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Sample records for temperature transients pwr

  1. Twenty-five years of transient counting experience in French PWR units

    Energy Technology Data Exchange (ETDEWEB)

    Barthelet, B. [Electricite de France (EDF DPN), 93 - Saint-Denis (France); Savoldelli, D.; Fritz, R. [Electricite de France (EDF DPN), 93 - Noisy le Grand (France)

    2001-07-01

    For nearly twenty five years, EDF has been checking that the actual operating transients are neither more severe nor more numerous than the design basis transients. This activity of transient cycle counting and bookkeeping has enabled EDF to own a database of more than 800 reactor.years for the PWR units. The current method of transient cycle counting is presented. In the paper, we will point out the main results of transient cycle counting and lessons learned. In general, the frequencies of transients are lower than the design frequencies. In few cases, they are higher, such as the transient frequencies of the RCS lines connected to auxiliary systems often due to operating procedures or particular periodic testing. Few periodic tests were not taken into account in the design basis transient file ; they have been detected thanks to the transient cycle counting. In the last 1980's, we achieved the first updating of the design basis transient file for the PWR 900 MWe series. In the early 1990's, we updated the design basis transient file of the PWR 1300 MWe series. In fact, since design and start-up, the operating conditions have been modified (fuel cycle with stretch-out, modification of the hot leg and cold leg temperatures for the PWR 1300 MWe,...). This was the cause of many unclassified transients. In the new design basis transient file, we have created new transients and increased the frequencies of some of them. This has enabled to consider the updated design basis transient file more representative of actual operating transients. For some years, we have increasingly associated the operators with the transient cycle counting concern. We noticed progress (decreased frequencies of most transients). (authors)

  2. Analysis of transient heat conduction in a PWR fuel rod by an improved lumped parameter approach

    Energy Technology Data Exchange (ETDEWEB)

    Dourado, Eneida Regina G. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Cotta, Renato M. [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Mecanica; Jian, Su, E-mail: eneidadourado@gmail.com, E-mail: sujian@nuclear.ufrj.br, E-mail: cotta@mecanica.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-07-01

    This paper aims to analyze transient heat conduction in a nuclear fuel rod by an improved lumped parameter approach. One-dimensional transient heat conduction is considered, with the circumferential symmetry assumed and the axial conduction neglected. The thermal conductivity and specific heat in the fuel pellet are considered temperature dependent, while the thermophysical properties of the cladding are considered constant. Hermite approximation for integration is used to obtain the average temperature and heat flux in the radial direction. Significant improvement over the classical lumped parameter formulation has been achieved. The proposed model can be also used in dynamic analysis of PWR and nuclear power plant simulators. (author)

  3. Transient performance and design aspects of low boron PWR cores with increased utilization of burnable absorbers

    Energy Technology Data Exchange (ETDEWEB)

    Papukchiev, Angel [GRS mbH Forschungsinstitute, Garching (Germany); Schaefer, Anselm [ISaR GmbH, Garching (Germany)

    2008-07-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As high boron concentrations have significant impact on reactivity feedback properties and core transient behaviour, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In order to assess the potential advantages of such strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 (Gd) and 805 (Er) ppm. An innovative low boron core design methodology was implemented combining a simplified reactivity balance search procedure with a core design approach based on detailed 3D diffusion calculations. Fuel cross sections needed for nuclear libraries were generated using the 2D lattice code HELIOS [2] and full core configurations were modelled with the 3D diffusion code QUABOX/CUBBOX [3]. For dynamic 3D calculations, the coupled code system ATHLET - QUABOX/CUBBOX was used [4]. The new cores meet German acceptance criteria regarding stuck rod, departure from nucleate boiling ratio (DNBR), shutdown margin, and maximal linear power. For the assessment of potential safety advantages of the new cores, comparative analyses were performed for three PWR core designs: the already mentioned two low boron designs and a standard design. The improved safety performance of the low boron cores in anticipated transients without scram (ATWS), boron dilution scenarios and beyond design basis accidents (BDBA) has already been reported in [1, 2 and 3]. This paper gives a short reminder on the results obtained. Moreover, it deals not only with the potential advantages, but also addresses the drawbacks of the new PWR configurations - complex core design, increased power

  4. Solution of the stationary state of the PWR MOX/UO-2 core transient benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Seubert, A.; Langenbuch, S.; Zwermann, W. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS mbH, Forschungsinstitute, D-85748 Garching (Germany)

    2006-07-01

    The multi-group Discrete Ordinates transport code DORT is applied to solve the stationary state of the OECD/NEA PWR MOX/UO-2 Core Transient Benchmark. Pin cell homogenised cross sections in 16 energy groups and P{sub 1} scattering order have been obtained by fuel assembly burn-up calculations using HELIOS. In this paper, we report on the details of our calculations for this benchmark problem and show our results to be in good agreement with an MCNP Monte Carlo solution with nuclear point data and a multi-group DeCART Method of Characteristics solution. (authors)

  5. Characterization of Decommissioned PWR Vessel Internals Materials Samples: Material Certification, Fluence, and Temperature (Nonproprietary Version)

    Energy Technology Data Exchange (ETDEWEB)

    M. Krug; R. Shogan; A. Fero; M. Snyder

    2004-11-01

    Pressurized water reactor (PWR) cores, operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs require detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel. This report contains basic material characterization information of the as-installed samples of reactor internals material which were harvested from a decommissioned PWR.

  6. TRAC analyses of severe overcooling transients for the Oconee-1 PWR

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    Ireland, J R [comp.

    1985-05-01

    This report describes the results of several Transient Reactor Analysis Code (TRAC)-PF1 calculations of overcooling transients in a Babcock and Wilcox lowered-loop, pressurized water reactor (Oconee-1). The purpose of this study is to provide detailed input on thermal-hydraulic data to Oak Ridge National Laboratory for pressurized thermal-shock analyses. The transient calculations performed were plant specific in that details of the primary system, the secondary system, and the plant-integrated control system of Oconee-1 were included in the TRAC input model. The results of the calculations indicate that the turbine-bypass valve failure transient was the most severe in terms of resulting in relatively cold liquid temperatures in the downcomer region of the vessel. The power-operated relief valve loss-of-coolant accident transient was the least severe in terms of downcomer liquid temperatures because of vent-valve fluid mixing and near-saturated conditions in the primary system. It is recommended that future calculations consider a wider range of operator actions to cover the spectra of overcooling transient sequences more completely. 6 refs., 287 figs., 32 tabs.

  7. Effects of generation and optimization of libraries of effective sections in the analysis of transient in PWR reactors; Efectos de generacion y optimizacion de librerias de secciones eficaces en el analisis de transitorios en reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Cervera, S.; Garcia Herranz, N.; Cuervo, D.; Ahnert, C.

    2014-07-01

    In this paper evaluates the impact that has a certain mesh on a transient in a PWR reactor in the expulsion of a control bar. Have been used for this purpose the coupled codes neutronic and Thermo-hydraulic COBAYA3/COBRA-TF. This objective has been chosen the OECD/NEA PWR MOX/UO{sub 2} rod ejection transient benchmark provides isotopic compositions and defined geometric configurations that allow the use of codes lattice to generate own bookstores. The code used for this transport has been the code APOLLO2.8. The results show large discrepancies when using the benchmark library or libraries own by comparing them to the other participants solutions. The source of these discrepancies is the nodal effective sections provided in the benchmark. (Author)

  8. PANTHER solution to the NEA-NSC 3-D PWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.

    1994-10-01

    This report contains the results of PANTHER calculations for the ``NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power``. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.).

  9. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  10. TAPINS: A THERMAL-HYDRAULIC SYSTEM CODE FOR TRANSIENT ANALYSIS OF A FULLY-PASSIVE INTEGRAL PWR

    Directory of Open Access Journals (Sweden)

    YEON-GUN LEE

    2013-08-01

    Full Text Available REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility. Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.

  11. Metallurgical and mechanical behaviours of PWR fuel cladding tube oxidised at high temperature; Comportements metallurqigue et mecanique des materiaux de gainage du combustible REP oxydes a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Stern, A

    2007-12-15

    Zirconium alloys are used as cladding materials in Pressurized Water Reactors (PWR). As they are submitted to very extreme conditions, it is necessary to check their behaviour and especially to make sure they meet the safety criteria. They are therefore studied under typical in service-loadings but also under accidental loadings. In one of these accidental scenarios, called Loss of Coolant Accident (LOCA) the cladding temperature may increase above 800 C, in a steam environment, and decrease before a final quench of the cladding. During this temperature transient, the cladding is heavily oxidised, and the metallurgical changes lead to a decrease of the post quench mechanical properties. It is then necessary to correlate this drop in residual ductility to the metallurgical evolutions. This is the problem we want to address in this study: the oxidation of PWR cladding materials at high temperature in a steam environment and its consequences on post quench mechanical properties. As oxygen goes massively into the metallic part - a zirconia layer grows at the same time - during the high temperature oxidation, the claddings tubes microstructure shows three different phases that are the outer oxide layer (zirconia) and the inner metallic phases ({alpha}(O) and 'ex {beta}') - with various mechanical properties. In order to reproduce the behaviour of this multilayered material, the first part of this study consisted in creating samples with different - but homogeneous in thickness - oxygen contents, similar to those observed in the different phases of the real cladding. The study was especially focused on the {beta}-->{alpha} phase transformation upon cooling and on the resulting microstructures. A mechanism was proposed to describe this phase transformation. For instance, we conclude that for our oxygen enriched samples, the phase transformation kinetics upon cooling are ruled by the oxygen partitioning between the two allotropic phases. Then, these materials

  12. Numerical Predictions of Transient Hydraulic Response of PWR Steam Generator Secondary Side to a Feedwater Pipe Break Using Flashing Flow Model

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    In this paper, CFD analyses of the transient flow field inside the SG secondary side of a PWR during blowdown following a FWLB accident were performed for two different outlet boundary condition models of the flashing flow through the broken pipe. The prediction results for the two different outlet boundary condition models were compared with each other to examine their applicability to the practical regulatory confirmation calculations. Based on the present CFD analysis results, it seems that both boundary condition models are generally acceptable for the application to the numerical prediction of the blowdown loading. However, it was confirmed that Model 2 extending the simulation domain to an atmospheric space surrounding the pipe broken end could yield more realistic simulation results of the present blowdown problem than the simpler Model 1 assuming a step pressure change at the broken pipe end.

  13. High-temperature compatibility between liquid metal as PWR fuel gap filler and stainless steel and high-density concrete

    Science.gov (United States)

    Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

    2014-08-01

    In conventional nuclear fuel rods for light-water reactors, a helium-filled as-fabricated gap between the fuel and the cladding inner surface accommodates fuel swelling and cladding creep down. Because helium exhibits a very low thermal conductivity, it results in a large temperature rise in the gap. Liquid metal (LM; 1/3 weight portion each of lead, tin, and bismuth) has been proposed to be a gap filler because of its high thermal conductivity (∼100 times that of He), low melting point (∼100 °C), and lack of chemical reactivity with UO2 and water. With the presence of LM, the temperature drop across the gap is virtually eliminated and the fuel is operated at a lower temperature at the same power output, resulting in safer fuel, delayed fission gas release and prevention of massive secondary hydriding. During normal reactor operation, should an LM-bonded fuel rod failure occurs resulting in a discharge of liquid metal into the bottom of the reactor pressure vessel, it should not corrode stainless steel. An experiment was conducted to confirm that at 315 °C, LM in contact with 304 stainless steel in the PWR water chemistry environment for up to 30 days resulted in no observable corrosion. Moreover, during a hypothetical core-melt accident assuming that the liquid metal with elevated temperature between 1000 and 1600 °C is spread on a high-density concrete basement of the power plant, a small-scale experiment was performed to demonstrate that the LM-concrete interaction at 1000 °C for as long as 12 h resulted in no penetration. At 1200 °C for 5 h, the LM penetrated a distance of ∼1.3 cm, but the penetration appeared to stop. At 1400 °C the penetration rate was ∼0.7 cm/h. At 1600 °C, the penetration rate was ∼17 cm/h. No corrosion based on chemical reactions with high-density concrete occurred, and, hence, the only physical interaction between high-temperature LM and high-density concrete was from tiny cracks generated from thermal stress. Moreover

  14. Investigation of feedback on neutron kinetics and thermal hydraulics from detailed online fuel behavior modeling during a boron dilution transient in a PWR with the two-way coupled code system DYN3D-TRANSURANUS

    Energy Technology Data Exchange (ETDEWEB)

    Holt, L., E-mail: lars.holt@tuev-sued.de [TÜV SÜD Energietechnik GmbH Baden-Württemberg, Gottlieb-Daimler-Str. 7, 70794 Filderstadt (Germany); Technical University München, Department of Nuclear Engineering, Boltzmannstr. 15, D-85748 Garching bei München (Germany); Rohde, U.; Kliem, S.; Baier, S. [Helmholtz-Zentrum Dresden—Rossendorf, Reactor Safety Division, PO Box 510119, D-01314 Dresden (Germany); Seidl, M. [E.ON Kernkraft GmbH, Tresckowstr. 5, D-30457 Hannover (Germany); Van Uffelen, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Macián-Juan, R. [Technical University München, Department of Nuclear Engineering, Boltzmannstr. 15, D-85748 Garching bei München (Germany)

    2016-02-15

    Highlights: • General coupling interface was developed for the fuel performance code TRANSURANUS. • With this new tool simplified fuel behavior models in codes can be replaced. • The reactor dynamics code DYN3D was coupled to TRANSURANUS at assembly level. • The feedback from detailed online fuel behavior modeling is analyzed for reactivity initiated accident (RIA). • The thermal hydraulics can be affected strongly even in fresh fuel assemblies. - Abstract: Recently the reactor dynamics code DYN3D (including an internal fuel behavior model) was coupled to the fuel performance code TRANSURANUS at assembly level. The coupled code system applies the new general TRANSURANUS coupling interface, hence it can be used for one-way or two-way coupling. In the coupling, DYN3D provides process time, time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in case of the two-way coupling approach replaces completely the internal DYN3D fuel behavior model and transfers parameters like radial fuel temperature distribution and cladding temperature back to DYN3D. For the first time results of the coupled code system are presented for a post-critical-heat-flux heat transfer. The corresponding heat transfer regime is mostly film boiling, where the cladding temperature can rise several hundreds of degrees. The simulated boron dilution transient assumed an injection of a 36 m{sup 3} slug of under-borated coolant into a German pressurized water reactor (PWR) core initiated from a sub-critical reactor state (extreme reactivity initiated accident (RIA) conditions). The feedback from detailed fuel behavior modeling was found negligible on the neutron kinetics and thermal hydraulics during the first power rise. In a later phase of the transient, the node injected energy can differ 25 J/g, even still around 20 J/g for nodes without film boiling. Furthermore, the thermal hydraulics can be affected strongly even in fresh fuel assemblies, where film boiling

  15. Thermal-hydraulics analysis of a PWR reactor using zircaloy and carbide silicon reinforced with type S fibers as fuel claddings: Simulation of a channel blockage transient

    Energy Technology Data Exchange (ETDEWEB)

    Matuck, Vinicius; Ramos, Mario C.; Faria, Rochkhudson B.; Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: matuck747@gmail.com, E-mail: patricialire@yahoo.com.br, E-mail: marc5663@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    A detailed thermal-hydraulic reactor model using as reference data from the Angra 2 Final Safety Analysis Report (FSAR) has been developed and SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding. The goal is to compare its behavior from the thermal viewpoint with the Zircaloy, at the steady- state and transient conditions. The RELAP-3D was used to perform the thermal-hydraulic analysis and a blockage transient has been investigated at full power operation. The transient considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

  16. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume I. RELAP4/MOD5 description. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1976-09-01

    RELAP4 is a computer program written in FORTRAN IV for the digital computer analysis of nuclear reactors and related systems. It is primarily applied in the study of system transient response to postulated perturbations such as coolant loop rupture, circulation pump failure, power excursions, etc. The program was written to be used for water-cooled (PWR and BWR) reactors and can be used for scale models such as LOFT and SEMISCALE. Additional versatility extends its usefulness to related applications, such as ice condenser and containment subcompartment analysis. Specific options are available for reflood (FLOOD) analysis and for the NRC Evaluation Model.

  17. Effect of irradiation temperature in PWR RPV materials and its inclusion in semi-mechanistic model

    Energy Technology Data Exchange (ETDEWEB)

    Debarberis, L. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Acosta, B. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: beatriz.acosta-iborra@jrc.nl; Zeman, A. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Sevini, F. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Ballesteros, A. [Tecnatom, Avd. Montes de Oca 1, San Sebasitan de los Reyes, E-28709 Madrid (Spain); Kryukov, A. [Russian Research Centre Kurchatov Institute, Kurchatov Square 1, 123182 Moscow (Russian Federation); Gillemot, F. [AEKI Atomic Research Institute, Konkoly Thege M. ut 29-33, 1121 Budapest (Hungary); Brumovsky, M. [NRI, Nuclear Research Institute, Husinec, Rez 130, 25068 Rez (Czech Republic)

    2005-09-15

    The irradiation temperature is a very important parameter in radiation damage kinetics. In this article the challenge of including temperature into a general semi-mechanistic model for radiation embrittlement is presented. In this manner the model allows data obtained at different temperatures, both in surveillance programmes and in research reactors, to be understood.

  18. Transient instability upon temperature quench in weakly ordered block copolymers

    Energy Technology Data Exchange (ETDEWEB)

    Qi, Shuyan [Department of Chemical Engineering, University of California, Berkeley, California 94610 (United States); Wang, Zhen-Gang [Division of Chemistry and Chemical Engineering, California Institute of Technology, Pasadena, California 91125 (United States)

    1999-12-15

    We report a novel transient instability upon temperature quench in weakly ordered block copolymer microphases possessing a soft direction or directions, such as the lamellar and hexagonal cylinder (HEX) phases. We show that reequilibration of the order parameter is accompanied by transient long wavelength undulation of the layers or cylinders--with an initial wavelength that depends on the depth of the temperature quench--that eventually disappears as the structure reaches its equilibrium at the new temperature. Such undulation leads to a transient transverse broadening of the scattering peaks near the Bragg positions. We argue that this instability might be responsible for the experimentally observed unusual ordering dynamics of the HEX phase of a diblock copolymer after quenching from the disordered state. (c) 1999 American Institute of Physics.

  19. Transient response of high temperature PEM fuel cell

    Science.gov (United States)

    Peng, J.; Shin, J. Y.; Song, T. W.

    A transient three-dimensional, single-phase and non-isothermal numerical model of polymer electrolyte membrane (PEM) fuel cell with high operating temperature has been developed and implemented in computational fluid dynamic (CFD) code. The model accounts for transient convective and diffusive transport, and allows prediction of species concentration. Electrochemical charge double-layer effect is considered. Heat generation according to electrochemical reaction and ohmic loss are involved. Water transportation across membrane is ignored due to low water electro-osmosis drag force of polymer polybenzimidazole (PBI) membrane. The prediction shows transient in current density which overshoots (undershoots) the stabilized state value when cell voltage is abruptly decreased (increased). The result shows that the peak of overshoot (undershoot) is related with cathode air stoichiometric mass flow rate instead of anode hydrogen stoichiometric mass flow rate. Current is moved smoothly and there are no overshoot or undershoot with the influence of charge double-layer effect. The maximum temperature is located in cathode catalyst layer and both fuel cell average temperature and temperature deviation are increased with increasing of current load.

  20. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume III. Checkout applications. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    Obenchain, C. F.; Ramsthaler, J. H.; Eales, E. P.; Charlton, T. R.; Childs, F. W.; Giles, M. M.; Good, E. G.; Gruen, G. E.; Guttman, J.; Johnsen, G. W.; Katsma, K. R.; Keeler, C. D.; Lawford, T. W.; Mohr, C. M.; Singer, G. L.; Townsend, W. C.

    1976-09-01

    Checkout problems presented include the following: PWR large cold leg break; PWR small cold leg break; PWR intermediate sized cold leg break; BWR large recirculation line break; BWR small recirculation line break; INEL Semiscale small cold leg break; INEL LOFT large cold leg break and INEL Semiscale large cold leg break. Also included is Update 2 of the RELAP 4/M0D5 code.

  1. SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

    1978-02-01

    The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.

  2. Transient analysis and burnout of high temperature superconducting current leads

    Science.gov (United States)

    Seol, S. Y.; Hull, J. R.

    The transient behaviour of high-temperature superconductor (HTS) current leads operated between liquid helium and liquid nitrogen temperatures is analysed for burnout conditions upon transition of the HTS into the normal state. Leads composed of HTS only and of HTS sheathed by pure silver or silver alloy are investigated numerically for temperature-dependent properties and analytically for temperature-independent properties. For lower values of shape factor (current density times length), the lead can be operated indefinitely without burnout. At higher values of shape factor, the lead reaches burnout in a finite time. With high current densities, the leads heat adiabatically. For a fixed shape factor, low current densities are desired to achieve long burnout times. To achieve a low helium boil-off rate in the superconducting state without danger of burnout, there is a preferred temperature dependence for thermal conductivity, and silver alloy sheaths are preferred to pure silver sheaths. However, for a given current density, pure silver sheaths take longer to burn out.

  3. A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Antonio Carlos Pinto Dias

    1993-12-31

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

  4. Numerical Analysis of Transient Temperature Response of Soap Film

    Science.gov (United States)

    Tanaka, Seiichi; Tatesaku, Akihiro; Dantsuka, Yuki; Fujiwara, Seiji; Kunimine, Kanji

    2015-11-01

    Measurements of thermophysical properties of thin liquid films are important to understand interfacial phenomena due to film structures composed of amphiphilic molecules in soap film, phospholipid bilayer of biological cell and emulsion. A transient hot-wire technique for liquid films less than 1 \\upmu m thick such as soap film has been proposed to measure the thermal conductivity and diffusivity simultaneously. Two-dimensional heat conduction equations for a solid cylinder with a liquid film have been solved numerically. The temperature of a thin wire with liquid film increases steeply with its own heat generation. The feasibility of this technique is verified through numerical experiments for various thermal conductivities, diffusivities, and film thicknesses. Calculated results indicate that the increase in the volumetric average temperature of the thin wire sufficiently varies with the change of thermal conductivity and diffusivity of the soap film. Therefore, the temperature characteristics could be utilized to evaluate both the thermal conductivity and diffusivity using the Gauss-Newton method.

  5. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2014-01-01

    Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

  6. PWR-blowdown heat transfer separate effects program

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, D.G.

    1976-01-01

    The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results are obtained from the Thermal-Hydraulic Test Facility (THTF). Supporting experiments are carried out in several additional test loops - the Forced Convection Test Facility (FCTF), an air-water loop, a transient steam-water loop, and a low-temperature water mockup of the THTF heater rod bundle. The studies to date are described.

  7. WRAP-PWR verification studies

    Energy Technology Data Exchange (ETDEWEB)

    Gregory, M V; Ames, P L; Beranek, F; Kuehn, N H; Parks, P B

    1980-01-01

    A modular computational system known as the Water Reactor Analysis Package - Evaluation Model (WRAP-EM) was developed for the Nuclear Regulatory Commission (NRC) to interpret and evaluate reactor vendor EM methods and computed results. A subset of the system (WRAP-PWR-EM) provides the computational tools to perform a complete analysis of loss-of-coolant accidents (LOCA's) in pressurized water reactors (PWR's). A set of calculations modeling experimental tests in the Semiscale and LOFT facilities, and calculations of a large break in a typical four-loop Westinghouse PWR plant have verified that the WRAP-PWR-EM system is functioning as intended.

  8. Thermoelectric energy harvesting from small ambient temperature transients

    Energy Technology Data Exchange (ETDEWEB)

    Moser, Andre

    2012-07-01

    Wireless sensor networks (WSNs) represent a key technology, used, for instance, in structural health monitoring, building automation systems, or traffic surveillance. Supplying power to a network of spatially distributed sensor nodes, especially at remote locations, is a large challenge: power grids are reliable but costly to install, whereas batteries provide a high flexibility in the installation but have a limited lifetime. This dilemma can be overcome by micro energy harvesting which offers both: reliability and flexibility. Micro energy harvesters are able to convert low grade ambient energy into useful electrical energy and thus provide power for wireless sensor networks or other electronic devices - in-situ, off-grid, and with an almost unlimited lifetime. Thermal energy is an omnipresent source of ambient energy: The day-night-cycle of the sun causes a temperature variation in the ambient air as well as arbitrary solids (soil, building walls, etc.). Unlike the air, solids have a large thermal inertia which dampens the temperature variation. This physical process leads to a temperature difference {Delta}T = T{sub air} - T{sub solid} between air and solid that can be converted directly into electrical energy by a thermoelectric generator (TEG). Thermal and electrical interfaces are necessary to connect the TEG to the thermal energy source (T{sub air}, T{sub solid}) and the electrical load (WSN). Reliable operation of the WSN may only be ensured if the harvester provides sufficient electrical energy, i.e. operates at its maximum power point. The goal of this thesis is to study, design, and test thermoelectric harvesters generating electrical energy from small ambient temperature transients in order to self-sufficiently power a WSN. Current research into thermoelectric energy harvesting, especially analytical modeling and application in the field are treated insufficiently. Therefore, a time-dependent analytical model of the harvester's output power is set

  9. Transient energy growth modulation by temperature dependent transport properties in a stratified plane Poiseuille flow

    NARCIS (Netherlands)

    Rinaldi, E.; Boersma, B.J.; Pecnik, R.

    2015-01-01

    We investigate the effect of temperature dependent thermal conductivity ? and isobaric specific heat c_P on the transient amplification of perturbations in a thermally stratified laminar plane Poiseuille flow. It is shown that for decreasing thermal conductivity the maximum transient energy growth

  10. Simulation and experiment on transient temperature field of a magnetorheological clutch for vehicle application

    Science.gov (United States)

    Wang, Daoming; Zi, Bin; Zeng, Yishan; Qian, Sen; Qian, Jun

    2017-09-01

    The unpredictable power fluctuation due to severe heating has been demonstrated to be a critical bottleneck technique restricting the application of magnetorheological (MR) clutches in vehicle industry. The aim of this study is to introduce a low-cost transient simulation approach for evaluating the heat build-up and dissipation of a liquid-cooled MR vehicle clutch. This paper firstly performs a detailed description of the developed MR clutch in terms of operation principle, material selection and configuration. Subsequently, transient temperature simulations are carried out under various conditions to reveal the distribution, variation and impact factors of the transient temperature field. Following these, an experimental setup is established for heating tests of the clutch prototype. Experimental results concerning the temperature variation of magnetorheological fluids and the maximum allowable transient slip power of the clutch prototype are presented, which in return verify the correctness and feasibility of the simulation.

  11. Study of the influence of temperature and time on the electroplating nickel layer in Inconel 718 strips used in spacer grid of Pressurized Water Cooled nuclear reactors (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Rezende, Renato; Abati, Amanda; Verne, Júlio; Panossian, Zehbour, E-mail: amanda.abati@marinha.mil.br, E-mail: jvernegropp@gmail.com, E-mail: renato.rezende@marinha.mil.br, E-mail: zep@ipt.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil). Laboratório de Desenvolvimento e Instrumentação de Combustível Nuclear; Instituto de Pesquisas Tecnológicas (IPT), São Paulo, SP (Brazil)

    2017-07-01

    The Inconel 718 (UNS N07718: Ni-{sup 19}Cr-{sup 18}Fe-{sup 5}Nb-3 Mo) is a precipitation hardenable nickel alloy that has good corrosion resistance and high mechanical strength. These strips are used for assembling the spacer grid of fuel element of pressurized water cooled nuclear reactors (PWR). The spacer grid is a structural component of fundamental importance in fuel elements of PWR reactors, maintaining the position and necessary spacing of the fuel rods within the arrangement of the fuel element. The spacer grid is formed by joining the points of intersection of the strips, by a joint process called brazing. For this process, these strips are stamped and plated with a thin layer of nickel by means of electroplating in order to protect against oxidation and allow a better flowability and wettability of the addition metal in the strips during brazing. Oxidation at the surface of the base material harms wettability and inhibits spreading of the liquid addition metal on the substrate surface during the brazing process. The use of coatings such as nickel plating is used to ensure such conditions. The results showed that there is a process of diffusion de some chemical elements such as chromium, iron, titanium and aluminum from the substrate to the nickel layer and nickel from the layer to the substrate. These chemical elements are responsible for the oxidation at the surface of the strip. (author)

  12. Optical-Thickness Corrections to Transient Ece Temperature-Measurements in Tokamak and Stellarator Plasmas

    NARCIS (Netherlands)

    Peters, M.; Gorini, G.; Mantica, P.

    1995-01-01

    The conditions are examined under which optical thickness (tau) corrections to electron cyclotron emission (ECE) measurements of electron temperature (T-e) can be neglected. By means of simple algebra it is demonstrated that for measurements of T-e transients the ECE radiation temperature (T-rad)

  13. Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal at Zero Power Benchmark Cases with NODAL3 Code

    Directory of Open Access Journals (Sweden)

    Tagor Malem Sembiring

    2017-01-01

    Full Text Available The in-house coupled neutronic and thermal-hydraulic (N/T-H code of BATAN (National Nuclear Energy Agency of Indonesia, NODAL3, based on the few-group neutron diffusion equation in 3-dimensional geometry using the polynomial nodal method, has been verified with static and transient PWR benchmark cases. This paper reports the verification of NODAL3 code in the NEA-NSC PWR uncontrolled control rods withdrawal at zero power benchmark. The objective of this paper is to determine the accuracy of NODAL3 code in solving the continuously slow and fast reactivity insertions due to single and group of control rod bank withdrawn while the power and temperature increment are limited by the Doppler coefficient. The benchmark is chosen since many organizations participated using various methods and approximations, so the calculation results of NODAL3 can be compared to other codes’ results. The calculated parameters are performed for the steady-state, transient core averaged, and transient hot pellet results. The influence of radial and axial nodes number was investigated for all cases. The results of NODAL3 code are in very good agreement with the reference solutions if the radial and axial nodes number is 2 × 2 and 2 × 18 (total axial layers, respectively.

  14. Feasibility Assessment of Thermal Barrier Seals for Extreme Transient Temperatures

    Science.gov (United States)

    Steinetz, Bruce M.; Dunlap, Patrick H., Jr.

    1998-01-01

    The assembly joints of modem solid rocket motor cases are generally sealed using conventional O-ring type seals. The 5500+ F combustion gases produced by rocket motors are kept a safe distance away from the seals by thick layers of phenolic insulation. Special compounds are used to fill insulation gaps leading up to the seals to prevent a direct flowpath to them. Design criteria require that the seals should not experience torching or charring during operation, or their sealing ability would be compromised. On limited occasions, NASA has observed charring of the primary O-rings of the Space Shuttle solid rocket nozzle assembly joints due to parasitic leakage paths opening up in the gap-fill compounds during rocket operation. NASA is investigating different approaches for preventing torching or charring of the primary O-rings. One approach is to implement a braided rope seal upstream of the primary O-ring to serve as a thermal barrier that prevents the hot gases from impinging on the O-ring seals. This paper presents flow, resiliency, and thermal resistance for several types of NASA rope seals braided out of carbon fibers. Burn tests were performed to determine the time to burn through each of the seals when exposed to the flame of an oxyacetylene torch (5500 F), representative of the 5500 F solid rocket motor combustion temperatures. Rope seals braided out of carbon fibers endured the flame for over six minutes, three times longer than solid rocket motor burn time. Room and high temperature flow tests are presented for the carbon seals for different amounts of linear compression. Room temperature compression tests were performed to assess seal resiliency and unit preloads as a function of compression. The thermal barrier seal was tested in a subscale "char" motor test in which the seal sealed an intentional defect in the gap insulation. Temperature measurements indicated that the seal blocked 2500 F combustion gases on the upstream side with very little temperature

  15. Temperature and distortion transients in gas tungsten-arc weldments

    Energy Technology Data Exchange (ETDEWEB)

    Glickstein, S.S.; Friedman, E.

    1979-10-01

    An analysis and test program to develop a fundamental understanding of the gas tungsten-arc welding process has been undertaken at the Bettis Atomic Power Laboratory to develop techniques to determine and control the various welding parameters and weldment conditions so as to result in optimum weld response characteristics. These response characteristics include depth of penetration, weld bead configuration, weld bead sink and roll, distortion, and cracking sensitivity. The results are documented of that part of the program devoted to analytical and experimental investigations of temperatures, weld bead dimensions, and distortions for moving gas tungsten-arc welds applied to Alloy 600 plates.

  16. Power-cooling mismatch test series. Test PCM-2A; test results report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cawood, G.W.; Holman, G.W.; Martinson, Z.R.; Legrand, B.L.

    1976-09-01

    The report describes the results of an in-pile experimental investigation of pre- and postcritical heat flux (CHF) behavior of a single 36-inch-long, pressurized water reactor (PWR) type, UO/sub 2/-fueled, zircaloy-clad fuel rod. The nominal coolant conditions for pressure and temperature were representative of those found in a commercial PWR. Nine separate departures from nucleate boiling (DNB) cycles were performed by either of two different methods: (a) decreasing the coolant flow rate while the fuel rod power was held constant, or (b) increasing the fuel rod power while the coolant flow rate was held constant. DNB was obtained during eight of the nine cycles performed. For the final flow reduction, the mass flux was decreased to 6.1 x 10/sup 5/ lb/hr-ft/sup 2/ at a constant peak linear heat generation rate of 17.8 kW/ft. The fuel rod was allowed to remain in film boiling for about 210 seconds during this final flow reduction. The fuel rod remained intact during the test. Results of on-line measurements of the fuel rod behavior are presented together with discussion of instrument performance. A comparison of the data with Fuel Rod Analysis Program-Transient 2 (FRAP-T2) computer code calculations is included.

  17. A Coupled Phase-Temperature Model for Dynamics of Transient Neuronal Signal in Mammals Cold Receptor

    Directory of Open Access Journals (Sweden)

    Firman Ahmad Kirana

    2016-01-01

    Full Text Available We propose a theoretical model consisting of coupled differential equation of membrane potential phase and temperature for describing the neuronal signal in mammals cold receptor. Based on the results from previous work by Roper et al., we modified a nonstochastic phase model for cold receptor neuronal signaling dynamics in mammals. We introduce a new set of temperature adjusted functional parameters which allow saturation characteristic at high and low steady temperatures. The modified model also accommodates the transient neuronal signaling process from high to low temperature by introducing a nonlinear differential equation for the “effective temperature” changes which is coupled to the phase differential equation. This simple model can be considered as a candidate for describing qualitatively the physical mechanism of the corresponding transient process.

  18. Transient deformational properties of high temperature alloys used in solid oxide fuel cell stacks

    DEFF Research Database (Denmark)

    Tadesse Molla, Tesfaye; Kwok, Kawai; Frandsen, Henrik Lund

    2017-01-01

    Stresses and probability of failure during operation of solid oxide fuel cells (SOFCs) is affected by the deformational properties of the different components of the SOFC stack. Though the overall stress relaxes with time during steady state operation, large stresses would normally appear through...... transients in operation including temporary shut downs. These stresses are highly affected by the transient creep behavior of metallic components in the SOFC stack. This study investigates whether a variation of the so-called Chaboche's unified power law together with isotropic hardening can represent...... to describe the high temperature inelastic deformational behaviors of Crofer 22 APU used for metallic interconnects in SOFC stacks....

  19. Measured versus simulated transients of temperature logs—a test of borehole climatology

    Science.gov (United States)

    Majorowicz, Jacek; Safanda, Jan

    2005-12-01

    We report the results of repeated temperature, T, measurements with depth (z) for two borehole sites located in the Western Canadian Sedimentary Basin, in central Alberta and south central Saskatchewan. These were logged at three different times within the time period of 1986 AD to 2004 AD. Subsurface temperature transient changes of 0.1 to 0.4 °C observed between the repeated temperature logs over the last two decades agree only partially with the changes derived from the synthetic profiles in which surface temperature time series were used as forcing signals. The surface temperature forcing is responsible for the majority of the observed deviation of temperature with depth. In some cases, differences higher than the error of measurement are observed between the model and measurements. This can be an indication that factors other than the surface temperature change also influence the subsurface thermal regime.

  20. 3D transient temperature measurement in homogeneous solid material with THz waves

    Science.gov (United States)

    Romano, M.; Sommier, A.; Batsale, J.-C.; Pradere, C.

    2016-04-01

    The first imaging system that is able to measure transient temperature phenomena taking place inside a bulk by 3D tomography is presented. This novel technique combines the power of terahertz waves and the high sensitivity of infrared imaging. The tomography reconstruction is achieved by the 3D motion of the sample at several angular positions followed by inverse Radon transform processing to retrieve the 3D transient temperatures. The aim of this novel volumetric imaging technique is to locate defects within the whole target body as well as to measure the temperature in the whole volume of the target. This new-fashioned thermal tomography will revolutionize the non-invasive monitoring techniques for volume inspection and in-situ properties estimations.

  1. Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)

    1995-08-01

    The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)

  2. Two-dimensional transient far-field analysis for the excess temperature from an arbitrary source

    Energy Technology Data Exchange (ETDEWEB)

    Witten, A.J.; Long, E.C.

    1978-07-01

    An analytic solution is presented for the two-dimensional time-dependent advective diffusion equation governing the distribution of excess temperature in a river of uniform width, depth, and downstream flow. The solution is also applicable to a straight coastline with uniform longshore flow. Exact solutions are obtained for a point heat source and a particular line heat source, while an approximate representation is given for an arbitrary time-varying heat source. These solutions are incorporated into a computer program which calculates excess temperature and time rate-of-change of excess temperature in a river or coast as a result of waste heat discharged from various transient sources.

  3. Transient receptor potential melastatin 8 (TRPM8) channels are involved in body temperature regulation.

    Science.gov (United States)

    Gavva, Narender R; Davis, Carl; Lehto, Sonya G; Rao, Sara; Wang, Weiya; Zhu, Dawn X D

    2012-05-09

    Transient receptor potential cation channel subfamily M member 8 (TRPM8) is activated by cold temperature in vitro and has been demonstrated to act as a 'cold temperature sensor' in vivo. Although it is known that agonists of this 'cold temperature sensor', such as menthol and icilin, cause a transient increase in body temperature (Tb), it is not known if TRPM8 plays a role in Tb regulation. Since TRPM8 has been considered as a potential target for chronic pain therapeutics, we have investigated the role of TRPM8 in Tb regulation. We characterized five chemically distinct compounds (AMG0635, AMG2850, AMG8788, AMG9678, and Compound 496) as potent and selective antagonists of TRPM8 and tested their effects on Tb in rats and mice implanted with radiotelemetry probes. All five antagonists used in the study caused a transient decrease in Tb (maximum decrease of 0.98°C). Since thermoregulation is a homeostatic process that maintains Tb about 37°C, we further evaluated whether repeated administration of an antagonist attenuated the decrease in Tb. Indeed, repeated daily administration of AMG9678 for four consecutive days showed a reduction in the magnitude of the Tb decrease Day 2 onwards. The data reported here demonstrate that TRPM8 channels play a role in Tb regulation. Further, a reduction of magnitude in Tb decrease after repeated dosing of an antagonist suggests that TRPM8's role in Tb maintenance may not pose an issue for developing TRPM8 antagonists as therapeutics.

