-East (ANL). The objective of the LSDP is to demonstrate potentially beneficial D and D technologies in comparison with current baseline technologies. As part of the LSDP, roto Peen with captive shot was demonstrated March 17--20, 1997, to treat a 20 x 25 ft area of radioactively contaminated concrete floor on the service level of the CP-5 building

  7. Graphite for high-temperature reactors

    International Nuclear Information System (INIS)

    Hammer, W.; Leushacke, D.F.; Nickel, H.; Theymann, W.

    1976-01-01

    The different graphites necessary for HTRs are being developed, produced and tested within the Federal German ''Development Programme Nuclear Graphite''. Up to now, batches of the following graphite grades have been manufactured and fully characterized by the SIGRI Company to demonstrate reproducibility: pitch coke graphite AS2-500 for the hexagonal fuel elements and exchangeable reflector blocks; special pitch coke graphite ASI2-500 for reflector blocks of the pebble-bed reactor and as back-up material for the hexagonal fuel elements; graphite for core support columns. The material data obtained fulfill most of the requirements under present specifications. Production of large-size blocks for the permanent side reflector and the core support blocks is under way. The test programme covers all areas important for characterizing and judging HTR-graphites. In-pile testing comprises evaluation of the material for irradiation-induced changes of dimensions, mechanical and thermal properties - including behaviour under temperature cycling and creep behaviour - as well as irradiating fuel element segments and blocks. Testing out-of-pile includes: evaluation of corrosion rates and influence of corrosion on strength; strength measurements; including failure criteria. The test programme has been carried out extensively on the AS2-graphite, and the results obtained show that this graphite is suitable as HTGR fuel element graphite. (author)

  8. Proceedings of the third specialist meeting on sodium/fuel interaction in fast reactors

    International Nuclear Information System (INIS)

    1976-01-01

    This specialist meeting, sponsored by the OECD-NEA and organized by the Power Reactor and Nuclear Fuel Development Corporation, was attended by 56 delegates from 6 countries and the CEC (Commission of the European Communities). The purpose of the meeting was to bring together and discuss in depth the Fuel-Sodium Interaction, a phenomenon of major importance in the assessment of the Hypothetical Core Disruptive Accident in the Liquid Metal Fast Breeder Reactor. The meeting was essentially a follow-up of an earlier meeting held at Ispra in December 1973. In all, 29 papers were presented, covering the following topics: 1. Current perspective on sodium-fuel interaction in LMFBR safety; 2. Basic experimental and theoretical studies including other materials; 3. In-pile and out-of-pile experimental studies on sodium-fuel interaction; 4. Theoretical models for the interpretation of experiments and for application to reactor situations. The meeting is considered useful in narrowing down the chain of events necessary to get energetic interaction, large work potential, but many points are being clarified on the gap between the basic vapor explosions and the real fuel sodium interactions in the HCDA scenario of LMFBR. Finally another meeting of the same nature as this one has been recommended

  9. Characteristics of thermal neutron calibration fields using a graphite pile

    International Nuclear Information System (INIS)

    Uchita, Yoshiaki; Saegusa, Jun; Kajimoto, Yoichi; Tanimura, Yoshihiko; Shimizu, Shigeru; Yoshizawa, Michio

    2005-03-01

    The Facility of Radiation Standards of Japan Atomic Energy Research Institute is equipped with thermal neutron fields for calibrating area and personal neutron dosemeters. The fields use moderated neutrons leaked from a graphite pile in which radionuclide sources are placed. In January 2003, we have renewed the pile with some modifications in its size. In accordance with the renewal, we measured and calculated thermal neutron fluence rates, neutron energy distributions and angular distributions of the fields. The thermal neutron fluence rates of the ''inside-pile fields'' and the outside-pile fields'' were determined by the gold foil activation method. The neutron energy distributions of the outside-pile fields were also measured with the Bonner multi-sphere spectrometer system. The contributions of epithermal and fast neutrons to the total dose-equivalents were 9% in the southern outside-pile field and 12% in the western outside-pile field. The personal dose-equivalents, H p,slab (10, α), in the outside-pile fields are evaluated by considering the calculated angular distributions of incoming neutrons. The H p,slab (10, α) was found to be about 40% higher than the value in assuming the unidirectional neutron between the pile and the test point. (author)

  10. Analysis of radon protection cover on uranium tailings pile

    International Nuclear Information System (INIS)

    Zhang Zhe

    1993-01-01

    The average radon emanation rate of the whole surface over one year was used for evaluating the radon release of uranium tailings pile. The effective of radon protection cover depends on the shape and property of the tailings pile, the properties of covering and the control of air vadose in the pile. It was indicated that the covering with low diffusion coefficient, small porosity and bad permeability was suitable to cover the pile. The analytical formula of the covering layer thickness was given

  11. Chicago's Dearborn Observatory: a study in survival

    Science.gov (United States)

    Bartky, Ian R.

    2000-12-01

    The Dearborn Observatory, located on the Old University of Chicago campus from 1863 until 1888, was America's most promising astronomical facility when it was founded. Established by the Chicago Astronomical Society and directed by one of the country's most gifted astronomers, it boasted the largest telescope in the world and virtually unlimited operating funds. The Great Chicago Fire of 1871 destroyed its funding and demolished its research programme. Only via the sale of time signals and the heroic efforts of two amateur astronomers did the Dearborn Observatory survive.

  12. 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

  13. Aerial sampling of emissions from biomass pile burns in ...

    Science.gov (United States)

    Emissions from burning piles of post-harvest timber slash in Grande Ronde, Oregon were sampled using an instrument platform lofted into the plume using a tether-controlled aerostat or balloon. Emissions of carbon monoxide, carbon dioxide, methane, particulate matter (PM2.5 µm), black carbon, ultraviolet absorbing PM, elemental/organic carbon, semi-volatile organics (polycyclic aromatic hydrocarbons and polychlorinated dibenzodioxins/dibenzofurans), filter-based metals, and volatile organics were sampled for determination of emission factors. The effect on emissions from covering or not covering piles with polyethylene sheets to prevent fuel wetting was determined. Results showed that the uncovered (“wet”) piles burned with lower combustion efficiency and higher emissions of volatile organic compounds. Results for other pollutants will also be discussed. This work determined the emissions from open burning of forest slash wood, with and without plastic sheeting. The foresters advocate the use of plastic to keep the slash wood dry and aid in the controlled combustion of the slash to reduce fuel loading. Concerns about the emissions from the burning plastic prompted this work which conducted an extensive characterization of dry, wet, and dry with plastic slash pile emissions.