  4. Evaluation of the heat transfer in a geological repository concept containing PWR, VHTR and hybrid ads-fission spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jonusan, Raoni A.S.; Pereira, Fernando; Velasquez, Carlos E.; Salome, Jean A.D.; Cardoso, Fabiano; Pereira, Claubia; Fortini, Angela, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    The investigation of the thermal behavior of spent fuel (SF) materials is essential to determining appropriate potential sites to accommodate geological repositories as well as the design of canisters, considering their potential risk to people health and of environmental contamination. This work presents studies of the temperature in a canister containing spent fuels discharged from Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) reactor systems in a geological repository concept. The thermal analyses were performed with the software ANSYS, which is widely used to solve engineering problems through the Finite Element Method. The ANSYS Transient Thermal module was used. The spent nuclear fuels were set as heat sources using data of previous studies derived from decay heat curves. The studies were based on comparison of the mean temperature on a canister surface along the time under geological disposal conditions, for a same amount of each type of spent nuclear fuel evaluated. The results conclude that fuels from VHTR and ADS systems are inappropriate to be disposed in a standardized PWR canister, demanding new studies to determine the optimal amount of spent fuel and new internal canister geometries. It is also possible to conclude that the hypothetical situation of a single type of canister being used to accommodate different types of spent nuclear fuels is not technically feasible. (author)

  5. Numerical Analysis on the Effects of Initial Temperature of the PCM in Transient Direct Contact Melting

    Science.gov (United States)

    Saito, Akio; Kumano, Hiroyuki; Okawa, Seiji; Ohta, Shinro

    The effects of the initial temperature of the solid PCM on a transient direct contact melting were invesigated, analytically. Four types of boundary conditions, that is, the constant temperature at the top, the constant heat flux at the top, the constant heat flux at the bottom, and the constant heat transfer coefficient and the constant temperature of the medium at the bottom of the heating plate, respectively, were examined. The effects of the initial temperature for each boundary condition were investigated through the transient process of the direct contact melting, where the heat transfer phenomena within the liquid film formed between the solid and the heating plate was unsteady, for the case of initial temperature lower than the melting point. It was found, as the results, that the time required until to approach the quasi-steady state and the melting speed at the time depended seriously on the initial temperature. It was understood, at the same time, that the heat flux flowing into the solid was not negligible, even in the quasi-steady state. Then, effects of the non-dimensional parameters, appearing in relation to the initial temperature variation of the solid, were examined, qualitatively.

  6. Strength and reliability of low temperature transient liquid phase bonded Cu-Sn-Cu interconnects

    DEFF Research Database (Denmark)

    Brincker, Mads; Söhl, Stefan; Eisele, Ronald

    2017-01-01

    As power electronic devices have tendencies to operate at higher temperatures and current densities, the demand for reliable and efficient packaging technologies are ever increasing. This paper reports the studies on application of transient liquid phase (TLP) bonding of CuSnCu systems as a poten......As power electronic devices have tendencies to operate at higher temperatures and current densities, the demand for reliable and efficient packaging technologies are ever increasing. This paper reports the studies on application of transient liquid phase (TLP) bonding of CuSnCu systems....... Micro-sectioning and optical microscopy studies of the samples reveal that the TLP bonds show good homogeneity with a small number of voids at the interface. Energy dispersive X-ray analysis is applied to examine at what rates Sn is converted into CuSn intermetallics since a full conversion is critical...

  7. VHTR, ADS, and PWR spent nuclear fuel analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salome, J.A.D.; Cardoso, F.; Velasquez, C.E.; Pereira, F.; Pereira, C. [Departamento de Engenharia Nuclear - Escola de Engenharia Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627, Pampulha, Belo Horizonte MG, CEP: 31270-901 (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores - CNPq, Rio de Janeiro (Brazil); Barros, G.P. [Comissao Nacional de Energia Nuclear - CNEN, Rua General Severiano 82, Botafogo, Rio de Janeiro, RJ, CEP: 22290-040 (Brazil)

    2016-07-01

    The aim of this study is to analyze and compare the discharged-spent fuel of 3 types of nuclear systems: a Very High-Temperature Gas Reactor (VHTR), a lead-cooled Accelerator-Driven System (ADS) and a standard Pressurized Water Reactor (PWR). The two first systems, VHTR, and ADS were designed to use reprocessed fuels. UREX+ and GANEX techniques were used for the reprocessing processes respectively. The fuel burnup simulated for the systems in other works have been used to obtain the final composition of the spent fuel discharged. After discharge, the radioactivity, the radiotoxicity, and the decay heat were evaluated through the ORIGEN 2.1 code until 10{sup 7} years and compared to the literature. The spent nuclear waste (SNF) coming from reprocessing techniques and burned up in advanced reactors show that the radiotoxicity decreases below a conventional SNF from a typical PWR for the time studied. The VHTR and ADs have higher values of radioactivity, radiotoxicity and decay heat, because of the greater concentrations of plutonium and curium in these reactors than in the PWR. Fission products have the greatest contribution for the first 25 years over the parameters studied for a PWR. The most harmful fission products are: Ba{sup 137m}, Tc{sup 99}, I{sup 129} and Nb{sup 93m} and for actinides is the plutonium and curium.

  8. Thermal behavior analysis of PWR fuel during RIA at various fuel burnups using modified theatre code

    Directory of Open Access Journals (Sweden)

    Nawaz Amjad

    2016-01-01

    Full Text Available The fuel irradiation and burnup causes geometrical and dimensional changes in the fuel rod which affects its thermal resistance and ultimately affects the fuel rod behavior during steady-state and transient conditions. The consistent analysis of fuel rod thermal performance is essential for precise evaluation of reactor safety in operational transients and accidents. In this work, analysis of PWR fuel rod thermal performance is carried out under steady-state and transient conditions at different fuel burnups. The analysis is performed by using thermal hydraulic code, THEATRe. The code is modified by adding burnup dependent fuel rod behavior models. The original code uses as-fabricated fuel rod dimensions during steady-state and transient conditions which can be modified to perform more consistent reactor safety analysis. AP1000 reactor is considered as a reference reactor for this analysis. The effect of burnup on steady-state fuel rod parameters has been investigated. For transient analysis, hypothetical reactivity initiated accident was simulated by considering a triangular power pulse of variable pulse height (relative to the full power reactor operating conditions and pulse width at different fuel burnups which corresponds to fresh fuel, low and medium burnup fuels. The effect of power pulse height, pulse width and fuel burnup on fuel rod temperatures has been investigated. The results of reactivity initiated accident analysis show that the fuel failure mechanisms are different for fresh fuel and fuel at different burnup levels. The fuel failure in fresh fuel is expected due to fuel melting as fuel temperature increases with increase in pulse energy (pulse height. However, at relatively higher burnups, the fuel failure is expected due to cladding failure caused by strong pellet clad mechanical interaction, where, the contact pressure increases beyond the cladding yield strength.

  9. Ag-In transient liquid phase bonding for high temperature stainless steel micro actuators

    OpenAIRE

    Andersson, Martin

    2013-01-01

    A stainless steel, high temperature, phase change micro actuator has been demonstrated using the solid-liquid phase transition of mannitol at 168°C and In-Ag transient liquid phase diffusion bonding. Joints created with this bonding technique can sustain temperatures up to 695°C, while being bonded at only 180°C, and have thicknesses between 1.4 to 6.0 μm. Physical vapour deposition, inkjet printing and electroplating have been evaluated as deposition methods for bond layers. For actuation, c...

  10. Transient performance and temperature field of a natural convection air dehumidifier loop

    Science.gov (United States)

    Fazilati, Mohammad Ali; Sedaghat, Ahmad; Alemrajabi, Ali-Akbar

    2017-07-01

    In this paper, transient performance of the previously introduced natural convection heat and mass transfer loop is investigated for an air dehumidifier system. The performance of the loop is studied in different conditions of heat source/heat sink temperature and different startup desiccant concentrations. Unlike conventional loops, it is observed that natural convection of the fluid originates from the heat sink towards the heat source. The proper operation of the cycle is highly dependent on the heat sink/heat source temperatures. To reduce the time constant of the system, a proper desiccant concentration should be adopted for charge of the loop.

  11. Improved analysis of transient temperature data from permanent down-hole gauges (PDGs)

    Science.gov (United States)

    Zhang, Yiqun; Zheng, Shiyi; Wang, Qi

    2017-08-01

    With the installation of permanent down-hole gauges (PDGs) during oil field development, large volumes of high resolution and continuous down-hole information are obtainable. The interpretation of these real-time temperature and pressure data can optimize well performance, provide information about the reservoir and continuously calibrate the reservoir model. Although the dynamic temperature data have been interpreted in practice to predict flow profiling and provide characteristic information of the reservoir, almost all of the approaches rely on established non-isothermal models which depend on thermodynamic parameters. Another problem comes from the temperature transient analysis (TTA), which is underutilized compared with pressure transient analysis (PTA). In this study, several model-independent methods of TTA were performed. The entire set of PDG data consists of many flow events. By utilizing the wavelet transform, the exact points of flow-rate changes can be located. The flow regime changes, for example, from early time linear flow to later time pseudo-radial flow, among every transient period with constant flow-rate. For the early time region (ETR) that is caused by flow-rate change operations, the TTA, along with the PTA can greatly reduce the uncertainties in flow regime diagnosis. Then, the temperature variations during ETR were examined to infer the true reservoir temperature history, and the relationships between the wavelet detailed coefficients and the flow-rate changes were analysed. For the scenarios with constant reservoir-well parameters, the detailed flow-rate history can be generated by calculating the coefficient of relationship in advance. For later times, the flow regime changes to pseudo-radial flow. An analytical solution was introduced to describe the sand-face temperature. The formation parameters, such as permeability and skin factor, were estimated with the previously calculated flow-rate. It is necessary to analyse temperature

  12. Length determination on industrial polymer parts from measurement performed under transient temperature conditions

    DEFF Research Database (Denmark)

    Dalla Costa, Giuseppe; Madruga, Daniel González; De Chiffre, Leonardo

    2016-01-01

    A way to reduce the cost of metrology in manufacturing is to perform dimensional verification directly in the production environment, avoiding a long and expensive acclimatization phase. In this work the effect of a transient temperature state, typical of the production environment...... the cooling phase, from 27 °C to 20 °C approximately. The length variation was measured by means of an inductive probe and the temperature with an RTD surface sensor. The frame of the system was composed by elements in Zerodur and Invar to minimize the thermal deformations of the structure. Uniform...

  13. Investigation of transient temperature's influence on damage of high-speed sliding electrical contact rail surface

    Science.gov (United States)

    Zhang, Yuyan; Sun, Shasha; Guo, Quanli; Yang, Degong; Sun, Dongtao

    2016-11-01

    In the high speed sliding electrical contact with large current, the temperature of contact area rises quickly under the coupling action of the friction heating, the Joule heating and electric arc heating. The rising temperature seriously affects the conductivity of the components and the yield strength of materials, as well affects the contact state and lead to damage, so as to shorten the service life of the contact elements. Therefore, there is vital significance to measure the temperature accurately and investigate the temperature effect on damage of rail surface. Aiming at the problem of components damage in high speed sliding electrical contact, the transient heat effect on the contact surface was explored and its influence and regularity on the sliding components damage was obtained. A kind of real-time temperature measurement method on rail surface of high speed sliding electrical contact is proposed. Under the condition of 2.5 kA current load, based on the principle of infrared radiation non-contact temperature sensor was used to measure the rail temperature. The dynamic distribution of temperature field was obtained through the simulation analysis, further, the connection between temperature changes and the rail surface damage morphology, the damage volume was analyzed and established. Finally, the method to reduce rail damage and improve the life of components by changing the temperature field was discussed.

  14. Overview of PWR chemistry options

    Energy Technology Data Exchange (ETDEWEB)

    Nordmann, F.; Stutzmann, A.; Bretelle, J.L. [Electricite de France, Central Labs. (France)

    2002-07-01

    EDF Central Laboratories, in charge of engineering in chemistry, of defining the chemistry specifications and studying the operation feedback and improvement for 58 PWR units, have the opportunity to evaluate many options of operation developed and applied all around the world. Thanks to these international relationships and to the benefit of a large feedback from many units, some general evaluation of the various options is discussed in this paper. (authors)

  15. Heat exchanger temperature response for duty-cycle transients in the NGNP/HTE.

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R. B.; Nuclear Engineering Division

    2009-03-12

    Control system studies were performed for the Next Generation Nuclear Plant (NGNP) interfaced to the High Temperature Electrolysis (HTE) plant. Temperature change and associated thermal stresses are important factors in determining plant lifetime. In the NGNP the design objective of a 40 year lifetime for the Intermediate Heat Exchanger (IHX) in particular is seen as a challenge. A control system was designed to minimize temperature changes in the IHX and more generally at all high-temperature locations in the plant for duty-cycle transients. In the NGNP this includes structures at the reactor outlet and at the inlet to the turbine. This problem was approached by identifying those high-level factors that determine temperature rates of change. First are the set of duty cycle transients over which the control engineer has little control but which none-the-less must be addressed. Second is the partitioning of the temperature response into a quasi-static component and a transient component. These two components are largely independent of each other and when addressed as such greater understanding of temperature change mechanisms and how to deal with them is achieved. Third is the manner in which energy and mass flow rates are managed. Generally one aims for a temperature distribution that minimizes spatial non-uniformity of thermal expansion in a component with time. This is can be achieved by maintaining a fixed spatial temperature distribution in a component during transients. A general rule of thumb for heat exchangers is to maintain flow rate proportional to thermal power. Additionally the product of instantaneous flow rate and heat capacity should be maintained the same on both sides of the heat exchanger. Fourth inherent mechanisms for stable behavior should not be compromised by active controllers that can introduce new feedback paths and potentially create under-damped response. Applications of these principles to the development of a plant control strategy for

  16. Transient receptor potential melastatin 8 (TRPM8 channels are involved in body temperature regulation

    Directory of Open Access Journals (Sweden)

    Gavva Narender R

    2012-05-01

    Full Text Available Abstract Background Transient receptor potential cation channel subfamily M member 8 (TRPM8 is activated by cold temperature in vitro and has been demonstrated to act as a ‘cold temperature sensor’ in vivo. Although it is known that agonists of this ‘cold temperature sensor’, such as menthol and icilin, cause a transient increase in body temperature (Tb, it is not known if TRPM8 plays a role in Tb regulation. Since TRPM8 has been considered as a potential target for chronic pain therapeutics, we have investigated the role of TRPM8 in Tb regulation. Results We characterized five chemically distinct compounds (AMG0635, AMG2850, AMG8788, AMG9678, and Compound 496 as potent and selective antagonists of TRPM8 and tested their effects on Tb in rats and mice implanted with radiotelemetry probes. All five antagonists used in the study caused a transient decrease in Tb (maximum decrease of 0.98°C. Since thermoregulation is a homeostatic process that maintains Tb about 37°C, we further evaluated whether repeated administration of an antagonist attenuated the decrease in Tb. Indeed, repeated daily administration of AMG9678 for four consecutive days showed a reduction in the magnitude of the Tb decrease Day 2 onwards. Conclusions The data reported here demonstrate that TRPM8 channels play a role in Tb regulation. Further, a reduction of magnitude in Tb decrease after repeated dosing of an antagonist suggests that TRPM8’s role in Tb maintenance may not pose an issue for developing TRPM8 antagonists as therapeutics.

  17. Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Yan Wang

    2017-01-01

    Full Text Available After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator.

  18. A study on thimble plug removal for PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Dong Soo; Lee, Chang Sup; Lee, Jae Yong; Jun, Hwang Yong [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for 8 Westinghouse type PWR plants in Korea as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increases approximately by 1.2%. The resulting DNBR penalties can be covered with the current DNBR margin. Accident analyses are also investigated that the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation. 8 refs., 5 tabs. (Author)

  19. TEMP: a computer code to calculate fuel pin temperatures during a transient. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Bard, F E; Christensen, B Y; Gneiting, B C

    1980-04-01

    The computer code TEMP calculates fuel pin temperatures during a transient. It was developed to accommodate temperature calculations in any system of axi-symmetric concentric cylinders. When used to calculate fuel pin temperatures, the code will handle a fuel pin as simple as a solid cylinder or as complex as a central void surrounded by fuel that is broken into three regions by two circumferential cracks. Any fuel situation between these two extremes can be analyzed along with additional cladding, heat sink, coolant or capsule regions surrounding the fuel. The one-region version of the code accurately calculates the solution to two problems having closed-form solutions. The code uses an implicit method, an explicit method and a Crank-Nicolson (implicit-explicit) method.

  20. SCAR - Post-Accident Simulator SIPA with safety analysis code CATHARE-2 and PWR cold shutdown state simulation

    Energy Technology Data Exchange (ETDEWEB)

    Farvacque, M.; Faydide, B. [CEA Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique; Iffenecker, F.; Pentori, B. [Electricite de France, 75 - Paris (France); Dufeil, Ph.; Raimond, E. [Institut de Radioprotection et de Surete Nucleaire (IRSN), 92 - Clamart (France)

    2003-07-01

    The use of Cathare in the simulators of pressurized water reactors has been effective since the beginning of the nineties. Scar project is the second stage of the Cathare strategy for the simulators, its main objective is the extension of the field of simulation to the accident situations in cold shutdown states. Work was carried out in 3 major areas: modelling, optimization and integration in the simulator. Throughout the project, the developments were part of a 3 stages validation strategy: -) elementary tests of the developments of new model on the N4 (1450 MW PWR); -) analytical tests and systems to ensure non regression of the validation of the physical laws of the Cathare code during the modifications carried out within the optimization stage; and -) overall tests of the SIPA-CP1 (900 MW PWR) simulator, controlled automatically by programmed scenarios including the transients which are carried out in PWR, the transients of the Regulatory Guides and the accident transients.

  1. A High Temperature Experimental Characterization Procedure for Oxide-Based Thermoelectric Generator Modules under Transient Conditions

    Directory of Open Access Journals (Sweden)

    Elena Anamaria Man

    2015-11-01

    Full Text Available The purpose of this study is to analyze the steady-state and transient behavior of the electrical and thermal parameters of thermoelectric generators (TEGs. The focus is on the required wait-time between measurements in order to reduce measurement errors which may appear until the system reaches steady-state. By knowing this waiting time, the total characterization time can also be reduced. The experimental characterization process is performed on a test rig known as TEGeta, which can be used to assess the output characteristics of TEG modules under different load values and temperature conditions. The setup offers the possibility to control the hot side temperature up to 1000 °C with a load variation range value between 0.22–8.13 Ω. A total of ten thermocouples are placed in the setup with the purpose of measuring the temperature in specific points between the heater and the heat sink. Based on the readings, the temperature on the hot and cold side of the modules can be extrapolated. This study provides quantitative data on the minimum waiting time of the temperatures in the surrounding system to reach equilibrium. Laboratory tests are performed on a calcium-manganese oxide module at temperatures between 400 and 800 °C to explore the high temperatures features of the setup.

  2. Temperature and Voltage Coupling to Channel Opening in Transient Receptor Potential Melastatin 8 (TRPM8)*♦

    Science.gov (United States)

    Raddatz, Natalia; Castillo, Juan P.; Gonzalez, Carlos; Alvarez, Osvaldo; Latorre, Ramon

    2014-01-01

    Expressed in somatosensory neurons of the dorsal root and trigeminal ganglion, the transient receptor potential melastatin 8 (TRPM8) channel is a Ca2+-permeable cation channel activated by cold, voltage, phosphatidylinositol 4,5-bisphosphate, and menthol. Although TRPM8 channel gating has been characterized at the single channel and macroscopic current levels, there is currently no consensus regarding the extent to which temperature and voltage sensors couple to the conduction gate. In this study, we extended the range of voltages where TRPM8-induced ionic currents were measured and made careful measurements of the maximum open probability the channel can attain at different temperatures by means of fluctuation analysis. The first direct measurements of TRPM8 channel temperature-driven conformational rearrangements provided here suggest that temperature alone is able to open the channel and that the opening reaction is voltage-independent. Voltage is a partial activator of TRPM8 channels, because absolute open probability values measured with fully activated voltage sensors are less than 1, and they decrease as temperature rises. By unveiling the fast temperature-dependent deactivation process, we show that TRPM8 channel deactivation is well described by a double exponential time course. The fast and slow deactivation processes are temperature-dependent with enthalpy changes of 27.2 and 30.8 kcal mol−1. The overall Q10 for the closing reaction is about 33. A three-tiered allosteric model containing four voltage sensors and four temperature sensors can account for the complex deactivation kinetics and coupling between voltage and temperature sensor activation and channel opening. PMID:25352597

  3. Transient turbid water mass reduces temperature-induced coral bleaching and mortality in Barbados.

    Science.gov (United States)

    Oxenford, Hazel A; Vallès, Henri

    2016-01-01

    Global warming is seen as one of the greatest threats to the world's coral reefs and, with the continued rise in sea surface temperature predicted into the future, there is a great need for further understanding of how to prevent and address the damaging impacts. This is particularly so for countries whose economies depend heavily on healthy reefs, such as those of the eastern Caribbean. Here, we compare the severity of bleaching and mortality for five dominant coral species at six representative reef sites in Barbados during the two most significant warm-water events ever recorded in the eastern Caribbean, i.e., 2005 and 2010, and describe prevailing island-scale sea water conditions during both events. In so doing, we demonstrate that coral bleaching and subsequent mortality were considerably lower in 2010 than in 2005 for all species, irrespective of site, even though the anomalously warm water temperature profiles were very similar between years. We also show that during the 2010 event, Barbados was engulfed by a transient dark green turbid water mass of riverine origin coming from South America. We suggest that reduced exposure to high solar radiation associated with this transient water mass was the primary contributing factor to the lower bleaching and mortality observed in all corals. We conclude that monitoring these episodic mesoscale oceanographic features might improve risk assessments of southeastern Caribbean reefs to warm-water events in the future.

  4. Temperature-dependent transformation thermotics for unsteady states: Switchable concentrator for transient heat flow

    Energy Technology Data Exchange (ETDEWEB)

    Li, Ying, E-mail: 13110290008@fudan.edu.cn [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Shen, Xiangying, E-mail: 13110190068@fudan.edu.cn [Department of Physics, State Key Laboratory of Surface Physics, and Collaborative Innovation Center of Advanced Microstructures, Fudan University, Shanghai 200433 (China); Huang, Jiping, E-mail: jphuang@fudan.edu.cn [Department of Physics, State Key Laboratory of Surface Physics, and Collaborative Innovation Center of Advanced Microstructures, Fudan University, Shanghai 200433 (China); Ni, Yushan, E-mail: niyushan@fudan.edu.cn [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)

    2016-04-22

    For manipulating heat flow efficiently, recently we established a theory of temperature-dependent transformation thermotics which holds for steady-state cases. Here, we develop the theory to unsteady-state cases by considering the generalized Fourier's law for transient thermal conduction. As a result, we are allowed to propose a new class of intelligent thermal metamaterial — switchable concentrator, which is made of inhomogeneous anisotropic materials. When environmental temperature is below or above a critical value, the concentrator is automatically switched on, namely, it helps to focus heat flux in a specific region. However, the focusing does not affect the distribution pattern of temperature outside the concentrator. We also perform finite-element simulations to confirm the switching effect according to the effective medium theory by assembling homogeneous isotropic materials, which bring more convenience for experimental fabrication than inhomogeneous anisotropic materials. This work may help to figure out new intelligent thermal devices, which provide more flexibility in controlling heat flow, and it may also be useful in other fields that are sensitive to temperature gradient, such as the Seebeck effect. - Highlights: • Established the unsteady-state temperature dependent transformation thermotics. • A thermal concentrator with switchable functionality. • An effective-medium design for experimental realization.

  5. Room temperature screening of thermal conductivity by means of thermal transient measurements

    Science.gov (United States)

    García-Cañadas, Jorge; Cheng, Shudan; Márquez-García, Lourdes; Prest, Martin J.; Akbari-Rahimabadi, Ahmad; Min, Gao

    2016-10-01

    A proof of concept of the possibility to estimate thermal conductivity of bulk disc samples at room temperature by means of thermal decays is demonstrated. An experimental set-up was designed and fabricated, which is able to perform thermal transient measurements by using a specially designed multifunctional probe that has the ability to measure temperature at its tip. Initially, the probe is heated by a heater coil located in its interior until the tip temperature reaches a steady state. Then, the probe is contacted with a disc sample which produces a temperature decay until a new state is reached. The difference between the initial and final states temperatures shows a correlation with the thermal conductivity of the sample. Employing a calibration equation, obtained using reference materials, the thermal conductivity can be calculated. Reasonably good random and systematic errors (<13% and ~9% respectively) are obtained. Theoretical simulations performed using COMSOL show a good qualitative agreement with experimental results. This new method involves an inexpensive and simple set-up which can be especially useful for thermal conductivity screening and high-throughput measurements.

  6. Low temperature transient response and electroluminescence characteristics of OLEDs based on Alq{sub 3}

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Chao [Key Laboratory of Semiconductor Material Sciences, Beijing Key Laboratory of Low Dimensional Semiconductor Materials and Devices, Institute of Semiconductors, Chinese Academy of Sciences, Beijing 100083 (China); College of Materials Science and Optoelectronic Devices, University of Chinese Academy of Sciences, Beijing 100049 (China); Guan, Min, E-mail: guanmin@semi.ac.cn [Key Laboratory of Semiconductor Material Sciences, Beijing Key Laboratory of Low Dimensional Semiconductor Materials and Devices, Institute of Semiconductors, Chinese Academy of Sciences, Beijing 100083 (China); Zhang, Yang [Key Laboratory of Semiconductor Material Sciences, Beijing Key Laboratory of Low Dimensional Semiconductor Materials and Devices, Institute of Semiconductors, Chinese Academy of Sciences, Beijing 100083 (China); College of Materials Science and Optoelectronic Devices, University of Chinese Academy of Sciences, Beijing 100049 (China); Li, Yiyang; Liu, Shuangjie [Key Laboratory of Semiconductor Material Sciences, Beijing Key Laboratory of Low Dimensional Semiconductor Materials and Devices, Institute of Semiconductors, Chinese Academy of Sciences, Beijing 100083 (China); Zeng, Yiping [Key Laboratory of Semiconductor Material Sciences, Beijing Key Laboratory of Low Dimensional Semiconductor Materials and Devices, Institute of Semiconductors, Chinese Academy of Sciences, Beijing 100083 (China); College of Materials Science and Optoelectronic Devices, University of Chinese Academy of Sciences, Beijing 100049 (China)

    2017-08-15

    Highlights: • The dependency relation between transmission rate and electron transport layer is revealed. • The critical temperature points for the influence of luminescent materials and injection barriers on delay time are found. • The influence of light-emitting material and injection layer on carrier accumulation is quantified. - Abstract: In this work, the organic light-emitting diodes (OLEDs) based on Alq{sub 3} are fabricated. In order to make clear the transport mechanism of carriers in organic light-emitting devices at low temperature, detailed electroluminescence transient response and the current-voltage–luminescence (I–V–L) characteristics under different temperatures in those OLEDs are investigated. It founds that the acceleration of brightness increases with increasing temperature is maximum when the temperature is 200 K and it is mainly affected by the electron transport layer (Alq{sub 3}). The MoO{sub 3} injection layer and the electroluminescent layer have great influence on the delay time when the temperature is 200 K. Once the temperature is greater than 250 K, the delay time is mainly affected by the MoO{sub 3} injection layer. On the contrary, the fall time is mainly affected by the electroluminescent material. The V{sub f} is the average growth rate of fall time when the temperature increases 1 K which represents the accumulation rate of carriers. The difference between V{sub f} caused by the MoO{sub 3} injection layer is 0.52 us/K and caused by the electroluminescent material Ir(ppy){sub 3} is 0.73 us/K.

  7. Fuel-rod temperature transients during LWR degraded-core accidents

    Energy Technology Data Exchange (ETDEWEB)

    Briscoe, F; Rivard, J B; Young, M F

    1982-01-01

    Heat transfer models of fuel rods and coolant have been developed in support of LWR damaged fuel studies underway at Sandia National Laboratories for the NRC. A one-dimensional, full-length model simulates a PWR fuel rod; a two-dimensional, 0.5 m model simulates 9-rod bundle experiments to be performed in the Annular Core Research Reactor. The models include zircaloy oxidation, heat transfer by convecting steam/hydrogen flow, and radiation between surfaces through an absorbing/emitting gas. Characteristics of the one-dimensional reactor fuel rod model for two types of accident sequence are reported, as well as comparisons with MARCH code results.

  8. Transient studies of low temperature catalysts for methane conversion. Final report, [September 1992--March 1996

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, E.E.

    1996-09-30

    The objective of this project is to use transient techniques to study gas surface interactions during the oxidative conversion of methane. Two groups of catalysts were studied: a double oxide of vanadium and phosphate or VPO, and double oxides of Ni, Co and Rh and lanthana. The objective of the studies involving the VPO catalyst was to understand gas-surface interactions leading to the formation of formaldehyde. In the second group of catalysts, involving metallo-oxides, the main objective was to study the gas-surface interactions that determine the selectivity to C{sub 2} hydrocarbons or synthesis gas. Transient techniques were used to study the methane-surface interactions and the role of lattice oxygen. The selection of the double oxides was made on the hypothesis that the metal oxide would provide an increase interaction with methane whereas the phosphate or lanthanide would provide the sites for oxygen adsorption. The hypothesis behind this selection of catalysts was that increasing the methane interaction with the catalysts would lower the reaction temperature and thus increase the selectivity to the desired products over the total oxidation reaction. In both groups of catalysts the role of Li as a modifier of the selectivity was also studied in detail.

  9. 3D transient model to predict temperature and ablated areas during laser processing of metallic surfaces

    Directory of Open Access Journals (Sweden)

    Babak. B. Naghshine

    2017-02-01

    Full Text Available Laser processing is one of the most popular small-scale patterning methods and has many applications in semiconductor device fabrication and biomedical engineering. Numerical modelling of this process can be used for better understanding of the process, optimization, and predicting the quality of the final product. An accurate 3D model is presented here for short laser pulses that can predict the ablation depth and temperature distribution on any section of the material in a minimal amount of time. In this transient model, variations of thermal properties, plasma shielding, and phase change are considered. Ablation depth was measured using a 3D optical profiler. Calculated depths are in good agreement with measured values on laser treated titanium surfaces. The proposed model can be applied to a wide range of materials and laser systems.

  10. Transient analytical solution of temperature distribution and fracture limits in pulsed solid-state laser rod

    Directory of Open Access Journals (Sweden)

    Shibib Khalid S.

    2017-01-01

    Full Text Available The exact analytical solution of axis-symmetry transient temperature and Tresca failure stress in pulsed mode solid-state laser rod is derived using integral transform method. The result obtained from this work is compared with previously published data and good agreement is found. The effect of increasing period is studied, and it is found that at constant pulse width as the period is increased, the allowable pumping power is increased too. Furthermore, the effect of changing pulse width with a constant period is studied, and it is found that as the pulse width is increased, the allowable pumping power is decreased. The effect of duty cycle is studied also and it is found that as duty cycle is increased the allowable pumping power is decreased. This work permits proper selection of pulse width, period and duty cycle to avoid laser rod fracture while obtaining maximum output laser power in the designing of laser system.

  11. Transient temperature rise in a mouse due to low-frequency regional hyperthermia

    Energy Technology Data Exchange (ETDEWEB)

    Trakic, Adnan; Liu Feng; Crozier, Stuart [School of Information Technology and Electrical Engineering, University of Queensland, Brisbane, Qld 4072 (Australia)

    2006-04-07

    A refined nonlinear heat transfer model of a mouse has been developed to simulate the transient temperature rise in a neoplastic tumour and neighbouring tissue during regional hyperthermia using a 150 kHz inductive coil. In this study, we incorporate various bio-energetic enhancements to the heat transfer equation and numerical validations based on experimental findings for the mouse, in terms of nonlinear metabolic heat production, homeothermy, blood perfusion parameters, thermoregulation, psychological and physiological effects. The discretized bio-heat transfer equation has been validated with the commercial software FEMLAB on a canonical multi-sphere object before applying the scheme to the inhomogeneous mouse voxel phantom. The time-dependent numerical results of regional hyperthermia of mouse thigh have been compared with the available experimental temperature results with only a few small disparities. During the first 20 min of local unfocused heating, the temperature in the tumour and the surrounding tissue increased by around 7.5 deg. C. The objective of this preliminary study was to develop a validated electrothermal numerical scheme for inductive hyperthermia of a small mammal with the intention of expanding the model into a complete numerical solution involving ferromagnetic nanoparticles for targeted heating of tumours at low frequencies. In addition, the numerical scheme herein could assist in optimizing and tailoring of focused electromagnetic fields for hyperthermia.

  12. Probabilistic Fracture Mechanics Analysis of Boling Water Reactor Vessel for Cool-Down and Low Temperature Over-Pressurization Transients

    Directory of Open Access Journals (Sweden)

    Jeong Soon Park

    2016-04-01

    Full Text Available The failure probabilities of the reactor pressure vessel (RPV for low temperature over-pressurization (LTOP and cool-down transients are calculated in this study. For the cool-down transient, a pressure–temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition (RTNDT. The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

  13. Nuclear data uncertainties by the PWR MOX/UO{sub 2} core rod ejection benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Pasichnyk, I.; Klein, M.; Velkov, K.; Zwermann, W.; Pautz, A. [Boltzmannstr. 14, D-85748 Garching b. Muenchen (Germany)

    2012-07-01

    Rod ejection transient of the OECD/NEA and U.S. NRC PWR MOX/UO{sub 2} core benchmark is considered under the influence of nuclear data uncertainties. Using the GRS uncertainty and sensitivity software package XSUSA the propagation of the uncertainties in nuclear data up to the transient calculations are considered. A statistically representative set of transient calculations is analyzed and both integral as well as local output quantities are compared with the benchmark results of different participants. It is shown that the uncertainties in nuclear data play a crucial role in the interpretation of the results of the simulation. (authors)

  14. Modeling of thermoelectric module operation in inhomogeneous transient temperature field using finite element method

    Directory of Open Access Journals (Sweden)

    Nikolić Radovan H.

    2014-01-01

    Full Text Available This paper is the result of research and operation modeling of the new systems for cooling of cutting tools based on thermoelectric module. A copper inlay with thermoelectric module on the back side was added to a standard turning tool for metal processing. For modeling and simulating the operation of thermoelectric module, finite element method was used as a method for successful solving the problems of inhomogeneous transient temperature field on the cutting tip of lathe knives. Developed mathematical model is implemented in the software package PAK-T through which numerical results are obtained. Experimental research was done in different conditions of thermoelectric module operation. Cooling of the hot module side was done by a heat exchanger based on fluid using automatic temperature regulator. After the calculation is done, numerical results are in good agreement with experimental. It can be concluded that developed mathematical model can be used successfully for modeling of cooling of cutting tools. [Projekat Ministarstva nauke Republike Srbije, br. TR32036

  15. Effects of transient high temperature treatment on the intestinal flora of the silkworm Bombyx mori.

    Science.gov (United States)

    Sun, Zhenli; Kumar, Dhiraj; Cao, Guangli; Zhu, Liyuan; Liu, Bo; Zhu, Min; Liang, Zi; Kuang, Sulan; Chen, Fei; Feng, Yongjie; Hu, Xiaolong; Xue, Renyu; Gong, Chengliang

    2017-06-13

    The silkworm Bombyx mori is a poikilotherm and is therefore sensitive to various climatic conditions. The influence of temperature on the intestinal flora and the relationship between the intestinal flora and gene expression in the silkworm remain unknown. In the present study, changes of the intestinal flora at 48, 96 and 144 h following transient high temperature treatment (THTT) of 37 °C for 8 h were investigated. According to principal component analysis, the abundances of Enterococcus and Staphylococcus showed a negative correlation with other dominant genera. After THTT, the gene expression levels of spatzle-1 and dicer-2 were increased and decreased, respectively, which suggested that the Toll and RNAi pathways were activated and suppressed, respectively. The species-gene expression matrix confirmed that the spatzle-1 and dicer-2 gene expression levels were negatively and positively correlated, respectively, with the abundance of Enterococcus and Staphylococcus in the control. The abundance of Variovorax post-THTT was positively correlated with the spatzle-1 gene expression level, whereas the community richness of Enterococcus was negatively correlated with the spatzle-1 gene expression level and positively correlated with the dicer-2. The results of the present investigation provide new evidence for understanding the relationships among THTT, intestinal flora and host gene expression.

  16. Development and validation of the 3-D PWR core dynamics SIMTRAN code

    Energy Technology Data Exchange (ETDEWEB)

    Merino, F. (Inst. Fusion Nuclear, Univ. Politecnica de Madrid (Spain)); Ahnert, C. (Inst. Fusion Nuclear, Univ. Politecnica de Madrid (Spain)); Aragones, J.M. (Inst. Fusion Nuclear, Univ. Politecnica de Madrid (Spain))

    1993-04-01

    We discuss the main features and results of the SIMTRAN development and validation work. Included in the first are the extension of the nodal neutronic solution to account for intranodal shape and spectrum, due to both heterogeneities and flux gradients, the implicit scheme for spatial kinetics with six delayed neutron precursors and the integration of the neutronic and thermohydraulic solutions on an staggered time mesh. Validation results are discussed for the NEACRP 3-D PWR Core Transient Benchmark and an actual transient with sudden increase of core flow occurred in the Vandellos-II 3-loop PWR NPP. Agreement with the reference numerical solution and measured plant data is shown for both problems. (orig./DG)

  17. ANALISIS SENSITIVITAS TURBULENSI ALIRAN PADA KANAL BAHAN BAKAR PWR BERBASIS CFD

    Directory of Open Access Journals (Sweden)

    Endiah Puji Hastuti

    2015-04-01

    yang sangat lama dan membutuhkan memori yang besar. Kata kunci: aliran turbulen, kanal PWR, CFD, tunak, transien   Coolant flow turbulence on heat transfer process serves to enhance the heat transfer coefficient, likewise flow in the fuel sub channel. Computational fluid dynamic program, FLUENT is a computational program based on finite element, that is able to predict and analyze the dynamics of fluid flow phenomena, accurately. CFD calculation program is selected in this study because of its accurately and it also can provide good visualization. Purpose of this research was to understand the characteristics of heat transfer, mass and momentum of the fuel rod to the coolant visually on: the temperature field, pressure field, and the kinetic energy field, as a function of the flow dynamics within fuel channel, on steady state and transient condition. Analysis of flow dynamics in the fuel channel base on CFD was done by using the PWR sample data with reactor power of 1000 MWe on 17x17 array of fuel. To examine the sensitivity of the flow equation in accordance with the model of turbulent flow on fuel channel, the turbulence equation model of k-omega (Ƙ-ω, k-epsilon (Ƙ-ε, and Reynold stress model (RSM for steady state was used, while for transient turbulence model DES and LES are applied. In the sensitivity analysis of turbulent flow, hexahedral mesh model of three cell geometry each are 0.5 mm, 0.2 mm and 0.15 mm, was selected. The analysis shows that there are similar results of turbulen model Ƙ-ε and Ƙ-ω standard, on steady state analysis. Comparing with Dittus Boelter criteria for Nusselt number, the Reynolds stress model (RSM is recommended. Sensitivity analysis of mesh geometry between cell size 0.5 mm, 0.2 mm and 0.15 mm, indicating that the cell size of 0.5 mm was sufficient. Developed flow already reached on DES and LES model, however only for short time (3 seconds for transient condition. LES model need very long computation time and big memory

  18. Nonlinear Fuzzy Model Predictive Control for a PWR Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Xiangjie Liu

    2014-01-01

    Full Text Available Reliable power and temperature control in pressurized water reactor (PWR nuclear power plant is necessary to guarantee high efficiency and plant safety. Since the nuclear plants are quite nonlinear, the paper presents nonlinear fuzzy model predictive control (MPC, by incorporating the realistic constraints, to realize the plant optimization. T-S fuzzy modeling on nuclear power plant is utilized to approximate the nonlinear plant, based on which the nonlinear MPC controller is devised via parallel distributed compensation (PDC scheme in order to solve the nonlinear constraint optimization problem. Improved performance compared to the traditional PID controller for a TMI-type PWR is obtained in the simulation.

  19. New instrumentation of reactor water level for PWR; Nueva Instrumentacion de nivel de agua del reactor para PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kaercher, S.