  14. Creating a marketplace for green roofs in Chicago

    International Nuclear Information System (INIS)

    Vitt Sale, L.; Berkshire, M.

    2004-01-01

    Since 2003, the Chicago Department of Planning and Development has been encouraging city developers to consider installing green roofs on buildings in Chicago, with the belief that this practice results in mitigation of the urban heat island effect, cleaner runoff leaving green roofs, sound attenuation, aesthetic value, oxygen production, and mitigation of carbon dioxide emissions. However, the benefits to developers, which include reduced stormwater runoff, extended roof life and energy savings, in total do not offset the first cost premium of a green roof. Despite this, and with no mandate requiring green roofs, the marketplace is growing. After seeing green roofs on a tour in Europe, the mayor of Chicago encouraged the first design and installation of a 20,300 square foot demonstration green roof in Chicago, and other city-sponsored pilot projects followed shortly after. Since then, the number of green roofs in Chicago has grown to over one million square feet. A map of Chicago showing locations of most of the projects was presented. It was suggested that lower prices for green roofs, higher energy costs and an inclination to invest in long-term strategies would accelerate the market. In an effort to engage the public in dialogue, the Department of Planning and Development held seminars to promote the benefits of green roofs . Participants had many questions about the applicability of green roofs to Chicago, expressing skepticism that Chicago's climate would provide the same benefits as in Europe. Other concerns were expressed regarding the devaluation of property values resulting from placing green roofs on buildings; doubts about roof leaks; maintenance practices; and, bugs and mold. Since the first cost premium of the system remains a question, most participants expressed interest in some kind of incentive program, but remained open-minded if benefits could be proved. 6 figs

  15. Creating a marketplace for green roofs in Chicago

    Energy Technology Data Exchange (ETDEWEB)

    Vitt Sale, L. [Wright and Co. Chicago, IL (United States); Berkshire, M. [City of Chicago, IL (United States)

    2004-07-01

    Since 2003, the Chicago Department of Planning and Development has been encouraging city developers to consider installing green roofs on buildings in Chicago, with the belief that this practice results in mitigation of the urban heat island effect, cleaner runoff leaving green roofs, sound attenuation, aesthetic value, oxygen production, and mitigation of carbon dioxide emissions. However, the benefits to developers, which include reduced stormwater runoff, extended roof life and energy savings, in total do not offset the first cost premium of a green roof. Despite this, and with no mandate requiring green roofs, the marketplace is growing. After seeing green roofs on a tour in Europe, the mayor of Chicago encouraged the first design and installation of a 20,300 square foot demonstration green roof in Chicago, and other city-sponsored pilot projects followed shortly after. Since then, the number of green roofs in Chicago has grown to over one million square feet. A map of Chicago showing locations of most of the projects was presented. It was suggested that lower prices for green roofs, higher energy costs and an inclination to invest in long-term strategies would accelerate the market. In an effort to engage the public in dialogue, the Department of Planning and Development held seminars to promote the benefits of green roofs . Participants had many questions about the applicability of green roofs to Chicago, expressing skepticism that Chicago's climate would provide the same benefits as in Europe. Other concerns were expressed regarding the devaluation of property values resulting from placing green roofs on buildings; doubts about roof leaks; maintenance practices; and, bugs and mold. Since the first cost premium of the system remains a question, most participants expressed interest in some kind of incentive program, but remained open-minded if benefits could be proved. 6 figs.

  16. Measurements of pile driving noise from control piles and noise-reduced piles at the Vashon Island ferry dock.

    Science.gov (United States)

    2017-04-01

    As part of the Washington State Department of Transportation (WSDOT) pile attenuation test program, : researchers from the University of Washington Applied Physics Laboratory (APL-UW) conducted underwater sound : measurements on 7 and 8 December 2015...

  17. Experimental Comparison of Statically and Cyclically Loaded Non-Slender Piles in Sand

    DEFF Research Database (Denmark)

    Sørensen, Søren Peder Hyldal; Ibsen, Lars Bo

    rigid form of motion. The Winkler model approach, employing p-y curves to describe the soil-pile interaction, is often employed as the design method for laterally loaded piles. The p-y curve formulation, currently recommended by the American Petroleum Institute and Det Norske Veritas, is based on tests...... on slender piles with length to diameter ratios larger than ten and outer pile diameters less than two meters. Hence, the pile tests that form the basis of the currently recommended p-y curve formulation are conducted with use of piles that exhibits a flexible behaviour, which is in contrast to the piles...... used as foundation for modern offshore wind energy converters. The aim of the present work is to investigate the pile behaviour for non-slender piles by means of small-scale testing. The pile behaviour is investigated and compared for both static and cyclic loading. When conducting small-scale tests...

  18. Theoretical study of short pile effect in tunnel excavation

    Science.gov (United States)

    Tian, Xiao-yan; Liu, Jing; Gao, Xiao-mei; Li, Yuan

    2017-09-01

    The Misaki Sato Go ideal elastoplastic model is adopted and the two stage analysis theory is used to study the effect of tunnel excavation on short pile effect in this paper. In the first stage, the free field vertical displacement of the soil at the corresponding pile location is obtained by using empirical formula. In the second stage, the displacement is applied to the corresponding pile location. The equilibrium condition of micro physical differential equation settlement of piles. Then through logical deduction and the boundary condition expressions of the settlement calculation, obtain the pile side friction resistance and axial force of the week. Finally, an engineering example is used to analyze the influence of the change of main parameters on their effects.

  19. Analysis of static and dynamic pile-soil-jacket behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Azadi, Mohammad Reza Emami

    1998-12-31

    In the offshore industry, recent extreme storms, severe earthquakes and subsidence of the foundation of jacket platforms have shown that new models and methods must take into account the jacket- pile-soil foundation interaction as well as the non-linear dynamic performance/loading effects. This thesis begins with a review of the state of art pile-soil interaction model, recognizing that most existing pile-soil models have been established based on large diameter pile tests on specific sites. The need for site independent and mechanistic pile-soil interaction models led to the development of new (t-z) and (p-y) disk models. These are validated using the available database from recent large diameter pile tests in the North Sea and Gulf of Mexico. The established static disk models are applied for non-linear static analysis of the jacket-pile-soil system under extreme wave loading. Dynamic pile-soil interaction is studied and a new disk-cone model is developed for the non-linear and non-homogeneous soils. This model is applied to both surface and embedded disks in a soil layer with non-linear properties. Simplified non-linear as well as more complex analysis methods are used to study the dynamic response of the jacket platform under extreme sea and seismic loading. Ductility spectra analysis is introduced and used to study the dynamic performance of the jacket systems near collapse. Case studies are used to illustrate the effects of structural, foundation failure characteristics as well as dynamic loading effects on the overall performance of the jacket-pile-soil systems near ultimate collapse. 175 refs., 429 figs., 70 tabs.

  20. Land of California?: The ambiguities of sweet home Chicago.

    Science.gov (United States)

    Kimsey, John

    2005-01-01

    This essay examines some historical questions and cultural constructions surrounding the song Sweet Home Chicago and its composer Robert Johnson. Noting that while the song has enjoyed long life, Johnson's lyric (describing Chicago as a land of California) has not, the essay critiques primitivist readings of Johnson while posing an African American cultural myth-Chicago as promised land of the Great Migration-as the subtext of his puzzling line. Finally, it considers whether mundane-sounding revisions of Johnson's lyric indicate a reduction in Chicago's mythic status, from safe haven to same old place.

  1. Recuperation of the energy released in the G-1, an air-cooled graphite reactor core; Recuperation de l'energie degagee dans G 1 pile a graphite refroidie a l'air

    Energy Technology Data Exchange (ETDEWEB)

    Chambadal, P [Electricite de France (EDF), 75 - Paris (France); Pascal, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    The CEA (in his five-year setting plan) has objective among others, the realization of the two first french reactors moderated with graphite. The construction of the G-1 reactor in Marcoule, first french plutonic core, is achieved so that it will diverge in the beginning of 1956 and reach its full power in the beginning of the second semester of the same year. In this report we will detail the specificities of the reactor and in particular its cooling and energy recuperation system. The G-1 reactor being essentially intended to allow the french technicians to study the behavior of an energy installation supply taking its heat in a nuclear source as early as possible. (M.B.) [French] Le Commissariat a l'Energie Atomique (dans le cadre du plan quinquennal) a entre autres objectifs, la realisation des deux premiers reacteurs francais moderes au graphite. La construction du reacteur G-1 a Marcoule, premiere pile plutonigene francaise, est realise afin qu'il puisse diverger au debut de 1956 et atteindre sa pleine puissance au debut du second semestre de la meme annee. Dans ce rapport nous detaillerons les specificites du reacteur et en particulier son systeme de refroidissement et de recuperation d'energie. Le reacteur G-1 etant essentielement destine a permettre aux techniciens francais d'etudier le plus tot possible le comportement d'une installation productrice d'energie empruntant sa chaleur a une source nucleaire. (M.B.)

  2. Out-pile test of the capsule with cone shape bottom structures

    Energy Technology Data Exchange (ETDEWEB)

    Choi, M. H.; Kang, Y. H.; Cho, M. S.; Choo, K. N.; Kim, B. G.; Son, J. M.; Park, S. J.; Shin, Y. T.; Oh, J. M

    2004-01-01

    The design modification of bottom guide structures for the instrumented capsule which is used for the irradiation test in the research reactor, HANARO is done because of the cutting trouble of the bottom guide arm's pin. The previous structure of the 3-pin arm shape is changed into one body of the cone shape. The specimens of the bottom end cap ring with three different sizes ({phi}68mm, {phi}70mm, {phi}72mm) are designed and manufactured. The out-pile test for the capsule with previous 3-pin arm and new three bottom structures of the cone shape is performed using the one-channel flow test facilities. In order to estimate the compatibility with HANARO, the structural stability and integrity of the capsule, the out-pile test such as a loading/unloading test, a pressure drop test, a thermal performance test, a displacement measurement due to a vibration and an endurance test etc. is conducted, and the outer diameter of the bottom end cap ring to meet the HANARO requirements is selected. From out-pile test results the capsule with cone shape bottom structures is evaluated as to have the structural stability and the benefit from the fluid's flow respect. Also the size satisfied various requirements among three kinds of bottom end cap rings is 70mm in diameter. It is expected that the new bottom structures of the cone shape with 70mm in diameter will be applicable to all material and special capsules which will be designed and manufactured for the purpose of irradiation tests in the future.

  3. Chicago exploration days

    International Nuclear Information System (INIS)

    Zeitler, Elmar

    2012-01-01

    Single heavy atoms supported on thin carbon film were first imaged by Crewe, Wall and Langmore with their dark-field STEM. This glimpse into a hitherto invisible world we owe undeniably to Crewe's vision and determination, and to his gift to electrify, engage and encourage talented students. Since this successful event happened during my sabbatical stay in Crewe's group, the editors of this memorial volume asked me to write an account of its early history, which I gladly composed mostly from memory. The circumstances that led to my collaboration with Albert Crew in Chicago are reviewed, and the main project that we jointly embarked on the Chicago 1 MeV STEM is described. It is shown that the project was nearing completion and would have likely been successful, had funding been continued. The paper concludes with a tribute to Albert I wrote many years ago. -- Highlights: ►► Reasons and motivations for Crewe's interest in electron microscopy are reviewed. ► Reasons and motivations for Crewe's interest in electron microscopy are reviewed. ► Early theoretical work on STEM imaging is summarized. The design of the Chicago 1 MeV Scanning Transmission Electron Microscope is described. ► Construction details are illustrated. Reasons for the project not reaching a successful conclusion are given. ► Tribute is paid to Albert Crewe.

  4. Pile-up correction by Genetic Algorithm and Artificial Neural Network

    Science.gov (United States)

    Kafaee, M.; Saramad, S.

    2009-08-01

    Pile-up distortion is a common problem for high counting rates radiation spectroscopy in many fields such as industrial, nuclear and medical applications. It is possible to reduce pulse pile-up using hardware-based pile-up rejections. However, this phenomenon may not be eliminated completely by this approach and the spectrum distortion caused by pile-up rejection can be increased as well. In addition, inaccurate correction or rejection of pile-up artifacts in applications such as energy dispersive X-ray (EDX) spectrometers can lead to losses of counts, will give poor quantitative results and even false element identification. Therefore, it is highly desirable to use software-based models to predict and correct any recognized pile-up signals in data acquisition systems. The present paper describes two new intelligent approaches for pile-up correction; the Genetic Algorithm (GA) and Artificial Neural Networks (ANNs). The validation and testing results of these new methods have been compared, which shows excellent agreement with the measured data with 60Co source and NaI detector. The Monte Carlo simulation of these new intelligent algorithms also shows their advantages over hardware-based pulse pile-up rejection methods.

  5. 40 CFR 761.347 - First level sampling-waste from existing piles.

    Science.gov (United States)

    2010-07-01

    ... existing piles. 761.347 Section 761.347 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY... from existing piles. (a) General. Sample piles that are either specifically configured for sampling... alternate sampling plan in accordance with § 761.62(c). (b) Specifically configured piles. A specifically...