    2005-07-01

    Today, many PWR reactors are equipped with a reactor water level instrumentation system based on different measurement methods. Due to obsolescence issues, FRAMATOME ANP started to develop and quality a new water level measurement system using heated und unheated thermocouple measurements. the measuring principle is based on the fact that the heat transfer in water is considerably higher than in steam. The electronic cabinet for signal processing is based on a proven technology already developed, qualified and installed by FRAMATOME ANP in several NPPs. It is equipped with and advanced temperature measuring transducer for acquisition and processing of thermocouple signals. (Author)

  20. High-temperature deformation field measurement by combining transient aerodynamic heating simulation system and reliability-guided digital image correlation

    Science.gov (United States)

    Pan, Bing; Wu, Dafang; Xia, Yong

    2010-09-01

    To determine the full-field high-temperature thermal deformation of the structural materials used in high-speed aerospace flight vehicles, a novel non-contact high-temperature deformation measurement system is established by combining transient aerodynamic heating simulation device with the reliability-guided digital image correlation (RG-DIC). The test planar sample with size varying from several mm 2 to several hundreds mm 2 can be heated from room temperature to 1100 °C rapidly and accurately using the infrared radiator of the transient aerodynamic heating simulation system. The digital images of the test sample surface at various temperatures are recorded using an ordinary optical imaging system. To cope with the possible local decorrelated regions caused by black-body radiation within the deformed images at the temperatures over 450 °C, the RG-DIC technique is used to extract full-field in-plane thermal deformation from the recorded images. In validation test, the thermal deformation fields and the values of coefficient of thermal expansion (CTEs) of a chromiumnickel austenite stainless steel sample from room temperature to 550 °C is measured and compared with the well-established handbook value, confirming the effectiveness and accuracy of the proposed technique. The experimental results reveal that the present system using an ordinary optical imaging system, is able to accurately measure full-field thermal deformation of metals and alloys at temperatures not exceeding 600 °C.

  1. Assessment of void swelling in austenitic stainless steel PWR core internals.

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H. M.; Energy Technology

    2006-01-31

    As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling

  2. Approximate analytic solutions of transient nonlinear heat conduction with temperature-dependent thermal diffusivity

    OpenAIRE

    Mustafa, M.T.; Arif, A.F.M.; Masood, Khalid

    2014-01-01

    A new approach for generating approximate analytic solutions of transient nonlinear heat conduction problems is presented. It is based on an effective combination of Lie symmetry method, homotopy perturbation method, finite element method, and simulation based error reduction techniques. Implementation of the proposed approach is demonstrated by applying it to determine approximate analytic solutions of real life problems consisting of transient nonlinear heat conduction in semi-infinite bars...

  3. Full MOX core design for advanced PWR

    Energy Technology Data Exchange (ETDEWEB)

    Tochihara, H.; Komano, Y.; Ishida, M.; Mukai, H. [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan)

    1995-12-31

    A full MOX core is an attractive option for large consumption of the recycled plutonium from reprocessed LWR fuel. The feasibility is verified of a full MOX core in a PWR with only small hardware modifications. The advanced PWR has been selected for this purpose. The full MOX core is feasible by increasing the number of control rods and adopting the enriched {sup 10}B in the soluble boron of reactor coolant system. The full MOX cores can be designed using one Pu` content per assembly and without any burnable absorbers. (author) 2 refs.

  4. Computational methods and implementation of the 3-D PWR core dynamics SIMTRAN code for online surveillance and prediction

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J.M.; Ahnert, C. [Universidad Politecnica de Madrid (Spain)

    1995-12-31

    New computational methods have been developed in our 3-D PWR core dynamics SIMTRAN code for online surveillance and prediction. They improve the accuracy and efficiency of the coupled neutronic-thermalhydraulic solution and extend its scope to provide, mainly, the calculation of: the fission reaction rates at the incore mini-detectors; the responses at the excore detectors (power range); the temperatures at the thermocouple locations; and the in-vessel distribution of the loop cold-leg inlet coolant conditions in the reflector and core channels, and to the hot-leg outlets per loop. The functional capabilities implemented in the extended SIMTRAN code for online utilization include: online surveillance, incore-excore calibration, evaluation of peak power factors and thermal margins, nominal update and cycle follow, prediction of maneuvers and diagnosis of fast transients and oscillations. The new code has been installed at the Vandellos-II PWR unit in Spain, since the startup of its cycle 7 in mid-June, 1994. The computational implementation has been performed on HP-700 workstations under the HP-UX Unix system, including the machine-man interfaces for online acquisition of measured data and interactive graphical utilization, in C and X11. The agreement of the simulated results with the measured data, during the startup tests and first months of actual operation, is well within the accuracy requirements. The performance and usefulness shown during the testing and demo phase, to be extended along this cycle, has proved that SIMTRAN and the man-machine graphic user interface have the qualities for a fast, accurate, user friendly, reliable, detailed and comprehensive online core surveillance and prediction.

  5. Transient flowing-fluid temperature modeling in reservoirs with large drawdowns

    National Research Council Canada - National Science Library

    N Chevarunotai; A R Hasan; C S Kabir; R Islam

    2017-01-01

    Modern downhole temperature measurements indicate that bottomhole fluid temperature can be significantly higher or lower than the original reservoir temperature, especially in reservoirs where high...

  6. Thermal conductance of interfaces with molecular layers - low temperature transient absorption study on gold nanorods supported on self assembled monolayers

    Science.gov (United States)

    Wang, Wei; Huang, Jingyu; Murphy, Catherine; Cahill, David; University of Illinois At Urbana Champaign, Department of Materials Science; Engineering Team; Department Collaboration

    2011-03-01

    While heat transfer via phonons across solid-solid boundary has been a core field in condense matter physics for many years, vibrational energy transport across molecular layers has been less well elucidated. We heat rectangular-shaped gold nanocrystals (nanorods) with Ti-sapphire femtosecond pulsed laser at their longitudinal surface plasmon absorption wavelength to watch how their temperature evolves in picoseconds transient. We observed single exponential decay behavior, which suggests that the heat dissipation is only governed by a single interfacial conductance value. The ``RC'' time constant was 300ps, corresponding to a conductance value of 95MW/ m 2 K. This interfacial conductance value is also a function of ambient temperature since at temperatures as low as 80K, which are below the Debye temperature of organic layers, several phonon modes were quenched, which shut down the dominating channels that conduct heat at room temperature.

  7. Temperature shock, injury and transient sensitivity to nisin in Gram negatives.

    Science.gov (United States)

    Boziaris, I S; Adams, M R

    2001-10-01

    The effect of thermal stresses on survival, injury and nisin sensitivity was investigated in Salmonella Enteritidis PT4, PT7 and Pseudomonas aeruginosa. Heating at 55 degrees C, rapid chilling to 0.5 degrees C or freezing at -20 degrees C produced transient sensitivity to nisin. Cells were only sensitive if nisin was present during stress. Resistance recovered rapidly afterwards, though some cells displayed residual injury. Injury was assessed by SDS sensitivity, hydrophobicity changes, lipopolysaccharide release and NPN uptake. LPS release and hydrophobicity were not always associated with transient nisin sensitivity. Uptake of NPN correlated better but persisted longer after treatment. Thermal shocks produce transient injury to the outer membrane, allowing nisin access. After treatment, the permeability barrier is rapidly restored by a process apparently involving reorganization rather than biosynthetic repair. Inclusion of nisin during food treatments that impose sub-lethal stress on Gram negatives could increase process lethality, enhancing microbiological safety and stability.

  8. Full MOX core design for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Komano, Y.; Tochihara, H.; Ishida, M. [Mitsubishi Heavy Industries Ltd., Tokyo (Japan)

    1999-12-01

    Full MOX core design for APWR was analyzed in nuclear design, fuel integrity analysis, thermal hydraulic design and safety analysis et. al. Feasibility of Full MOX core was confirmed from these analyses without any large modifications. Full MOX PWR core has very good characteristics in which single Pu content in an assembly, burnable poison free, higher burnup and longer cycle operation are feasible. (author)

  9. Approximate Analytic Solutions of Transient Nonlinear Heat Conduction with Temperature-Dependent Thermal Diffusivity

    Directory of Open Access Journals (Sweden)

    M. T. Mustafa

    2014-01-01

    Full Text Available A new approach for generating approximate analytic solutions of transient nonlinear heat conduction problems is presented. It is based on an effective combination of Lie symmetry method, homotopy perturbation method, finite element method, and simulation based error reduction techniques. Implementation of the proposed approach is demonstrated by applying it to determine approximate analytic solutions of real life problems consisting of transient nonlinear heat conduction in semi-infinite bars made of stainless steel AISI 304 and mild steel. The results from the approximate analytical solutions and the numerical solution are compared indicating good agreement.

  10. 21-PWR Waste Package Side and End Impacts

    Energy Technology Data Exchange (ETDEWEB)

    T. Schmitt

    2005-08-29

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1.

  11. Calcium regulation by temperature-sensitive transient receptor potential channels in human uveal melanoma cells.

    Science.gov (United States)

    Mergler, Stefan; Derckx, Raissa; Reinach, Peter S; Garreis, Fabian; Böhm, Arina; Schmelzer, Lisa; Skosyrski, Sergej; Ramesh, Niraja; Abdelmessih, Suzette; Polat, Onur Kerem; Khajavi, Noushafarin; Riechardt, Aline Isabel

    2014-01-01

    Uveal melanoma (UM) is both the most common and fatal intraocular cancer among adults worldwide. As with all types of neoplasia, changes in Ca(2+) channel regulation can contribute to the onset and progression of this pathological condition. Transient receptor potential channels (TRPs) and cannabinoid receptor type 1 (CB1) are two different types of Ca(2+) permeation pathways that can be dysregulated during neoplasia. We determined in malignant human UM and healthy uvea and four different UM cell lines whether there is gene and functional expression of TRP subtypes and CB1 since they could serve as drug targets to either prevent or inhibit initiation and progression of UM. RT-PCR, Ca(2+) transients, immunohistochemistry and planar patch-clamp analysis probed for their gene expression and functional activity, respectively. In UM cells, TRPV1 and TRPM8 gene expression was identified. Capsaicin (CAP), menthol or icilin induced Ca(2+) transients as well as changes in ion current behavior characteristic of TRPV1 and TRPM8 expression. Such effects were blocked with either La(3+), capsazepine (CPZ) or BCTC. TRPA1 and CB1 are highly expressed in human uvea, but TRPA1 is not expressed in all UM cell lines. In UM cells, the CB1 agonist, WIN 55,212-2, induced Ca(2+) transients, which were suppressed by La(3+) and CPZ whereas CAP-induced Ca(2+) transients could also be suppressed by CB1 activation. Identification of functional TRPV1, TRPM8, TRPA1 and CB1 expression in these tissues may provide novel drug targets for treatment of this aggressive neoplastic disease. © 2013.

  12. Transient Modeling and Analysis of a Metabolic Heat-Regenerated Temperature Swing Adsorption (MTSA) System for a PLSS

    Science.gov (United States)

    Iacomini, Christie; Powers, Aaron; Speight, Garland; Padilla, Sebastian; Paul, Heather L.

    2009-01-01

    A Metabolic heat-regenerated Temperature Swing Adsorption (MTSA) system is being developed for carbon dioxide, water and thermal control in a lunar and martian portable life support system (PLSS). A previous system analysis was performed to evaluate the impact of MTSA on PLSS design. That effort was Mars specific and assumed liquid carbon dioxide (LCO2) coolant made from martian resources. Transient effects were not considered but rather average conditions were used throughout the analysis. This effort takes into further consideration the transient effects inherent in the cycling MTSA system as well as assesses the use of water as coolant. Standard heat transfer, thermodynamic, and heat exchanger methods are presented to conduct the analysis. Assumptions and model verification are discussed. The tool was used to perform various system studies. Coolant selection was explored and takes into account different operational scenarios as the minimum bed temperature is driven by the sublimation temperature of the coolant (water being significantly higher than LCO2). From this, coolant mass is sized coupled with sorbent bed mass because MTSA adsorption performance decreases with increasing sublimation temperature. Reduction in heat exchanger performance and even removal of certain heat exchangers, like a recuperative one between the two sorbent beds, is also investigated. Finally, the coolant flow rate is varied over the cycle to determine if there is a more optimal means of cooling the bed from a mass perspective. Results of these studies and subsequent recommendations for system design are presented.

  13. Evolution of vertebrate transient receptor potential vanilloid 3 channels: opposite temperature sensitivity between mammals and western clawed frogs.

    Directory of Open Access Journals (Sweden)

    Shigeru Saito

    2011-04-01

    Full Text Available Transient Receptor Potential (TRP channels serve as temperature receptors in a wide variety of animals and must have played crucial roles in thermal adaptation. The TRP vanilloid (TRPV subfamily contains several temperature receptors with different temperature sensitivities. The TRPV3 channel is known to be highly expressed in skin, where it is activated by warm temperatures and serves as a sensor to detect ambient temperatures near the body temperature of homeothermic animals such as mammals. Here we performed comprehensive comparative analyses of the TRPV subfamily in order to understand the evolutionary process; we identified novel TRPV genes and also characterized the evolutionary flexibility of TRPV3 during vertebrate evolution. We cloned the TRPV3 channel from the western clawed frog Xenopus tropicalis to understand the functional evolution of the TRPV3 channel. The amino acid sequences of the N- and C-terminal regions of the TRPV3 channel were highly diversified from those of other terrestrial vertebrate TRPV3 channels, although central portions were well conserved. In a heterologous expression system, several mammalian TRPV3 agonists did not activate the TRPV3 channel of the western clawed frog. Moreover, the frog TRPV3 channel did not respond to heat stimuli, instead it was activated by cold temperatures. Temperature thresholds for activation were about 16 °C, slightly below the lower temperature limit for the western clawed frog. Given that the TRPV3 channel is expressed in skin, its likely role is to detect noxious cold temperatures. Thus, the western clawed frog and mammals acquired opposite temperature sensitivity of the TRPV3 channel in order to detect environmental temperatures suitable for their respective species, indicating that temperature receptors can dynamically change properties to adapt to different thermal environments during evolution.

  14. CFD simulation of a four-loop PWR at asymmetric operation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Jian-Ping; Yan, Li-Ming; Li, Feng-Chen, E-mail: lifch@hit.edu.cn

    2016-04-15

    Highlights: • A CFD numerical simulation procedure was established for simulating RPV of VVER-1000. • The established CFD approach was validated by comparing with available data. • Thermal hydraulic characteristics under asymmetric operation condition were investigated. • Apparent influences of the shutdown loop on its neighboring loops were obtained. - Abstract: The pressurized water reactor (PWR) with multiple loops may have abnormal working conditions with coolant pumps out of running in some loops. In this paper, a computational fluid dynamics (CFD) numerical study of the four-loop VVER-1000 PWR pressure vessel model was presented. Numerical simulations of the thermohydrodynamic characteristics in the pressure vessel were carried out at different inlet conditions with four and three loops running, respectively. At normal stead-state condition (four-loop running), different parameters were obtained for the full fluid domain, including pressure losses across different parts, pressure, velocity and temperature distributions in the reactor pressure vessel (RPV) and mass flow distribution of the coolant at the inlet of reactor core. The obtained results for pressure losses matched with the experimental reference values of the VVER-1000 PWR at Tianwan nuclear power plant (NPP). For most fuel assemblies (FAs), the inlet flow rates presented a symmetrical distribution about the center under full-loop operation conditions, which accorded with the practical distribution. These results indicate that it is now possible to study the dynamic transition process between different asymmetric operation conditions in a multi-loop PWR using the established CFD method.

  15. HEMERA: a 3D computational chain for PWR safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Bruna, G.B.; Dubois, F.; Fouquet, F.; Mury, E.; Normand, B.; Sargeni, A.; Scarcelli, F.; Touillon, R. [Institut de Radioprotection et de Surete Nucleaire, BP 17 - 92262 Fontenay-aux-Roses Cedex (France); Le Pallec, J.C.; Hourcade, E.; Richebois, E.; Poinot-Salanon, C. [Commissariat a l' energie atomique, Nuclear Energy Division, Systems and structures modelling department, Reactors studies and applied mathematic service, Centre de Saclay, 91 191 Gif sur Yvette Cedex (France)

    2008-07-01

    In the framework of their collaboration to develop a system to study reactor transients in 'safety-representative conditions', IRSN and CEA have launched the development of a fully coupled 3D computational chain, called HEMERA (Highly Evolutionary Methods for Extensive Reactor Analyses), based on the French SAPHYR code system, composed by APOLLO2, CRONOS2 and FLICA4 codes, and the system code CATHARE. It includes cross sections generation, steady-state, depletion and transient computation capabilities in a consistent approach. Multi-level and multi-dimensional models are developed to account for neutronics, core thermal-hydraulics, fuel thermal analysis and system thermal-hydraulics, dedicated to best-estimate simulations and sensitivity analysis. Currently Control Rod Ejection (REA) and Main Steam Line Break (MSLB) accidents are investigated. The HEMERA system is presently applied to French PWR. The present paper outlines the main physical phenomena to be accounted for in such a coupled computational chain with significant time and space effects. A selection of results is presented along with a comparison of the available levels of simulation 3D, ranging from assembly-wise to pin-wise in the core. (authors)

  16. Analysis and sensitivity studies with CORETRAN and RETRAN-3D of the NEACRP PWR rod ejection benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Ferroukhi, H.; Coddington, P. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    2001-07-01

    The OECD/NEA PWR rod ejection benchmark has been analysed using the 3-D nodal spatial-kinetic codes CORETRAN and RETRAN-3D. The following results were obtained. A) The agreement in 3-D solution between CORETRAN and RETRAN-3D was found to be very good both during steady-state and transient conditions. In particular at HZP (hot zero power), an excellent agreement in the initial steady-state 3-D power distribution and with regard to the core power excursion during the super-prompt critical phase of the transient (i.e. when the negative reactivity feedback is still very weak) was found. This illustrates the consistency in the neutronic solution between both codes. B) At both HZP and FP (full power) conditions, the CORETRAN and RETRAN-3D results lie well within the range of the previous benchmark solutions. In particular at HZP, both codes predict a power excursion and an increase in maximum pellet temperature that are among the closest results to those obtained with the benchmark reference solution. It must here be emphasised that these analyses are by no means a validation of the codes. However, the good agreement of both CORETRAN and RETRAN-3D with other 3-D solutions provides confidence in the ability of these codes to analyse LWR (light water reactor) core transients. In addition, it was found appropriate to perform, for this well-defined international benchmark problem, some sensitivity studies in order to assess the impact of modelling options on the CORETRAN and RETRAN-3D results. (authors)

  17. Evolution of the seasonal temperature cycle in a transient Holocene simulation: orbital forcing and sea-ice

    Directory of Open Access Journals (Sweden)

    N. Fischer

    2011-11-01

    Full Text Available Changes in the Earth's orbit lead to changes in the seasonal and meridional distribution of insolation. We quantify the influence of orbitally induced changes on the seasonal temperature cycle in a transient simulation of the last 6000 years – from the mid-Holocene to today – using a coupled atmosphere-ocean general circulation model (ECHAM5/MPI-OM including a land surface model (JSBACH.

    The seasonal temperature cycle responds directly to the insolation changes almost everywhere. In the Northern Hemisphere, its amplitude decreases according to an increase in winter insolation and a decrease in summer insolation. In the Southern Hemisphere, the opposite is true.

    Over the Arctic Ocean, decreasing summer insolation leads to an increase in sea-ice cover. The insulating effect of sea ice between the ocean and the atmosphere leads to decreasing heat flux and favors more "continental" conditions over the Arctic Ocean in winter, resulting in strongly decreasing temperatures. Consequently, there are two competing effects: the direct response to insolation changes and a sea-ice insulation effect. The sea-ice insulation effect is stronger, and thus an increase in the amplitude of the seasonal temperature cycle over the Arctic Ocean occurs. This increase is strongest over the Barents Shelf and influences the temperature response over northern Europe.

    We compare our modeled seasonal temperatures over Europe to paleo reconstructions. We find better agreements in winter temperatures than in summer temperatures and better agreements in northern Europe than in southern Europe, since the model does not reproduce the southern European Holocene summer cooling inferred from the paleo reconstructions. The temperature reconstructions for northern Europe support the notion of the influence of the sea-ice insulation effect on the evolution of the seasonal temperature cycle.

  18. GAIA: AREVAs New PWR fuel assembly design

    Energy Technology Data Exchange (ETDEWEB)

    Vollmert, N.; Gentet, G.; Louf, P.H.; Mindt, M.; O' Brian, J.; Peucker, J.

    2015-07-01

    GAIA is the label of a new PWR Fuel Assembly design developed by AREVA with the objective to provide its customers an advanced fuel assembly design regarding both robustness and performance. Since 2012 GAIA lead fuel assemblies are under irradiation in a Swedish reactor and since 2015 in a U.S. reactor. Visual inspections and examinations carried out so far during the outages confirmed the intended reliability, robustness and the performance enhancement of the design. (Author)

  19. Frictional Behavior of Fe-based Cladding Candidates for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Ho; Kim, Hyung-Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Byun, Thak Sang [Oak Ridge National Lab., Oak Ridge (United States)

    2014-10-15

    After the recent nuclear disaster at Fukushima Daiichi reactors, there is a growing consensus on the development of new fuel systems (i.e., accident-tolerant fuel, ATF) that have high safety margins under design-basis accident (DBA) and beyond design-basis accident (BDBA). A common objective of various developing candidates is to archive the outstanding corrosion-resistance under severe accidents such as DBA and DBDA conditions for decreasing hydrogen production and increasing coping time to respond to severe accidents. ATF could be defined as new fuel/cladding system with enhanced accident tolerant to loss of active cooling in the core for a considerably longer time period under severe accidents while maintaining or improving the fuel performance during normal operations. This means that, in normal operating conditions, new fuel systems should be applicable to current operating PWRs for suppressing various degradation mechanisms of current fuel assembly without excessive design changes. When considering that one of the major degradation mechanisms of PWR fuel assemblies is a grid-to-rod fretting (GTRF), it is necessary to examine the tribological behavior of various ATF candidates at initial development stage. In this study, friction and reciprocating wear behavior of two kinds of Fe-based ATF candidates were examined with a reciprocating wear tests at room temperature (RT) air and water. The objective is to examine the compatibilities of these Fe-based alloys against current Zr-based alloy properties, which is used as major structural materials of PWR fuel assemblies. The reciprocating wear behaviors of Fe-based accident-tolerant fuel cladding candidates against current Zr-based alloy has been studied using a reciprocating sliding wear tester in room temperature air and water. Frictional behavior and wear depth were used for evaluating the applicability and compatibilities of Fe-based candidates without significant design changes of PWR fuel assemblies

  20. Incompatibility between Zircaloy-2 and Inconel X-750 during Temperature Transients

    DEFF Research Database (Denmark)

    Warren, M. R.; Rørbo, Kaj; Adolph, Eivind

    1975-01-01

    The incompatibility of Zircaloy-2 and Inconel X-750 has been investigated between 1000°C and 1200°C (1200°C being the currently allowable maximum temperature in the acceptance criteria for ECCS for water reactors). It has been found for the temperatures of 1000°C and 1200°C that oxide thicknesses...

  1. Nonlinear Fuzzy Model Predictive Control for a PWR Nuclear Power Plant

    OpenAIRE

    Liu, Xiangjie; Wang, Mengyue

    2014-01-01

    Reliable power and temperature control in pressurized water reactor (PWR) nuclear power plant is necessary to guarantee high efficiency and plant safety. Since the nuclear plants are quite nonlinear, the paper presents nonlinear fuzzy model predictive control (MPC), by incorporating the realistic constraints, to realize the plant optimization. T-S fuzzy modeling on nuclear power plant is utilized to approximate the nonlinear plant, based on which the nonlinear MPC controller is devised via pa...

  2. Two-wavelength Raman imaging for non-intrusive monitoring of transient temperature in microfluidic devices

    Science.gov (United States)

    Kuriyama, Reiko; Sato, Yohei

    2014-09-01

    The present study proposes a non-intrusive visualization technique based on two-wavelength Raman imaging for in-situ monitoring of the unsteady temperature field in microfluidic systems. The measurement principle relies on the contrasting temperature dependencies of hydrogen-bonded and non-hydrogen-bonded OH stretching modes of the water Raman band, whose intensities were simultaneously captured by two cameras equipped with corresponding bandpass filters. The temperature distributions were then determined from the intensity ratio of the simultaneously-obtained Raman images, which enables compensation for temporal fluctuation and spatial inhomogeneity of the excitation laser intensity. A calibration experiment exhibited a linear relationship between the temperature and the intensity ratio in the range 293-343 K and least-regression analysis gave an uncertainty of 1.43 K at 95% confidence level. By applying the calibration data, time series temperature distributions were quantitatively visualized in a Y-shaped milli-channel at a spatial resolution of 6.0  ×  6.0 µm2 with an acquisition time of 16.5 s. The measurement result clearly exhibited the temporal evolution of the temperature field and was compared with the values obtained by thermocouples. This paper therefore demonstrates the viability of employing the two-wavelength Raman imaging technique for temperature measurements in microfluidic devices.

  3. A High Temperature Experimental Characterization Procedure for Oxide-Based Thermoelectric Generator Modules under Transient Conditions

    DEFF Research Database (Denmark)

    Man, Elena Anamaria; Schaltz, Erik; Rosendahl, Lasse

    2015-01-01

    to 1000 °C with a load variation range value between [0.22 – 8.13] Ω. A total of ten thermocouples are placed in the setup with the purpose of measuring the temperature in specific points between the heater and the heat sink. Based on the readings, the temperature on the hot and cold side of the modules...... can be extrapolated. Additionally, the set of measurements can offer heat flux information on both hot and cold sides of the system. This study provides quantitative data on the minimum waiting time of the temperatures in the surrounding system to reach equilibrium. Laboratory test results...

  4. Numerical simulation of transient moisture and temperature distribution in polycarbonate and aluminum electronic enclosures

    DEFF Research Database (Denmark)

    Shojaee Nasirabadi, Parizad; Jabbaribehnam, Mirmasoud; Hattel, Jesper Henri

    2016-01-01

    The challenge of developing a reliable electronic product requires huge amounts of resources and knowledge. Temperature and thermal features directly affect the life of electronic products. Furthermore, moisture can be damaging for electronic components. Nowadays, computational fluid dynamics (CF...

  5. Reduced-Temperature Transient-Liquid-Phase Bonding of AluminaUsing a Ag-Cu-Based Brazing Alloy

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Moo; Glaeser, Andreas M.

    2005-12-19

    The mechanical properties and microstructural evolution ofmetal-ceramic bonds produced using a transient liquid phase (TLP) aredescribed. Alumina (Al2O3) was joined at 500 degrees C, 600 degrees C,and 700 degrees C using a multilayer In/Cusil-ABA (R) (commercialcopper-silver eutectic brazing alloy)/In interlayer. The introduction ofthin In cladding layers allows the system to bond at much lowertemperatures than those typically used for brazing with Cusil-ABA (R),thereby protecting temperature-sensitive components. After chemicalhomogenization, the interlayers retain an operating temperature rangesimilar to that of the brazed joints. TLP bonds made at 500 degrees C,600 degrees C, and 700 degrees C with holding times ranging from as lowas 1.5 h to 24 h had average fracture strengths above 220 MPa. Theeffects of bonding temperature and time on fracture strength aredescribed. Preliminary analysis of the interlayers shows that the Ag-Inor Cu-In intermetallic phases do not form. Considerations unique tosystems with two-phase core layers are discussed. Experiments usingsingle-crystal sapphire indicate rapid formation of a reaction layer at700 degrees C, suggesting the possibility of making strong bonds usinglower temperatures and/or shorter processing times.

  6. WHTSubmersible: a simulator for estimating transient circulation temperature in offshore wells with the semi-submersible platform

    Science.gov (United States)

    Song, Xun-cheng; Liu, Yong-wang; Guan, Zhi-chuan

    2015-10-01

    Offshore wellbore temperature field is significant to drilling fluids program, equipment selection, evaluations on potential risks caused by casing thermal stress, etc. This paper mainly describes the theoretical basis, module structure and field verification of the simulator WHTSubmersible. This computer program is a useful tool for estimating transient temperature distribution of circulating drilling fluid on semi-submersible platform. WHTSubmersible is based on a mathematical model which is developed to consider radial and axial two-dimensional heat exchange of the inner drill pipe, the annulus, the drill pipe wall, the sea water and the formation in the process of drilling fluid circulation. The solution of the discrete equations is based on finite volume method with an implicit scheme. This scheme serves to demonstrate the numerical solution procedure. Besides, the simulator also considers the heating generated by drilling fluid circulation friction, drill bit penetrating rocks, friction between the drill column and the borehole wall, and the temperature effect on thermal physical properties and rheology of the drilling fluid. These measures ensure more accurate results. The simulator has been programmed as a dynamic link library using Visual C++, the routine interface is simple, which can be connected with other computer programs conveniently. The simulator is validated with an actual well temperature filed developed on a semi-submersible platform in South China, and the error is less than 5 %.

  7. Thermophysical Property Estimation by Transient Experiments: The Effect of a Biased Initial Temperature Distribution

    Directory of Open Access Journals (Sweden)

    Federico Scarpa

    2015-01-01

    Full Text Available The identification of thermophysical properties of materials in dynamic experiments can be conveniently performed by the inverse solution of the associated heat conduction problem (IHCP. The inverse technique demands the knowledge of the initial temperature distribution within the material. As only a limited number of temperature sensors (or no sensor at all are arranged inside the test specimen, the knowledge of the initial temperature distribution is affected by some uncertainty. This uncertainty, together with other possible sources of bias in the experimental procedure, will propagate in the estimation process and the accuracy of the reconstructed thermophysical property values could deteriorate. In this work the effect on the estimated thermophysical properties due to errors in the initial temperature distribution is investigated along with a practical method to quantify this effect. Furthermore, a technique for compensating this kind of bias is proposed. The method consists in including the initial temperature distribution among the unknown functions to be estimated. In this way the effect of the initial bias is removed and the accuracy of the identified thermophysical property values is highly improved.

  8. Correlating measured transient temperature rises with damage rate processes in cultured cells

    Science.gov (United States)

    Denton, Michael L.; Tijerina, Amanda J.; Gonzalez, Cherry C.; Gamboa, B. Giovana; Noojin, Gary D.; Ahmed, Elharith M.; Rickman, John M.; Dyer, Phillip H.; Rockwell, Benjamin A.

    2017-02-01

    Thermal damage rate processes in biological tissues are usually characterized by a kinetics approach. This stems from experimental data that show how the transformation of a specified biological property of cells or biomolecule (plating efficiency for viability, change in birefringence, tensile strength, etc.) is dependent upon both time and temperature. Here, two disparate approaches were used to study thermal damage rate processes in cultured retinal pigment epithelial cells. Laser exposure (photothermal) parameters included 2-μm laser exposure of non-pigmented cells and 532-nm exposures of cells possessing a variety of melanosome particle densities. Photothermal experiments used a mid-IR camera to record temperature histories with spatial resolution of about 8 μm, while fluorescence microscopy of the cell monolayers identified threshold damage at the boundary between live and dead cells. Photothermal exposure durations ranged from 0.05-20 s, and the effects of varying ambient temperature were investigated. Temperature during heat transfer using a water-jacketed cuvette was recorded with a fast microthermister, while damage and viability of the suspended cells were determined as percentages. Exposure durations for the heat transfer experiments ranged from 50- 60 s. Empirically-determined kinetic parameters for the two heating methods were compared with each other, and with values found in the literature.

  9. Fatigue Life of Stainless Steel in PWR Environments with Strain Holding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taesoon; Kim, Kyuhyung [KHNP CRI, Daejeon (Korea, Republic of); Seo, Myeonggyu; Jang, Changheui [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    Many components and structures of nuclear power plants are exposed to the water chemistry conditions during the operation. Recently, as design life of nuclear power plant is expanded over 60 years, the environmentally assisted fatigue (EAF) due to these water chemistry conditions has been considered as one of the important damage mechanisms of the safety class 1 components. Therefore, many studies to evaluate the effect of light water reactor (LWR) coolant environments on fatigue life of materials have been conducted. Many EAF test results including Argonne National Laboratory’s consistently indicated the substantial reduction of fatigue life in the light water reactor environments. However, there is a discrepancy between laboratory test data and plant operating experience regarding the effects of environment on fatigue: while laboratory test data suggest huge accumulation of fatigue damage, very limited experience of cracking caused by the low cycle fatigue in light water reactor. These hold-time effect tests are preformed to characterize the effects of strain holding on the fatigue life of austenitic stainless steels in PWR environments in comparison with the existing fixed strain rate results. Low cycle fatigue life tests were conducted for the type 316 stainless steel in 310℃ air and PWR environments with triangular strain. In agreement with the previous reports, the LCF life was reduced in PWR environments. Also for the slower strain rate, the reduction of LCF life was greater than the faster strain rate. The LCF test conditions for the hold-time effects were determined by the references and consideration of actual plant transient. To simulate the heat-up and cooldown transient, sub-peak strain holding during the down-hill of strain amplitude was chosen instead of peak strain holding which used in the previous researches.

  10. Fast transient infrared studies in material science: development of a novel low dead-volume, high temperature DRIFTS cell.

    Science.gov (United States)

    Dal Santo, V; Dossi, C; Fusi, A; Psaro, R; Mondelli, C; Recchia, S

    2005-04-30

    A prototype DRIFTS flow reaction chamber was designed and developed in order to find analytical application in the study of heterogeneous catalysts operating at high temperatures under fast transient gas feed conditions. Minimisation of dead-volumes allows gas replacement in 8-10s at 10mLmin(-1) total flow. To overcome problems related to the reactivity of the cell walls under alternating oxidizing/reducing gases, the cell was built with Inconel 600trade mark, which was tested to be very inert even at high temperatures. The sample holder, which was developed to closely resemble a micro plug-flow reactor, poses some problems in terms of heat transfer to the outer body of the cell (limiting then the maximum reachable temperature) and of the correct measurement of the actual sample temperature. These problems were solved with a careful re-design of the upper part of the cell. The second prototype thus derived is able to reach temperatures up to 803K and allows gas replacement in less than 4s at 10mLmin(-1). The cell is inserted in a MCT-FT-IR, which allows to collect high quality spectra with a 1s time-resolution. The downstream flow can be analysed by a quadrupole mass spectrometer equipped with an enclosed source and by a commercial GC. The performances of this prototype cell are presented showing some tests carried out with ceria-zirconia (Ce(x)Zr(1-x)O(2)) catalysts for CO abatement under real operando conditions.

  11. Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis

    Energy Technology Data Exchange (ETDEWEB)

    Fridman, E., E-mail: e.fridman@fzd.d [Institute of Safety Research, Forschungszentrum Dresden-Rossendorf, POB 51 01 19, Dresden 01314 (Germany); Kliem, S. [Institute of Safety Research, Forschungszentrum Dresden-Rossendorf, POB 51 01 19, Dresden 01314 (Germany)

    2011-01-15

    Research highlights: Detailed 3D 100% Th-MOX PWR core design is developed. Pu incineration increased by a factor of 2 as compared to a full MOX PWR core. The core controllability under steady state conditions is demonstrated. - Abstract: Current practice of Pu recycling in existing Light Water Reactors (LWRs) in the form of U-Pu mixed oxide fuel (MOX) is not efficient due to continuous Pu production from U-238. The use of Th-Pu mixed oxide (TOX) fuel will considerably improve Pu consumption rates because virtually no new Pu is generated from thorium. In this study, the feasibility of Pu recycling in a typical pressurized water reactor (PWR) fully loaded with TOX fuel is investigated. Detailed 3-dimensional 100% TOX and 100% MOX PWR core designs are developed. The full MOX core is considered for comparison purposes. The design stages included determination of Pu loading required to achieve 18-month fuel cycle assuming three-batch fuel management scheme, selection of poison materials, development of the core loading pattern, optimization of burnable poison loadings, evaluation of critical boron concentration requirements, estimation of reactivity coefficients, core kinetic parameters, and shutdown margin. The performance of the MOX and TOX cores under steady-state condition and during selected reactivity initiated accidents (RIAs) is compared with that of the actual uranium oxide (UOX) PWR core. Part I of this paper describes the full TOX and MOX PWR core designs and reports the results of steady state analysis. The TOX core requires a slightly higher initial Pu loading than the MOX core to achieve the target fuel cycle length. However, the TOX core exhibits superior Pu incineration capabilities. The significantly degraded worth of control materials in Pu cores is partially addressed by the use of enriched soluble boron and B{sub 4}C as a control rod absorbing material. Wet annular burnable absorber (WABA) rods are used to flatten radial power distribution. The

  12. Investigation of Natural Circulation Instability and Transients in Passively Safe Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, Mamoru [Purdue Univ., West Lafayette, IN (United State

    2016-11-30

    The NEUP funded project, NEUP-3496, aims to experimentally investigate two-phase natural circulation flow instability that could occur in Small Modular Reactors (SMRs), especially for natural circulation SMRs. The objective has been achieved by systematically performing tests to study the general natural circulation instability characteristics and the natural circulation behavior under start-up or design basis accident conditions. Experimental data sets highlighting the effect of void reactivity feedback as well as the effect of power ramp-up rate and system pressure have been used to develop a comprehensive stability map. The safety analysis code, RELAP5, has been used to evaluate experimental results and models. Improvements to the constitutive relations for flashing have been made in order to develop a reliable analysis tool. This research has been focusing on two generic SMR designs, i.e. a small modular Simplified Boiling Water Reactor (SBWR) like design and a small integral Pressurized Water Reactor (PWR) like design. A BWR-type natural circulation test facility was firstly built based on the three-level scaling analysis of the Purdue Novel Modular Reactor (NMR) with an electric output of 50 MWe, namely NMR-50, which represents a BWR-type SMR with a significantly reduced reactor pressure vessel (RPV) height. The experimental facility was installed with various equipment to measure thermalhydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests were performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The control system and data acquisition system were programmed with LabVIEW to realize the realtime control and data storage. The thermal-hydraulic and nuclear coupled startup transients were performed to investigate the flow instabilities at low pressure and low power conditions for NMR-50. Two different power ramps were chosen to study the effect of startup

  13. Transient analysis of nuclear graphite oxidation for high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Wei, E-mail: wxu12@mails.tsinghua.edu.cn; Shi, Lei; Zheng, Yanhua

    2016-09-15

    Graphite is widely used as moderator, reflector and structural materials in the high temperature gas-cooled reactor pebble-bed modular (HTR-PM). In normal operating conditions or water/air ingress accident, the nuclear graphite in the reactor may be oxidized by air or steam. Oxidation behavior of nuclear graphite IG-110 which is used as the structural materials and reflector of HTR-PM is mainly researched in this paper. To investigate the penetration depth of oxygen in IG-110, this paper developed the one dimensional spherical oxidation model. In the oxidation model, the equations considered graphite porosity variation with the graphite weight loss. The effect of weight loss on the effective diffusion coefficient and the oxidation rate was also considered in this model. Based on this theoretical model, this paper obtained the relative concentration and local weight loss ratio profile in graphite. In addition, the local effective diffusion coefficient and oxidation rate in the graphite were also investigated.

  14. Conformational relaxation dynamics of a poly(N-isopropylacrylamide) aqueous solution measured using the laser temperature jump transient grating method.