  6. Some particular aspects of control in nuclear power reactors; Conception de la surete en france et influence des imperatifs de surete sur la conception des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Vathaire, F de; Vernier, Ph; Pascouet, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    This paper reviews the experience acquired in France on the question, of reactor safety. Since a special paper is being presented on reactors of the graphite gas type, the safety of the other types studied in France is discussed here: - heavy water-gas reactors, - fast neutron reactors, - water research reactors of the swimming-pool and tank types. The safety rules peculiar to the different types are explained, with emphasis on their influence on the reactor designs and on the power limits they impose. The corresponding safety studies are presented, particular stress being placed on the original work developed in these fields. Special mention is made of the experimental systems constructed for these studies: the reactor CABRI, pile loop for depressurization tests, loops outside the pile, mock-ups etc. (authors) [French] La presente communication propose une synthese de l'experience acquise en France en matiere de surete des reacteurs. Les reacteurs de la filiere graphite-gaz faisant l'objet d'une communication particuliere, on examine ici la surete des autres types de reacteurs etudies en France: - reacteurs eau lourde-gaz, - reacteurs a neutrons rapides, - reacteurs de recherche a eau des types piscines et tank. Les imperatifs de surete propres aux differentes filieres sont developpes, en mettant l'accent sur leur influence sur la conception des reacteurs et sur les limitations de puissance qu'ils entrainent. Les etudes de surete correspondantes sont presentees, en insistant plus particulierement sur les travaux originaux developpes dans ces domaines. On indique notamment les moyens d'essais qui ont ete construits pour ces etudes: le reacteur CABRI, boucle en pile pour essais de depressurisation, boucles hors pile, maquettes, etc. (auteurs)

  7. The impact of wind energy turbine piles on ocean dynamics

    Science.gov (United States)

    Grashorn, Sebastian; Stanev, Emil V.

    2016-04-01

    The small- and meso-scale ocean response to wind parks has not been investigated in the southern North Sea until now with the help of high-resolution numerical modelling. Obstacles such as e.g. wind turbine piles may influence the ocean current system and produce turbulent kinetic energy which could affect sediment dynamics in the surrounding area. Two setups of the unstructured-grid model SCHISM (Semi-implicit Cross-scale Hydroscience Integrated System Model) have been developed for an idealized channel including a surface piercing cylindrical obstacle representing the pile and a more realistic test case including four exemplary piles. Experiments using a constant flow around the obstacles and a rotating M2 tidal wave are carried out. The resulting current and turbulence patterns are investigated to estimate the influence of the obstacles on the surrounding ocean dynamics. We demonstrate that using an unstructured ocean model provides the opportunity to embed a high-resolution representation of a wind park turbine pile system into a coarser North Sea setup, which is needed in order to perform a seamless investigation of the resulting geophysical processes.

  8. The Settlement Behavior of Piled Raft Interaction in Undrained Soil

    DEFF Research Database (Denmark)

    Ghalesari, Abbasali Taghavi; Barari, Amin; Amini, Pedram Fardad

    2013-01-01

    Offshore piled raft foundations are one of the most commonly used foundations in offshore structures. When a raft foundation alone does not satisfy the design requirements, the addition of piles may improve both the ultimate load capacity and the settlement performance of the raft. In this paper......, the behavior of a piled raft on undrained soil is studied based on a series of parametric studies on the average and differential settlement of piled raft using three-dimensional finite element analysis. The settlement behavior is found to be dependent on the number of piles and raft thickness....

  9. Fast reactor programme

    International Nuclear Information System (INIS)

    Plakman, J.C.

    1982-01-01

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

  10. 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)

  11. Greenhouse gas emissions during composting of dairy manure: Delaying pile mixing does not reduce overall emissions

    Science.gov (United States)

    The effect of the timing of pile mixing on greenhouse gas (GHG) emissions during dairy manure composting was determined using large flux chambers designed to completely cover replicate pilot-scale compost piles. GHG emissions from compost piles that were mixed at 2, 3, 4, or 5 weeks after initial c...

  12. Effects of Diameter on Initial Stiffness of P-Y Curves for Large-Diameter Piles in Sand

    DEFF Research Database (Denmark)

    Sørensen, Søren Peder Hyldal; Ibsen, Lars Bo; Augustesen, Anders Hust

    2010-01-01

    is developed for slender piles with diameters up to approximately 2.0 m. Hence, the method is not validated for piles with diameters of 4–6 m. The aim of the paper is to extend the p-y curve method to large-diameter non-slender piles in sand by considering the effects of the pile diameter on the soil-pile...... interaction. Hence, a modified expression for the p-y curves for statically loaded piles in sand is proposed in which the initial slope of the p-y curves depends on the depth below the soil surface, the pile diameter and the internal angle of friction. The evaluation is based on three-dimensional numerical...... analyses by means of the commercial program FLAC3D incorporating a Mohr-Coulomb failure criterion. The numerical model is validated with laboratory tests in a pressure tank at Aalborg University....

  13. TerraPower, Bill Gates' reactor

    International Nuclear Information System (INIS)

    Guidez, J.

    2016-01-01

    TerraPower is a traveling wave reactor, it means that the reactor gradually converts non fissile material into the fuel it needs and the active part of the core progressively moves through the core leaving spent fuel behind. The last design of the TerraPower shows that it will use depleted uranium as fuel and that its core will need reloading every 10 years. Re-arrangement of the nuclear fuel will have to be made every 18 months to keep the core reactive. Metallic nuclear fuels will be used as they allow the highest breeding rates. It appears that apart from the very specific configuration of the core, the TerraPower is a reactor very similar to sodium-cooled fast reactors. Neutron transport inside traveling wave reactor core is complex and simulations show that the piling-up of fission product tends to kill the chain reaction and a continuous neutron addition may be necessary to keep the reactor going. A large part of the TerraPower feasibility studies concerns neutron transport inside its core. (A.C.)

  14. Pile mixing increases greenhouse gas emissions during composting of dairy manure

    Science.gov (United States)

    The effect of pile mixing on greenhouse gas (GHG) emissions from stored dairy manure was determined using large flux chambers designed to completely cover pilot-scale manure piles. GHG emissions from piles that were mixed four times during the 80 day trial were about 20% higher than unmixed piles. ...