    Science.gov (United States)

    Inoue, Hayato; Katayama, Kenji; Iwai, Kaoru; Miura, Atsushi; Masuhara, Hiroshi

    2012-04-28

    We observed phase transition and phase relaxation processes of a poly(N-isopropylacrylamide) (PNIPAM) aqueous solution using the heterodyne transient grating (HD-TG) method combined with the laser temperature jump technique. The sample temperature was instantaneously raised by about 1.0 K after irradiation of a pump pulse to crystal violet (CV) molecules for heating, and the phase transition was induced for the sample with an initial temperature just below the lower critical solution temperature (LCST); the following phase relaxation dynamics was observed. Turbidity relaxation was observed in both the turbidity and HD-TG responses, while another relaxation process was observed only in the HD-TG response, namely via the refractive index change. It is suggested that this response is due to formation of globule molecules or their assemblies since they would have nothing to do with turbidity change but would affect the refractive index, which is dependent on the molar volume of a chemical species. Furthermore, the grating spacing dependence of the HD-TG responses suggests that the response was caused by the counter propagating diffusion of the coil molecules as a reactant species and the globule molecules as a product species and the lifetime of the globule molecules ranged from 1.5 to 5 seconds. Thus, we conclude that the turbidity reflects the dynamics of aggregate conditions, not molecular conditions. The coil and globule sizes were estimated from the obtained diffusion coefficient. The sizes of the coil molecules did not change at the initial temperatures below the LCST but increased sharply as it approaches LCST. We propose that the coil-state molecules associate due to hydrophobic interaction when the initial temperature was higher than LCST minus 0.5 K and that the globule-state molecules generated from the coil-state molecules showed a similar trend in temperature. The phase transition was also induced by heating under a microscope, and the relaxation process

  15. TEMLOPI/V.2: a computer program for estimation of fully transient temperatures in geothermal wells during circulation and shut-in

    Science.gov (United States)

    Espinosa-Paredes, G.; Garcia, A.; Santoyo, E.; Hernandez, I.

    2001-04-01

    This paper describes the development, validation and application of the TEMLOPI/V.2 computer program. This program is a useful tool for estimating in-situ the transient temperature distribution of the fluids employed for drilling geothermal wells. TEMLOPI/V.2 is based on a mathematical model which is developed to consider two-dimensional transient heat transfer during drilling and shut-in conditions in and around a geothermal well. The solution of the partial differential equations is based on the finite-difference technique with an implicit scheme. This scheme serves to demonstrate the numerical solution procedure. Each radial grid node is placed in a different thermal region: flow inside the pipe, metal pipe wall, flow inside annulus, and the surrounding formation. The program was written in FORTRAN 77 using modular programming and runs on most IBM compatible personal computers. The software code, its architecture, input and output files, the solution algorithm, flow diagrams and source programs are described in detail. From validation tests, computed temperatures differ by less than 5°C from analytically obtained temperatures. Comparison of results from the fully transient TEMLOPI/V.2 simulator and the pseudo-transient version, TEMLOPI/V.1, with measured data shows that the fully transient model provides better results. Application of TEMLOPI/V.2 is demonstrated in a practical application study of well EAZ-2 from Los Azufres Mexican geothermal field.

  16. Study of safety relief valve operation under ATWS conditions. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hutmacher, E.S.; Nesmith, B.J.; Brukiewa, J.B.

    1979-06-25

    A literature survey and analysis project has been performed to determine if recent (since mid-1975) data has been reported which could influence the current approach to predicting PWR relief valve capacity under ATWS conditions. This study was conducted by the Energy Technology Engineering Center for NRC. Results indicate that the current relief valve capacity model tends to predict less capacity than actually obtains; however, no experimental verification at PWR ATWS conditions was found. Other project objectives were to establish the availability of methods for evaluating reaction forces and back pressure effects on relief valve capacity, and to determine if facilities exist which are capable of testing PWR relief valves at ATWS conditions.

  17. A novel alpha-amylase gene is transiently upregulated during low temperature exposure in apple fruit.

    Science.gov (United States)

    Wegrzyn, T; Reilly, K; Cipriani, G; Murphy, P; Newcomb, R; Gardner, R; MacRae, E

    2000-03-01

    An alpha-amylase gene product was isolated from apple fruit by reverse-transcriptase PCR using redundant primers, followed by 5' and 3' RACE. The gene is a member of a small gene family. It encodes a putative 46.9 kDa protein that is most similar to an alpha-amylase gene from potato (GenBank accession M79328). In apple fruit this new gene was expressed at low levels, as detected by reverse-transcriptase PCR, in a number of plant tissues and during fruit development. Highest levels of mRNA for this transcript were observed 3 to 9 days after placing apple fruit at 0.5 degrees C. Phylogenetic analysis of amino acid sequence places the potato and apple proteins as a distinct and separate new subgroup within the plant alpha-amylases, which appears to have diverged prior to the split between monocotyledonous and dicotyledonous plants. These two divergent alpha-amylases lack the standard signal peptide structures found in all other plant alpha-amylases, and have sequence differences within the B-domain and C-domain. However, comparisons with structures of known starch hydrolases suggest that these differences are unlikely to affect the enzymatic alpha-1,4-amylase function of the protein. This is the first report of upregulation of a dicotyledonous alpha-amylase in response to low temperature, and confirms the presence of a new family of alpha-amylases in plants.

  18. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C., E-mail: sabrinapral@gmail.com, E-mail: amir@cdtn.brm, E-mail: hcr@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

  19. Features of chlorophyll fluorescence transients can be used to investigate low temperature induced effects on photosystem II of algal lichens from polar regions

    OpenAIRE

    Mishra Anamika; Hájek Josef; Tuháčková Tereza; Barták Miloš; Mishra Kumud Bandhu

    2015-01-01

    Chlorophyll fluorescence is an effective tool for investigating characteristics of any photosynthesizing organisms and its responses due to different stressors. Here, we have studied a short-term temperature response on three Antarctic green algal lichen species: Umbilicaria antarctica, Xanthoria elegans, and Rhizoplaca melanophtalma. We measured slow chlorophyll fluorescence transients in these Antarctic lichen species during slowely cooling of thallus temperature from 20°C to 5, 0 and -5°C ...

  20. Characterization of Factors affecting IASCC of PWR Core Internals

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Woo; Hwang, Seong Sik; Kim, Won Sam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-09-15

    A lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate IASCC in PWR, but the mechanism in PWR is not fully understood yet as compared with that in BWR due to a lack of data from laboratories and fields. Therefore it is strongly needed to review and analyse recent researches of IASCC in both BWR and PWR for establishing a proactive management technology for IASCC of core internals in Korean PWRs. This work is aimed to review mainly recent technical reports on IASCC of stainless steels for core internals in PWR. For comparison, the works on IASCC in BWR were also reviewed and briefly introduced in this report.

  1. Evaluation of tight-pitch PWR cores

    Energy Technology Data Exchange (ETDEWEB)

    Correa, F.; Driscoll, M.J.; Lanning, D.D.

    1979-08-01

    The impact of tight pinch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system - /sup 235/U/UO/sub 2/ : Pu/ThO/sub 2/ : /sup 233/U/ThO/sub 2/ - and the conventional recycle-mode uranium system - /sup 235/U/UO/sub 2/ : Pu/UO/sub 2/. The basic parameter varied was the fuel-to-moderator volume ratio (F/M) of the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices) < F/M < 4.0 are limited by the scarcity of experiments with F/M > 1.0, the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments.

  2. A pressure drop model for PWR grids

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dong Seok; In, Wang Ki; Bang, Je Geon; Jung, Youn Ho; Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development. 7 refs., 3 figs., 3 tabs. (Author)

  3. CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

    Energy Technology Data Exchange (ETDEWEB)

    Pena, C.; Pellacani, F.; Macian Juan, R., E-mail: carlos.pena@ntech.mw.tum.de, E-mail: pellacani@ntech.mw.tum.de, E-mail: macian@ntech.mw.tum.de [Technische Universitaet Muenchen, Garching (Germany). Ntech Lehrstuhl fuer Nukleartechnik; Chiva, S., E-mail: schiva@emc.uji.es [Universitat Jaume I, Castellon de la Plana (Spain). Dept. de Ingenieria Mecanica y Construccion; Barrachina, T.; Miro, R., E-mail: rmiro@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV) (Spain). Institute for Industrial, Radiophysical and Environmental Safety

    2011-07-01

    A computational code system based on coupling the 3D neutron diffusion code PARCS v2.7 and the Ansys CFX 13.0 Computational Fluid Dynamics (CFD) code has been developed as a tool for nuclear reactor systems simulations. This paper presents the coupling methodology between the CFD and the neutronic code. The methodology to simulate a 3D-neutronic problem coupled with 1D thermal hydraulics is already a mature technology, being part of the regular calculations performed to analyze different kinds of Reactivity Insertion Accidents (RIA) and asymmetric transients in Nuclear Power Plants, with state-of-the-art coupled codes like TRAC-B/NEM, RELAP5/PARCS, TRACE/PARCS, RELAP3D, RETRAN3D, etc. This work represents one of the first attempts to couple the multiphysics of a nuclear reactor core with a 3D spatial resolution in a computer code. This will open new possibilities regarding the analysis of fuel elements, contributing to a better understanding and design of the heat transfer process and specific fluid dynamics phenomena such as cross flow among fuel elements. The transient simulation of control rod insertion, boron dilution and cold water injection will be made possible with a degree of accuracy not achievable with current methodologies based on the use of system and/or subchannel codes. The transport of neutrons depends on several parameters, like fuel temperature, moderator temperature and density, boron concentration and fuel rod insertion. These data are calculated by the CFD code with high local resolution and used as input to the neutronic code to calculate a 3D nodal power distribution that will be returned and remapped to the CFD code control volumes (cells). Since two different nodalizations are used to discretized the same system, an averaging and interpolating procedure is needed to realize an effective data exchange. These procedures have been developed by means of the Ansys CFX 'User Fortran' interface; a library with several subroutines has

  4. Transient Simulation of the Multi-SERTTA Experiment with MAMMOTH

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, Sebastian [Idaho National Lab. (INL), Idaho Falls, ID (United States); deHart, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-11

    This work details the MAMMOTH reactor physics simulations of the Static Environment Rodlet Transient Test Apparatus (SERTTA) conducted at Idaho National Laboratory in FY-2017. TREAT static-environment experiment vehicles are being developed to enable transient testing of Pressurized Water Reactor (PWR) type fuel specimens, including fuel concepts with enhanced accident tolerance (Accident Tolerant Fuels, ATF). The MAMMOTH simulations include point reactor kinetics as well as spatial dynamics for a temperature-limited transient. The strongly coupled multi-physics solutions of the neutron flux and temperature fields are second order accurate both in the spatial and temporal domains. MAMMOTH produces pellet stack powers that are within 1.5% of the Monte Carlo reference solutions. Some discrepancies between the MCNP model used in the design of the flux collars and the Serpent/MAMMOTH models lead to higher power and energy deposition values in Multi-SERTTA unit 1. The TREAT core results compare well with the safety case computed with point reactor kinetics in RELAP5-3D. The reactor period is 44 msec, which corresponds to a reactivity insertion of 2.685% delta k/k$. The peak core power in the spatial dynamics simulation is 431 MW, which the point kinetics model over-predicts by 12%. The pulse width at half the maximum power is 0.177 sec. Subtle transient effects are apparent at the beginning insertion in the experimental samples due to the control rod removal. Additional difference due to transient effects are observed in the sample powers and enthalpy. The time dependence of the power coupling factor (PCF) is calculated for the various fuel stacks of the Multi-SERTTA vehicle. Sample temperatures in excess of 3100 K, the melting point UO$_2$, are computed with the adiabatic heat transfer model. The planned shaped-transient might introduce additional effects that cannot be predicted with PRK models. Future modeling will be focused on the shaped-transient by improving the

  5. Optimal design of passive containment cooling system for innovative PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Huiun; Lee, Sang Won; Kim, Hangon [Central Research Institute, Korea Hydro and Nuclear Power, Ltd., Daejeon (Korea, Republic of)

    2017-08-15

    Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  6. Optimal design of passive containment cooling system for innovative PWR

    Directory of Open Access Journals (Sweden)

    Huiun Ha

    2017-08-01

    Full Text Available Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS of an innovative pressurized water reactor (PWR. A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT geometry, PCCS heat exchanger (PCCX location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  7. Evaluation of steam generator U-tube integrity during PWR station blackout with secondary system depressurization

    Energy Technology Data Exchange (ETDEWEB)

    Hidaka, Akihide; Asaka, Hideaki; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ueno, Shingo; Yoshino, Takehito

    1999-12-01

    In PWR severe accidents such as station blackout, the integrity of steam generator U-tube would be threatened early at the transient among the pipes of primary system. This is due to the hot leg countercurrent natural circulation (CCNC) flow which delivers the decay heat of the core to the structures of primary system if the core temperature increases after the secondary system depressurization. From a view point of accident mitigation, this steam generator tube rupture (SGTR) is not preferable because it results in the direct release of primary coolant including fission products (FP) to the environment. Recent SCDAP/RELAP5 analyses by USNRC showed that the creep failure of pressurizer surge line which results in release of the coolant into containment would occur earlier than SGTR during the secondary system depressurization. However, the analyses did not consider the decay heat from deposited FP on the steam generator U-tube surface. In order to investigate the effect of decay heat on the steam generator U-tube integrity, the hot leg CCNC flow model used in the USNRC's calculation was, at first, validated through the analysis for JAERI's LSTF experiment. The CCNC model reproduced well the thermohydraulics observed in the LSTF experiment and thus the model is mostly reliable. An analytical study was then performed with SCDAP/RELAP5 for TMLB' sequence of Surry plant with and without secondary system depressurization. The decay heat from deposited FP was calculated by JAERI's FP aerosol behavior analysis code, ART. The ART analysis showed that relatively large amount of FPs may deposit on steam generator U-tube inlet mainly by thermophoresis. The SCDAP/RELAP5 analyses considering the FP decay heat predicted small safety margin for steam generator U-tube integrity during secondary system depressurization. Considering associated uncertainties in the analyses, the potential for SGTR cannot be ignored. Accordingly, this should be considered in the

  8. OPTIMASI PENDINGINAN EKSTERNAL PADA MODEL SUNGKUP PWR-1000 MENGGUNAKAN METODE ESTIMASI ANALITIK

    Directory of Open Access Journals (Sweden)

    Hendro Tjahjono

    2015-03-01

    Full Text Available Sungkup reaktor merupakan benteng terakhir dalam menahan terlepasnya zat-zat radioaktif ke lingkungan ketika terjadi suatu kecelakaan reaktor. Oleh karena itu integritasnya harus selalu dipertahankan yang antara lain dilakukan dengan cara mencegah dilampauinya batas desain tekanan dan temperatur yang bisa terjadi pada kondisi kecelakaan melalui pendinginan sungkup yang mencukupi. Pada reaktor generasi III+ yang menerapkan konsep pendinginan pasif seperti AP1000, sungkup didinginkan secara eksternal melalui konveksi alamiah pada celah udara dan guyuran air pendingin di permukaan luar sungkup. Karakteristik pendinginan eksternal ini akan diteliti secara eksperimental melalui model sungkup PWR1000 berskala 1/40. Tujuan dari penelitian ini adalah untuk mengetahui nilai debit optimal yang diperlukan dalam pendinginan model sungkup sebelum konfirmasi secara eksperimental dilakukan. Metode yang digunakan adalah dengan melakukan pemodelan analitis dan pemrograman berbasis Matlab yang mampu mengestimasi nilai-nilai parameter pendinginan eksternal seperti laju alir, temperatur dan daya kalor yang dievakuasi. Penerapan program pada sungkup AP1000 juga dilakukan untuk bisa dibandingkan dengan data desain. Hasilnya menunjukkan kesesuaian dengan data desain sungkup AP1000 dengan debit optimal sebesar 9,5 liter/detik yang mampu mengevakuasi kalor sebesar 21,6 MW. Sedangkan pada model sungkup diperoleh debit optimal sebesar 22 cc/detik yang mampu mengevakuasi kalor sebesar 37 KW. Disimpulkan bahwa dengan penelitian ini karakteristik pendinginan eksternal sungkup reaktor PWR mampu diestimasi dan bersamaan dengan itu dapat diketahui nilai optimal dari debit pendingin yang diperlukan. Kata kunci: pendinginan eksternal, sungkup PWR, estimasi analitik, AP1000.   Reactor containment is the last barrier in avoiding the release of radioactive substances into the environment in the event of a reactor accident. Therefore, its integrity must always be maintained, among

  9. Conceptual study on advanced PWR system

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Young; Chang, M. H.; Yu, K. J.; Lee, D. J.; Cho, B. H.; Kim, H. Y.; Yoon, J. H.; Lee, Y. J.; Kim, J. P.; Park, C. T.; Seo, J. K.; Kang, H. S.; Kim, J. I.; Kim, Y. W.; Kim, Y. H.

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode concepts : analysis and experiment were conducted. (8) Wet thermal insulator : tests for thin steel layers and assessment of materials. (9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs.

  10. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  11. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

  12. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  13. Seismic qualification of PWR plant auxiliary feedwater systems

    Energy Technology Data Exchange (ETDEWEB)

    Lu, S.C.; Tsai, N.C.

    1983-08-01

    The NRC Standard Review Plan specifies that the auxiliary feedwater (AFW) system of a pressurized water reactor (PWR) is a safeguard system that functions in the event of a Safe Shutdown Earthquake (SSE) to remove the decay heat via the steam generator. Only recently licensed PWR plants have an AFW system designed to the current Standard Review Plan specifications. The NRC devised the Multiplant Action Plan C-14 in order to make a survey of the seismic capability of the AFW systems of operating PWR plants. The purpose of this survey is to enable the NRC to make decisions regarding the need of requiring the licensees to upgrade the AFW systems to an SSE level of seismic capability. To implement the first phase of the C-14 plan, the NRC issued a Generic Letter (GL) 81-14 to all operating PWR licensees requesting information on the seismic capability of their AFW systems. This report summarizes Lawrence Livermore National Laboratory's efforts to assist the NRC in evaluating the status of seismic qualification of the AFW systems in 40 PWR plants, by reviewing the licensees' responses to GL 81-14.

  14. Impact of boron dilution accidents on low boron PWR safety

    Energy Technology Data Exchange (ETDEWEB)

    Papukchiev, A.; Liu, Y. [Dept. of Reactor Dynamics and Reactor Safety, Technical Univ. Munich, Walther Meissner-Str. 2, 85748 Garching (Germany); Schaefer, A. [ISaR Inst. for Safety and Reliability, Walther Meissner-Str. 2, 85748 Garching (Germany)

    2006-07-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As an inadvertent reduction of the boron concentration during a boron dilution accident could introduce positive reactivity and have a negative impact on PWR safety, design changes to reduce boron concentration in the reactor coolant are of general interest. In the framework of an investigation into the feasibility of low boron design, a PWR core configuration based on fuel with higher gadolinium (Gd) load has been developed which permits to reduce the natural boron concentration at begin of cycle (BOC) to 518 ppm. For the assessment of the potential safety advantages, a boron dilution accident due to small break loss-of-coolant-accident (SBLOCA) has been simulated with the system code ATHLET for two PWR core designs: a low boron design and a standard core design. The results from the comparative analyses showed that the impact of the boron dilution accident on the new PWR design safety is significantly lower in comparison with the standard design. The new reactor design provided at least 4, 4% higher reactivity margin to recriticality during the whole accident which is equivalent to the negative reactivity worth of additional 63% of all control rods fully inserted in to the core. (authors)

  15. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1989-01-01

    Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

  16. Thermal hydraulic investigations and optimization on the EVC system of a PWR by CFD simulation

    Energy Technology Data Exchange (ETDEWEB)

    Xi, Mengmeng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Tang, Mao [China Nuclear Power Design Engineering Co., Ltd., 518124 Shenzhen (China); Wang, Chenglong; Zheng, Meiyin; Qiu, Suizheng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China)

    2015-08-15

    Highlights: • This study constructs a full CFD model for the EVC system of a PWR. • The complex fluid and solid coupling is treated in the computation. • Primary characteristics of the velocity, pressure and temperature distributions in the EVC system are investigated. • The optimization of the EVC system with different inlet boundaries are performed. - Abstract: In order to optimize the design of Reactor Pit Ventilation (EVC) system in a Pressurized Water Reactor (PWR), it is necessary to study the characteristics of the velocity, pressure and temperature fields in the EVC system. A full computational fluid dynamics (CFD) model for the EVC system is constructed by a commercial CFD code, where the complex fluid and solid coupling is treated. The Shear Stress Transport (SST) model is adopted to perform the turbulence calculation. This paper numerically investigates the characteristics of the velocity, pressure and temperature distributions in the EVC system. In particular, the effects of inlet air parameters on the thermal hydraulic characteristics and the reactor pit structure are also discussed for the EVC system optimization. Simulations are carried out with different mesh sizes and boundary conditions for sensitivity analysis. The computational results are important references to optimize the design and verify the rationality of the EVC system.

  17. Experimental Analysis and Modeling of the Electrical and Thermal Transients of the Diode-By-pass for the LHC- Magnet Protection at Cryogenic Temperatures

    CERN Document Server

    Denz, R

    1998-01-01

    For the protection of the LHC superconducting lattice magnets cold bypass diodes will be installed inside the magnet cryostat, subjecting them to superfluid helium temperatures and radiation. During a magnet quench, the power generated in the diode must be dissipated in the adjacent heat sinks of copper that are part of the diode package. Results from endurance tests on the diode package are presented. A simple thermo-electric model has been developed to simulate the thermal and electrical transients in the diode package during the endurance pulse. Simulation results are in good agreement with the measured temperatures.

  18. One-dimensional modeling of radial heat removal during depressurized heatup transients in modular pebble-bed and prismatic high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Savage, M.G.

    1984-07-01

    A one-dimensional computational model was developed to evaluate the heat removal capabilities of both prismatic-core and pebble-bed modular HTGRs during depressurized heatup transients. A correlation was incorporated to calculate the temperature- and neutron-fluence-dependent thermal conductivity of graphite. The modified Zehner-Schluender model was used to determine the effective thermal conductivity of a pebble bed, accounting for both conduction and radiation. Studies were performed for prismatic-core and pebble-bed modular HTGRs, and the results were compared to analyses performed by GA and GR, respectively. For the particular modular reactor design studied, the prismatic HTGR peak temperature was 2152.2/sup 0/C at 38 hours following the transient initiation, and the pebble-bed peak temperature was 1647.8/sup 0/C at 26 hours. These results compared favorably with those of GA and GE, with only slight differences caused by neglecting axial heat transfer in a one-dimensional radial model. This study found that the magnitude of the initial power density had a greater effect on the temperature excursion than did the initial temperature.

  19. The advanced main control console for next japanese PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Tsuchiya, A. [Hokkaido Electric Power Co., Inc., Sapporo (Japan); Ito, K. [Mitsubishi Heavy Industries, Ltd., Nuclear Energy Systems Engineering Center, Yokohama (Japan); Yokoyama, M. [Mitsubishi Electric Corporation, Energy and Industrial Systems Center, Kobe (Japan)

    2001-07-01

    The purpose of the improvement of main control room designing in a nuclear power plant is to reduce operators' workload and potential human errors by offering a better working environment where operators can maximize their abilities. In order to satisfy such requirements, the design of main control board applied to Japanese Pressurized Water Reactor (PWR) type nuclear power plant has been continuously modified and improved. the Japanese Pressurized Water Reactor (PWR) Utilities (Electric Power Companies) and Mitsubishi Group have developed an advanced main control board (console) reflecting on the study of human factors, as well as using a state of the art electronics technology. In this report, we would like to introduce the configuration and features of the Advanced Main Control Console for the practical application to the next generation PWR type nuclear power plants including TOMARI No.3 Unit of Hokkaido Electric Power Co., Inc. (author)

  20. Steam generator tube integrity program. Phase I report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Alzheimer, J.M.; Clark, R.A.; Morris, C.J.; Vagins, M.

    1979-09-01

    The results are presented of the pressure tests performed as part of Phase I of the Steam Generator Tube Integrity (SGTI) program at Battelle Pacific Northwest Laboratory. These tests were performed to establish margin-to-failure predictions for mechanically defected Pressurized Water Reactor (PWR) steam generator tubing under operating and accident conditions. Defect geometries tested were selected because they simulate known or expected defects in PWR steam generators. These defect geometries are Electric Discharge Machining (EDM) slots, elliptical wastage, elliptical wastage plus through-wall slot, uniform thinning, denting, denting plus uniform thinning, and denting plus elliptical wastage. All defects were placed in tubing representative of that currently used in PWR steam generators.

  1. Ex-Stream: A MATLAB program for calculating fluid flux through sediment-water interfaces based on steady and transient temperature profiles

    Science.gov (United States)

    Swanson, Travis E.; Cardenas, M. Bayani

    2011-10-01

    Temperature is a useful environmental tracer for quantifying movement and exchange of water and heat through and near sediment-water interfaces (SWI). Heat tracing involves analyzing temperature time series or profiles from temperature probes deployed in sediments. Ex-Stream is a MATLAB program that brings together two transient and two steady one-dimensional coupled heat and fluid flux analytical models. The program includes a graphical user interface, a detailed user manual, and postprocessing capabilities that enable users to extract fluid fluxes from time-series temperature observations. Program output is written to comma-separated values files, displayed within the MATLAB command window, and may be optionally plotted. The models that are integrated into Ex-Stream can be run collectively, allowing for direct comparison, or individually.

  2. Leak before break application in French PWR plants under operation

    Energy Technology Data Exchange (ETDEWEB)

    Faidy, C. [EDF SEPTEN, Villeurbanne (France)

    1997-04-01

    Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

  3. Structures and Materials of Reactor Internals for PWR in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. W.; Kim, W. S.; Kwon, S. C.; Kwon, J. H.; Kim, Y. S.; Kim, H. P.; Yoo, C. S.; Lee, S. R.; Jung, M. K.; Hwang, S. S

    2007-10-15

    Nuclear reactor types in Korea are PWR type reactor (Westinghouse, Combustion Engineering, Farmatome type) and CANDU type reactor. Structures and Materials for reactor internal of PWR type were investigated. Reactor internal was composed of lower core support structure, upper core support assembly, incore instrumentation support structure. Lower core support structure of these structures is the most important. The major material for the reactor internal is type 304 and 316 stainless steel and radial support clevis bolts are made of Inconel. The main damage mechanism for reactor internal was IASCC and the effect of IASCC on reactor internal was investigated. The accident for reactor internal was also investigate.

  4. Advanced ion exchange resins for PWR condensate polishing

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, B. [Rohm and Haas Co. (United States); Tsuzuki, S. [Rohm and Haas Co. (Japan)

    2002-07-01

    The severe chemical and mechanical requirements of a pressurized water reactor (PWR) condensate polishing plant (CPP) present a major challenge to the design of ion exchange resins. This paper describes the development and initial operating experience of improved cation and anion exchange resins that were specifically designed to meet PWR CPP needs. Although this paper focuses specifically on the ion exchange resins and their role in plant performance, it is also recognized and acknowledged that excellent mechanical design and operation of the CPP system are equally essential to obtaining good results. (authors)

  5. Multi-domain simulation of transient junction temperatures and resulting stress-strain behavior of power switches for long-term mission profiles

    Energy Technology Data Exchange (ETDEWEB)

    Drofenik, U.; Kovacevic, I.; Kolar, J. W. [Swiss Federal Institute of Technology, Power Electronic Systems Laboratory, Zuerich (Switzerland); Schmidt, R. [ABB Switzerland Ltd., Corporate Research, Baden-Daettwil (Switzerland)

    2008-07-01

    For lifetime estimation of power converters in traction applications, one method is to calculate numerically the stress-strain hysteresis curves of the interfaces silicon-solder-DCB and/or DCB-solder-baseplate inside the power modules. This can only be achieved if the transient junction temperatures in these layers are known for a defined mission profile. Therefore, one has to couple circuit simulation with thermal simulation and stress-strain computation. The second challenge of this problem is to perform this transient simulation taking into account switching losses in the {mu}s-range for mission profiles over a couple of minutes. In this paper we employ a new multi-domain simulation software to achieve results with reasonable computational effort. (author)

  6. Safety and licensing issues that are being addressed by the Power Burst Facility test programs. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    McCardell, R.K.; MacDonald, P.E.

    1980-01-01

    This paper presents an overview of the results of the experimental program being conducted in the Power Burst Facility and the relationship of these results to certain safety and licensing issues. The safety issues that were addressed by the Power-Cooling-Mismatch, Reactivity Initiated Accident, and Loss of Coolant Accident tests, which comprised the original test program in the Power Burst Facility, are discussed. The resolution of these safety issues based on the results of the thirty-six tests performed to date, is presented. The future resolution of safety issues identified in the new Power Burst Facility test program which consists of tests which simulate BWR and PWR operational transients, anticipated transients without scram, and severe fuel damage accidents, is described.

  7. Training simulater for core physics test of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hoizumi, Atsushi

    1989-03-01

    The core physics tests are done at the start-up test and the annual inspection of PWR in order to confirm the safety and the design. In these tests, some special equipement/operations are required. Therefore, in order to maintain these special techniques, we developed the training simulator for core physical tests using a digital computer.

  8. Computational fluid dynamics (CFD) round robin benchmark for a pressurized water reactor (PWR) rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Shin K., E-mail: paengki1@tamu.edu; Hassan, Yassin A.

    2016-05-15

    Highlights: • The capabilities of steady RANS models were directly assessed for full axial scale experiment. • The importance of mesh and conjugate heat transfer was reaffirmed. • The rod inner-surface temperature was directly compared. • The steady RANS calculations showed a limitation in the prediction of circumferential distribution of the rod surface temperature. - Abstract: This study examined the capabilities and limitations of steady Reynolds-Averaged Navier–Stokes (RANS) approach for pressurized water reactor (PWR) rod bundle problems, based on the round robin benchmark of computational fluid dynamics (CFD) codes against the NESTOR experiment for a 5 × 5 rod bundle with typical split-type mixing vane grids (MVGs). The round robin exercise against the high-fidelity, broad-range (covering multi-spans and entire lateral domain) NESTOR experimental data for both the flow field and the rod temperatures enabled us to obtain important insights into CFD prediction and validation for the split-type MVG PWR rod bundle problem. It was found that the steady RANS turbulence models with wall function could reasonably predict two key variables for a rod bundle problem – grid span pressure loss and the rod surface temperature – once mesh (type, resolution, and configuration) was suitable and conjugate heat transfer was properly considered. However, they over-predicted the magnitude of the circumferential variation of the rod surface temperature and could not capture its peak azimuthal locations for a central rod in the wake of the MVG. These discrepancies in the rod surface temperature were probably because the steady RANS approach could not capture unsteady, large-scale cross-flow fluctuations and qualitative cross-flow pattern change due to the laterally confined test section. Based on this benchmarking study, lessons and recommendations about experimental methods as well as CFD methods were also provided for the future research.

  9. Methodology for the LABIHS PWR simulator modernization

    Energy Technology Data Exchange (ETDEWEB)

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.b, E-mail: mvitor@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  10. Amorphous and Nanocrystalline High Temperature Magnetic Material for PWR

    Science.gov (United States)

    2006-03-01

    in collaboration with Magnetics, Inc. has produced nanopowders of the HITPERM materials. The work was extended to include study of...the interfacial stresses between the substrate and coating that arises during the coating processes. Alumina , Beryllia, Forsterite and Pt were...trial was performed to evaluate the efficacy of plasma synthesized ferrite coatings. NiZn ferrites were sprayed onto Alumina substrates using the

  11. Corrosion fatigue initiation behaviour of wrought austenitic stainless pipe steels under simulated BWR/HWC and PWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Leber, H.J.; Ritter, S.; Seifert, H.P [Paul Scherrer Institute, Nuclear Energy and Safety Research Department, Laboratory for Nuclear Materials, 5232 Villigen PSI (Switzerland)

    2011-07-01

    The corrosion fatigue (CF) initiation and short crack growth behavior of different low-carbon and stabilized austenitic stainless steels was characterized under simulated BWR and primary PWR conditions by cyclic fatigue tests with sharply notched fracture mechanics specimens in the temperature range from 70 to 320 C. Environmental reduction of fatigue initiation life was observed in all stainless steels at strain rates {<=} 0.1 %/s in BWR and PWR environment. The stationary short crack CF crack growth rates after crack advances of 50 to 300 {mu}m from the notch-root were in the typical range of corresponding results from tests with long cracks (pre-cracked specimens) and also showed the same system parameter response. The effect of environment on the initiation process ({Delta}a = 10 {mu}m) was relevantly stronger than on the subsequent stationary short crack growth. Both, under BWR/HWC and PWR conditions, a relevant environmental reduction of fatigue initiation life occurred for the combination of temperatures {>=} 100 C, notch strain rates {<=} 0.1 %/s and notch strain amplitudes {>=} 0.3 %. If these conjoint threshold conditions were simultaneously satisfied, the environmental enhancement increased with decreasing strain rate and increasing temperature. Material and water chemistry parameters usually only had a little effect. Sensitization affected the CF behavior under highly oxidizing BWR/NWC conditions only. Preliminary block loading experiments did not reveal significant static load hold period effects on the technical corrosion fatigue initiation life. If the critical requirements were satisfied, the BWR/HWC and PWR environments usually resulted in acceleration of short fatigue crack growth by a factor of 5 to 20 with respect to air. Solution annealed steels showed slightly shorter CF initiation lives, but also lower stationary short CF crack growth rates under BWR/HWC and PWR conditions with low ECPs than under highly oxidizing BWR/NWC conditions. A very

  12. A procedure for the transient expression of genes by agroinfiltration above the permissive threshold to study temperature-sensitive processes in plant-pathogen interactions.

    Science.gov (United States)

    Del Toro, Francisco; Tenllado, Francisco; Chung, Bong-Nam; Canto, Tomas

    2014-10-01

    Localized expression of genes in plants from T-DNAs delivered into plant cells by Agrobacterium tumefaciens is an important tool in plant research. The technique, known as agroinfiltration, provides fast, efficient ways to transiently express or silence a desired gene without resorting to the time-consuming, challenging stable transformation of the host, the use of less efficient means of delivery, such as bombardment, or the use of viral vectors, which multiply and spread within the host causing physiological alterations themselves. A drawback of the agroinfiltration technique is its temperature dependence: early studies have shown that temperatures above 29 °C are nonpermissive to tumour induction by the bacterium as a result of failure in pilus formation. However, research in plant sciences is interested in studying processes at these temperatures, above the 25 °C experimental standard, common to many host-environment and host-pathogen interactions in nature, and agroinfiltration is an excellent tool for this purpose. Here, we measured the efficiency of agroinfiltration for the expression of reporter genes in plants from T-DNAs at the nonpermissive temperature of 30 °C, either transiently or as part of viral amplicons, and envisaged procedures that allow and optimize its use for gene expression at this temperature. We applied this technical advance to assess the performance at 30 °C of two viral suppressors of silencing in agropatch assays [Potato virus Y helper component proteinase (HCPro) and Cucumber mosaic virus 2b protein] and, within the context of infection by a Potato virus X (PVX) vector, also assessed indirectly their effect on the overall response of the host Nicotiana benthamiana to the virus. © 2014 BSPP AND JOHN WILEY & SONS LTD.

  13. A Comparison of Simple Methods to Incorporate Material Temperature Dependency in the Green's Function Method for Estimating Transient Thermal Stresses in Thick-Walled Power Plant Components.

    Science.gov (United States)

    Rouse, James; Hyde, Christopher

    2016-01-06

    The threat of thermal fatigue is an increasing concern for thermal power plant operators due to the increasing tendency to adopt "two-shifting" operating procedures. Thermal plants are likely to remain part of the energy portfolio for the foreseeable future and are under societal pressures to generate in a highly flexible and efficient manner. The Green's function method offers a flexible approach to determine reference elastic solutions for transient thermal stress problems. In order to simplify integration, it is often assumed that Green's functions (derived from finite element unit temperature step solutions) are temperature independent (this is not the case due to the temperature dependency of material parameters). The present work offers a simple method to approximate a material's temperature dependency using multiple reference unit solutions and an interpolation procedure. Thermal stress histories are predicted and compared for realistic temperature cycles using distinct techniques. The proposed interpolation method generally performs as well as (if not better) than the optimum single Green's function or the previously-suggested weighting function technique (particularly for large temperature increments). Coefficients of determination are typically above 0 . 96 , and peak stress differences between true and predicted datasets are always less than 10 MPa.

  14. Rapid Ag/Sn/Ag transient liquid phase bonding for high-temperature power devices packaging by the assistance of ultrasound.

    Science.gov (United States)

    Shao, Huakai; Wu, Aiping; Bao, Yudian; Zhao, Yue; Liu, Lei; Zou, Guisheng

    2017-07-01

    Rapid transient liquid phase (TLP) bonding process on Ag/Sn/Ag system is achieved in air by the assistance of ultrasonic, which has great potential to be applied to high-temperature power devices packaging. In this study, the influence of ultrasonic effect on the morphology and growth kinetics of Ag3Sn grains, and the joint microstructure, mechanical property and thermal reliability were systematically investigated. Experimental results indicated that the rapid consumption of the "dynamic" transient liquid phase was attributed to the accelerated dissolution of Ag substrate and the extrusion of liquid Sn, which were entirely induced by the complex sonochemical effects on the liquid/solid intermetallic compounds (IMCs) interface. An elongated scallop-like morphology of Ag3Sn grains was developed during Ag/Sn interfacial reaction with ultrasonic, accompanied by widening of grooves between neighbored grains. This phenomenon is called as a strengthening thermal grooving, in which the grooves at grain boundaries provide stable molten channels for Ag atoms diffusion from the substrate. Consequently, the improved elemental diffusion was evaluated through the growth kinetics of Ag3Sn IMCs, with conservative estimation of 6-16.5 times faster than the traditional TLP process. In addition, both excellent mechanical property and thermal reliability of the Ag-Sn intermetallic joint were experimentally verified by shear test and high-temperature storage test, respectively. Copyright © 2017 Elsevier B.V. All rights reserved.

  15. Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Min; Sabundjian, Gaianê, E-mail: smlee@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-11-01

    The aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input. (author)

  16. MELCOR 1.8.2 assessment: Surry PWR TMLB` (with a DCH study)

    Energy Technology Data Exchange (ETDEWEB)

    Kmetyk, L.N.; Cole, R.K. Jr.; Smith, R.C.; Summers, R.M.; Thompson, S.L.

    1994-02-01

    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC. This code models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze a station blackout transient in Surry, a three-loop Westinghouse PWR. Basecase results obtained with MELCOR 1.8.2 are presented, and compared to earlier results for the same transient calculated using MELCOR 1.8.1. The effects of new models added in MELCOR 1.8.2 (in particular, hydrodynamic interfacial momentum exchange, core debris radial relocation and core material eutectics, CORSOR-Booth fission product release, high-pressure melt ejection and direct containment heating) are investigated individually in sensitivity studies. The progress in reducing numeric effects in MELCOR 1.8.2, compared to MELCOR 1.8.1, is evaluated in both machine-dependency and time-step studies; some remaining sources of numeric dependencies (valve cycling, material relocation and hydrogen burn) are identified.

  17. Control of the neutronic and thermohydraulic conditions of power ramps in an irradiation loop for PWR fuel rod; Controle des conditions neutroniques et thermohydrauliques des rampes de puissance dans une boucle d`irradiation de combustibles de reacteur a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Moulin, D.J.F.

    1993-09-10

    In order to study the power transients effects on PWR fuel rod clad, ramp tests in a pressurized water loop, are carried out at OSIRIS reactor. The present thesis deals with the on-line control of the device, during power ramp and conditioning irradiation. Based on a convolution-type resolution of the kinetics equations, a dynamic compensation of the Silver self-powered neutron detector was developed. With this method, the uncertainty of the ramp end-point is lower than 1%, thus it is very suited for monitoring both transient, as well as steady state conditions. Furthermore, a thermohydraulic model of the irradiation device is described: heat transfer equations, including gamma heating in materials, are solved to obtain temperatures and thermal fluxes of steady states. Results from the model and temperature measurements of the coolant are used together for fuel power determination, in real time. The clad external temperature profile is also calculated and displayed, to improve the irradiation monitoring. (author), 51 refs., 12 annexes, 66 figs.