  15. Performance of water cooled nuclear power reactor fuels in India – Defects, failures and their mitigation

    International Nuclear Information System (INIS)

    Ganguly, Chaitanyamoy

    2015-01-01

    Water cooled and moderated nuclear power reactors account for more than 95% of the operating reactors in the world today. Light water reactors (LWRs) consisting of pressurized water reactor (PWR), their Russian counterpart namely VVER and boiling water reactor (BWR) will continue to dominate the nuclear power market. Pressurized heavy water reactor (PHWR), also known as CANDU, is the backbone of the nuclear power program in India. Updates on LWR and PHWR fuel performance are being periodically published by IAEA, OECD-NEA and the World Nuclear Association (WNA), highlighting fuel failure rate and the mitigation of fuel defects and failures. These reports clearly indicate that there has been significant improvement in in – pile fuel performance over the years and the present focus is to achieve zero fuel failure in high burn up and high performance fuels. The present paper summarizes the status of PHWR and LWR fuel performance in India, highlighting the manufacturing and the related quality control and inspection steps that are being followed at the PHWR fuel fabrication plant in order to achieve zero manufacturing defect which could contribute to achieving zero in – pile failure rate in operating and upcoming PHWR units in India. (author)

  16. Kinematic seismic response of piles in layered soil profile

    International Nuclear Information System (INIS)

    Ahmad, I.; Khan, A.N.

    2006-01-01

    This paper is aimed at highlighting the importance of Kinematic Seismic Response of Piles, a phenomenon often ignored in dynamic analysis. A case study is presented where the end bearing pile is embedded in two layer soil system of highly contrasting stiffnesses; a typical case where kinematic loading plays important role. The pile soil system is modeled as continuous system and as discrete parameter system; both are based on BDWF (Beam on Dynamic Winkler Foundation) formulation. For discrete parameter system, a finite element software SAP2000 is used and the modeling technique of kinematic interaction in finite element software is discussed. For pile soil system modeled as continuous system, a general MATLAB code is developed capable of performing elastic site response analysis in two layer soil system, solving differential equation governing kinematic interaction, and giving as output the maximum ground displacement, maximum pile displacement, rotation, moment and shear distribution along pile length. The paper concludes that kinematic seismic actions must be evaluated particularly at the interface of soil layers of significantly differing soil stiffnesses. (author)

  17. Design phase identification of high pile rebound soils : final report

    Science.gov (United States)

    2010-12-15

    An engineering problem has occurred when installing displacement piles in certain soils. During driving, piles are rebounding excessively during each hammer blow, causing delay and as a result may not achieve the required design capacities. Piles dri...

  18. Backfilling of a Scour Hole around a Pile in Waves and Current

    DEFF Research Database (Denmark)

    Sumer, B. Mutlu; Petersen, Thor Ugelvig; Locatelli, Luca

    2013-01-01

    This paper presents the results of an experimental investigation of the backfilling of scour holes around circular piles. Scour holes around a pile are generated either by a current or a wave. Subsequently, the flow climate is changed from current to wave, combined waves and current, or wave...... around the pile for the same wave (or combined waves and current) climate. The time scale of backfilling has been determined as a function of three parameters, namely, (1) the Keulegan-Carpenter number of the initial wave or current (which generates the initial scour hole); (2) that of the subsequent...

  19. Power reactor core safety research

    International Nuclear Information System (INIS)

    Rim, C.S.; Kim, W.C.; Shon, D.S.; Kim, J.

    1981-01-01

    As a part of nuclear safety research program, a project was launched to develop a model to predict fuel failure, to produce the data required for the localizaton of fuel design and fabrication technology, to establish safety limits for regulation of nuclear power plants and to develop reactor operation method to minimize fuel failure through the study of fuel failure mechanisms. During 1980, the first year of this project, various fuel failure mechanisms were analyzed, an experimental method for out-of-pile tests to study the stress corrosion cracking (SCC) behaviour of Zircaloy cladding underiodine environment was established, and characteristics of PWR and CANDU Zircaloy specimens were examined. Also developed during 1980 were the methods and correlations to evaluate fuel failures in the reactor core based on operating data from power reactors

  20. The Japanese aerial attack on Hanford Engineer Works

    Science.gov (United States)

    Clark, Charles W.

    The day before the Pearl Harbor attack, December 6, 1941, the University of Chicago Metallurgical Laboratory was given four goals: design a plutonium (Pu) bomb; produce Pu by irradiation of uranium (U); extract Pu from the irradiated U; complete this in time to be militarily significant. A year later the first controlled nuclear chain reaction was attained in Chicago Pile 1 (CP-1). In January 1943, Hanford, WA was chosen as the site of the Pu factory. Neutron irradiation of 238U was to be used to make 239Pu. This was done by a larger version of CP-1, Hanford Reactor B, which went critical in September 1944. By July 1945 it had made enough Pu for two bombs: one used at the Trinity test in July; the other at Nagasaki, Japan in August. I focus on an ironic sidelight to this story: disruption of hydroelectric power to Reactor B by a Japanese fire balloon attack on March 10, 1945. This activated the costly coal-fired emergency backup plant to keep the reactor coolant water flowing, thwarting disaster and vindicating the conservative design of Hanford Engineer Works. Management of the Hanford Engineer Works in World War II, H. Thayer (ASCE Press 1996).

  1. Spontaneous vegetation on overburden piles in the Coal Basin of Santa Catarina, Brazil

    Energy Technology Data Exchange (ETDEWEB)

    dos Santos, R.; Citadini-Zanette, V.; Leal-Filho, L.S.; Hennies, W.T. [University of Extremo Sul Catarinense, Criciuma (Brazil)

    2008-09-15

    The objective of this work was to select indigenous vegetal species for restoration programs aiming at the regeneration of ombrophilous dense forest. Thirty-five spoil piles located in the county of Sideropolis, Santa Catarina, that received overburden disposal for 39 years (1950-1989) were selected for study because they exhibited remarkable spontaneous regrowth of trees compared to surrounding spoil piles. Floristic inventory covered the whole area of the 35 piles, whereas survey on phytosociology and natural regeneration studies were conducted in 70 plots distributed along the 35 piles. Floristic inventory recorded 83 species from 28 botanical families. Herbaceous terricolous plants constituted the predominant species (47.0%), followed by shrubs (26.5%), trees (19.3%), and vines (7.2%). Severe chemical (acidic pH and lack of nutrients) and physical (coarse substrate and slope angle of 40-50{sup o} characteristics displayed by the overburden piles constituted limitations to floristic diversity and size of indigenous trees, indicating the need for substrate reclamation prior to forest restoration.

  2. Processing Satellite Imagery To Detect Waste Tire Piles

    Science.gov (United States)

    Skiles, Joseph; Schmidt, Cynthia; Wuinlan, Becky; Huybrechts, Catherine

    2007-01-01

    A methodology for processing commercially available satellite spectral imagery has been developed to enable identification and mapping of waste tire piles in California. The California Integrated Waste Management Board initiated the project and provided funding for the method s development. The methodology includes the use of a combination of previously commercially available image-processing and georeferencing software used to develop a model that specifically distinguishes between tire piles and other objects. The methodology reduces the time that must be spent to initially survey a region for tire sites, thereby increasing inspectors and managers time available for remediation of the sites. Remediation is needed because millions of used tires are discarded every year, waste tire piles pose fire hazards, and mosquitoes often breed in water trapped in tires. It should be possible to adapt the methodology to regions outside California by modifying some of the algorithms implemented in the software to account for geographic differences in spectral characteristics associated with terrain and climate. The task of identifying tire piles in satellite imagery is uniquely challenging because of their low reflectance levels: Tires tend to be spectrally confused with shadows and deep water, both of which reflect little light to satellite-borne imaging systems. In this methodology, the challenge is met, in part, by use of software that implements the Tire Identification from Reflectance (TIRe) model. The development of the TIRe model included incorporation of lessons learned in previous research on the detection and mapping of tire piles by use of manual/ visual and/or computational analysis of aerial and satellite imagery. The TIRe model is a computational model for identifying tire piles and discriminating between tire piles and other objects. The input to the TIRe model is the georeferenced but otherwise raw satellite spectral images of a geographic region to be surveyed

  3. Thermal-hydraulic tests with out-of-pile test facility for BOCA development

    International Nuclear Information System (INIS)

    Kitagishi, Shigeru; Aoyama, Masashi; Tobita, Masahiro; Inaba, Yoshitomo; Yamaura, Takayuki

    2012-01-01

    The fuel transient test facility was prepared for power ramping tests of light-water-reactor (LWR) fuels in the Japan Materials Testing Reactor (JMTR) under a contract project with the Nuclear Industrial Safety Agent (NISA) of the Ministry of Economy, Trade and Industry (METI). It is necessary to develop high accuracy analysis procedure for power ramping tests after restart of the JMTR. The out-of-pile test facility to simulate thermal-hydraulic conditions of the fuel transient test facility was therefore developed. Applicability of the analysis code ACE-3D was examined for thermal-hydraulic analysis of power ramping tests for 10x10 BWR fuels by the fuel transient test facility. As the results, the calculated temperature was 304°C in comparison with measured value of 304.9-317.4°C in the condition of 600 W/cm. There is a bright prospect of high accuracy power ramping tests by the fuel transient test facility in JMTR. (author)

  4. Optimal Design of Sheet Pile Wall Embedded in Clay

    Science.gov (United States)

    Das, Manas Ranjan; Das, Sarat Kumar

    2015-09-01

    Sheet pile wall is a type of flexible earth retaining structure used in waterfront offshore structures, river protection work and temporary supports in foundations and excavations. Economy is an essential part of a good engineering design and needs to be considered explicitly in obtaining an optimum section. By considering appropriate embedment depth and sheet pile section it may be possible to achieve better economy. This paper describes optimum design of both cantilever and anchored sheet pile wall penetrating clay using a simple optimization tool Microsoft Excel ® Solver. The detail methodology and its application with examples are presented for cantilever and anchored sheet piles. The effects of soil properties, depth of penetration and variation of ground water table on the optimum design are also discussed. Such a study will help professional while designing the sheet pile wall penetrating clay.

  5. 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.)

  6. A performance case study of energy pile foundation at Rosborg Gymnasium (Denmark)

    DEFF Research Database (Denmark)

    Pagola, Maria Alberdi; Jensen, Rasmus Lund; Poulsen, Søren Erbs

    2016-01-01

    The Rosborg Gymnasium building in Vejle (Denmark) is partially founded on 200 foundation pile heat exchangers (energy piles). The thermo-active foundation has supplemented the heating and free cooling needs of the building since 2011 (4,000 m2 living area). Operational data from the ground source...... ground heat extraction/injection activity , this paper provides a performance study of the energy pile-based ground source heat pump installation utilising operational data. The study demonstrates that the measured seasonal performance factors so far are lower than expected: 2.7 in heating mode and 4.......2 in cooling mode. Nevertheless, there is room for improvement if novel energy management strategies are applied. This highlights the relevance of considering the daily heating/cooling requirements of the building during the design phase of the heating and cooling system. Moreover, this study demonstrates...

  7. Design criteria of out-pile system of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1997-07-01

    The objective of HANARO aims at the development and localization of nuclear technologies through the engineering tests. Thus it is very important the design and installation of the irradiation test facilities to be installed at the irradiation hole for verification test of the fuel performance are in connection with maximization of the utilization of HANARO. The principle subjects of this study are to presend and informed the detail design criteria and technical specification of out-pile system of HANARO fuel test loop for the developing of the fuel and reactor material. This results will become guidance for the planning of the irradiation testing using the HANARO fuel test loop. (author). 16 refs., 31 tabs., 9 figs.

  8. Chicago exploration days

    Energy Technology Data Exchange (ETDEWEB)

    Zeitler, Elmar, E-mail: zr@fhi-berlin.mpg.de [Fritz-Haber-Institut der Max-Planck-Gesellschaft Faradayweg 4-6 D-1000 Berlin 33 (Dahlem) (Germany)

    2012-12-15

    Single heavy atoms supported on thin carbon film were first imaged by Crewe, Wall and Langmore with their dark-field STEM. This glimpse into a hitherto invisible world we owe undeniably to Crewe's vision and determination, and to his gift to electrify, engage and encourage talented students. Since this successful event happened during my sabbatical stay in Crewe's group, the editors of this memorial volume asked me to write an account of its early history, which I gladly composed mostly from memory. The circumstances that led to my collaboration with Albert Crew in Chicago are reviewed, and the main project that we jointly embarked on the Chicago 1 MeV STEM is described. It is shown that the project was nearing completion and would have likely been successful, had funding been continued. The paper concludes with a tribute to Albert I wrote many years ago. -- Highlights: Black-Right-Pointing-Pointer Black-Right-Pointing-Pointer Reasons and motivations for Crewe's interest in electron microscopy are reviewed. Black-Right-Pointing-Pointer Reasons and motivations for Crewe's interest in electron microscopy are reviewed. Black-Right-Pointing-Pointer Early theoretical work on STEM imaging is summarized. The design of the Chicago 1 MeV Scanning Transmission Electron Microscope is described. Black-Right-Pointing-Pointer Construction details are illustrated. Reasons for the project not reaching a successful conclusion are given. Black-Right-Pointing-Pointer Tribute is paid to Albert Crewe.

  9. Pile group program for full material modeling and progressive failure.

    Science.gov (United States)

    2008-12-01

    Strain wedge (SW) model formulation has been used, in previous work, to evaluate the response of a single pile or a group of piles (including its : pile cap) in layered soils to lateral loading. The SW model approach provides appropriate prediction f...