  18. Studi Operasi Resin Penukar Ion Dalam Sistem Purifikasi Air Primer Pwr

    OpenAIRE

    Biyantoro, Dwi; Basuki, Kris Tri; Subagiono, Subagiono

    2006-01-01

    STUDI OPERASI RESIN PENUKAR ION DALAM SISTEM PURIFIKASI AIR PRIMER PWR. Telah dilakukan studioperasi resin penukar ion dalam sistem purifikasi air primer PWR. Air pendingin reaktor yang pada awalnya sesuaidengan persyaratan setelah pengoperasian reaktor sering kualitasnya berubah, sehingga harus dimurnikan. Unsurunsurpengotor dalam air primer PWR diidentifikasi sebagai penyebab pengotor seperti korosi, pelepasan produk fisi(Cs137, Sr90, Co60,C14, Tc99), dan pelepasan kembali unsur oleh resin ...

  19. Numerical simulation of transient temperature profiles for canned apple puree in semi-rigid aluminum based packaging during pasteurization.

    Science.gov (United States)

    Shafiekhani, Soraya; Zamindar, Nafiseh; Hojatoleslami, Mohammad; Toghraie, Davood

    2016-06-01

    Pasteurization of canned apple puree was simulated for a 3-D geometry in a semi-rigid aluminum based container which was heated from all sides at 378 K. The computational fluid dynamics code Ansys Fluent 14.0 was used and the governing equations for energy, momentum, and continuity were computed using a finite volume method. The food model was assumed to have temperature-dependent properties. To validate the simulation, the apple puree was pasteurized in a water cascading retort. The effect of the mesh structures was studied for the temperature profiles during thermal processing. The experimental temperature in the slowest heating zone in the container was compared with the temperature predicted by the model and the difference was not significant. The study also investigated the impact of head space (water-vapor) on heat transfer.

  20. IPSN expert appraisal programme on the chooz A 300 MWe PWR. Lessons learned by IPSN

    Energy Technology Data Exchange (ETDEWEB)

    Morlent, O.; Reuchet, J. [CEA Fontenay-aux-Roses, Inst. de Protection et de Surete Nucleaire, 92 (France)

    2001-07-01

    The closure of Chooz A PWR provided an opportunity to take samples of items that had aged in situ in conditions close to those encountered in PWR in operation over a period of 140.000 hours, which is far longer than the usual time-spans of simulated laboratory tests. 4 topics have been studied: 1) effect of radiation on reactor vessel internals, 2) dissimilar metal joints of reactor coolant system: pressurizer surge line, 3) cast parts of austeno-ferritic steel: hot and cold leg primary valves, and 4) ageing of cables in high temperatures and under irradiation. The examination of the lower internals on some baffle angle bracket and core shroud screws, subjected to varying amounts of irradiation, did not reveal any cracking or corrosion, and confirmed the saturation effect between 4 and 10 dpa for the hardening of 304 austenitic steel in the low temperature range. Expert appraisal of the dissimilar metal joints on the pressurizer surge line confirmed the existence of small fabrication defects due to high temperature cracking. Expert appraisal of the 3 valve body samples from the main section of the coolant system confirmed that -) thermal ageing of the valve body on the hot leg was more advanced than that of the cold leg valve, -) the material of the valve housing on the cold leg which, in theory, was not sensitive to ageing phenomena, exhibited unexpectedly low impact strength values. As for cables, measurements confirmed that their mechanical and electrical properties remained sufficient for them to carry out their functions. (A.C.)

  1. Evaluation of the environmental quality of cables at PWR plants in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Nakanishi, S. [Kansai Electric Power Co., INC., Osaka (Japan); Mito, K. [Hokkaido Electric Power Co., INC., Sapporo (Japan); Kikuchi, K. [Shikoku Electric Power Co., INC., Takamatsu (Japan); Konishi, M. [Kyushu Electric Power Co., INC., Fukuoka (Japan); Suzuki, S. [The Japan Atomic Power Co., INC., Tokyo (Japan); Yamamoto, T. [Mitsubishi Heavy Industries, LTD., Kobe (Japan)

    2001-07-01

    Safety-related cables used at PWR plants in Japan must pass the environmental qualification test in accordance with IEEE Std. 323 and 383. IEEE Std.383 describes the methods of accelerated aging of cables that corresponds to the aging during the in-service period. However, IEEE Std. 775, etc. raise questions about the process in which: 1) conditions for the accelerated thermal aging are estimated by extrapolating activation energy, which is evaluated by the Arrhenius plot obtained in relatively high temperature range, 2) irradiation at high dose rate (<10 kGy/h) is permitted, and 3) irradiation is carried out sequentially after the thermal treatment. As a result, it was found that the activation energies obtained through the thermal aging test in a relatively low temperature range were considerably smaller than those obtained through the thermal aging test in a high temperature range. As to the method of the superposition of time dependent data proposed in IEC1244-2, we obtained good prospects that it would be possible to predict deterioration in an irradiation atmosphere at a very low dose rate of an actual plant with a considerably high accuracy, though it is necessary to obtain more data. Furthermore, we conducted LOCA exposure tests after the long-term thermal/radiation aging tests on EP rubber insulated cables and confirmed that cable soundness was maintained even after the LOCA. As a result of evaluating the long-term thermal/radiation aging period under the environmental conditions of an actual PWR plant by means of the shift factors obtained by the method of the superposition of time dependent data, it was evaluated that it would be equivalent to the aging of maximum 145 years.

  2. Winter stream temperature in the rain-on-snow zone of the Pacific Northwest: influences of hillslope runoff and transient snow cover

    Directory of Open Access Journals (Sweden)

    J. A. Leach

    2014-02-01

    Full Text Available Stream temperature dynamics during winter are less well studied than summer thermal regimes, but the winter season thermal regime can be critical for fish growth and development in coastal catchments. The winter thermal regimes of Pacific Northwest headwater streams, which provide vital winter habitat for salmonids and their food sources, may be particularly sensitive to changes in climate because they can remain ice-free throughout the year and are often located in rain-on-snow zones. This study examined winter stream temperature patterns and controls in small headwater catchments within the rain-on-snow zone at the Malcolm Knapp Research Forest, near Vancouver, British Columbia, Canada. Two hypotheses were addressed by this study: (1 winter stream temperatures are primarily controlled by advective fluxes associated with runoff processes and (2 stream temperatures should be depressed during rain-on-snow events, compared to rain-on-bare-ground events, due to the cooling effect of rain passing through the snowpack prior to infiltrating the soil or being delivered to the stream as saturation-excess overland flow. A reach-scale energy budget analysis of two winter seasons revealed that the advective energy input associated with hillslope runoff overwhelms vertical energy exchanges (net radiation, sensible and latent heat fluxes, bed heat conduction, and stream friction and hyporheic energy fluxes during rain and rain-on-snow events. Historical stream temperature data and modelled snowpack dynamics were used to explore the influence of transient snow cover on stream temperature over 13 winters. When snow was not present, daily stream temperature during winter rain events tended to increase with increasing air temperature. However, when snow was present, stream temperature was capped at about 5 °C, regardless of air temperature. The stream energy budget modelling and historical analysis support both of our hypotheses. A key implication is that

  3. Report on the PWR-radiation protection/ALARA Committee

    Energy Technology Data Exchange (ETDEWEB)

    Malone, D.J. [Consumers Power Co., Covert, MI (United States)

    1995-03-01

    In 1992, representatives from several utilities with operational Pressurized Water Reactors (PWR) formed the PWR-Radiation Protection/ALARA Committee. The mission of the Committee is to facilitate open communications between member utilities relative to radiation protection and ALARA issues such that cost effective dose reduction and radiation protection measures may be instituted. While industry deregulation appears inevitable and inter-utility competition is on the rise, Committee members are fully committed to sharing both positive and negative experiences for the benefit of the health and safety of the radiation worker. Committee meetings provide current operational experiences through members providing Plant status reports, and information relative to programmatic improvements through member presentations and topic specific workshops. The most recent Committee workshop was facilitated to provide members with defined experiences that provide cost effective ALARA performance.

  4. Conservative approach for PWR MOX Burnup Credit implementation

    Energy Technology Data Exchange (ETDEWEB)

    Jutier, Ludyvine; Checiak, Benoit; Raby, Jerome; Aguiar, Luis; Le Bars, Igor [IRSN, Fontenay-aux-Roses (France)

    2008-07-01

    Burnup Credit allows considering the reactivity decrease due to fuel irradiation in criticality studies for the nuclear fuel cycle. Its implementation requires to carefully analyze the validity of the assumption made to: define the axial profile of the burnup, determine the composition of the irradiated fuel and compute the criticality simulation. In the framework of Burnup Credit implementation for PWR mixed oxide fuels (MOX), this paper focus on the determination of a conservative inventory of the irradiated fuel. The studies presented in this paper concern: the influence of irradiation conditions and of the MOX fuel initial composition on the irradiated MOX fuel reactivity. Criticality calculations are also performed for PWR MOX fuel industrial applications in order to get Burnup Credit gain estimations. (authors)

  5. Condensate polishing guidelines for PWR and BWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Robbins, P.; Crinigan, P.; Graham, B.; Kohlmann, R.; Crosby, C.; Seager, J.; Bosold, R.; Gillen, J.; Kristensen, J.; McKeen, A.; Jones, V.; Sawochka, S.; Siegwarth, D.; Keeling, D.; Polidoroff, T.; Morgan, D.; Rickertsen, D.; Dyson, A.; Mills, W.; Coleman, L.

    1993-03-01

    Under EPRI sponsorship, an industry committee, similar in form and operation to other guideline committees, was created to develop Condensate Polishing Guidelines for both PWR and BWR systems. The committee reviewed the available utility and water treatment industry experience on system design and performance and incorporated operational and state-of-the-art information into document. These guidelines help utilities to optimize present condensate polisher designs as well as be a resource for retrofits or new construction. These guidelines present information that has not previously been presented in any consensus industry document. The committee generated guidelines that cover both deep bed and powdered resin systems as an integral part of the chemistry of PWR and BWR plants. The guidelines are separated into sections that deal with the basis for condensate polishing, system design, resin design and application, data management and performance and management responsibilities.

  6. A concept of PWR using plate and shell heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir; Andrade, Delvonei Alves de, E-mail: luciano.ondir@gmail.com, E-mail: delvonei@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    In previous work it was verified the physical possibility of using plate and shell heat exchangers for steam generation in a PWR for merchant ships. This work studies the possibility of using GESMEX commercial of the shelf plate and shell heat exchanger of series XPS. It was found it is feasible for this type of heat exchanger to meet operational and accidental requirements for steam generation in PWR. Additionally, it is proposed an arrangement of such heat exchangers inside the reactor pressure vessel. Such arrangement may avoid ANSI/ANS51.1 nuclear class I requirements on those heat exchangers because they are contained in the reactor coolant pressure barrier and play no role in accidental scenarios. Additionally, those plates work under compression, preventing the risk of rupture. Being considered non-nuclear safety, having a modular architecture and working under compression may turn such architectural choice a must to meet safety objectives with improved economics. (author)

  7. Isotopic depletion of soluble boron in a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J.M.; Ahnert, C.; Crespo, A.; Leon, J.R.

    1988-01-01

    The purpose of the work reported in this paper is to determine the isotopic depletion of the soluble boron in the primary of a pressurized water reactor (PWR) along cycle operation under the limiting condition of continuous boron dilution without fresh boron feeding, which maximizes the boron isotopic depletion effect. Presented here are the results for cycle 4 of the C.N. Almaraz-II PWR, which has been operated close to these continuous dilution conditions at rated power throughout most of the cycle. The limiting continuous boron dilution model with the /sup 10/B depletion calculations based on the COBAYA code is in quite good agreement with the measured boron concentrations for this cycle, explaining very well the differences found with the expected design letdown curve, which increased up to +60 ppm at middle of cycle when this work was done.

  8. Assessment of PWR plutonium burners for nuclear energy centers

    Energy Technology Data Exchange (ETDEWEB)

    Frankel, A J; Shapiro, N L

    1976-06-01

    The purpose of the study was to explore the performance and safety characteristics of PWR plutonium burners, to identify modifications to current PWR designs to enhance plutonium utilization, to study the problems of deploying plutonium burners at Nuclear Energy Centers, and to assess current industrial capability of the design and licensing of such reactors. A plutonium burner is defined to be a reactor which utilizes plutonium as the sole fissile addition to the natural or depleted uranium which comprises the greater part of the fuel mass. The results of the study and the design analyses performed during the development of C-E's System 80 plant indicate that the use of suitably designed plutonium burners at Nuclear Energy Centers is technically feasible.

  9. Application of the Relap5-3D to phase 1 and 3 of the OECD-CSNI/NSC PWR MSLB benchmark related to TMI-1

    Energy Technology Data Exchange (ETDEWEB)

    D' Auria, F.; Galassi, G.; Spadoni, A. [Pisa Univ., DMNP (Italy); Hassan, Y. [Texas A and M University, Dept. of Nuclear Engineering, College Station, Texas (United States)

    2001-07-01

    The Relap5-3D, the latest in the series of the Relap5 code, distinguishes from the previous versions by the fully integrated, multi-dimensional thermalhydraulic and kinetic modeling capability. It has been applied to Phase I and III of OECD-CSNI/ NSC PWR MSLB Benchmark adopting the same thermalhydraulic input deck already used with Relap5/Parcs and Relap5/Quabbox coupled codes during the previous MSLB analysis. The OECD jointly with the US NRC proposed the PWR MSLB Benchmark in order to gather a common understanding about the coupling between thermal hydraulics and neutronics, and evaluating the behavior of this transient with different coupled codes, giving emphasis to the 3-D modeling. This paper deals with the application of Relap5-3D code to phase I and III of the PWR MSLB Benchmark. The Relap5-3D is a thermal hydraulics-neutronics internally coupled code, the thermal hydraulics module is the INEEL version of Relap and the neutronics module is derived from NESTLE multi-dimension kinetics code. (author)

  10. Transient, three-dimensional heat transfer model for the laser assisted machining of silicon nitride: 1. Comparison of predictions with measured surface temperature histories

    Energy Technology Data Exchange (ETDEWEB)

    Rozzi, J.C.; Pfefferkorn, F.E.; Shin, Y.C. [Purdue University, (United States). Laser Assisted Materials Processing Laboratory, School of Mechanical Engineering; Incropera, F.P. [University of Notre Dame, (United States). Aerospace and Mechanical Engineering Department

    2000-04-01

    Laser assisted machining (LAM), in which the material is locally heated by an intense laser source prior to material removal, provides an alternative machining process with the potential to yield higher material removal rates, as well as improved control of workpiece properties and geometry, for difficult-to-machine materials such as structural ceramics. To assess the feasibility of the LAM process and to obtain an improved understanding of governing physical phenomena, experiments have been performed to determine the thermal response of a rotating silicon nitride workpiece undergoing heating by a translating CO{sub 2} laser and material removal by a cutting tool. Using a focused laser pyrometer, surface temperature histories were measured to determine the effect of the rotational and translational speeds, the depth of cut, the laser-tool lead distance, and the laser beam diameter and power on thermal conditions. The measurements are in excellent agreement with predictions based on a transient, three-dimensional numerical solution of the heating and material removal processes. The temperature distribution within the unmachined workpiece is most strongly influenced by the laser power and laser-tool lead distance, as well as by the laser/tool translational velocity. A minimum allowable operating temperature in the material removal region corresponds to the YSiAlON glass transition temperature, below which tool fracture may occur. In a companion paper, the numerical model is used to further elucidate thermal conditions associated with laser assisted machining. (author)

  11. Study on thermal-hydraulics during a PWR reflood phase

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation behavior observed during a combined injection PWR reflood phase. Further, author made analytical models for these specified issues, and succeeded to predict reflood behaviors at representative types of PWRs, i.e.cold leg injection PWRs and Combined injection PWRs, in good accuracy. Above results were incorporated into REFLA code which is developed at JAERI, and they improved accuracy in prediction and enlarged applicability of the code. In the present study, models were intended to be utilized in a practical use, and hence these models are simplified ones. However, physical understanding on the specified issues in the present study is basic and principal for reflood behavior, and then it is considered to be used in a future advanced code development and improvement. (author). 110 refs.

  12. Propagation of nuclear data Uncertainties for PWR core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cabellos, O.; Castro, E.; Ahnert, C.; Holgado, C. [Dept. of Nuclear Engineering, Universidad Politecnica de Madrid, Madrid (Spain)

    2014-06-15

    An uncertainty propagation methodology based on the Monte Carlo method is applied to PWR nuclear design analysis to assess the impact of nuclear data uncertainties. The importance of the nuclear data uncertainties for {sup 235,238}U, {sup 239}Pu, and the thermal scattering library for hydrogen in water is analyzed. This uncertainty analysis is compared with the design and acceptance criteria to assure the adequacy of bounding estimates in safety margins.

  13. PROPAGATION OF NUCLEAR DATA UNCERTAINTIES FOR PWR CORE ANALYSIS

    Directory of Open Access Journals (Sweden)

    O. CABELLOS

    2014-06-01

    Full Text Available An uncertainty propagation methodology based on the Monte Carlo method is applied to PWR nuclear design analysis to assess the impact of nuclear data uncertainties. The importance of the nuclear data uncertainties for 235,238U, 239Pu, and the thermal scattering library for hydrogen in water is analyzed. This uncertainty analysis is compared with the design and acceptance criteria to assure the adequacy of bounding estimates in safety margins.

  14. Safety Analysis Report for the PWR Spent Fuel Canister

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui Joo; Choi, Jong Won; Cho, Dong Keun; Chun, Kwan Sik; Lee, Jong Youl; Kim, Seong Ki; Kim, Seong Soo; Lee, Yang

    2005-11-15

    This report outlined the results of the safety assessment of the canisters for the PWR spent fuels which will be used in the KRS. All safety analyses including criticality and radiation shielding analyses, mechanical analyses, thermal analyses, and containment analyses were performed. The reference PWR spent fuels were in the 17x17 and determined to have 45,000 MWD/MTU burnup. The canister consists of copper outer shell and nodular cast iron inner structure with diameter of 102 cm and height of 483 cm. Criticality safety was checked for normal and abnormal conditions. It was assumed that the integrity of engineered barriers is preserved and saturated with water of 1.0g/cc for normal condition. For the abnormal condition container and bentonite was assumed to disappear, which allows the spent fuel to be surrounded by water with the most reactive condition. In radiation shielding analysis it was investigated that the absorbed dose at the surface of the canister met the safety limit. The structural analysis was conducted considering three load conditions, normal, extreme, and rock movement condition. Thermal analysis was carried out for the case that the canister with four PWR assemblies was deposited in the repository 500 meter below the surface with 40 m tunnel spacing and 6 m deposition hole spacing. The results of the safety assessment showed that the proposed KDC-1 canister met all the safety limits.

  15. Validation of gadolinium burnout using PWR benchmark specification

    Energy Technology Data Exchange (ETDEWEB)

    Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2014-07-01

    Graphical abstract: - Highlights: • We present methodology for validation of gadolinium burnout in PWR. • We model 17 × 17 PWR fuel assembly using MCB code. • We demonstrate C/E ratios of measured and calculated concentrations of Gd isotopes. • The C/E for Gd154, Gd156, Gd157, Gd158 and Gd160 shows good agreement of ±10%. • The C/E for Gd152 and Gd155 shows poor agreement below ±10%. - Abstract: The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17 × 17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor K{sub inf}, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157.

  16. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

  17. Timing analysis of PWR fuel pin failures

    Energy Technology Data Exchange (ETDEWEB)

    Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Straka, M. (Halliburton NUS, Idaho Falls, ID (United States))

    1992-09-01

    Research has been conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin bumup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PFL/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design bumup. Using peaking factors commensurate widi actual bumups would result in longer intervals for both reactor designs. This document also contains appendices A through J of this report.

  18. Optimization of the distribution of bars with gadolinium oxide in reactor fuel elements PWR; Optimizacion de la distribucion de barras con oxido de gadolinio en elementos combustibles para reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Melgar Santa Cecilia, P. A.; Velazquez, J.; Ahnert Iglesias, C.

    2014-07-01

    In the schemes of low leakage, currently used in the majority of PWR reactors, it makes use of absorbent consumables for the effective control of the factors of peak, the critical concentration of initial boron and the moderator temperature coefficient. One of the most used absorbing is the oxide of gadolinium, which is integrated within the fuel pickup. Occurs a process of optimization of fuel elements with oxide of gadolinium, which allows for a smaller number of configurations with a low peak factor for bar. (Author)

  19. Three dimensional calculations of the primary coolant flow in a 900 MW PWR vessel. Numerical simulation of the accurate RCP start-up flow rate

    Energy Technology Data Exchange (ETDEWEB)

    Martin, A.; Alvarez, D.; Cases, F.; Stelletta, S. [Electricite de France (EDF), 78 - Chatou (France). Lab. National d`Hydraulique

    1997-06-01

    This report explains the last results about the mixing in the 900 MW PWR vessels. The accurate fluid flow transient, induced by the RCP starting-up, is represented. In a first time, we present the Thermalhydraulic Finite Element Code N3S used for the 3D numerical computations. After that, results obtained for one reactor operation case are given. This case is dealing with the transient mixing of a clear plug in the vessel when one primary pump starts-up. A comparison made between two injection modes; a steady state fluid flow conditions or the accurate RCP transient fluid flow conditions. The results giving the local minimum of concentration and the time response of the mean concentration at the core inlet are compared. The results show the real importance of the unsteadiness characteristics of the fluid flow transport of the clear water plug. (author) 12 refs.

  20. [Effect of traditional Chinese medicines with different properties on thermoregulation and temperature-sensitive transient receptor potentialion channel protein of rats with yeast-induced fever].

    Science.gov (United States)

    Wan, Hong-Ye; Kong, Xiang-Ying; Li, Xiao-Min; Zhu, Hong-Wei; Su, Xiao-Hui; Lin, Na

    2014-10-01

    To compare the intervention effects of four traditional Chinese medicines (TCMs) with typical cold or hot property on body temperature and temperature-sensitive transient receptor potential ion channel proteins (TRPs) of rats with yeast-induced fever. The pyrexia model was induced by injecting yeast suspension subcutaneously. Totally 108 male SD rats were randomly divided into the normal group, the model group, the Rhei Radix et Rhizoma treated group, the Coptidis Rhizoma treated group, the Euodiae Fructus treated group, and the Alpiniae Officinarum Rhizoma treated group, with 18 rats in each group. At the 4 h, 8 h and 12 h after injection of yeast, the rats were sacrificed to collect their hypothalamus and dorsal root ganglion. The expressions of TRPV1 and TRPM8 were detected by immunohistochemistry and Western blot method. Compared with the normal group, after injection of yeast, the temperature of rats in the model group notably increased, and reached the peak at 8 h (P < 0.01). The TRPV1 level in hypothalamus and dorsal root ganglia (DRG) of the model group significantly increased, whereas the TRPM8 level significantly reduced. Compared with the model group, the Rhei Radix et Rhizoma group and the Coptidis Rhizoma group showed significant decrease in the high body temperature of rats caused by yeast, down-regulation in the expression of TRPV1, and up-regulation in the expression of TRPM8 (P < 0.05 or P < 0.01). Euodiae Fructus and Alpiniae Officinarum Rhizoma had no significant effect on either temperature or TRPs of fever rats. Rhei Radix et Rhizoma and Coptidis Rhizoma, both are TCMs with cold property, can reduce the temperature of fever rats induced by yeast, which may be related to their effective regulation of TRPV1 and TRPM8 in hypothalamus and DRG, while Euodiae Fructus and Alpiniae Officinarum Rhizoma had no relevant effect.

  1. Demonstration of Uncertainty Quantification and Sensitivity Analysis for PWR Fuel Performance with BISON

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin; Ladd, Jacob; Zhao, Haihua; Zou, Ling; Burns, Douglas

    2015-11-01

    BISON is an advanced fuels performance code being developed at Idaho National Laboratory and is the code of choice for fuels performance by the U.S. Department of Energy (DOE)’s Consortium for Advanced Simulation of Light Water Reactors (CASL) Program. An approach to uncertainty quantification and sensitivity analysis with BISON was developed and a new toolkit was created. A PWR fuel rod model was developed and simulated by BISON, and uncertainty quantification and sensitivity analysis were performed with eighteen uncertain input parameters. The maximum fuel temperature and gap conductance were selected as the figures of merit (FOM). Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis.

  2. Pressure loss tests for DR-BEP of fullsize 17 x 17 PWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Moon Ki; Chun, Se Young; Chang, Seok Kyu; Won, Soon Youn; Cho, Young Rho; Kim, Bok Deuk; Min, Kyoung Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-01-01

    This report describes the conditions, procedure and results in the pressure loss tests carried out for a double grid type debris resistance bottom end piece (DR-BEP) designed by KAERI. In this test, the pressure loss coefficients of the full size 17 x 17 PWR simulated fuel assembly with DR-BET and with standard-BEP were measured respectively, and the pressure loss coefficients of DR-BEP were compared with the coefficients of STD-BET. The test conditions fall within the ranges of loop pressure from 5.2 to 45 bar, loop temperature from 27 to 221 deg C and Reynolds number in fuel bundle from 2.17 x 10{sup 4} to 3.85 x 10{sup 5}. (Author) 5 refs., 18 figs., 5 tabs.

  3. Modeling and Simulation of the Transient Response of Temperature and Relative Humidity Sensors with and without Protective Housing

    Science.gov (United States)

    Rocha, Keller Sullivan Oliveira; Martins, José Helvecio; Martins, Marcio Arêdes; Ferreira Tinôco, Ilda de Fátima; Saraz, Jairo Alexander Osorio; Filho, Adílio Flauzino Lacerda; Fernandes, Luiz Henrique Martins

    2014-01-01

    Based on the necessity for enclosure protection of temperature and relative humidity sensors installed in a hostile environment, a wind tunnel was used to quantify the time that the sensors take to reach equilibrium in the environmental conditions to which they are exposed. Two treatments were used: (1) sensors with polyvinyl chloride (PVC) enclosure protection, and (2) sensors with no enclosure protection. The primary objective of this study was to develop and validate a 3-D computational fluid dynamics (CFD) model for analyzing the temperature and relative humidity distribution in a wind tunnel using sensors with PVC enclosure protection and sensors with no enclosure protection. A CFD simulation model was developed to describe the temperature distribution and the physics of mass transfer related to the airflow relative humidity. The first results demonstrate the applicability of the simulation. For verification, a sensor device was successfully assembled and tested in an environment that was optimized to ensure fast change conditions. The quantification setup presented in this paper is thus considered to be adequate for testing different materials and morphologies for enclosure protection. The results show that the boundary layer flow regime has a significant impact on the heat flux distribution. The results indicate that the CFD technique is a powerful tool which provides a detailed description of the flow and temperature fields as well as the time that the relative humidity takes to reach equilibrium with the environment in which the sensors are inserted. PMID:24851994

  4. Modeling and simulation of the transient response of temperature and relative humidity sensors with and without protective housing.

    Science.gov (United States)

    Rocha, Keller Sullivan Oliveira; Martins, José Helvecio; Martins, Marcio Arêdes; Tinôco, Ilda de Fátima Ferreira; Saraz, Jairo Alexander Osorio; Lacerda Filho, Adílio Flauzino; Fernandes, Luiz Henrique Martins

    2014-01-01

    Based on the necessity for enclosure protection of temperature and relative humidity sensors installed in a hostile environment, a wind tunnel was used to quantify the time that the sensors take to reach equilibrium in the environmental conditions to which they are exposed. Two treatments were used: (1) sensors with polyvinyl chloride (PVC) enclosure protection, and (2) sensors with no enclosure protection. The primary objective of this study was to develop and validate a 3-D computational fluid dynamics (CFD) model for analyzing the temperature and relative humidity distribution in a wind tunnel using sensors with PVC enclosure protection and sensors with no enclosure protection. A CFD simulation model was developed to describe the temperature distribution and the physics of mass transfer related to the airflow relative humidity. The first results demonstrate the applicability of the simulation. For verification, a sensor device was successfully assembled and tested in an environment that was optimized to ensure fast change conditions. The quantification setup presented in this paper is thus considered to be adequate for testing different materials and morphologies for enclosure protection. The results show that the boundary layer flow regime has a significant impact on the heat flux distribution. The results indicate that the CFD technique is a powerful tool which provides a detailed description of the flow and temperature fields as well as the time that the relative humidity takes to reach equilibrium with the environment in which the sensors are inserted.

  5. Measurement of Transient Tool Internal Temperature Fields by Novel Micro Thin Film Sensors Embedded in Polycrystalline Cubic Boron Nitride Cutting Inserts

    Science.gov (United States)

    Werschmoeller, Dirk

    Monitoring and control of thermomechanical phenomena in tooling are imperative for advancing fundamental understanding, enhancing reliability, and improving workpiece quality in material removal processes. Polycrystalline cubic boron nitride (PCBN) tools are being used heavily in numerous machining processes, e.g., machining of hardened low carbon steel and superalloys. These processes are very sensitive to variations in local cutting conditions at, or close to, the tool-workpiece interface, but lack a thorough understanding of fundamental transient thermo-mechanical phenomena present. As a result, abrupt catastrophic tool failures and degraded machined surfaces frequently occur. Existing sensors are not suitable for process control and monitoring, as they are either destructively embedded and/or do not possess the necessary spatial and temporal resolution to provide relevant data during machining. This research presents a novel approach for obtaining thermomechanical data from the close vicinity (i.e., 10s of micrometers) of the tool-workpiece interface. Arrays of micro thin film thermocouples with junction size 5 x 5 mum were fabricated by standard microfabrication methods and have been successfully embedded into PCBN using diffusion bonding. Electron microscopy and X-ray spectroscopy were employed to examine material interactions at the bonding interface and to determine optimal bonding parameters. Static and dynamic sensor performances have been characterized. The sensors exhibit excellent linearity up to 1300 °C, fast rise time of 150 ns, and possess good sensitivity. The inserts instrumented with embedded thin film C-type thermocouples were successfully applied to measure internal tool temperatures as close as 70 mum to the cutting edge while machining aluminum and hardened steel workpieces at industrially relevant cutting parameters. Acquired temperature data follow theoretical trends very well. Correlations between temperature and cutting parameters have

  6. Chemical reaction and radiation effects on the transient MHD free convection flow of dissipative fluid past an infinite vertical porous plate with ramped wall temperature

    Directory of Open Access Journals (Sweden)

    V. RAJESH

    2011-06-01

    Full Text Available A finite-difference analysis is performed to study the effects of thermal radiation and chemical reaction on the transient MHD free convection and mass transform flow of a dissipative fluid past an infinite vertical porous plate subject to ramped wall temperature. The fluid considered here is a gray, absorbing/ /emitting radiation but a non-scattering medium. The dimensionless governing equations are unsteady, coupled and non-linear partial differential equations. An analytical method fails to give a solution. Hence an implicit finite difference scheme of Crank-Nicolson method is employed. The effect of the magnetic parameter (M, chemical reaction parameter (K, radiation parameter (F, buoyancy ratio parameter (N, Schmidt number (Sc on the velocity field and skin friction for both air (Pr = 0.71 and water (Pr = 7 in the presence of both aiding (N>0 and opposing (N<0 flows are extensively discussed with the help of graphs.

  7. PSA LEVEL 3 DAN IMPLEMENTASINYA PADA KAJIAN KESELAMATAN PWR

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-03-01

    Full Text Available Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian deterministik. Metodologi kajian menggunakan Probabilistic Safety Assessment (PSA Level 3 diperlukan terutama untuk estimasi kecelakaan parah atau kecelakaan luar dasar desain PLTN. Metode ini banyak dilakukan setelah kejadian kecelakaan Fukushima. Dalam penelitian ini dilakukan implementasi PSA Level 3 pada kajian keselamatan PWR, postulasi kecelakan luar dasar desain PWR AP-1000 dan disimulasikan di contoh tapak Bangka Barat. Rangkaian perhitungan yang dilakukan adalah: menghitung suku sumber dari kegagalan teras yang terjadi, pemodelan kondisi meteorologi tapak dan lingkungan, pemodelan jalur paparan, analisis dispersi radionuklida dan transportasi fenomena di lingkungan, analisis deposisi radionuklida, analisis dosis radiasi, analisis perlindungan & mitigasi, dan analisis risiko. Kajian menggunakan rangkaian subsistem pada perangkat lunak PC Cosyma. Hasil penelitian membuktikan bahwa implementasi metode kajian keselamatan PSA Level 3 sangat efektif dan komprehensif terhadap estimasi dampak, konsekuensi, risiko, kesiapsiagaan kedaruratan nuklir (nuclear emergency preparedness, dan manajemen kecelakaan reaktor terutama untuk kecelakaan parah atau kecelakaan luar dasar desain PLTN. Hasil kajian dapat digunakan sebagai umpan balik untuk kajian keselamatan PSA Level 1 dan PSA Level 2. Kata kunci: PSA level 3, kecelakaan, PWR   Reactor safety assessment of nuclear power plants using probabilistic assessment methodology is most important in addition to the deterministic assessment. The methodology of Level 3 Probabilistic Safety Assessment (PSA is especially required to estimate severe accident or beyond design basis accidents of nuclear power plants. This method is carried out after the Fukushima accident. In this research, the postulations beyond design basis accidentsof PWR AP - 1000 would be taken, and simulated at West Bangka sample site. The

  8. ORINC: a one-dimensional implicit approach to the inverse heat conduction problem. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ott, L.J.; Hedrick, R.A.

    1977-10-18

    The report develops an implicit solution technique to determine both the transient surface temperature and the transient surface heat flux of electrically heated rods given the power input and an ''indicated'' internal temperature during a simulated loss-of-coolant accident. A digital computer program ORINC (ORNL Inverse Code) is developed which solves a one-dimensional, transient, lumped parameter, implicit formulation of the conduction equation at each bundle thermocouple position in the Thermal-Hydraulic Test Facility (THTF).

  9. NUMERICAL MULTIGROUP TRANSIENT ANALYSIS OF SLAB NUCLEAR REACTOR WITH THERMAL FEEDBACK

    Directory of Open Access Journals (Sweden)

    Filip Osuský

    2016-12-01

    Full Text Available The paper describes a new numerical code for multigroup transient analyses with thermal feedback. The code is developed at Institute of Nuclear and Physical Engineering. It is necessary to carefully investigate transient states of fast neutron reactors, due to recriticality issues after accident scenarios. The code solves numerical diffusion equation for 1D problem with possible neutron source incorporation. Crank-Nicholson numerical method is used for the transient states. The investigated cases are describing behavior of PWR fuel assembly inside of spent fuel pool and with the incorporated neutron source for better illustration of thermal feedback.

  10. Acetosyringone, pH and temperature effects on transient genetic transformation of immature embryos of Brazilian wheat genotypes by Agrobacterium tumefaciens

    Directory of Open Access Journals (Sweden)

    Ernandes Manfroi

    2015-01-01

    Full Text Available AbstractLow transformation efficiency is one of the main limiting factors in the establishment of genetic transformation of wheat via Agrobacterium tumefaciens. To determine more favorable conditions for T-DNA delivery and explant regeneration after infection, this study investigated combinations of acetosyringone concentration and pH variation in the inoculation and co-cultivation media and co-culture temperatures using immature embryos from two Brazilian genotypes (BR 18 Terena and PF 020037. Based on transient expression of uidA, the most favorable conditions for T-DNA delivery were culture media with pH 5.0 and 5.4 combined with co-culture temperatures of 22 °C and 25 °C, and a 400 μM acetosyringone supplement. These conditions resulted in blue foci in 81% of the embryos. Media with more acidic pH also presented reduced A. tumefaciensovergrowth during co-culture, and improved regeneration frequency of the inoculated explants. BR 18 Terena was more susceptible to infection by A. tumefaciens than PF 020037. We found that it is possible to improve T-DNA delivery and explant regeneration by adjusting factors involved in the early stages of A. tumefaciens infection. This can contribute to establishing a stable transformation procedure in the future.

  11. Singular deposit formation in PWR due to electrokinetic phenomena - application to SG clogging

    Energy Technology Data Exchange (ETDEWEB)

    Guillodo, M.; Muller, T.; Barale, M.; Foucault, M. [AREVA NP SAS, Technical Centre (France); Clinard, M.-H.; Brun, C.; Chahma, F. [AREVA NP SAS, Chemistry and Radiochemistry Group (France); Corredera, G.; De Bouvier, O. [Electricite de France, Centre d' Expertise de I' inspection dans les domaines de la Realisation et de l' Exploitation (France)

    2009-07-01

    The deposits which cause clogging of the 'foils' of the tube support plates (TSP) in Steam Generators (SG) of PWR present two characteristics which put forward that the mechanism at the origin of their formation is different from the mechanism that drives the formation of homogeneous deposits leading to the fouling of the free spans of SG tubes. Clogging occurs near the leading edge of the TSP and the deposits appear as diaphragms localized between both TSP and SG tubing materials, while the major part of the tube/TSP interstice presents little or no significant clogging. This type of deposit seems rather comparable to the ones which were reproduced in Lab tests to explain the flow rate instabilities observed on a French unit during hot shutdown in the 90's. The deposits which cause TSP clogging are owed to a discontinuity of the streaming currents in the vicinity of a surface singularity (orifices, scratches ...) which, in very low conductivity environment, produce local potential variations and/or current loop in the metallic pipe material due to electrokinetic effects. Deposits can be built by two mechanisms which may or not coexist: (i) accumulation of particles stabilized by an electrostatic attraction due to the local variation of electrokinetic potential, and (ii) crystalline growth of magnetite produced by the oxidation of ferrous ions on the anodic branch of a current loop. Lab investigations carried out by AREVA NP Technical Centre since the end of the 90's showed that this type of deposit occurs when the redox potential is higher than a critical value, and can be gradually dissolved when the potential becomes lower than this value which depends on the 'Material - Chemistry' couple. Special emphasis will be given in this paper to the TSP clogging of SG in PWR secondary coolant dealing particularly with the potential strong effect of electrokinetic phenomena in low conductive environment and in high temperature conditions

  12. Test requirements for the integral effect test to simulate Korean PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K

    2001-02-01

    In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time.

  13. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Matias, R.; Fernandez, K.; Justo, D.; Bocanegra, R.; Mena, L.; Queral, C.

    2015-07-01

    During a severe accident in a PWR, the hydrogen generated may be distributed in the containment atmosphere and reach the combustion conditions that can cause the containment failure. In this research project, a preliminary study has been done about the capacities of ANSYS Fluent 15.0 and GOTHIC 8.0 to tri dimensional distribution of the hydrogen in a PWR containment during a severe accident. (Author)

  14. Transient Heat Conduction

    DEFF Research Database (Denmark)

    Rode, Carsten

    1998-01-01

    Analytical theory of transient heat conduction.Fourier's law. General heat conducation equation. Thermal diffusivity. Biot and Fourier numbers. Lumped analysis and time constant. Semi-infinite body: fixed surface temperature, convective heat transfer at the surface, or constant surface heat flux...

  15. Experimental observation of transient δ18O interaction between snow and advective airflow under various temperature gradient conditions

    Directory of Open Access Journals (Sweden)

    P. P. Ebner

    2017-07-01

    Full Text Available Stable water isotopes (δ18O obtained from snow and ice samples of polar regions are used to reconstruct past climate variability, but heat and mass transport processes can affect the isotopic composition. Here we present an experimental study on the effect of airflow on the snow isotopic composition through a snow pack in controlled laboratory conditions. The influence of isothermal and controlled temperature gradient conditions on the δ18O content in the snow and interstitial water vapour is elucidated. The observed disequilibrium between snow and vapour isotopes led to the exchange of isotopes between snow and vapour under non-equilibrium processes, significantly changing the δ18O content of the snow. The type of metamorphism of the snow had a significant influence on this process. These findings are pertinent to the interpretation of the records of stable isotopes of water from ice cores. These laboratory measurements suggest that a highly resolved climate history is relevant for the interpretation of the snow isotopic composition in the field.