  10. Safety and dose management during decommissioning of a fire damaged nuclear reactor

    International Nuclear Information System (INIS)

    Pomfret, D.G.

    2000-01-01

    Windscale Piles 1 and 2 in Cumbria in the UK were constructed in the early 1950s. They were not intended to produce electricity but were for military purposes only. They were graphite-moderated, air-cooled reactors with horizontal fuel channels fuelled with uranium rods clad in finned aluminium. In October 1957, Windscale Pile 1 suffered a core fire during a planned release of Wigner energy and both Piles were subsequently closed down. Following the fire, Pile 2 was defuelled entirely and as much fuel as was possible was removed from Pile 1. However, it is estimated that up to 15 Te of fuel remains in the core, a large proportion of which is located in the central Fire Affected Zone (FAZ). The condition of the fuel and graphite moderator in the FAZ is not known. It is possible that the moderator could contain voidage, greatly reduced density graphite and fused materials in a disordered matrix. It is probable that there is residual Wigner energy in the graphite. Use of water as part of the attempt to extinguish the fire, together with the possibility that the solidification of molten materials or other local sealing mechanisms could have excluded air mean that uranium hydrides, carbides and other pyrophorics may be present within the core. Pile 1 is now being decommissioned and a considerable amount of preparatory work has already been carried out during Phase 1. Phase 2 decommissioning, which will remove the residual fuel, the moderator and associated steelwork, condition the wastes and place them in a purpose built store is now underway. This work will be carried out for the site licensee, the United Kingdom Energy Authority, by a consortium of British Nuclear Fuels plc, Rolls Royce and NUKEM. This paper will describe the methods to be used to decommission Pile 1 and the systems and procedures which will be used to ensure that it is done safely and with the lowest reasonably practicable environmental impact. It will also describe the methods which will be used to

  11. Piles, tabs and overlaps in navigation among documents

    DEFF Research Database (Denmark)

    Jakobsen, Mikkel Rønne; Hornbæk, Kasper

    2010-01-01

    Navigation among documents is a frequent, but ill supported activity. Overlapping or tabbed documents are widespread, but they offer limited visibility of their content. We explore variations on navigation support: arranging documents with tabs, as overlapping windows, and in piles. In an experim......Navigation among documents is a frequent, but ill supported activity. Overlapping or tabbed documents are widespread, but they offer limited visibility of their content. We explore variations on navigation support: arranging documents with tabs, as overlapping windows, and in piles....... In an experiment we compared 11 participants’ navigation with these variations and found strong task effects. Overall, overlapping windows were preferred and their structured layout worked well with some tasks. Surprisingly, tabbed documents were efficient in tasks requiring simply finding a document. Piled...... on document navigation and its support by piling....

  12. Pile-Reinforcement Behavior of Cohesive Soil Slopes: Numerical Modeling and Centrifuge Testing

    Directory of Open Access Journals (Sweden)

    Liping Wang

    2013-01-01

    Full Text Available Centrifuge model tests were conducted on pile-reinforced and unreinforced cohesive soil slopes to investigate the fundamental behavior and reinforcement mechanism. A finite element analysis model was established and confirmed to be effective in capturing the primary behavior of pile-reinforced slopes by comparing its predictions with experimental results. Thus, a comprehensive understanding of the stress-deformation response was obtained by combining the numerical and physical simulations. The response of pile-reinforced slope was indicated to be significantly affected by pile spacing, pile location, restriction style of pile end, and inclination of slope. The piles have a significant effect on the behavior of reinforced slope, and the influencing area was described using a continuous surface, denoted as W-surface. The reinforcement mechanism was described using two basic concepts, compression effect and shear effect, respectively, referring to the piles increasing the compression strain and decreasing the shear strain of the slope in comparison with the unreinforced slope. The pile-soil interaction induces significant compression effect in the inner zone near the piles; this effect is transferred to the upper part of the slope, with the shear effect becoming prominent to prevent possible sliding of unreinforced slope.

  13. Design Optimization of Piles for Offshore Wind Turbine Jacket Foundations

    DEFF Research Database (Denmark)

    Sandal, Kasper; Zania, Varvara

    Numerical methods can optimize the pile design. The aim of this study is to automatically design optimal piles for offshore wind turbine jacket foundations (Figure 1). Pile mass is minimized with constraints on axial and lateral capacity. Results indicate that accurate knowledge about soil...

  14. Characterization of Odorant Compounds from Mechanical Aerated Pile Composting and Static Aerated Pile Composting.

    Science.gov (United States)

    Kumari, Priyanka; Lee, Joonhee; Choi, Hong-Lim

    2016-04-01

    We studied airborne contaminants (airborne particulates and odorous compounds) emitted from compost facilities in South Korea. There are primarily two different types of composting systems operating in Korean farms, namely mechanical aerated pile composting (MAPC) and aerated static pile composting (SAPC). In this study, we analyzed various particulate matters (PM10, PM7, PM2.5, PM1, and total suspended particles), volatile organic compounds and ammonia, and correlated these airborne contaminants with microclimatic parameters, i.e., temperature and relative humidity. Most of the analyzed airborne particulates (PM7, PM2.5, and PM1) were detected in high concentration at SAPC facilities compered to MAPC; however these differences were statistically non-significant. Similarly, most of the odorants did not vary significantly between MAPC and SAPC facilities, except for dimethyl sulfide (DMS) and skatole. DMS concentrations were significantly higher in MAPC facilities, whereas skatole concentrations were significantly higher in SAPC facilities. The microclimate variables also did not vary significantly between MAPC and SAPC facilities, and did not correlate significantly with most of the airborne particles and odorous compounds, suggesting that microclimate variables did not influence their emission from compost facilities. These findings provide insight into the airborne contaminants that are emitted from compost facilities and the two different types of composting agitation systems.

  15. OSIRIS reactor radioprotection, radioprotection measurements performed during the power rise and the first 50 megawatt operation; Radioprotection de la pile OSIRIS, mesures de radioprotection effectuees au cours de la montee en puissance et des premiers fonctionnements a 50 megawatts

    Energy Technology Data Exchange (ETDEWEB)

    Fanton, B.; Lebouleux, P

    1967-12-01

    The authors supply the results of the measurements that have been made near the Osiris reactor during the power increase and during the first functioning at 50 megawatts. The measurements relate to the absorbed dose rates in the premises, the water activation and the atmospheric contamination. The influence of the heat layer of water movements and the water rate in the core chimney on the absorbed dose rate at the footbridge level overhanging the pile core has been studied. The modifications to the protection devices that have been proposed after the measurements and the effect of these modifications on the results of the measures are given then. The regeneration process of a water purification chain has been examined from the radiation protection point of view. It has been possible to make some twenty radionuclides obvious in the produced effluents and to determine the volume activity of these effluents for each radionuclide. The whole of results show that in a general way, the irradiation levels are low during the usual reactor functioning. [French] Les auteurs fournissent les resultats des mesures de radioprotection oui ont ete effectuees aupres de la pile Osiris pendant la montee en puissance et au cours des premiers fonctionnements a 50 megawatts. Les mesures portent sur les debits de dose absorbee dans les locaux, l'activation de l'eau et la contamination atmospherique. L'influence de la couche chaude des mouvements d'eau et du debit d'eau dans la cheminee du coeur sur le debit de dose absorbee au niveau de la passerelle surplombant le coeur de la pile, a ete etudiee. Les modifications aux dispositifs de protection, qui ont ete proposees a la suite des mesures, et l'effet de ces modifications sur les resultats des mesures sont indiques ensuite. Le processus de regeneration d'une chaine d'epuration de l'eau a ete examine sous l'angle de la radioprotection. Il a ete possible de mettre en evidence une vingtaine

  16. Finite Element Investigations on the Interaction between a Pile and Swelling Clay

    DEFF Research Database (Denmark)

    Kaufmann, Kristine Lee; Nielsen, Benjaminn Nordahl; Augustesen, Anders Hust

    of Little Belt Clay. The case study involves a circular concrete pile installed in clay immediately after an excavation. The influence of the swelling soil on the soil–pile interaction and the internal pile forces are analysed by solely observing the upper pile part positioned in the swelling zone...... of the surrounding soil implies upward shear stresses at the soil–pile interface leading to tensile vertical stresses in the pile. In the current case, they exceed the tensile strength of concrete. The tensile vertical stresses peak after 35-50 years. However, the heave of the soil continues for additional 300 years....... It appears that the development of plastic interface implies the shrinkage of the pile....

  17. Some particular problems put by operating experimental reactors

    International Nuclear Information System (INIS)

    Candiotti, C.; Mabeix, R.; Uguen, R.

    1960-01-01

    On basis of a six years experience in operating research reactors, the authors explain, first, the difference in their utilization between these piles and another similar ones and, after, in consequence, they set off corresponding servitudes. These servitudes put very particular problems in operating itself, maintenance, modifications or additions on these apparatus. (author) [fr

  18. CONDUCTING AND ANALYZING THE RESULTS OF THE EXPERIMENTAL BOX TEST OF RETAINING WALL MODELS WITHOUT PILES AND ON THE PILE FOUNDATION

    Directory of Open Access Journals (Sweden)

    M. A. Lisnevskyi

    2015-08-01

    Full Text Available Purpose. Taking into consideration that the bearing capacity of the foundation may be insufficient, in the study it is assumed that pile foundation can be used to reduce the impact of the construction of new retaining structures on roads and railways near the existing buildings or in areas of dense urban development and ensure the stability of the foundation. To reduce the volume of excavation it is necessary to choose the economic structure of the retaining wall. To do this, one should explore stress-strain state (SSS of the retaining walls, to develop methods to improve their strength and stability, as well as to choose the most appropriate method of their analysis. Methodology. In the design of retaining walls foundation mat and piles are considered as independent elements. Since the combined effect of the retaining wall, piles and foundation mat as well as the effect of soil or rock foundation on the structure are considered not fully, so there are some limitations in the existing design techniques. To achieve the purpose the box tests of retaining walls models without piles and with piles for studying their interaction with the surrounding soil massif were conducted. Findings. Laboratory simulation of complex systems «surrounding soil – retaining wall – pile» was carried out and on the basis of the box test results were analyzed strains and its main parameters of the stress-strain state. Analysis of the results showed that the structure of a retaining wall with piles is steady and stable. Originality. So far, in Ukraine has not been carried out similar experimental box tests with models of retaining walls in such combinations. In the article has been presented unique photos and test results, as well as their analysis. Practical value. Using the methodology of experimental tests of the retaining wall models with piles and without them gives a wider opportunity to study stress-strain state of such structures.

  19. Assessing the Extent of Sediment Contamination Around Creosote-treated Pilings Through Chemical and Biological Analyses

    Science.gov (United States)

    Stefansson, E. S.

    2008-12-01

    Creosote is a common wood preservative used to treat marine structures, such as docks and bulkheads. Treated dock pilings continually leach polycyclic aromatic hydrocarbons (PAHs) and other creosote compounds into the surrounding water and sediment. Over time, these compounds can accumulate in marine sediments, reaching much greater concentrations than those in seawater. The purpose of this study was to assess the extent of creosote contamination in sediments, at a series of distances from treated pilings. Three pilings were randomly selected from a railroad trestle in Fidalgo Bay, WA and sediment samples were collected at four distances from each: 0 meters, 0.5 meters, 1 meter, and 2 meters. Samples were used to conduct two bioassays: an amphipod bioassay (Rhepoxynius abronius) and a sand dollar embryo bioassay. Grain size and PAH content (using a fluorometric method) were also measured. Five samples in the amphipod bioassay showed significantly lower effective survival than the reference sediment. These consisted of samples closest to the piling at 0 and 0.5 meters. One 0 m sample in the sand dollar embryo bioassay also showed a significantly lower percentage of normal embryos than the reference sediment. Overall, results strongly suggest that creosote-contaminated sediments, particularly those closest to treated pilings, can negatively affect both amphipods and echinoderm embryos. Although chemical data were somewhat ambiguous, 0 m samples had the highest levels of PAHs, which corresponded to the lowest average survival in both bioassays. Relatively high levels of PAHs were found as far as 2 meters away from pilings. Therefore, we cannot say how far chemical contamination can spread from creosote-treated pilings, and at what distance this contamination can still affect marine organisms. These results, as well as future research, are essential to the success of proposed piling removal projects. In addition to creosote-treated pilings, contaminated sediments must

  20. Hybrid pulse pile-up rejection system as applied to Rutherford backscattering

    International Nuclear Information System (INIS)

    Boie, R.A.; Wildnauer, K.R.

    1977-01-01

    The problems of pulse on pulse pile-up and noise limited pile-up rejectors are considered in detail for Rutherford backscattering spectra. The forms of these spectra allow the distortions from pile-up and the residual pile-up after rejection to be understood via a simple model. Extended calculations allow us to predict the effects quite accurately. A new pile-up rejection system is described. The ''linear'' rejection method is implemented with peak stretchers and advantageously combined with an event counting rejector to provide a versatile high performance system