  16. Experimental observation of transient δ18O interaction between snow and advective airflow under various temperature gradient conditions

    Science.gov (United States)

    Ebner, Pirmin Philipp; Steen-Larsen, Hans Christian; Stenni, Barbara; Schneebeli, Martin; Steinfeld, Aldo

    2017-07-01

    Stable water isotopes (δ18O) obtained from snow and ice samples of polar regions are used to reconstruct past climate variability, but heat and mass transport processes can affect the isotopic composition. Here we present an experimental study on the effect of airflow on the snow isotopic composition through a snow pack in controlled laboratory conditions. The influence of isothermal and controlled temperature gradient conditions on the δ18O content in the snow and interstitial water vapour is elucidated. The observed disequilibrium between snow and vapour isotopes led to the exchange of isotopes between snow and vapour under non-equilibrium processes, significantly changing the δ18O content of the snow. The type of metamorphism of the snow had a significant influence on this process. These findings are pertinent to the interpretation of the records of stable isotopes of water from ice cores. These laboratory measurements suggest that a highly resolved climate history is relevant for the interpretation of the snow isotopic composition in the field.

  17. Estimating probable flaw distributions in PWR steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Gorman, J.A.; Turner, A.P.L. [Dominion Engineering, Inc., McLean, VA (United States)

    1997-02-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses.

  18. Assessment of cold composite fuels for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Coulon-Picard, E.; Agard, M.; Boulore, A.; Castelier, E.; Chabert, C.; Conti, A.; Frayssines, P.E.; Lechelle, J.; Maillard, S.; Matheron, P.; Pelletier, M.; Phelip, M.; Piluso, P.; Vaudano, A

    2009-06-15

    This study is devoted to evaluation of a new innovative micro structured fuel for future pressurized water reactor. This fuel would have potential to increase the safety margins, lowering fuel temperatures by adding a small fraction of a high conductivity second phase material in the oxide fuel phase. The behavior of this fuel in a standard rod has been modeled with finite element codes and was assessed for different aspects of the cycle as neutronic studies, thermal behavior, reprocessing and economics. Feasibility of fuels has been investigated with the fabrication and characterizations of the microstructure of composite fuels with powder metallurgy and HIP processes. First, a CERCER (Ceramic = UO{sub 2}- Ceramic matrix made of silicon carbide, SiC) fuel type has been investigated, the advantages of a ceramic being generally its transparency to neutrons and its high melting temperature. A first design of kernel type fuel was first chosen with a gap between the UO{sub 2} particles and the second phase material in order to avoid mechanical interaction between the two components. Due to lowering thermal conductivity of SiC under irradiation, this CERCER fuel did not allow a temperature gain compared to current fuel. No ceramic material seems to exhibit all required properties. Even beryllium oxide (BeO), which conductivity does not decrease with irradiation according to the literature, induces difficulties with ({alpha}, n) reactions and toxicity. The study then focused on Cermet fuels (Ceramic-Metal). The metal matrix must be transparent to neutrons and have a good thermal conductivity. Several materials have been considered such as zirconium alloys, austenitic and ferritic stainless steals and chromium based alloys. The heterogeneous composite fuels were modeled using the 3D/CASTM finite element code. From an economical and neutron point of view, it was important to keep a low fraction of metal phase, i.e. less than 10 % of Zr for example. However, the fuel

  19. Simulation of steam generator plugging tubes in a PWR to analyze the operating impact

    Energy Technology Data Exchange (ETDEWEB)

    Pla, Patricia, E-mail: patricia.pla-freixa@ec.europa.eu [Nuclear Reactor Safety Assessment Unit, Institute for Energy and Transport, Joint Research Centre (JRC) of the European Commission, Petten (Netherlands); Reventos, Francesc, E-mail: francesc.reventos@upc.edu [Technical University of Catalonia (UPC), Barcelona (Spain); Martin Ramos, Manuel, E-mail: manuel.martin-ramos@ec.europa.eu [Nuclear Safety and Security Coordination Unit, Policy Support Coordination, Joint Research Centre of the European Commission, Brussels (Belgium); Sol, Ismael, E-mail: isol@anacnv.com [Asociación Nuclear Ascó-Vandellós-II (ANAV), Tarragona (Spain); Strucic, Miodrag, E-mail: miodrag.strucic@ec.europa.eu [Nuclear Reactor Safety Assessment Unit, Institute for Energy and Transport, Joint Research Centre (JRC) of the European Commission, Petten (Netherlands)

    2016-08-15

    Highlights: • Plugging a fraction of the SG tubes does not affect power output of the plant. • There is a limit to SG plugging in the range of 10–15%. • The rupture of a SG tube in a 12% plugged SG has shown no significant differences in operator actions. • A SBLOCA in a 12% plugged SG has shown no significant differences in operator actions. - Abstract: A number of nuclear power plants (NPPs) with pressurized water reactors (PWR) in the world have replaced their steam generators (SG) due to degradation of the SG tubes caused by different problems. Several methods were attempted to correct the defects of the tubes, but eventually the only permanent solution was to plug them. The consequences of plugging the tubes are the decrease of heat transfer surface, the reduction of the flow area and subsequent reduction of the primary system mass flow and for a fraction of plugged tubes higher than a given value, the reduction of reactor output and economic losses. The objective of this paper is to analyze whether steam generator tube plugging has an impact in the effectiveness of accident management actions. An analysis with Relap5 Mod 3.3 patch03 for the Spanish reactor Ascó-2, a 3-loop 2940.6 MWth Westinghouse PWR, in which plugging of steam generator tubes are simulated, is presented in order to find the limit for the adequate operation of the plant. Several steady state calculations were performed with different fractions of plugged SG tubes, by modeling the reduction of the primary to secondary heat transfer surface and the reduction of the primary coolant mass flow area in the tubes as well. The results of the analysis yield that plugging 12% of the SG tubes is around the limit for optimal reactor operation. To complete the study two events, in which the steam generators are used to cooldown the plant, were simulated to find out if the plugging of SGs tubes could influence the efficiency of the operator actions described in the emergency operating

  20. Modelling guidelines for core exit temperature simulations with system codes

    Energy Technology Data Exchange (ETDEWEB)

    Freixa, J., E-mail: jordi.freixa-terradas@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Martínez-Quiroga, V., E-mail: victor.martinez@nortuen.com [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Zerkak, O., E-mail: omar.zerkak@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Reventós, F., E-mail: francesc.reventos@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain)

    2015-05-15

    Highlights: • Core exit temperature is used in PWRs as an indication of core heat up. • Modelling guidelines of CET response with system codes. • Modelling of heat transfer processes in the core and UP regions. - Abstract: Core exit temperature (CET) measurements play an important role in the sequence of actions under accidental conditions in pressurized water reactors (PWR). Given the difficulties in placing measurements in the core region, CET readings are used as criterion for the initiation of accident management (AM) procedures because they can indicate a core heat up scenario. However, the CET responses have some limitation in detecting inadequate core cooling and core uncovery simply because the measurement is not placed inside the core. Therefore, it is of main importance in the field of nuclear safety for PWR power plants to assess the capabilities of system codes for simulating the relation between the CET and the peak cladding temperature (PCT). The work presented in this paper intends to address this open question by making use of experimental work at integral test facilities (ITF) where experiments related to the evolution of the CET and the PCT during transient conditions have been carried out. In particular, simulations of two experiments performed at the ROSA/LSTF and PKL facilities are presented. The two experiments are part of a counterpart exercise between the OECD/NEA ROSA-2 and OECD/NEA PKL-2 projects. The simulations are used to derive guidelines in how to correctly reproduce the CET response during a core heat up scenario. Three aspects have been identified to be of main importance: (1) the need for a 3-dimensional representation of the core and Upper Plenum (UP) regions in order to model the heterogeneity of the power zones and axial areas, (2) the detailed representation of the active and passive heat structures, and (3) the use of simulated thermocouples instead of steam temperatures to represent the CET readings.

  1. Life management plants at nuclear power plants PWR; Planes de gestion de vida en centrales nucleares PWR

    Energy Technology Data Exchange (ETDEWEB)

    Esteban, G.

    2014-10-01

    Since in 2009 the CSN published the Safety Instruction IS-22 (1) which established the regulatory framework the Spanish nuclear power plants must meet in regard to Life Management, most of Spanish nuclear plants began a process of convergence of their Life Management Plants to practice 10 CFR 54 (2), which is the current standard of Spanish nuclear industry for Ageing Management, either during the design lifetime of the plant, as well as for Long-Term Operation. This article describe how Life Management Plans are being implemented in Spanish PWR NPP. (Author)

  2. Beta and gamma dose calculations for PWR and BWR containments

    Energy Technology Data Exchange (ETDEWEB)

    King, D.B.

    1989-07-01

    Analyses of gamma and beta dose in selected regions in PWR and BWR containment buildings have been performed for a range of fission product releases from selected severe accidents. The objective of this study was to determine the radiation dose that safety-related equipment could experience during the selected severe accident sequences. The resulting dose calculations demonstrate the extent to which design basis accident qualified equipment could also be qualified for the severe accident environments. Surry was chosen as the representative PWR plant while Peach Bottom was selected to represent BWRs. Battelle Columbus Laboratory performed the source term release analyses. The AB epsilon scenario (an intermediate to large LOCA with failure to recover onsite or offsite electrical power) was selected as the base case Surry accident, and the AE scenario (a large break LOCA with one initiating event and a combination of failures in two emergency cooling systems) was selected as the base case Peach Bottom accident. Radionuclide release was bounded for both scenarios by including spray operation and arrested sequences as variations of the base scenarios. Sandia National Laboratories used the source terms to calculate dose to selected containment regions. Scenarios with sprays operational resulted in a total dose comparable to that (2.20 /times/ 10/sup 8/ rads) used in current equipment qualification testing. The base case scenarios resulted in some calculated doses roughly an order of magnitude above the current 2.20 /times/ 10/sup 8/ rad equipment qualification test region. 8 refs., 23 figs., 12 tabs.

  3. VERA Core Simulator Methodology for PWR Cycle Depletion

    Energy Technology Data Exchange (ETDEWEB)

    Kochunas, Brendan [University of Michigan; Collins, Benjamin S [ORNL; Jabaay, Daniel [University of Michigan; Kim, Kang Seog [ORNL; Graham, Aaron [University of Michigan; Stimpson, Shane [University of Michigan; Wieselquist, William A [ORNL; Clarno, Kevin T [ORNL; Palmtag, Scott [Core Physics, Inc.; Downar, Thomas [University of Michigan; Gehin, Jess C [ORNL

    2015-01-01

    This paper describes the methodology developed and implemented in MPACT for performing high-fidelity pressurized water reactor (PWR) multi-cycle core physics calculations. MPACT is being developed primarily for application within the Consortium for the Advanced Simulation of Light Water Reactors (CASL) as one of the main components of the VERA Core Simulator, the others being COBRA-TF and ORIGEN. The methods summarized in this paper include a methodology for performing resonance self-shielding and computing macroscopic cross sections, 2-D/1-D transport, nuclide depletion, thermal-hydraulic feedback, and other supporting methods. These methods represent a minimal set needed to simulate high-fidelity models of a realistic nuclear reactor. Results demonstrating this are presented from the simulation of a realistic model of the first cycle of Watts Bar Unit 1. The simulation, which approximates the cycle operation, is observed to be within 50 ppm boron (ppmB) reactivity for all simulated points in the cycle and approximately 15 ppmB for a consistent statepoint. The verification and validation of the PWR cycle depletion capability in MPACT is the focus of two companion papers.

  4. Degradation of fastener in reactor internal of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. W.; Ryu, W. S.; Jang, J. S.; Kim, S. H.; Kim, W. G.; Chung, M. K.; Han, C. H

    2000-03-01

    Main component degraded in reactor internal structure of PWR is fastener such as bolts, stud, cap screw, and pins. The failure of these components may damage nuclear fuel and limits the operation of nuclear reactor. In foreign reactors operated more than 10 years, an increasing number of incidents of degraded thread fasteners have been reported. The degradation of these components impair the integrity of reactor internal structure and limit the life extension of nuclear power plant. To solve the problem of fastener failure, the incidents of failure and main mechanisms should be investigated. the purpose of this state-of-the -art report is to investigate the failure incidents and mechanisms of fastener in foreign and domestic PWR and make a guide to select a proper materials. There is no intent to describe each event in detail in this report. This report covers the failures of fastener and damage mechanisms reported by the licensees of operating nuclear power plants and the applications of plants constructed after 1964. This information is derived from pertinent licensee event report, reportable occurrence reports, operating reactor event memoranda, failure analysis reports, and other relevant documents. (author)

  5. Resistance of Incoloy 800 steam generator tube to pitting corrosion in PWR secondary water at 250°C

    Energy Technology Data Exchange (ETDEWEB)

    Schvartzman, Mônica M.A.M. [Pontifícia Universidade Católica de Minas Gerais (PUC-Minas), Belo Horizonte, MG (Brazil); Albuquerque, Adriana Silva de; Esteves, Luiza; Rabello, Emerson G.; Mansur, Fábio Abud, E-mail: monicacdtn@gmail.com, E-mail: asa@cdtn.br, E-mail: luiza.esteves@cdtn.br, E-mail: egr@cdtn.br, E-mail: fametalurgica@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The steam generator (SG) is one of the main components of a PWR, so the performance of this type of nuclear power plant depends to a large extent on the trouble-free operation of SGs. Its degradation significantly affects the overall plant performance. Alloy 800NG (Incoloy® 800) is a nickel-iron-chromium alloy used for steam generator tubes in PWRs due to their high strength, good workability and resistance to corrosion. This behavior is attributed to the protective oxide film formed on the metal surface by contact with the high temperature pressurized water. However, chloride is one of major SG impurities that cause the breakdown of the passive film and initiate localized corrosion in passive metals as Alloy 800NG. The aim of this study is to provide information about the pitting corrosion behavior of the Incoloy® 800 steam generator tube under normal secondary circuit parameters (250 deg C and 5 MPa) and abnormal conditions of operation (presence of chloride ions in the secondary water). For this, optical microscopy, XRD and EDS analysis and electrochemical tests have been carried out under simulated PWR secondary water operating conditions. The susceptibility to pitting corrosion was evaluated using electrochemical tests and the oxide layer formed on material was examined by means of scanning electron microscopy (SEM) and energy dispersive X-ray spectrometer (EDS) analyses. (author)

  6. Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

    2015-07-01

    Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

  7. VENUS critical facility for study of MOX fuel lattices: VIPEX-PWR programme

    Energy Technology Data Exchange (ETDEWEB)

    Van der Meer, K.; Marloye, D.; D`hondt, P.; Ait Abderrahim, H.; Minsart, G. [Centre d`Etude de l`Energie Nucleaire, Mol (Belgium); Maldague, T.; Basselier, J. [Belgonucleaire SA, Brussels (Belgium)

    1996-12-31

    The VENUS critical facility is a water-moderated zero-power reactor. It consists of an open (non-pressurized) stainless-steel cylindrical vessel including a set of grids which maintain fuel rods in a vertical position. In the paper, the uncertainties of parameters, measured at the VENUS reactor, are discussed together with their implications to the benchmarking of the reactor codes. The current VIPEX-PWR programme aims at determining parameters for 17 by 17 MOX PWR fuel assemblies, mainly of interest for reactor operation, e.g. Beff AM effect, control rod worth, etc. The experimental methods used for the VIPEX-PWR programme are described in detail. With the VIPEX-PWR programme a new step is taken towards measuring important parameters and validating reactor codes for reactor operation purposes.

  8. Modeling local chemistry in PWR steam generator crevices

    Energy Technology Data Exchange (ETDEWEB)

    Millett, P.J. [EPRI, Palo Alto, CA (United States)

    1997-02-01

    Over the past two decades steam generator corrosion damage has been a major cost impact to PWR owners. Crevices and occluded regions create thermal-hydraulic conditions where aggressive impurities can become highly concentrated, promoting localized corrosion of the tubing and support structure materials. The type of corrosion varies depending on the local conditions, with stress corrosion cracking being the phenomenon of most current concern. A major goal of the EPRI research in this area has been to develop models of the concentration process and resulting crevice chemistry conditions. These models may then be used to predict crevice chemistry based on knowledge of bulk chemistry, thereby allowing the operator to control corrosion damage. Rigorous deterministic models have not yet been developed; however, empirical approaches have shown promise and are reflected in current versions of the industry-developed secondary water chemistry guidelines.

  9. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  10. Review of PWR-related thermal-shock studies

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.; Iskander, S.K.; Ball, D.G.

    1986-01-01

    Flaw behavior trends associated with pressurized-thermal-shock (PTS) loading of PWR pressure vessels have been under investigation at ORNL for approx.12 years. During that time, eight thermal-shock experiments with thick-walled steel cylinders were conducted as a part of the investigations. These experiments demonstrated, in good agreement with linear elastic fracture mechanics (LEFM), crack initiation and arrest, a series of initiation-arrest events with deep penetration of the wall, long crack jumps without significant dynamic effects at arrest, arrest in a rising K/sub I/ field, extensive surface extension of an initially short and shallow flaw, and warm prestressing with K/sub I/ equal to or less than 0. This information was used in the development of a fracture-mechanics model that is being used extensively in the evaluation of the PTS issue.

  11. PWR steam generator chemical cleaning, Phase I. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Rothstein, S.

    1978-07-01

    United Nuclear Industries (UNI) entered into a subcontract with Consolidated Edison Company of New York (Con Ed) on August 8, 1977, for the purpose of developing methods to chemically clean the secondary side tube to tube support crevices of the steam generators of Indian Point Nos. 1 and 2 PWR plants. This document represents the first reporting on activities performed for Phase I of this effort. Specifically, this report contains the results of a literature search performed by UNI for the purpose of determining state-of-the-art chemical solvents and methods for decontaminating nuclear reactor steam generators. The results of the search sought to accomplish two objectives: (1) identify solvents beyond those proposed at present by UNI and Con Ed for the test program, and (2) confirm the appropriateness of solvents and methods of decontamination currently in use by UNI.

  12. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  13. PWR and BWR spent fuel assembly gamma spectra measurements

    Energy Technology Data Exchange (ETDEWEB)

    Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2016-10-11

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  14. Advanced PFBC transient analysis

    Energy Technology Data Exchange (ETDEWEB)

    White, J.S. [Parsons Power Group, Inc., Reading, PA (United States); Bonk, D.L.; Rogers, L. [USDOE Morgantown Energy Technology Center, WV (United States)

    1996-12-31

    Transient modeling and analysis of Advanced Pressurized Fluidized Bed Combustion (PFBC) systems is a research area that is currently under investigative study by the United States Department of Energy`s Morgantown Energy Technology Center (METC). The object of the effort is to identify key operating parameters affecting plant performance and then quantify the basic response of major sub-systems to changes in operating conditions. PC-TRAX, a commercially available dynamic software program, was chosen and applied in this modeling and analysis effort. This paper summarizes and describes the development of a series of TRAX-based transient models of Advanced PFBC power plants. These power plants generate a high temperature flue gas by burning coal or other suitable fuel in a PFBC. The high temperature flue gas supports low-Btu fuel gas or natural gas combustion in a gas turbine topping combustor. When utilized, low-Btu fuel gas is produced in a bubbling bed carbonizer. High temperature, high pressure combustion products exiting the topping combustor are expanded in a modified gas turbine to generate electrical power. Waste heat from the system is used to generate and superheat steam for a reheat steam turbine bottoming cycle that generates additional electrical power. Basic control/instrumentation models were developed and modeled in PC-TRAX and used to investigate off-design plant performance. System performance for various transient conditions and control philosophies was studied.

  15. BEACON TSM application system to the operation of PWR reactors; Aplicacion del Sistema BEACON TSM a la operacion de reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lozano, J. A.; Mildrum, C.; Serrano, J. F.

    2012-07-01

    BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

  16. Fatigue-crack growth behavior of Type 347 stainless steels under simulated PWR water conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seokmin; Min, Ki-Deuk; Yoon, Ji-Hyun; Kim, Min-Chul; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Fatigue crack growth rate (FCGR) curve of stainless steel exists in ASME code section XI, but it is still not considering the environmental effects. The longer time nuclear power plant is operated, the more the environmental degradation issues of materials pop up. There are some researches on fatigue crack growth rate of S304 and S316, but researches of FCGR of S347 used in Korea nuclear power plant are insufficient. In this study, the FCGR of S347 stainless steel was evaluated in the PWR high temperature water conditions. The FCGRs of S347 stainless steel under pressurized-water conditions were measured by using compact-tension (CT) specimens at different levels of dissolved oxygen (DO) and frequency. 1. FCGRs of SS347 were slower than that in ASME XI and environmental effect did not occur when frequency was higher than 1Hz. 2. Fatigue crack growth is accelerated by corrosion fatigue and it is more severe when frequency is slower than 0.1Hz. 3. Increase of crack tip opening time increased corrosion fatigue and it deteriorated environmental fatigue properties.

  17. Study on nuclear physics of high burnup full MOX PWR core

    Energy Technology Data Exchange (ETDEWEB)

    Kugo, Teruhiko; Shimada, Syoichiro; Okubo, Tsutomu; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    As one of options for future light water reactors, we have been studying a new concept of a high burnup full MOX PWR core. We have proposed a core of 600 MWe to ensure discharged burnup of 100 GWd/t by increasing a moderator to fuel volume ratio to 2.6 with enlarged fuel pin pitch of 13.8 mm, and have investigated its feasibility in neutronics. A plutonium fissile content of 12% is needed for this core. A soluble boric acid with B-10 enrichment of 40% is able to control burnup reactivity without increasing a capacity of boron tanks. A control rod cluster with use of natural boron carbide (B{sub 4}C) per three fuel assemblies ensures a shutdown margin of more than 2%dk/kk`. A moderator temperature and void coefficients are negative through an operating cycle. Although the use of burnable poisons like Gd{sub 2}O{sub 3} and Er{sub 2}O{sub 3} is not necessary to reduce an excess reactivity, it can lower a radial power peaking factor by about 0.1. (author)

  18. The stress corrosion cracking behavior of alloys 690 and 152 WELD in a PWR environment.

    Energy Technology Data Exchange (ETDEWEB)

    Alexandreanu, B.; Chopra, O. K.; Shack, W. J. (Nuclear Engineering Division); ( EVS); ( ESE)

    2009-01-01

    Alloys 690 and 152 are the replacement materials of choice for Alloys 600 and 182, respectively. The latter two alloys are used as structural materials in pressurized water reactors (PWRs) and have been found to undergo stress corrosion cracking (SCC). The objective of this work is to determine the crack growth rates (CGRs) in a simulated PWR water environment for the replacement alloys. The study involved Alloy 690 cold-rolled by 26% and a laboratory-prepared Alloy 152 double-J weld in the as-welded condition. The experimental approach involved pre-cracking in a primary water environment and monitoring the cyclic CGRs to determine the optimum conditions for transitioning from the fatigue transgranular to intergranular SCC fracture mode. The cyclic CGRs of cold-rolled Alloy 690 showed significant environmental enhancement, while those for Alloy 152 were minimal. Both materials exhibited SCC of 10{sup -11} m/s under constant loading at moderate stress intensity factors. The paper also presents tensile property data for Alloy 690TT and Alloy 152 weld in the temperature range 25--870 C.

  19. Advanced PFBC transient analysis

    Energy Technology Data Exchange (ETDEWEB)

    White, J.S. [Parsons Power Group, Inc., Reading, PA (United States); Bonk, D.L. [USDOE Federal Energy Technology Center, Morgantown, WV (United States)

    1997-05-01

    Transient modeling and analysis of advanced Pressurized Fluidized Bed Combustion (PFBC) systems is a research area that is currently under investigation by the US Department of Energy`s Federal Energy Technology Center (FETC). The object of the effort is to identify key operating parameters that affect plant performance and then quantify the basic response of major sub-systems to changes in operating conditions. PC-TRAX{trademark}, a commercially available dynamic software program, was chosen and applied in this modeling and analysis effort. This paper describes the development of a series of TRAX-based transient models of advanced PFBC power plants. These power plants burn coal or other suitable fuel in a PFBC, and the high temperature flue gas supports low-Btu fuel gas or natural gas combustion in a gas turbine topping combustor. When it is utilized, the low-Btu fuel gas is produced in a bubbling bed carbonizer. High temperature, high pressure combustion products exiting the topping combustor are expanded in a modified gas turbine to generate electrical power. Waste heat from the system is used to raise and superheat steam for a reheat steam turbine bottoming cycle that generates additional electrical power. Basic control/instrumentation models were developed and modeled in PC-TRAX and used to investigate off-design plant performance. System performance for various transient conditions and control philosophies was studied.

  20. Penetrative Internal Oxidation from Alloy 690 Surfaces and Stress Corrosion Crack Walls during Exposure to PWR Primary Water

    Science.gov (United States)

    Olszta, Matthew J.; Schreiber, Daniel K.; Thomas, Larry E.; Bruemmer, Stephen M.

    Analytical electron microscopy and three-dimensional atom probe tomography (ATP) examinations of surface and near-surface oxidation have been performed on Ni-30%Cr alloy 690 materials after exposure to high-temperature, simulated PWR primary water. The oxidation nanostructures have been characterized at crack walls after stress-corrosion crack growth tests and at polished surfaces of unstressed specimens for the same alloys. Localized oxidation was discovered for both crack walls and surfaces as continuous filaments (typically dislocations were oxidized in non-deformed materials, while the oxidation path appeared to be along more complex dislocation substructures in heavily deformed materials. This paper will highlight the use of high resolution scanning and transmission electron microscopy in combination with APT to better elucidate the microstructure and microchemistry of the filamentary oxidation.

  1. A nonlinear analytic function expansion nodal method for transient calculations

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Han Gyn; Park, Sang Yoon; Cho, Byung Oh; Zee, Sung Quun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The nonlinear analytic function expansion nodal (AFEN) method is applied to the solution of the time-dependent neutron diffusion equation. Since the AFEN method requires both the particular solution and the homogeneous solution to the transient fixed source problem, the derivation of the solution method is focused on finding the particular solution efficiently. To avoid complicated particular solutions, the source distribution is approximated by quadratic polynomials and the transient source is constructed such that the error due to the quadratic approximation is minimized, In addition, this paper presents a new two-node solution scheme that is derived by imposing the constraint of current continuity at the interface corner points. The method is verified through a series of application to the NEACRP PWR rod ejection benchmark problems. 6 refs., 2 figs., 1 tab. (Author)

  2. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L. [Rempe and Associates, LLC, Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lutz, R. J. [Lutz Nuclear Safety Consultant, LLC, Asheville, NC (United States)

    2015-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

  3. Application of the BEACON-TSM system to the operation of PWR reactors; Aplicacion del sistema Beacon TSM a la operacion de reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lozano, J. A.; Mildrum, C.; Serrano, J. F.

    2011-07-01

    BEACON-TSM is an advanced system of the operation support of PWR reactors that combines the capabilities of an advanced nodal neutronic model and the measures of the instrumentation available in plant to determine, accurately and continuously, the distribution of power in the core and the available margins to the limits of the beak factors.

  4. Reference neutron transport calculation note for Korea nuclear power plants with 3-loop PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byung Cheol; Chang, Ki Oak

    1997-05-01

    Reactor pressure vessel (RPV) steels are subjected to neutron irradiation at a temperature of about 290 deg C. This radiation exposure alters the mechanical properties, leading to a shift of the brittle-to-ductile transition temperature toward higher temperatures and to a diminution of the rupture energy as determined by Charpy V-notch tests. This radiation embrittlement is one of the important aging factors of nuclear power plants. U.S. NRC recommended the basic requirements for the determination of the pressure vessel fluence by regulatory guide DG-1025 in order to reduce the uncertainty in the determination of neutron fluence calculation and measurements. The determination of the pressure vessel fluence is based on both calculations and measurements. The fluence prediction is made with a calculation and the measurements are used to qualify the calculational methodology. Because of the importance and the difficulty of these calculations, the method`s qualification by comparison to measurement must be made to ensure a reliable and accurate vessel fluence determination. This reference calculation note is to provide a series of forward and adjoint neutron transport calculations for use in the evaluation of neutron dosimetry from surveillance capsule irradiations at 3-loop PWR reactor as well as for use in the determination of the neutron exposure of the reactor vessel wall in accordance with U.S Regulatory Guide DG-1025 requirements. The calculations of the pressure vessel fluence consist of the following steps; (1) Determination of the geometrical and material input data, (2) Determination of the core neutron source, and (3) Propagation of the neutron fluence from the core to the vessel and into the cavity. (author). 12 tabs., 3 figs., 7 refs.

  5. Transient-Absorption Spectroscopy of Cis-Trans Isomerization of N,N-dimethyl-4,4'-Azodianiline with 3D-Printed Temperature-Controlled Sample Holder

    Science.gov (United States)

    Kosenkov, Dmytro; Shaw, James; Zuczek, Jennifer; Kholod, Yana

    2016-01-01

    The laboratory unit demonstrates a project based approach to teaching physical chemistry laboratory where upper-division undergraduates carry out a transient-absorption experiment investigating the kinetics of cis-trans isomerization of N,N-dimethyl-4,4'-azodianiline. Students participate in modification of a standard flash-photolysis spectrometer…

  6. Integral Test Facility PKL: Experimental PWR Accident Investigation

    Directory of Open Access Journals (Sweden)

    Klaus Umminger

    2012-01-01

    Full Text Available Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs arranged symmetrically around the reactor pressure vessel (RPV. The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

  7. Applicability of oxygenated water chemistry for PWR secondary systems

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H.P. [Studsvik Nuclear AB, Nykoeping (Sweden); Takiguchi, H.; Otoha, K. [Japan Atomic Power Co., Tokyo (Japan)

    2002-07-01

    Introduction of oxygenated water chemistry (OWC) in PWR secondary side is considered as a means to reduce the transportation of corrosion products into the steam generator and thus also minimizing crevice deposits and subsequent materials problems. One main concern, however, is the risk of inter-granular attack (IGA) in crevices. In order to study effects on crevice tube IGA by OWC, a series of experiments were performed in a steam generator (SG) simulating loop. This comprised a SG tube and a tube support plate (TSP) together forming the crevice. The over-all objective of the work accounted here was to demonstrate that it is possible to operate the steam generator secondary side with OWC without causing intolerable IGA or other types of attack on the tube in the crevice area. Tubes of sensitized Alloy 600 were exposed during a total of nine experiments in an autoclave using a TSP/tube arrangement with an asymmetric crevice design. Experiments were performed at high and low pH and potential under open and packed crevice conditions. The aggressiveness of the crevice environment was also further increased by addition of carbonate and chloride. Furthermore the tube was pressurized. Experimental parameters were monitored on the primary side as well as in the secondary bulk phase and in the crevice. (authors)

  8. Qualification tests for PWR control element drive mechanism

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Yong; Jin, Choon Eon; Choi Suhn [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    It is necessary to perform the qualification test for the magnetic jack type CEDM to show the design compatibility because the CEDM is composed of many mechanical and electrical components complicatedly. ABB-CE performed various qualification tests during the development of the System80 CEDM to which Korea Standard Nuclear Plant (KSNP) CEDM referred. The qualification test for the CEDM is classified into the performance test and the dynamic test. The performance test is to verify operability of the CEDM, and the dynamic test is to find dynamic characteristics and to verify the structural integrity if the CEDM for the seismic accidents. Described in this report are the test requirements, the test facilities and the test methods for the performance and the dynamic qualification tests of the PWR magnetic jack type CEDM. The impacts of the design changes in the Korea Next Generation Reactor (KNGR) on the KSNP CEDM were analyzed to present the necessity for the tests. This report also proposes the facilities to perform the tests in KAERI including reasonable schedule for the tests. Attached to this report is the summary of qualification tests of System 80 CEDM performed by ABB-CE. 20 figs., 16 tabs., 21 refs. (Author) .new.

  9. Mitsubishi PWR nuclear fuel with advanced design features

    Energy Technology Data Exchange (ETDEWEB)

    Kaua Goe, Toshiy Uki; Nuno kawa, Koi Chi [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)

    2008-10-15

    In the last few decades, the global warming has been a big issue. As the breakthrough in this crisis, advanced operations of the water reactor such as higher burn up, longer cycle, and up rating could be effective ways. From this viewpoint, Mitsubishi Heavy Industries (MHI) has developed the fuel for burn up extension, whose assembly burn-up limit is 55GWd/t(A), with the original and advanced designs such as corrosion resistant cladding material MDA, and supplied to Japanese PWR utilities. On the other hand, MHI intends to supply more advanced fuel assemblies not only to domestic market but to the global market. Actually MHI has submitted the application for standard design certification of USA . Advanced Pressurized Water Reactor on Jan. 2nd 2008. The fuel assembly for US APWR is 17x17 type with active fuel length of 14ft, characterized with three features, to {sup E}nhance Fuel Economy{sup ,} {sup E}nable Flexible Core Operation{sup ,} and to {sup I}mprove Reliability{sup .} MHI has also been conducting development activities for more advanced products, such as 70GWd/t(A) burn up limit fuel with cladding, guide thimble and spacer grid made from M-MDATM alloy that is new material with higher corrosion resistance, such as 12ft and 14ft active length fuel, such as fuel with countermeasure against grid fretting, debris fretting, and IRI. MHI will present its activities and advanced designs.

  10. Reactor transient

    Energy Technology Data Exchange (ETDEWEB)

    Menegus, R.L.

    1956-05-31

    The authors are planning a calculation to be done on the Univac at the Louviers Building to estimate the effect of xenon transients, a high reactor power. This memorandum outlines the reasons why they prefer to do the work at Louviers rather than at another location, such as N.Y.U. They are to calculate the response of the reactor to a sudden change in position of the half rods. Qualitatively, the response will be a change in the rooftop ratio of the neutron flux. The rooftop ratio may oscillate with high damping, or, instead, it may oscillate for many cycles. It has not been possible for them to determine this response by hand calculation because of the complexity of the problem, and yet it is important for them to be certain that high power operation will not lead us to inherently unstable operation. Therefore they have resorted to machine computation. The system of differential equations that describes the response has seven dependent variables; therefore there are seven equations, each coupled with one or more of the others. The authors have discussed the problem with R.R. Haefner at the plant, and it is his opinion that the IBM 650 cannot adequately handle the system of seven equations because the characteristic time constants vary over a range of about 10{sup 8}. The Univac located at the Louviers Building is said to be satisfactory for this computation.

  11. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

    2017-06-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  12. Experiment data report for Semiscale Mod-1 tests S-05-6 and S-05-7 (alternate ECC injection tests). [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Feldman, E. M.; Sackett, K. E.

    1977-06-01

    Recorded test data are presented for Tests S-05-6 and S-05-7 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Tests S-05-6 and S-05-7 were conducted from initial conditions of 2263 psia and 550/sup 0/F and 2253 psia and 551/sup 0/F, respectively, to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. The specific objective for these tests was to investigate the effectiveness of low pressure injection system (LPIS) coolant injection into the upper plenum when combined with cold leg intact loop injection.

  13. Trap behaviours characterization of AlGaN/GaN high electron mobility transistors by room-temperature transient capacitance measurement

    Directory of Open Access Journals (Sweden)

    Bin Dong

    2016-09-01

    Full Text Available In this paper, the trap behaviours in AlGaN/GaN high electron mobility transistors (HEMTs are investigated using transient capacitance measurement. By measuring the transient gate capacitance variance (ΔC with different pulse height, the gate pulse induced trap behaviours in SiNX gate dielectric layer or at the SiNX/AlGaN interface is revealed. Based on the results, a model on electron traps in AlGaN/GaN HEMTs is proposed. The threshold voltage (Vth instability in AlGaN/GaN HEMTs is believed to be correlated with the presence of these traps in SiNX gate dielectric layer or at the SiNX/AlGaN interface. Furthermore, trap density before and after step-stress applied on drain electrode is quantitatively analyzed based on ΔC measurement.

  14. Transient tachypnea - newborn

    Science.gov (United States)

    TTN; Wet lungs - newborns; Retained fetal lung fluid; Transient RDS; Prolonged transition; Neonatal - transient tachypnea ... Newborns with transient tachypnea have breathing problems soon after birth, most often within 1 to 2 hours. ...

  15. Conceptual study of advanced PWR systems. A study of passive and inherent safety design concepts for advanced light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; No, Hee Cheon; Baek, Won Pil; Shim Young Jae; Lee, Goung Jin; Na, Man Gyun; Lee, Jae Young; Kim, Han Gon; Kang, Ki Sig; Moon, Sang Ki; Kim, Yun Il; Park, Jae Wook; Yang, Soo Hyung; Kim, Soo Hyung; Lee, Seong Wook; Kim, Hong Che; Park, Hyun Sik; Jeong, Ji Hwan; Lee, Sang Il; Jung, Hae Yong; Kim, Hyong Tae; Chae, Kyung Sun; Moon, Ki Hoon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1995-08-01

    The five thermal-hydraulic concepts chosen for advanced PWR have been studied as follows: (1) Critical Heat Flux: Review of previous works, analysis of parametric trends, analysis of transient CHF characteristics, extension of the CHF date bank, survey and assessment of correlations, design of a intermediate-pressure CHF test loop have been performed. (2) Passive Cooling Concepts for Concrete Containment system: Review of condensation phenomena with noncondensable gases, selection of a promising concept (i.e., use of external condensers), design of test loop according to scaling laws have been accomplished. and computer programs based on the control-volume approach, and the conceptual design of test loop have been accomplished. (4) Fluidic Diode Concepts: Review of previous applications of the concept, analysis major parameters affecting the performance, development of a computational code, and conceptual investigation of the verification test loop have been performed. (5) Wet Thermal Insulator: Review of previous works, selection of promising methods ( i.e. ceramic fiber in a steel case and mirror-type insulator), and conceptual design of the experimental loop have been performed. (author). 9 refs.

  16. A study on the direct use of spent PWR fuel in CANDU reactors -Fuel management and safety analysis-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Boh Wook; Choi, Hang Bok; Lee, Yung Wook; Cho, Jae Sun; Huh, Chang Wook [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The reference DUPIC fuel composition was determined based on the reactor safety, thermal-hydraulics, economics, and refabrication aspects. The center pin of the reference DUPIC fuel bundle is poisoned with natural dysprosium. The worst LOCA analysis has shown that the transient power and heat deposition of the reference DUPIC core are the same as those of natural uranium CANDU core. The intra-code comparison has shown that the accuracy of DUPIC physics code system is comparable to the current CANDU core design code system. The sensitivity studies were performed for the refuelling schemes of DUPIC core and the 2-bundle shift refuelling scheme was selected as the standard refuelling scheme of the DUPIC core. The application of 4-bundle shift refuelling scheme will be studied in parallel as the auto-refuelling method is improved and the reference core parameters of the heterogeneous DUPIC core are defined. The heterogeneity effect was analyzed in a preliminary fashion using 33 fuel types and the random loading strategy. The refuelling simulation has shown that the DUPIC core satisfies the current CANDU 6 operating limits of channel and bundle power regardless of the fuel composition heterogeneity. The 33 fuel types used in the heterogeneity analysis was determined based on the initial enrichment and discharge burnup of the PWR fuel. 90 figs, 62 tabs, 63 refs. (Author).

  17. Development the Technique for Laboratory-Degraded Alloy 600 Steam Generator tubing and Electrochemical Impedance Spectroscopy of Oxide Film on at PWR environment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Tae Hyun; Nam, Hyo On; Bahn, Chi Bum; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    2006-07-01

    For the integrity management of SG tubes, nondestructive- evaluation performed using eddy-current test (ECT) is necessary in the assessment. The reliability of ECT evaluation is dependent on the accuracy of ECT for various kinds of defects. For basic calibration and qualification of these techniques, cracked SG tube specimens having mechanical and micro-structural characteristics of intergranular cracks in the field, are needed. To produce libraries of Laboratory-Degraded- SG-tubes (LDT) with intergranular cracks, a radial denting method and direct tension method were explored for generating ID and OD axial cracks and circumferential crack by three-dimensional finite element- analysis and experimental demonstration, respectively. The technique is proven to be applicable for generating axial cracks with long and shallow geometry as opposed to the semi-circular cracks typically obtained by the internal-pressurization method. In addition, a DCPD method with array probes was developed for accurate monitoring and controlling of crack size and shape. By these methods, long and shallow intergranular axial and circumferential cracks more typical of actual degraded SG tubes were successfully produced. As a final step, the developed Laboratory-Degraded Alloy 600 Steam Generator tubing specimens were exposed to typical PWR water with 1,200 ppm boron and 2 ppm lithium at a pressure of 15 MPa and temperatures of 320 degree C. Dissolved hydrogen concentration was varied from 0 to 30 cm{sup 3} (STP)/kg to study its effect on the surface oxide film composition of alloy 600 in PWR water environment. As an optional experiment, Alloy 600 SG tubing specimens were exposed to PWR secondary water chemistry environment to understanding the properties of oxide films formed on alloy surface in PWR secondary environment. This study is focused on the on-line monitoring method on crack generation using DCPD method and the properties of oxide films formed on Alloy 600 in high temperature

  18. Rapid formation of Ni3Sn4 joints for die attachment of SiC-based high temperature power devices using ultrasound-induced transient liquid phase bonding process.

    Science.gov (United States)

    Li, Z L; Dong, H J; Song, X G; Zhao, H Y; Feng, J C; Liu, J H; Tian, H; Wang, S J

    2017-05-01

    High melting point Ni3Sn4 joints for the die attachment of SiC-based high temperature power devices was successfully achieved using an ultrasound-induced transient liquid phase (TLP) bonding process within a remarkably short bonding time of 8s. The formed intermetallic joints, which are completely composed of the refined equiaxial Ni3Sn4 grains with the average diameter of 2μm, perform the average shear strength of 26.7MPa. The sonochemical effects of ultrasonic waves dominate the mechanism and kinetics of the rapid formation of Ni3Sn4 joints. Copyright © 2016 Elsevier B.V. All rights reserved.

  19. A study of high-temperature heat pipes with multiple heat sources and sinks. I - Experimental methodology and frozen startup profiles. II - Analysis of continuum transient and steady-state experimental data with numerical predictions

    Science.gov (United States)

    Faghri, A.; Cao, Y.; Buchko, M.

    1991-01-01

    Experimental profiles for heat pipe startup from the frozen state were obtained, using a high-temperature sodium/stainless steel pipe with multiple heat sources and sinks to investigate the startup behavior of the heat pipe for various heat loads and input locations, with both low and high heat rejection rates at the condensor. The experimental results of the performance characteristics for the continuum transient and steady-state operation of the heat pipe were analyzed, and the performance limits for operation with varying heat fluxes and location are determined.

  20. Planning of operational maneuvers with the 3-D PWR core dynamics SIMTRAN-online code

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J.M.; Ahnert, C.; Cano, D.; Garcia-Herranz, N. [Universidad Politecnica de Madrid (Spain)

    1996-09-01

    In this work we discuss the modelling capabilities developed in our 3-D PWR core dynamics SIMTRAN online code for the fast and accurate calculations required in the planning of optimal operational maneuvers; the validation results by comparison with actual operating data; and the systematic analysis of optimal maneuvers, along a typical PWR cycle, with the Constant Axial Offset Control (CAOC) technical specification; concluding with some relevant recommendations for the online planning of maneuvers and to relax the CAOC technical specification at low power. (author)

  1. A Comparison of Simple Methods to Incorporate Material Temperature Dependency in the Green’s Function Method for Estimating Transient Thermal Stresses in Thick-Walled Power Plant Components

    Science.gov (United States)

    Rouse, James; Hyde, Christopher

    2016-01-01

    The threat of thermal fatigue is an increasing concern for thermal power plant operators due to the increasing tendency to adopt “two-shifting” operating procedures. Thermal plants are likely to remain part of the energy portfolio for the foreseeable future and are under societal pressures to generate in a highly flexible and efficient manner. The Green’s function method offers a flexible approach to determine reference elastic solutions for transient thermal stress problems. In order to simplify integration, it is often assumed that Green’s functions (derived from finite element unit temperature step solutions) are temperature independent (this is not the case due to the temperature dependency of material parameters). The present work offers a simple method to approximate a material’s temperature dependency using multiple reference unit solutions and an interpolation procedure. Thermal stress histories are predicted and compared for realistic temperature cycles using distinct techniques. The proposed interpolation method generally performs as well as (if not better) than the optimum single Green’s function or the previously-suggested weighting function technique (particularly for large temperature increments). Coefficients of determination are typically above 0.96, and peak stress differences between true and predicted datasets are always less than 10 MPa. PMID:28787827

  2. Current interruption transients calculation

    CERN Document Server

    Peelo, David F

    2014-01-01

    Provides an original, detailed and practical description of current interruption transients, origins, and the circuits involved, and how they can be calculated Current Interruption Transients Calculationis a comprehensive resource for the understanding, calculation and analysis of the transient recovery voltages (TRVs) and related re-ignition or re-striking transients associated with fault current interruption and the switching of inductive and capacitive load currents in circuits. This book provides an original, detailed and practical description of current interruption transients, origins,

  3. Numerical methods for solving the governing equations for a seriated continuum. [BWR and PWR

    Energy Technology Data Exchange (ETDEWEB)

    Narum, R.E.; Noble, C.; Mortensen, G.A.; McFadden, J.H.

    1976-09-01

    A desire to more accurately predict the behavior of transient two-phase flows has resulted in an investigation of the feasibility of computing unequal phase velocities and unequal phase temperatures. The finite difference forms of a set of equations governing a seriated continuum are presented along with two methods developed for solving the resulting systems of simultaneous nonlinear equations. Results from a one-dimensional computer code are presented to illustrate the capabilities of one of the solution methods.

  4. PWR Fuel Deposit Analysis at a B&W Plant with a 24 Month Fuel Cycle

    Science.gov (United States)

    Pop, Mike G.; Lamanna, Larry S.; Harne, Richard; Riddle, John

    The paper presents the crud analysis of a twice-burned fuel deposit using the AREVA crud sampling method in a domestic B&W PWR. The patented AREVA sampling method allowed the separation of deposit flakes that retain all the characteristics of unperturbed crud deposition.

  5. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based

  6. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  7. Neutron Collar Evolution and Fresh PWR Assembly Measurements with a New Fast Neutron Passive Collar

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, Howard Olsen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Geist, William H. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Root, Margaret A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rael, Carlos D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Belian, Anthony P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-11-02

    The passive neutron collar approach removes the effect of poison rods when using a 1mm Gd liner. This project sets out to solve the following challenges: BWR fuel assemblies have less mass and less neutron multiplication than PWR; and effective removal of cosmic ray spallation neutron bursts needed via QC tests.

  8. Determination of the activation energy for SCC crack growth for Alloy 182 weld in a PWR environment

    Energy Technology Data Exchange (ETDEWEB)

    Alexandreanu, B.; Chopra, O.K.; Shack, W.J. [Argonne National Lab., Nuclear Engineering Div., Argonne, Illinois (United States)

    2007-07-01

    The objective of this work was to determine the activation energy for stress corrosion cracking growth rates in a simulated PWR water environment for Alloy 182 weld metals. For this purpose, the crack growth rates (CGRs) of two heats of Alloy 182 were measured as a function of temperature between 290{sup o}C and 350{sup o}C. The difference in electrochemical potential between the specimen and the Ni/NiO line was maintained constant at each temperature by adjusting the hydrogen overpressure on the water supply tank. The CGR data as a function of temperature yielded activation energies of 252 kJ/mol for a double-J weld and 189 kJ/mol for a deep-groove weld. The data reported here and those in the literature suggest that the average activation energy for Alloy 182 welds is on the order of 220-230 kJ/mol, higher than the 130 kJ/mol commonly used for Alloy 600. The consequences of using a larger value of activation energy for SCC CGR data analysis are discussed. (author)

  9. Crack growth rates of nickel alloy welds in a PWR environment.

    Energy Technology Data Exchange (ETDEWEB)

    Alexandreanu, B.; Chopra, O. K.; Shack, W. J.; Energy Technology

    2006-05-31

    In light water reactors (LWRs), vessel internal components made of nickel-base alloys are susceptible to environmentally assisted cracking. A better understanding of the causes and mechanisms of this cracking may permit less conservative estimates of damage accumulation and requirements on inspection intervals. A program is being conducted at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated LWR coolant environments. This report presents crack growth rate (CGR) results for Alloy 182 shielded-metal-arc weld metal in a simulated pressurized water reactor (PWR) environment at 320 C. Crack growth tests were conducted on 1-T compact tension specimens with different weld orientations from both double-J and deep-groove welds. The results indicate little or no environmental enhancement of fatigue CGRs of Alloy 182 weld metal in the PWR environment. The CGRs of Alloy 182 in the PWR environment are a factor of {approx}5 higher than those of Alloy 600 in air under the same loading conditions. The stress corrosion cracking for the Alloy 182 weld is close to the average behavior of Alloy 600 in the PWR environment. The weld orientation was found to have a profound effect on the magnitude of crack growth: cracking was found to propagate faster along the dendrites than across them. The existing CGR data for Ni-alloy weld metals have been compiled and evaluated to establish the effects of key material, loading, and environmental parameters on CGRs in PWR environments. The results from the present study are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of the specific Ni-alloy weld to environmentally enhanced cracking.

  10. Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, William [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-09-01

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in September 2015 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2015 report we presented a baseline mechanistic finite element model of a two-loop pressurized water reactor (PWR) for systemlevel heat transfer analysis and subsequent thermal-mechanical stress analysis and fatigue life estimation under reactor thermal-mechanical cycles. In the present report, we provide tensile and fatigue test data for 508 low-alloy steel (LAS) base metal, 508 LAS heat-affected zone metal in 508 LAS–316 stainless steel (SS) dissimilar metal welds, and 316 SS-316 SS similar metal welds. The test was conducted under different conditions such as in air at room temperature, in air at 300 oC, and under PWR primary loop water conditions. Data are provided on materials properties related to time-independent tensile tests and time-dependent cyclic tests, such as elastic modulus, elastic and offset strain yield limit stress, and linear and nonlinear kinematic hardening model parameters. The overall objective of this report is to provide guidance to estimate tensile/fatigue hardening parameters from test data. Also, the material models and parameters reported here can directly be used in commercially available finite element codes for fatigue and ratcheting evaluation of reactor components under in-air and PWR water conditions.

  11. Modeling of PWR fuel at extended burnup; Estudo de modelos para o comportamento a altas queimas de varetas combustiveis de reatores a agua leve pressurizada

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Raphael Mejias

    2016-11-01

    This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

  12. Transient heat transfer behavior of water spray evaporative cooling on a stainless steel cylinder with structured surface for safety design application in high temperature scenario

    Science.gov (United States)

    Aamir, Muhammad; Liao, Qiang; Hong, Wang; Xun, Zhu; Song, Sihong; Sajid, Muhammad

    2017-02-01

    High heat transfer performance of spray cooling on structured surface might be an additional measure to increase the safety of an installation against any threat caused by rapid increase in the temperature. The purpose of present experimental study is to explore heat transfer performance of structured surface under different spray conditions and surface temperatures. Two cylindrical stainless steel samples were used, one with pyramid pins structured surface and other with smooth surface. Surface heat flux of 3.60, 3.46, 3.93 and 4.91 MW/m2 are estimated for sample initial average temperature of 600, 700, 800 and 900 °C, respectively for an inlet pressure of 1.0 MPa. A maximum cooling rate of 507 °C/s was estimated for an inlet pressure of 0.7 MPa at 900 °C for structured surface while for smooth surface maximum cooling rate of 356 °C/s was attained at 1.0 MPa for 700 °C. Structured surface performed better to exchange heat during spray cooling at initial sample temperature of 900 °C with a relative increase in surface heat flux by factor of 1.9, 1.56, 1.66 and 1.74 relative to smooth surface, for inlet pressure of 0.4, 0.7, 1.0 and 1.3 MPa, respectively. For smooth surface, a decreasing trend in estimated heat flux is observed, when initial sample temperature was increased from 600 to 900 °C. Temperature-based function specification method was utilized to estimate surface heat flux and surface temperature. Limited published work is available about the application of structured surface spray cooling techniques for safety of stainless steel structures at very high temperature scenario such as nuclear safety vessel and liquid natural gas storage tanks.

  13. New methodology for the analysis of the quality controls of ENUSA on PWR components; Nueva metodologia para el analisis de los controles de calidad de ENUSA sobre los componentes PWR

    Energy Technology Data Exchange (ETDEWEB)

    Munoz, A.; Navas, I. de; Prieto, M.

    2012-07-01

    For the manufacture of PWR fuel assemblies, ENUSA receives the components of Westinghouse, who ensures its quality. However, ENUSA carried out on these components various quality controls that increase reliability and give added value.

  14. Simulation model and methodology for calculating the damage by internal radiation in a PWR reactor; Modelo de simulacion y metodologia para el calculo del dano por irradiacion en los internos de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barreira Pereira, P.

    2012-07-01

    This study involves the development of the methodology and three-dimensional models to estimate the damage to the vessel internals of a commercial PWR reactor from irradiation history of operating cycles.

  15. DOSE RATES FOR WESTINGHOUSE 17X17 MOX PWR SNF IN A WASTE PACKAGE (SCPB: N/A)

    Energy Technology Data Exchange (ETDEWEB)

    T.L. Lotz

    1997-01-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to estimate the dose rate on and near the surface a Multi-Purpose Canister (MPC) PWR waste package (WP) which is loaded with Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel. The 21 PWR MPC WP is used to provide an upper bound for waste package designs since the 12 PWR MPC WP will have a smaller source term and an equivalent amount of shielding. the objectives of this evaluation are to calculate the requested dose rate(s) and document the calculation in a fashion to allow comparisons to other waste forms and WP designs at a future time.

  16. Transient charging and discharging current study in pure PVF and ...

    Indian Academy of Sciences (India)

    Keywords. PVF; PVDF; polyblend; fluoropolymers; transient current. Abstract. The transient current were analysed by considering the effect of variation of forming time, temperature, field and composition of blend specimens. Measurements indicated that transient charging and discharging currents exhibited thermally ...

  17. Transient Global Amnesia

    Science.gov (United States)

    ... sudden memory loss. Request an Appointment at Mayo Clinic Causes The underlying cause of transient global amnesia is unknown. There appears to be a link between transient global amnesia and a history of migraines, though the underlying factors that contribute ...

  18. Crack growth rates and metallographic examinations of Alloy 600 and Alloy 82/182 from field components and laboratory materials tested in PWR environments.

    Energy Technology Data Exchange (ETDEWEB)

    Alexandreanu, B.; Chopra, O. K.; Shack, W. J.

    2008-05-05

    In light water reactors, components made of nickel-base alloys are susceptible to environmentally assisted cracking. This report summarizes the crack growth rate results and related metallography for field and laboratory-procured Alloy 600 and its weld alloys tested in pressurized water reactor (PWR) environments. The report also presents crack growth rate (CGR) results for a shielded-metal-arc weld of Alloy 182 in a simulated PWR environment as a function of temperature between 290 C and 350 C. These data were used to determine the activation energy for crack growth in Alloy 182 welds. The tests were performed by measuring the changes in the stress corrosion CGR as the temperatures were varied during the test. The difference in electrochemical potential between the specimen and the Ni/NiO line was maintained constant at each temperature by adjusting the hydrogen overpressure on the water supply tank. The CGR data as a function of temperature yielded activation energies of 252 kJ/mol for a double-J weld and 189 kJ/mol for a deep-groove weld. These values are in good agreement with the data reported in the literature. The data reported here and those in the literature suggest that the average activation energy for Alloy 182 welds is on the order of 220-230 kJ/mol, higher than the 130 kJ/mol commonly used for Alloy 600. The consequences of using a larger value of activation energy for SCC CGR data analysis are discussed.

  19. Transient drainage summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    This report summarizes the history of transient drainage issues on the Uranium Mill Tailings Remedial Action (UMTRA) Project. It defines and describes the UMTRA Project disposal cell transient drainage process and chronicles UMTRA Project treatment of the transient drainage phenomenon. Section 4.0 includes a conceptual cross section of each UMTRA Project disposal site and summarizes design and construction information, the ground water protection strategy, and the potential for transient drainage.

  20. Features of chlorophyll fluorescence transients can be used to investigate low temperature induced effects on photosystem II of algal lichens from polar regions

    Czech Academy of Sciences Publication Activity Database

    Mishra, Anamika; Hajek, J.; Tuháčková, T.; Barták, M.; Mishra, Kumud

    2015-01-01

    Roč. 5, č. 1 (2015), s. 99-111 ISSN 1805-0689 R&D Projects: GA MŠk(CZ) LO1415 Institutional support: RVO:67179843 Keywords : Rhizoplaca melanophtalma * Umbilicaria antarctica * Xanthoria elegans * temperature stress Subject RIV: EH - Ecology, Behaviour

  1. PSH Transient Simulation Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Muljadi, Eduard [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-12-21

    PSH Transient Simulation Modeling presentation from the WPTO FY14 - FY16 Peer Review. Transient effects are an important consideration when designing a PSH system, yet numerical techniques for hydraulic transient analysis still need improvements for adjustable-speed (AS) reversible pump-turbine applications.

  2. Transient Ischemic Attack

    Medline Plus

    Full Text Available Transient Ischemic Attack TIA , or transient ischemic attack, is a "mini stroke" that occurs when a blood clot blocks an artery for a short time. The only ... TIA is that with TIA the blockage is transient (temporary). TIA symptoms occur rapidly and last a ...

  3. Feasibility Study on Thimble Plug Removal for Westinghouse Type PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang Sup; Lee, Jae Yong; Yoon, Duk Joo; Jun, Hwang Yong; Kim, Yoon Ho [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1996-12-31

    1. Abstract of Thimble Plug Removal- Thimble Plug Removal from the core increase core bypass flow few percent and may reduce DNBR Margin 2{approx}3%. In this feasibility study, the following analyses were performed in terms of the best estimate flow, bypass flow, DNBR margin etc. 2. Area of analysis and evaluations (a). Thermal Hydraulic (b). PCWG (c). Nuclear Design (d). Rod Performance (e). mechanical Design (f). Transient Analysis (g). LOCA Analysis. 3. Evaluation of Economic and Licensing 4. Detail analysis and design were performed for Youngkwang unit 1 as a sample plant. (author). 68 refs., figs.

  4. An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Massoud, M

    1987-01-01

    Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients.

  5. An efficient Neuro-Fuzzy approach to nuclear power plant transient identification

    Energy Technology Data Exchange (ETDEWEB)

    Gomes da Costa, Rafael [Instituto de Engenharia Nuclear - CNEN, Programa de Pos-Graduacao em Ciencia e Tecnologia Nucleares, Via Cinco, s/no, Cidade Universitaria, Rua Helio de Almeida, 75, Postal Box 68550, Zip Code 21941-906 Rio de Janeiro (Brazil); Abreu Mol, Antonio Carlos de, E-mail: mol@ien.gov.br [Instituto de Engenharia Nuclear - CNEN, Programa de Pos-Graduacao em Ciencia e Tecnologia Nucleares, Via Cinco, s/no, Cidade Universitaria, Rua Helio de Almeida, 75, Postal Box 68550, Zip Code 21941-906 Rio de Janeiro (Brazil); Instituto Nacional de C and T de Reatores Nucleares Inovadores (Brazil); Carvalho, Paulo Victor R. de, E-mail: paulov@ien.gov.br [Instituto de Engenharia Nuclear - CNEN, Programa de Pos-Graduacao em Ciencia e Tecnologia Nucleares, Via Cinco, s/no, Cidade Universitaria, Rua Helio de Almeida, 75, Postal Box 68550, Zip Code 21941-906 Rio de Janeiro (Brazil); Lapa, Celso Marcelo Franklin, E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear - CNEN, Programa de Pos-Graduacao em Ciencia e Tecnologia Nucleares, Via Cinco, s/no, Cidade Universitaria, Rua Helio de Almeida, 75, Postal Box 68550, Zip Code 21941-906 Rio de Janeiro (Brazil); Instituto Nacional de C and T de Reatores Nucleares Inovadores (Brazil)

    2011-06-15

    Highlights: > We investigate a Neuro-Fuzzy modeling tool use for able transient identification. > The prelusive transient type identification is done by an artificial neural network. > After, the fuzzy-logic system analyzes the results emitting reliability degree of it. > The research support was made in a PWR simulator at the Brazilian Nuclear Engineering Institute. > The results show the potential to help operators' decisions in a nuclear power plant. - Abstract: Transient identification in nuclear power plants (NPP) is often a computational very hard task and may involve a great amount of human cognition. The early identification of unexpected departures from steady state behavior is an essential step for the operation, control and accident management in NPPs. The bases for the transient identification relay on the evidence that different system faults and anomalies lead to different pattern evolution in the involved process variables. During an abnormal event, the operator must monitor a great amount of information from the instruments that represents a specific type of event. Recently, several works have been developed for transient identification. These works frequently present a non reliable response, using the 'don't know' as the system output. In this work, we investigate the possibility of using a Neuro-Fuzzy modeling tool for efficient transient identification, aiming to helping the operator crew to take decisions relative to the procedure to be followed in situations of accidents/transients at NPPs. The proposed system uses artificial neural networks (ANN) as first level transient diagnostic. After the ANN has done the preliminary transient type identification, a fuzzy-logic system analyzes the results emitting reliability degree of it. A validation of this identification system was made at the three loops Pressurized Water Reactor (PWR) simulator of the Human-System Interface Laboratory (LABIHS) of the Nuclear Engineering Institute

  6. Performance evaluation of passive containment cooling system of an advanced PWR using coupled RELAP5/GOTHIC simulation

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Zheng, E-mail: hzheng@kth.se; Ma, Weimin, E-mail: weimin@kth.se

    2016-12-15

    Highlights: • A coupled model was developed to simulate the simultaneous response of PCS and the containment during LOCA. • The designed PCS can effectively mitigate the accident for a long term. • The loop height has a significant effect on the performance and working mode. • A feed-and-bleed operation for the cooling tank is proposed to further mitigate the accident. • A simplified analytical model for the feed-and-bleed operation is developed. - Abstract: Motivated to investigate the thermal–hydraulic characteristics and performance of a passive containment cooling system (PCS) for a Generation III pressurized water reactor (PWR), a coupled RELAP5/GOTHIC model was developed, which was then employed to simultaneously simulate the transient responses of the PCS and the containment during a large break loss of coolant accident of the reactor. The results show that the PCS is capable of lowering the containment pressure to an acceptable level for a long period (up to 3 days). In a separate-effect study, it was found that the height of the PCS loop plays an important role in determining the flow characteristics and heat removal performance of the PCS. Within the range of the considered loop heights, phase change occurs in the riser of the loop after the height exceeds a specific value (between 13 m and 15 m), below which only single-phase flow takes place. With increasing height of the loop, the heat removal capability increases monotonically at first; however, it is no longer sensitive to the height after two-phase flow appears. Finally, a feed-and-bleed operation for the cooling tank of the PCS was proposed as an enhancement measure of the heat removal capacity, and the simulation results show it further mitigates the accident. Moreover, a simplified analytical model is developed to predict the impact of the feed-and-bleed flowrate on the PCS performance, which can be used in engineering design.

  7. Temperature dependence of internal motions of protein side-chain NH3(+) groups: insight into energy barriers for transient breakage of hydrogen bonds.

    Science.gov (United States)

    Zandarashvili, Levani; Iwahara, Junji

    2015-01-20

    Although charged side chains play important roles in protein function, their dynamic properties are not well understood. Nuclear magnetic resonance methods for investigating the dynamics of lysine side-chain NH3(+) groups were established recently. Using this methodology, we have studied the temperature dependence of the internal motions of the lysine side-chain NH3(+) groups that form ion pairs with DNA phosphate groups in the HoxD9 homeodomain-DNA complex. For these NH3(+) groups, we determined order parameters and correlation times for bond rotations and reorientations at 15, 22, 28, and 35 °C. The order parameters were found to be virtually constant in this temperature range. In contrast, the bond-rotation correlation times of the NH3(+) groups were found to depend strongly on temperature. On the basis of transition state theory, the energy barriers for NH3(+) rotations were analyzed and compared to those for CH3 rotations. Enthalpies of activation for NH3(+) rotations were found to be significantly higher than those for CH3 rotations, which can be attributed to the requirement of hydrogen bond breakage. However, entropies of activation substantially reduce the overall free energies of activation for NH3(+) rotations to a level comparable to those for CH3 rotations. This entropic reduction in energy barriers may accelerate molecular processes requiring hydrogen bond breakage and play a kinetically important role in protein function.

  8. Late Noachian Icy Highlands climate model: Exploring the possibility of transient melting and fluvial/lacustrine activity through peak annual and seasonal temperatures

    Science.gov (United States)

    Palumbo, Ashley M.; Head, James W.; Wordsworth, Robin D.

    2018-01-01

    The nature of the Late Noachian climate of Mars remains one of the outstanding questions in the study of the evolution of martian geology and climate. Despite abundant evidence for flowing water (valley networks and open/closed basin lakes), climate models have had difficulties reproducing mean annual surface temperatures (MAT) > 273 K in order to generate the ;warm and wet; climate conditions presumed to be necessary to explain the observed fluvial and lacustrine features. Here, we consider a ;cold and icy; climate scenario, characterized by MAT ∼225 K and snow and ice distributed in the southern highlands, and ask: Does the formation of the fluvial and lacustrine features require continuous ;warm and wet; conditions, or could seasonal temperature variation in a ;cold and icy; climate produce sufficient summertime ice melting and surface runoff to account for the observed features? To address this question, we employ the 3D Laboratoire de Météorologie Dynamique global climate model (LMD GCM) for early Mars and (1) analyze peak annual temperature (PAT) maps to determine where on Mars temperatures exceed freezing in the summer season, (2) produce temperature time series at three valley network systems and compare the duration of the time during which temperatures exceed freezing with seasonal temperature variations in the Antarctic McMurdo Dry Valleys (MDV) where similar fluvial and lacustrine features are observed, and (3) perform a positive-degree-day analysis to determine the annual volume of meltwater produced through this mechanism, estimate the necessary duration that this process must repeat to produce sufficient meltwater for valley network formation, and estimate whether runoff rates predicted by this mechanism are comparable to those required to form the observed geomorphology of the valley networks. When considering an ambient CO2 atmosphere, characterized by MAT ∼225 K, we find that: (1) PAT can exceed the melting point of water (>273 K) in

  9. Preliminary assessment of a combined passive safety system for typical 3-loop PWR CPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Zijiang; Shan, Jianqiang, E-mail: jqshan@mail.xjtu.edu.cn; Gou, Junli

    2017-03-15

    Highlights: • A combined passive safety system was placed on a typical 3-loop PWR CPR1000. • Three accident analyses show the three different accident mitigation methods of the passive safety system. • The three mitigation methods were proved to be useful. - Abstract: As the development of the nuclear industry, passive technology turns out to be a remarkable characteristic of advanced nuclear power plants. Since the 20th century, much effort has been given to the passive technology, and a number of evolutionary passive systems have developed. Thoughts have been given to upgrade the existing reactors with passive systems to meet stricter safety demands. In this paper, the CPR1000 plant, which is one kind of mature pressurized water reactor plants in China, is improved with some passive systems to enhance safety. The passive systems selected are as follows: (1) the reactor makeup tank (RMT); (2) the advanced accumulator (A-ACC); (3) the in-containment refueling water storage tank (IRWST); (4) the passive emergency feed water system (PEFS), which is installed on the secondary side of SGs; (5) the passive depressurization system (PDS). Although these passive components is based on the passive technology of some advanced reactors, their structural and trip designs are adjusted specifically so that it could be able to mitigate accidents of the CPR1000. Utilizing the RELAP5/MOD3.3 code, accident analyses (small break loss of coolant accident, large break loss of coolant accident, main feed water line break accident) of this improved CPR1000 plant were presented to demonstrate three different accident mitigation methods of the safety system and to test whether the passive safety system preformed its function well. In the SBLOCA, all components of the passive safety system were put into work sequentially, which prevented the core uncover. The LBLOCA analysis illustrates the contribution of the A-ACCs whose small-flow-rate injection can control the maximum cladding

  10. Technical improvement and maintainability upgrade of PWR instrumentation and control systems

    Energy Technology Data Exchange (ETDEWEB)

    Takashima, Makoto; Saitou, Minoru [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Nuclear Energy Systems Engineering Center; Nishimura, Yukitaka

    1995-05-01

    The digital control technology has been applied step by step to the recent PWR instrumentation and control (I and C) systems, and also it is planned to apply it to the all the I and C systems, including the safety grade systems related to the reactor safety for the Advanced PWR plant. The integrated digital I and C systems have also been updated to enhance the maintenance capabilities of the systems, to optimize maintenance activities and reduce work loads of maintenance personnel. As a result, the integrated digital I and C systems have been established which can be applied to the actual plants, and which can contribute to the enhancement of reliability, hardware reductions and maintenance workload reduction of about 20% in comparison with recent plants. (author).

  11. Eddy current NDT: a suitable tool to measure oxide layer thickness in PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Alencar, Donizete A.; Silva Junior, Silverio F. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)], e-mail: daa@cdtn.br, e-mail: silvasf@cdtn.br; Vieira, Andre L.P.S. [Industrias Nucleares do Brasil (INB S.A.), Resende, RJ (Brazil). Fabrica de Combustivel Nuclear], e-mail: andre@inb.gov.br; Soares, Adolpho [Technotest Consultoria e Acessoria Ltda., Belo Horizonte, MG (Brazil)], e-mail: adolpho@technotest.com.br

    2009-07-01

    Eddy current is a nondestructive test (NDT) widely used in industry to support integrity analysis of components and equipment. In the nuclear area it is frequently applied to inspect tubes installed in tube exchangers, such as steam generators and condensers in PWR plants, as well as turbine blades. Adequately assisted by means of robotic devices, that inspection method has been pointed as a suitable tool to perform accurate oxide layer thickness measurements in PWR fuel rods. This paper shows some theoretical aspects and physical operating principles of the inspection method, as well as test probes construction details, and the calibration reference standards fabrication processes. Furthermore, some data, experimentally obtained at INB laboratories and other technical information obtained from TECNATOM S.A. are presented, showing the accuracy and efficacy of such NDT method. (author)

  12. PENGARUH KONDISI ATMOSFERIK TERHADAP PERHITUNGAN PROBABILISTIK DAMPAK RADIOLOGI KECELAKAAN PWR 1000-MWe

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-10-01

    Full Text Available ABSTRAK PENGARUH KONDISI ATMOSFERIK TERHADAP PERHITUNGAN PROBABILISTIK DAMPAK RADIOLOGI KECELAKAAN PWR 1000-MWe.  Perhitungan dampak kecelakaan radiologi terhadap lepasan produk fisi akibat kecelakaan potensial yang mungkin terjadi di Pressurized Water Reactor (PWR diperlukan secara probabilistik. Mengingat kondisi atmosfer sangat berperan terhadap dispersi radionuklida di lingkungan, dalam penelitian ini akan dianalisis pengaruh kondisi atmosferik terhadap perhitungan probabilistik dari konsekuensi kecelakaan reaktor.  Tujuan penelitian adalah melakukan analisis terhadap pengaruh kondisi atmosfer berdasarkan model data input meteorologi terhadap dampak radiologi kecelakaan PWR 1000-MWe yang disimulasikan pada tapak yang mempunyai kondisi meteorologi yang berbeda. Simulasi menggunakan program PC-Cosyma dengan moda perhitungan probabilistik, dengan data input meteorologi yang dieksekusi secara cyclic dan stratified, dan disimulasikan di Tapak Semenanjung Muria dan Pesisir Serang. Data meteorologi diambil setiap jam untuk jangka waktu satu tahun. Hasil perhitungan menunjukkan bahwa frekuensi kumulatif  untuk model input yang sama untuk Tapak pesisir Serang lebih tinggi dibandingkan dengan Semenanjung Muria. Untuk tapak yang sama, frekuensi kumulatif model input cyclic lebih tinggi dibandingkan model stratified. Model cyclic memberikan keleluasan dalam menentukan tingkat ketelitian perhitungan dan tidak membutuhkan data acuan dibandingkan dengan model stratified. Penggunaan model cyclic dan stratified melibatkan jumlah data yang besar dan pengulangan perhitungan  akan meningkatkan  ketelitian nilai-nilai statistika perhitungan. Kata kunci: dampak kecelakaan, PWR 1000-MWe,  probabilistik,  atmosferik, PC-Cosyma   ABSTRACT THE INFLUENCE OF ATMOSPHERIC CONDITIONS TO PROBABILISTIC CALCULATION OF IMPACT OF RADIOLOGY ACCIDENT ON PWR-1000MWe. The calculation of the radiological impact of the fission products releases due to potential accidents

  13. Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

    Energy Technology Data Exchange (ETDEWEB)

    Debarberis, L. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: luigi.debarberis@cec.eu.int; Acosta, B. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: beatriz.acosta-iborra@jrc.nl; Zeman, A. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Sevini, F. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Ballesteros, A. [Tecnatom, Avd. Montes de Oca 1, San Sebasitan de los Reyes, E-28709 Madrid (Spain); Kryukov, A. [Russian Research Centre Kurchatov Institute, Kurchatov Square 1, 123182 Moscow (Russian Federation); Gillemot, F. [AEKI Atomic Research Institute, Konkoly Thege M. ut 29-33, 1121 Budapest (Hungary); Brumovsky, M. [NRI, Nuclear Research Institute, Husinec-Rez 130, 25068 Rez (Czech Republic)

    2006-04-15

    It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.

  14. Primary water chemistry improvement for radiation exposure reduction at Japanese PWR Plants

    Energy Technology Data Exchange (ETDEWEB)

    Nishizawa, Eiichi [Omiya Technical Institute, Saitama-ken (Japan)

    1995-03-01

    Radiation exposure during the refueling outages at Japanese Pressurized Water Reactor (PWR) Plants has been gradually decreased through continuous efforts keeping the radiation dose rates at relatively low level. The improvement of primary water chemistry in respect to reduction of the radiation sources appears as one of the most important contributions to the achieved results and can be classified by the plant operation conditions as follows

  15. Probabilistic analysis of PWR and BWR fuel rod performance using the code CASINO-SLEUTH

    Energy Technology Data Exchange (ETDEWEB)

    Bull, A.J.

    1987-05-01

    This paper presents a brief description of the Monte Carlo and response surface techniques used in the code, and a probabilistic analysis of fuel rod performance in PWR and BWR applications. The analysis shows that fission gas release predictions are very sensitive to changes in certain of the code's inputs, identifies the most dominant input parameters and compares their effects in the two cases.

  16. Chemical and radiochemical specifications - PWR power plants; Specifications chimiques et radiochimiques - Centrales REP

    Energy Technology Data Exchange (ETDEWEB)

    Stutzmann, A. [Electricite de France (EDF), 93 - Saint-Denis (France)

    1997-07-01

    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  17. EDF/CIDEN - ONECTRA: PWR decontamination; EDF/CIDEN - ONECTRA: assainissement REP

    Energy Technology Data Exchange (ETDEWEB)

    Fayolle, P. [EDFICIDEN, 35-37, rue Louis Guerin - B.P. 21212, 69611 Villeurbanne Cedex (France); Orcel, H. [ONECTRA, ZA les Tomples BP45, 26701 Pierrelatte Cedex (France); Wertz, L. [ONECTRA, Le Britannia, Allee C, 20 Bd Eugene Deruelle, 69432 Lyon Cedex 03 (France)

    2010-07-01

    In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination

  18. Effect of transplutonium doping on approach to long-life core in uranium-fueled PWR

    Energy Technology Data Exchange (ETDEWEB)

    Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Shmelev, Anatolii [Moscow Engineering Physics Institute, Moscow (Russian Federation)

    2002-08-01

    The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them {sup 241}Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides {sup 242}Cm and {sup 244}Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile {sup 239}Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of {sup 241}Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100 GWd/tHM with about 20% {sup 238}Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles. (author)

  19. Study on stress corrosion of the zone affected by the AISI 316L steel heat under PWR reactor environment at 325 deg Celsius; Estudo da corrosao sob tensao da zona afetada pelo calor do aco AISI 316L em ambiente de reator PWR a 325 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Satler Filho, Luiz F.; Schvartzman, Monica M.A.M.; Quinan, Marco A.D.; Soares, Antonio E.G., E-mail: aegs@cdtn.b, E-mail: fernandosatler@yahoo.com.b, E-mail: quinanm@cdtn.b, E-mail: monicas@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Lima, Luciana I.L., E-mail: lill@cdtn.b [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2009-07-01

    This paper evaluates the stress corrosion susceptibility of the HAZ (heat affected zone) of the AISI 316L stainless steel of a dissimilar welding done between the ASTM A-508 steel and the AISI 316L steel, using a nickel alloy, under a chemical environment similar to the PWR (Pressurized Water Reactor) nuclear reactor primary circuit. The nickel 82 and 182 alloys were used in the GTAW (Gas Tungsten Arc Welding) and SMAW (Shielded Metal Arc Welding) processes respectively. The test at slow deformation - SSRT (Slow Strain Rate Test) was applied, using a deformation rate of 3x10{sup -7} s{sup -1}, at a temperature of 325 degree Celsius and pressure of 12.5 MPa. The susceptibility under tress corrosion evaluation was performed comparing the resistance limit, the total deformation and the fracture time obtained at the inert medium (nitrogen) and at the PWR medium. Also, the fracture surfaces were observed under a scanning electron microscope, verifying the fragile fracture regions

  20. Inactivation of the gene katA or sodA affects the transient entry into the viable but non-culturable response of Staphylococcus aureus in natural seawater at low temperature.

    Science.gov (United States)

    Masmoudi, Salma; Denis, Michel; Maalej, Sami

    2010-12-01

    We have investigated the fate of Staphylococcus aureus by starving the cells and maintaining them in natural seawater at 22 and 4 °C. At 22 °C, cells developed a long-term survival state where about 0.037% of the initial population remained culturable over more than 7 months, whereas at 4 °C, bacteria lost culturability and transiently entered into the viable but non-culturable state (VBNC). However, after 22 days of entry into the VBNC state, the number of viable cells detected via the direct viable count method decreased significantly. We show here that mutational inactivation of catalase (KatA) or superoxide dismutase (SodA) rendered strains hypersensitive to seawater stress at 4 °C and consequently, part of the seawater lethality on S. aureus at low temperature is mediated by reactive oxygen species (ROS) during microcosm-survival process. Shifting the temperature from 4 to 22 °C of totally non-culturable wild-type cells induced a partial recovery of the population. However, deficiencies in catalase or superoxide dismutase prevent resuscitation ability. Copyright © 2010 Elsevier Ltd. All rights reserved.

  1. Dynamic modeling of primary and secondary systems of IRIS reactor for transient analysis using SIMULINK

    Energy Technology Data Exchange (ETDEWEB)

    Magalhaes, Mardson Alencar de Sa; Lira, Carlos Alberto Brayner de Oliveira; Silva, Mario Augusto Bezerra da, E-mail: cabol@ufpe.b [Universidade Federal de Pernambuco (DEN/UFPE), Recife, PE (Brazil). Dept. de Energia Nuclear; Lima, Fernando Roberto de Andrade, E-mail: falima@cnen.gov.b [Centro Regional de Ciencias Nucleares (CRCN-NE/CNEN-PE), Recife, PE (Brazil)

    2011-07-01

    The IRIS project has significantly advanced in the last few years in response to a demand for a new generation reactor, that could fulfill the essential requirements for a future nuclear power plant: better economics, safety-by-design, low proliferation risk and environmental sustainability. IRIS reactor is a integral type PWR in which all primary components are arranged inside the pressure vessel. This configuration involves important changes in relation to a conventional PWR. These changes require several studies to comply with the safe operational limits for the reactor. In this paper, a study has been conducted to develop a dynamic model (named MODIRIS) for transient analysis, implemented in the MATLAB'S software SIMULINK, allowing the analysis of IRIS behavior by considering the neutron point kinetics for power production. The methodology is based on generating a set of differential equations of neutronic and thermal-hydraulic balances which describes the dynamics of the primary circuit, as well as a set of differential equations describing the dynamics of secondary circuit. The equations and initialization parameters at full power were into the SIMULINK and the code was validated by the confrontation with RELAP simulations for a transient of feedwater reduction in the steam generators. (author)

  2. Transient diagnosis system using quantum-inspired computing and Minkowski distance

    Energy Technology Data Exchange (ETDEWEB)

    Nicolau, Andressa dos Santos; Schirru, Roberto, E-mail: andressa@lmp.ufrj.b, E-mail: schirru@lmp.ufrj.b [Federal University of Rio de Janeiro (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Nuclear Engineering Program

    2011-07-01

    This paper proposes a diagnosis system model for identification of transient in a PWR nuclear power plant, optimized by the Quantum Inspired Evolutionary Algorithm - QEA in order to help nuclear power plant operator reduce his cognitive load and increase his available time to maintain the plant operating in a safe condition. This method was developed in order to be able to recognize the normal condition and three accidents of the design basis list of the nuclear power plant Angra 2, postulated in the Final Safety Analysis Report (FSAR). This System compares the similarly distance between the set of variables of the anomalous event, in a given time t, and the centroids of the design-basis transient variables. The lower similarly distance indicates the class of the transient to which the anomalous event belongs. The QEA was then used to find the best position of the centroids of each class of the selected transients. Such positions maximize the number of the correct classifications. Unlike the diagnosis system proposed in the literature, Minkowski distance was employed to calculate the similarity distance. The signatures of four transients were submitted to 1% and 2% of noise, and tested with prototype vector found by QEA. The results showed that the present transient diagnostic system was successfully implemented in the nuclear accident identification problem and was compatible with the techniques presented in the literature. (author)

  3. HEMERA: a 3D coupled core-plant system for accidental reactor transient simulation

    Energy Technology Data Exchange (ETDEWEB)

    Bruna, G.B.; Fouquet, F.; Dubois, F. [Institut de Radioprotection et de Surete Nucleaire, 92 - Fontenay aux Roses (France); Le Pallec, J.C.; Richebois, E.; Hourcade, E.; Poinot-Salanon, C.; Royer, E. [CEA Saclay, Dir. de l' Energie Nucleaire (DEN/DM2S), 91 - Gif sur Yvette (France)

    2007-07-01

    In the framework of their collaboration to develop a system to study reactor transients in safety-representative conditions, IRSN (Radioprotection and Nuclear Safety Institute) and Cea have launched the development of a fully coupled 3-dimensional computational chain, called HEMERA (Highly Evolutionary Methods for Extensive Reactor Analyses), based on the French SAPHYR code system, composed of APOLLO-2, CRONOS-2 and FLICA-4 codes, and the system code CATHARE. It includes cross sections generation, steady-state, depletion and transient computation capabilities in a consistent approach. Multi-level and multi-dimensional models are developed to account for neutronics, core thermal-hydraulics, fuel thermal analysis and system thermohydraulics. Currently Control Rod Ejection (RIA) and Main Steam Line Break (MSLB) accidents are investigated. The HEMERA system is presently applied to French PWR. The present paper outlines the main physical phenomena to be accounted for in such a coupled computational chain with significant time and space effects.

  4. Validation Code Trailer COBRA-TF/PARCSv2.7 before a Transient Boron Injection; Validacion del Codigo Acoplado COBRA-TF/PARCSv2.7 ante un Transitorio de Inyeccion de Boro

    Energy Technology Data Exchange (ETDEWEB)

    Abarca, A.; Barrachina, T.; Miro, R.; Verdu, G.

    2013-07-01

    The object of this work is to validate the attached code COBRA-TF/PARCS in a scenario of asymmetric injection of boron in a PWR reactor of three loops. This is intended to check the newly developed model of transport of boron code COBRA-TF and the routines of interpolation of effective section based on the concentration of boron implemented Parcs v2.7, demonstrating its proper functioning and its validity for the analysis of transient in nuclear safety.

  5. Optimization of thermal efficiency of nuclear central power like as PWR; Otimizacao da eficiencia termica de uma usina nuclear do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lapa, Nelbia da Silva

    2005-10-15

    The main purpose of this work is the definition of operational conditions for the steam and power conservation of Pressurized Water Reactor (PWR) plant in order to increase its system thermal efficiency without changing any component, based on the optimization of operational parameters of the plant. The thermal efficiency is calculated by a thermal balance program, based on conservation equations for homogeneous modeling. The circuit coefficients are estimated by an optimization tool, allowing a more realistic thermal balance for the plans under analysis, as well as others parameters necessary to some component models. With the operational parameter optimization, it is possible to get a level of thermal efficiency that increase capital gain, due to a better relationship between the electricity production and the amount of fuel used, without any need to change components plant. (author)

  6. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Martinez, R. M.; Fernandez, K.; Morato, D. J.; Bocanegra Melian, R.; Mena, L.; Queral, C.

    2014-07-01

    During the development of a severe accident in a PWR reactor can be generated, large quantities of hydrogen by the oxidation of metals present in the nucleus, mainly the zirconium fuel pods. This hydrogen, along with steam and other gases, can be released to the atmosphere of contention by a leak or break in the primary circuit and achieving conditions in which combustion may occur. Combustion causes thermal and pressure loads that can damage the security systems and the integrity of the containment building, last barrier of confinement of radioactive materials. The main condition that defines the characteristics of the combustion is the concentration of species, detailed knowledge of the distribution of hydrogen is very important to correctly predict the possible damage to the containment in the event that there is combustion. (Author)

  7. Fission product transport analysis: task 2. Quarterly progress report, April--June 1976. [PWR; spent fuel casks

    Energy Technology Data Exchange (ETDEWEB)

    Gieseke, J.A.; Baybutt, P.; Denning, R.S.; Jordan, H.; Wooton, R.O.

    1976-09-07

    Analysis methods and computer models for predicting fission product transport and deposition under postulated loss of coolant accident conditions within water-cooled reactor primary systems and shipping casks are being developed. Existing thermal-hydraulic analyses (RELAP-EM) form the basis for defining the conditions within the primary systems and the transport flow from which fission products deposit. The analyses to be developed in this study will provide methods for more realistically calculating the rate and magnitude of fission product release to the containment. Efforts during the reporting period were directed toward postulating the types of conditions to be assumed in developing the model for failed shipping casks, completing the PWR transport model format design, beginning the PWR transport model assembly and checkout, and performing various tasks related to fission product deposition rates, fission product chemistry, and thermal-hydraulic condition specification in support of the PWR model development.

  8. On the performance of an artificial bee colony optimization algorithm applied to the accident diagnosis in a PWR nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Iona Maghali S. de; Schirru, Roberto; Medeiros, Jose A.C.C., E-mail: maghali@lmp.ufrj.b, E-mail: schirru@lmp.ufrj.b, E-mail: canedo@lmp.ufrj.b [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear

    2009-07-01

    The swarm-based algorithm described in this paper is a new search algorithm capable of locating good solutions efficiently and within a reasonable running time. The work presents a population-based search algorithm that mimics the food foraging behavior of honey bee swarms and can be regarded as belonging to the category of intelligent optimization tools. In its basic version, the algorithm performs a kind of random search combined with neighborhood search and can be used for solving multi-dimensional numeric problems. Following a description of the algorithm, this paper presents a new event classification system based exclusively on the ability of the algorithm to find the best centroid positions that correctly identifies an accident in a PWR nuclear power plant, thus maximizing the number of correct classification of transients. The simulation results show that the performance of the proposed algorithm is comparable to other population-based algorithms when applied to the same problem, with the advantage of employing fewer control parameters. (author)

  9. Image-Processing-Based Study of the Interfacial Behavior of the Countercurrent Gas-Liquid Two-Phase Flow in a Hot Leg of a PWR

    Directory of Open Access Journals (Sweden)

    Gustavo A. Montoya

    2012-01-01

    Full Text Available The interfacial behavior during countercurrent two-phase flow of air-water and steam-water in a model of a PWR hot leg was studied quantitatively using digital image processing of a subsequent recorded video images of the experimental series obtained from the TOPFLOW facility, Helmholtz-Zentrum Dresden-Rossendorf e.V. (HZDR, Dresden, Germany. The developed image processing technique provides the transient data of water level inside the hot leg channel up to flooding condition. In this technique, the filters such as median and Gaussian were used to eliminate the drops and the bubbles from the interface and the wall of the test section. A Statistical treatment (average, standard deviation, and probability distribution function (PDF of the obtained water level data was carried out also to identify the flow behaviors. The obtained data are characterized by a high resolution in space and time, which makes them suitable for the development and validation of CFD-grade closure models, for example, for two-fluid model. This information is essential also for the development of mechanistic modeling on the relating phenomenon. It was clarified that the local water level at the crest of the hydraulic jump is strongly affected by the liquid properties.

  10. PENGEMBANGAN MODEL UNTUK SIMULASI KESELAMATAN REAKTOR PWR 1000 MWe GENERASI III+ MENGGUNAKAN PROGRAM KOMPUTER RELAP5

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-04-01

    Full Text Available Reaktor daya PWR AP1000 yang didesain oleh Westinghouse adalah reaktor Generasi III+ pertama yang telah menerima persetujuan desain dari U.S. Nuclear Regulatory Commission (NRC. Saat ini utilitas China telah memulai pembangunan beberapa unit AP1000 di dua tapak terpilih untuk rencana operasi pada 2013-2015. AP1000 sebagai desain PWR berdasarkan teknologi teruji dari desain PWR lainnya yang dibuat oleh Westinghouse dengan penguatan pada sistem keselamatan pasif dengan demikian dapat dipertimbangkan untuk dibangun di Indonesia bila mengacu pada persyaratan pada PP 43/2006 mengenai Perijinan Reaktor Nuklir. Namun demikian, desain tersebut perlu diverifikasi oleh Technical Support Organization (TSO independen sebelum dapat dibangun di Indonesia. Verifikasi dapat dilakukan menggunakan paket program RELAP5 dalam bentuk analisis kecelakaan. Selama ini analisis kecelakaan PLTN dilakukan untuk tipe PWR 1000 MWe dari generasi II atau tipe konvensional. Mengingat saat ini referensi yang menggambarkan teknologi AP1000 yang menyertakan teknologi keselamatan pasif sudah tersedia maka dilakukan kegiatan pemodelan yang nantinya dapat digunakan untuk melakukan analisis kecelakaan. Metode pengembangan model mengacu pada pedoman IAEA yang terdiri dari pengumpulan data instalasi, pengembangan engineering data dan penyusunan input deck, verifikasi dan validasi data input. Model yang berhasil dikembangkan secara umum telah mewakili sistem AP1000 secara keseluruhan dan dianggap sebagai model dasar. Model tersebut telah diverifikasi dan divalidasi dengan data desain yang terdapat pada referensi dimana respon parameter termohidraulika menunjukkan perbedaan hasil ± 3% selain untuk parameter penurunan tekanan teras yang lebih rendah 13%. Sebagai model dasar, input deck yang diperoleh dapat dikembangkan lebih lanjut dengan mengintegrasikan pemodelan sistem keselamatan, sistem proteksi, dan sistem kendali yang spesifik AP1000 untuk keperluan simulasi keselamatan yang lebih

  11. Transient simulation of radiating flows

    Energy Technology Data Exchange (ETDEWEB)

    Selcuk, Nevin [Department of Chemical Engineering, Middle East Technical University, Inonu Bulvari, 06531 Ankara (Turkey)]. E-mail: selcuk@metu.edu.tr; Bilge Uygur, A. [Department of Chemical Engineering, Middle East Technical University, Inonu Bulvari, 06531 Ankara (Turkey); Ayranci, Isil [Department of Chemical Engineering, Middle East Technical University, Inonu Bulvari, 06531 Ankara (Turkey); Tarhan, Tanil [Department of Chemical Engineering, Middle East Technical University, Inonu Bulvari, 06531 Ankara (Turkey)

    2005-06-15

    Time-dependent Navier-Stokes equations are solved in conjunction with the radiative transfer equation by coupling a previously developed direct numerical simulation-based computational fluid dynamics code to an existing radiation code, both based on the method of lines approach. The temperature profiles predicted by the coupled code are validated against steady-state solutions available in the literature for laminar, axisymmetric, hydrodynamically developed flow of a gray, absorbing, emitting fluid in a heated pipe. Favorable comparisons show the predictive accuracy and reliability of the coupling strategy employed. Transient solutions for a more realistic heat transfer problem are also demonstrated for simultaneous hydrodynamic and thermal development.

  12. Comparison between HELIOS calculations and a PWR cell benchmark for actinides transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Guzman, Rafael [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico); Francois, Juan-Luis [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico)]. E-mail: jlfl@fi-b.unam.mx

    2007-01-15

    This paper shows a comparison between the results obtained with the HELIOS code and other similar codes used in the international community, with respect to the transmutation of actinides. To do this, the international benchmark: 'Calculations of Different Transmutation Concepts' of the Nuclear Energy Agency is analyzed. In this benchmark, two types of cells are analyzed: a small cell corresponding to a standard pressurized water reactor (PWR), and a wide cell corresponding to a highly moderated PWR. Two types of discharge burnup are considered: 33 GWd/tHM and 50 GWd/tHM. The following results are analyzed: the neutron multiplication factor as a function of burnup, the atomic density of the principal actinide isotopes, the radioactivity of selected actinides at reactor shutdown and cooling times from 7 until 50,000 years, the void reactivity and the Doppler reactivity. The results are compared with the following codes: KAPROS/KARBUS (FZK, Germany), SRAC95 (JAERI, Japan), TRIFON (ITTEP, Russian Federation) and WIMS (IPPE, Russian Federation). For the neutron multiplication factor, the results obtained with HELIOS show a difference of around 1% {delta}k/k. For the isotopic concentrations: {sup 241}Pu, {sup 242}Pu, and {sup 242m}Am, the results of all the institutions present a difference that increases at higher burnup; for the case of {sup 237}Np, the results of FZK diverges from the other results as the burnup increases. Regarding the activity, the difference of the results is acceptable, except for the case of {sup 241}Pu. For the Doppler coefficient, the results are acceptable, except for the cells with high moderation. In the case of the void coefficient, the difference of the results increases at higher void fractions, being the highest at 95%. In summary, for the PWR benchmark, the results obtained with HELIOS agree reasonably well within the limits of the multiple plutonium recycling established by the NEA working party on plutonium fuels and

  13. PWR circuit contamination assessment tool. Use of OSCAR code for engineering studies at EDF

    Directory of Open Access Journals (Sweden)

    Benfarah Moez

    2016-01-01

    Full Text Available Normal operation of PWR generates corrosion and wear products in the primary circuit which are activated in the core and constitute the major source of the radiation field. In addition, cases of fuel failure and alpha emitter dissemination in the coolant system could represent a significant radiological risk. Radiation field and alpha risks are the main constraints to carry out maintenance and to handle effluents. To minimize these risks and constraints, it is essential to understand the behavior of corrosion products and actinides and to carry out the appropriate measurements in PWR circuits and loop experiments. As a matter of fact, it is more than necessary to develop and use a reactor contamination assessment code in order to take into account the chemical and physical mechanisms in different situations in operating reactors or at design stage. OSCAR code has actually been developed and used for this aim. It is presented in this paper, as well as its use in the engineering studies at EDF. To begin with, the code structure is described, including the physical, chemical and transport phenomena considered for the simulation of the mechanisms regarding PWR contamination. Then, the use of OSCAR is illustrated with two examples from our engineering studies. The first example of OSCAR engineering studies is linked to the behavior of the activated corrosion products. The selected example carefully explores the impact of the restart conditions following a reactor mid-cycle shutdown on circuit contamination. The second example of OSCAR use concerns fission products and disseminated fissile material behavior in the primary coolant. This example is a parametric study of the correlation between the quantity of disseminated fuel and the variation of Iodine 134 in the primary coolant.

  14. Effect of co-free valve on activity reduction in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bahn, C.B.; Han, B.C.; Bum, J.S.; Hwang, I.S. [Department of Nuclear Engineering, Seoul National Univ. (Korea, Republic of); Lee, C.B. [Korea Atomic Energy Research Inst., Daejon (Korea, Republic of)

    2002-07-01

    Radioactive nuclei, such as {sup 68}Co and {sup 60}Co, deposited on out-of-core surfaces in a pressurized water reactor (PWR) primary coolant system, are major sources of occupational radiation exposure to plant maintenance personnel and act as costly impediment to prompt and effective repairs. Valve hardfacing alloys exposed to primary coolant are considered as one of the main Co sources. To evaluate the Co-free valve, such as NOREM 02 and Deloro 50, the candidates for the alternative to Stellite 6, in a simulated PWR primary condition, SNU corrosion test loop (SCOTL) was constructed. For gate valves hard-faced with made of NOREM 02 and Deloro 50 hot cycling tests were conducted for up to 2,000 on-off cycles with cold leak tests at 1,000 cycle interval. It was observed that the leak rate of NOREM 02 (Fe-base) did not satisfy the nuclear grade valve leak criteria. After 1000 cycles test, while there was no leakage in case of Deloro 50 (Ni-base). Also, Deloro 50 showed no leakage after 2000 cycles. To estimate the activity reduction effect, we modified CRUDSIM-MIT which modeled the effects of coolant chemistry on the crud transport and activity buildup in the primary system of PWR. In the new code, crud evaluation and assessment (CREAT), {sup 60}Co activity buildup prediction includes 1) Co-base valve replacement effect, 2) Co-base valve maintenance effect, and 3) control rod drive mechanism (CRDM) and main coolant pump (MCP) shaft contribution. CREAT predicted that the main contributor of Co activity buildup was the corrosion-induced release of Co from the steam generator (SG) tubing. With new SG's tubed with alloy 690, Korean Next Generation Reactor (APR-1400) is expected to have about 64% lower Co activity on SG surface. The use of all Co-free valves is expected to cut additional 8% of activity which is only marginal. (authors)

  15. Study on core physics characteristics of high burn-up full MOX PWR core. 2

    Energy Technology Data Exchange (ETDEWEB)

    Kugo, Teruhiko; Okubo, Tsutomu; Shimada, Syoichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-09-01

    As one of options for future light water reactors, we have been studying a new concept of a high burn-up full MOX PWR core with a discharge burn-up of 100 GWd/t and a 3-year operation cycle being based on the existing light water reactor technology. We have already confirmed the feasibility of the core, in which a moderator to fuel volume ratio(Vm/Vf) is increased to 2.6 with the same fuel pin diameter of 9.5 mm as in the current PWR but with the enlarged fuel pin pitch of 13.8 mm. In this report, to improve the neutronics and thermal hydraulic performance of the high burn-up core, we subsequently propose a 600 MWe core ensuring discharge burn-up of 100 GWd/t by increasing Vm/Vf to 3.0 with the same fuel pin pitch of 12.6 mm as in the current PWR and the smaller fuel rod diameter of 8.3 mm instead of 9.5 mm. We have investigated its core characteristics in neutronics and confirmed its feasibility. The core neutronics performance is compared between Vm/Vf = 2.6 and 3.0. From the comparison, it is found that the proposed core with Vm/Vf 3.0 has more promising characteristics than with Vm/Vf = 2.6 such as saving of a fissile plutonium content of 0.3wt%, improvement in a departure from nucleate boiling ratio (DNBR) and so on, except for a shortened cycle length by 9%. In addition, we have investigated a low-leakage refueling scheme for both types of high burn-up cores. Without modification to fuel material such as addition of burnable poison and/or transuranium isotopes, it can not be expected to improve the burn-up efficiency by the low-leakage refueling scheme. (author)

  16. [Transient epileptic amnesia].

    Science.gov (United States)

    Muramatsu, Kazuhiro; Yoshizaki, Takahito

    2016-03-01

    Transient amnesia is one of common clinical phenomenon of epilepsy that are encountered by physicians. The amnestic attacks are often associated with persistent memory disturbances. Epilepsy is common among the elderly, with amnesia as a common symptom and convulsions relatively uncommon. Therefore, amnesia due to epilepsy can easily be misdiagnosed as dementia. The term 'transient epileptic amnesia (TEA)' was introduced in the early 1990s by Kapur, who highlighted that amnestic attacks caused by epilepsy can be similar to those occurring in 'transient global amnesia', but are distinguished by features brevity and recurrence. In 1998, Zeman et al. proposed diagnostic criteria for TEA.

  17. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    Energy Technology Data Exchange (ETDEWEB)

    Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  18. NUCLEAR DATA UNCERTAINTY PROPAGATION FOR A TYPICAL PWR FUEL ASSEMBLY WITH BURNUP

    Directory of Open Access Journals (Sweden)

    D. ROCHMAN

    2014-06-01

    Full Text Available The effects of nuclear data uncertainties are studied on a typical PWR fuel assembly model in the framework of the OECD Nuclear Energy Agency UAM (Uncertainty Analysis in Modeling expert working group. The “Fast Total Monte Carlo” method is applied on a model for the Monte Carlo transport and burnup code SERPENT. Uncertainties on k∞, reaction rates, two-group cross sections, inventory and local pin power density during burnup are obtained, due to transport cross sections for the actinides and fission products, fission yields and thermal scattering data.

  19. Cavern/Vault Disposal Concepts and Thermal Calculations for Direct Disposal of 37-PWR Size DPCs

    Energy Technology Data Exchange (ETDEWEB)

    Hardin, Ernest [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Clayton, Daniel James [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-03-01

    This report provides two sets of calculations not presented in previous reports on the technical feasibility of spent nuclear fuel (SNF) disposal directly in dual-purpose canisters (DPCs): 1) thermal calculations for reference disposal concepts using larger 37-PWR size DPC-based waste packages, and 2) analysis and thermal calculations for underground vault-type storage and eventual disposal of DPCs. The reader is referred to the earlier reports (Hardin et al. 2011, 2012, 2013; Hardin and Voegele 2013) for contextual information on DPC direct disposal alternatives.

  20. Analysis of high moderation PWR MOX core MISTRAL-4 with SRAC and MVP

    Energy Technology Data Exchange (ETDEWEB)

    Tatsumi, Masahiro [Nuclear Fuel Industries, Ltd., Fuel Engineering and Development Department, Kumatori, Osaka (Japan); Kan, Taro [Mitsubishi Heavy Industries, Ltd., Nuclear Energy Systems Engineering Center, Yokohama, Kanagawa (Japan); Ishii, Kazuya [Hitachi, Ltd., Power and Industrial Systems R and D Lab., Hitachi, Ibaraki (Japan); Ando, Yoshihira [Toshiba Corp., Power and Industrial Systems Research and Development Center, Kawasaki, Kanagawa (Japan); Yamamoto, Toru; Iwata, Yutaka; Ueji, Masao [Nuclear Power Engineering Corp., Systems Safety Department, Tokyo (Japan)

    2002-08-01

    An extensive experimental program, MISTRAL, is undertaken by NUPEC and CEA in order to measure the main core physics parameters of high moderation 100% MOX LWRs. The analysis of the MISTRAL-4, a mock-up of a full MOX PWR core was carried out by diffusion/transport calculations with the SRAC code system and by continuous-energy Monte Carlo calculations with the MVP code. The calculation results agreed with the experimental results of both calculations within experimental uncertainties and calculation results showed no specific trend caused by heterogeneity in highly-moderated mock-up core configurations. (author)

  1. The stress corrosion cracking behavior of Alloys 690 and 152 weld in a PWR environment

    Energy Technology Data Exchange (ETDEWEB)

    Alexandreanu, B.; Chopra, O.; Shack, W. [Argonne National Lab., Argonne, Illinois (United States)

    2007-07-01

    'Full text:' Alloys 690 and 152 are the replacement materials of choice for Alloys 600 and, respectively, 182. The objective of this work was to determine the stress corrosion cracking (SCC) crack growth rates (CGRs) in a simulated PWR water environment for the two replacement alloys. The study involved cold-rolled Alloy 690 and a laboratory-prepared Alloy 152 double-V weld. In testing in primary water, both alloys sustained SCC cracking under constant loading conditions in the 10E-11 m/s range. (author)

  2. TRU Recycling in Thorium Fuel Cycle of PWR with various moderator-to-fuel volume ratio

    Energy Technology Data Exchange (ETDEWEB)

    Waris, Abdul; Su' ud, Zaki; Kurniadi, Rizal [Institut Teknologi Bandung, Jl. Ganesa No. 10, Bandung, West Java 40132 (Indonesia)

    2009-06-15

    Global attention to proliferation resistance has increased and as a result advanced nuclear core and fuel designs are required to possess a high level of proliferation resistance. Thorium fuel is well recognized for its inherently high proliferation resistance potential because of the low production rates of plutonium and minor actinides as compared with uranium fuel. Thorium has less radio-toxicity of spent fuel and also displays a good breeding ratio in thermal reactors. Therefore, study on thorium-based nuclear fuels has become interesting again. It is believed that by using a modern thorium-based fuel design in current reactor, a better economic performance can be achieved [1, 2]. Thorium can be utilized as fuel in nuclear reactors like uranium even though thorium is not fissile material, since {sup 232}Th is capable to capture thermal neutrons to produce {sup 233}U, a fissile isotope. Therefore, the common thorium fuel consists of {sup 232}Th and {sup 233}U for initial loading fuel. For this reason we need a thorium breeder to realize the thorium fuel reactor system, since {sup 233}U does not occur naturally. As a part of revisiting the thorium-based nuclear fuel for the present and future nuclear energy system, the present study focuses on analysis of trans-uranium (TRU) recycling in thorium fuel cycle of pressurized water reactor (PWR) to overcome the need of thorium breeder. PWR was chosen since it will still dominate the nuclear energy system up to 2050. For comprehensive study, an influence of moderator-to-fuel volume ratio (MFR) changes by changing the pin-pitch of fuel cell is also evaluated. The MFR ranges from 0.5 to 4.0. The result show that the standard PWR with MFR = 2.0 may be a good design for TRU recycling in thorium fuel cycle. The complete results will be given in the full paper. References: [1] A. Waris, H. Sekimoto and G. Kastchiev (2002), Influence of Moderator-to-Fuel Volume Ratio on Pu and MA Recycling in Equilibrium Fuel Cycle of PWR

  3. Application of LBB to high energy piping systems in operating PWR

    Energy Technology Data Exchange (ETDEWEB)

    Swamy, S.A.; Bhowmick, D.C. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    1997-04-01

    The amendment to General Design Criterion 4 allows exclusion, from the design basis, of dynamic effects associated with high energy pipe rupture by application of leak-before-break (LBB) technology. This new approach has resulted in substantial financial savings to utilities when applied to the Pressurized Water Reactor (PWR) primary loop piping and auxiliary piping systems made of stainless steel material. To date majority of applications pertain to piping systems in operating plants. Various steps of evaluation associated with the LBB application to an operating plant are described in this paper.

  4. Transient Ischemic Attack

    Medline Plus

    Full Text Available ... TIA , or transient ischemic attack, is a "mini stroke" that occurs when a blood clot blocks an ... a short time. The only difference between a stroke and TIA is that with TIA the blockage ...

  5. Transient tic disorder

    Science.gov (United States)

    ... makes 1 or many brief, repeated, movements or noises (tics). These movements or noises are involuntary (not on purpose). Causes Transient tic ... less than a year. Other disorders such as anxiety , attention deficit hyperactivity disorder ( ADHD ), uncontrollable movement ( myoclonus ), ...

  6. Transient Ischemic Attack

    Medline Plus

    Full Text Available ... Ischemic Attack TIA , or transient ischemic attack, is a "mini stroke" that occurs when a blood clot blocks an artery for a short time. The only difference between a stroke ...

  7. Transient Microcavity Sensor

    CERN Document Server

    Shu, Fang-Jie; Özdemir, Şahin Kaya; Yang, Lan; Guo, Guang-Can

    2015-01-01

    A transient and high sensitivity sensor based on high-Q microcavity is proposed and studied theoretically. There are two ways to realize the transient sensor: monitor the spectrum by fast scanning of probe laser frequency or monitor the transmitted light with fixed laser frequency. For both methods, the non-equilibrium response not only tells the ultrafast environment variance, but also enable higher sensitivity. As examples of application, the transient sensor for nanoparticles adhering and passing by the microcavity is studied. It's demonstrated that the transient sensor can sense coupling region, external linear variation together with the speed and the size of a nanoparticle. We believe that our researches will open a door to the fast dynamic sensing by microcavity.

  8. Transient multivariable sensor evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, Richard B.; Heifetz, Alexander

    2017-02-21

    A method and system for performing transient multivariable sensor evaluation. The method and system includes a computer system for identifying a model form, providing training measurement data, generating a basis vector, monitoring system data from sensor, loading the system data in a non-transient memory, performing an estimation to provide desired data and comparing the system data to the desired data and outputting an alarm for a defective sensor.

  9. THERMIT2. BWR & PWR Thermal-Hydraulic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kazimi, M.S.; Kao, S.P.; Kelly, J.E. [Massachusetts Institute of Technology, Cambridge, MA (United States)

    1992-02-27

    THERMIT2, the most recent release of THERMIT, is intended for thermal-hydraulic analysis of both boiling and pressurized water reactor cores. It solves the three-dimensional, two-fluid equations describing the two-phase flow and heat transfer dynamics in rectangular coordinates. The two-fluid model uses separate partial differential equations expressing conservation of mass, momentum, and energy for each fluid. By expressing the exchange of mass, momentum, and energy between the fluids with physically-based mathematical models, the relative motion and thermal non-equilibrium between the fluids can exist. THERMIT2 offers the choice of either pressure or velocity boundary conditions at the top and bottom of the core. THERMIT2 includes a two-phase turbulent mixing model which provides subchannel analysis capability. THERMIT2 also solves the radial heat conduction equations for fuel pin temperatures, and calculates the heat flux from fuel pin to coolant with appropriate heat transfer models described by a boiling curve.

  10. Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-07-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  11. Application of COBAYA3 code to the multiscale analysis with neutronic-thermohydraulic connection in PWR reactors; Aplicacion del codigo COBAY A3 al analisis multiescala con acoplamiento neutronico-termohidraulico en PWR

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, J.; Herrero, J. J.; Lozano, J. A.; Aragones, J. M.; Cuervo, D.; Garcia-Herranz, N.; Ahnert, C.

    2010-07-01

    This paper presents the application of COBAYA3 codes in PWR reactors cores analysis. The methodology for the breakdown in subcommands by means of alternated dissections has been implemented in the above-mentioned code system as part of the work done by the first two authors in their doctoral thesis.

  12. Comparison of BR3 Surveillance and Vessel Plates to the Surrogate Plates Representative of the Yankee Rowe PWR Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Fabry, A.; Chaouadi, R.; Puzzolante, J.L.; Van de Velde, J.; Biemiller, E.C.; Rosinski, S.T.; Carter, R.G

    1998-07-01

    The sister pressure vessels at the BR3 and Yankee Rowe PWR plants were operated at lower-than-usual temperature ( 260 degrees Celsius) and their plates were austenitized a higher-than-usual temperature (970 degrees Celsius) - a heat treatment leading to a coarser microstructure than is typical for the fine grain plates considered in development of USNRC Regulatory Guide 1.99. The surveillance programs provided by Westinghouse for the two plants were limited to the same A302-B plate representative of the Rowe vessel upper shell plate; this material displayed outlier behaviour characterized by a 41 J Charpy-V Notch shift significantly larger than predicted by Regulatory Guide 1.99. Because lower irradiation temperature and nickel alloying are generally considered detrimental to irradiation sensitivity, there was a major concern that the nickel-modified lower Rate plate and the nickel-modified BR3 plate may become too embrittled to satisfy the toughness requirements embodied in the PTS screening criterion. This paper compares free complementary studies undertaken to clarify these uncertainties: (1) The accelerated irradiation and test program launched in 1990 by Yankee Atomic Electric Company using typical vessel plate materials containing 0.24% copper at two nickel levels: YA1, 0.63 % (A533-B) and YA9, 0.19 (A302-B). These were heat-treated to produce the coarse and fine grain microstructures representative of the Yankee/BR3 and the Regulatory Guide plates, respectively; (2) The BR3 surveillance and vessel testing program: this vessel was wet-annealed in 1984, relicensed for operation till the plant shutdown in 1987, and was trepanned in early 1995; (3) The accelerated irradiations in the Belgian test reactor BR2 of the Yankee coarse grain plates YA1 and YA9 together with BR3 vessel specimens extracted at nozzle elevation, a location with negligible radiation exposure. It is contended that the PTS screening criterion was never attained by the BR3 and Rowe plates, and

  13. VOF Simulations of Countercurrent Gas-Liquid Flow in a PWR Hot Leg

    Directory of Open Access Journals (Sweden)

    Michio Murase

    2012-12-01

    Full Text Available In order to evaluate flow patterns and CCFL (countercurrent flow limitation characteristics in a PWR hot leg under reflux condensation, numerical simulations have been done using a two-fluid model and a VOF (volume of fluid method implemented in the CFD software, FLUENT6.3.26. The two-fluid model gave good agreement with CCFL data under low pressure conditions but did not give good results under high pressure steam-water conditions. On the other hand, the VOF method gave good agreement with CCFL data for tests with a rectangular channel but did not give good results for calculations in a circular channel. Therefore, in this paper, the computational grid and schemes were improved in the VOF method, numerical simulations were done for steam-water flows at 1.5 MPa under PWR full-scale conditions with the diameter of 0.75 m, and the calculated results were compared with the UPTF data at 1.5 MPa. As a result, the calculated flow pattern was found to be similar to the flow pattern observed in small-scale air-water tests, and the calculated CCFL characteristics agreed well with the UPTF data at 1.5 MPa except in the region of a large steam volumetric flux.

  14. Gas entrainment by one single French PWR spray, SARNET-2 spray benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, Saclay (France); Mimouni, S., E-mail: stephane.mimouni@edf.fr [Electricité de France, EDF MF2E, Chatou (France); Manzini, G., E-mail: giovanni.manzini@rse-web.it [RSE, Milano (Italy); Xiao, J., E-mail: jianjun.xiao@kit.edu [IKET, KIT, Karlsruhe (Germany); Vyskocil, L., E-mail: vyl@ujv.cz [UJV Rez (Czech Republic); Siccama, N.B., E-mail: siccama@nrg.eu [NRG, Safety and Power (Netherlands); Huhtanen, R., E-mail: risto.huhtanen@vtt.fi [VTT, PO Box 1000, FI-02044 VTT (Finland)

    2015-02-15

    Highlights: • This paper presents a benchmark performed in the frame of the SARNET-2 EU project. • It concerns momentum transfer between a PWR spray and the surrounding gas. • The entrained gas velocities can vary up to 100% from one code to another. • Simplified boundary conditions for sprays are generally used by the code users. • It is shown how these simplified conditions impact the gas entrainment. - Abstract: This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety.

  15. PWR Containment Shielding Calculations with SCALE6.1 Using Hybrid Deterministic-Stochastic Methodology

    Directory of Open Access Journals (Sweden)

    Mario Matijević

    2016-01-01

    Full Text Available The capabilities of the SCALE6.1/MAVRIC hybrid shielding methodology (CADIS and FW-CADIS were demonstrated when applied to a realistic deep penetration Monte Carlo (MC shielding problem of a full-scale PWR containment model. Automatic preparation of variance reduction (VR parameters is based on deterministic transport theory (SN method providing the space-energy importance function. The aim of this paper was to determine the neutron-gamma dose rate distributions over large portions of PWR containment with uniformly small MC uncertainties. The sources of ionizing radiation included fission neutrons and photons from the reactor and photons from the activated primary coolant. We investigated benefits and differences of FW-CADIS over CADIS methodology for the objective of the uniform MC particle density in the desired tally regions. Memory intense deterministic module was used with broad group library “v7_27n19g” opposed to the fine group library “v7_200n47g” used for final MC simulation. Compared with CADIS and with the analog MC, FW-CADIS drastically improved MC dose rate distributions. Modern shielding problems with large spatial domains require not only extensive computational resources but also understanding of the underlying physics and numerical interdependence between SN-MC modules. The results of the dose rates throughout the containment are presented and discussed for different volumetric adjoint sources.

  16. Burnable Absorber-Filled Annular UO{sub 2} Fuels for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Yahya, Mohd-Syukri; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of); Chung, ChangKyu [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    Its central annulus hole also provides an additional plenum for the fission gas release. In fact, annular UO{sub 2} fuels have successfully been used in commercial Russian's nuclear reactors for decades. It was upon this notion that a study was recently performed to re-investigate neutronic characteristics of the annular fuel in a rod-cell lattice. The said study also proposed an innovative integral burnable absorber (BA) concept by loading of a porous BA rod inside central hole of the annular fuel. This current work aims to extend the said investigation by characterizing neutronic performances of the BA-filled annular fuels in standard PWR 17x17 and 16x16 fuel assembly lattices. Preliminary results suggested promising potentials of the novel BA concept in managing the assembly lattice reactivity and power peaking. All calculations were performed using the Monte Carlo Serpent code with ENDF/B7.0 library. This paper demonstrates neutronic feasibilities of the BA-filled annular fuels in standard PWR 17x17 and 16x16 fuel assembly lattices. One notes that the BA-filled annular fuel-loaded lattice display comparable neutronic characteristics to the benchmarked commercial BA designs, especially in terms of reactivity and peaking factor management.

  17. Improvement of the fracture toughness in the PWR pressurizer surgeline material (Final report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jun Hwa; Lee, Bong Sang; Oh, Yong Jun; Yoon, Ji Hyun; Oh, Jong Myung; Park, Buk Gyun; Kim, Ju Hak; Kuk Il Hyun; Byun, Taek Sang [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-12-01

    Fracture toughness property of PWR surgeline pipe materials is one of the most important factor for Leak Before Break(LBB) analysis. In order to improve fracture toughness of the surgeline material (SA312-TP347 stainless steel), base on the evaluation and analysis of the commercial TP347 alloys, eleven TP347 model alloys were designed and manufactured. Tensile and fracture resistance properties of the model alloys, as well as microstructure, were evaluated. It is concluded that the nitrogen shall be added more than 0.1% for high tensile property and the carbon shall be in the range of 0.02 to 0.04% for high fracture resistance. In addition, four TP316N stainless steels were manufactured and evaluated to find out the applicability as a candidate material for PWR surgeline pipe. As a conclusion, TP316N stainless steels had an excellent property to be used for surgeline piping materials, substituting the present TP347 stainless steels. (author). 11 refs., 50 figs., 12 tabs.

  18. Regeneratively Cooled Liquid Oxygen/Methane Technology Development Between NASA MSFC and PWR

    Science.gov (United States)

    Robinson, Joel W.; Greene, Christopher B.; Stout, Jeffrey B.

    2012-01-01

    The National Aeronautics & Space Administration (NASA) has identified Liquid Oxygen (LOX)/Liquid Methane (LCH4) as a potential propellant combination for future space vehicles based upon exploration studies. The technology is estimated to have higher performance and lower overall systems mass compared to existing hypergolic propulsion systems. NASA-Marshall Space Flight Center (MSFC) in concert with industry partner Pratt & Whitney Rocketdyne (PWR) utilized a Space Act Agreement to test an oxygen/methane engine system in the Summer of 2010. PWR provided a 5,500 lbf (24,465 N) LOX/LCH4 regenerative cycle engine to demonstrate advanced thrust chamber assembly hardware and to evaluate the performance characteristics of the system. The chamber designs offered alternatives to traditional regenerative engine designs with improvements in cost and/or performance. MSFC provided the test stand, consumables and test personnel. The hot fire testing explored the effective cooling of one of the thrust chamber designs along with determining the combustion efficiency with variations of pressure and mixture ratio. The paper will summarize the status of these efforts.

  19. The Shape of Things to Come: Estimating Northern-Hemisphere (NH) Transient Climate Response Through Hindcasting and Forecasting the Frequency Distribution of Earth's NH Land Temperature Anomalies for the Period 1951-2071

    Science.gov (United States)

    Leclerc, D. F.

    2016-12-01

    Northern-hemisphere (NH) heatwaves, during which temperatures rise 5 standard deviations (SD), sigma, above the historical mean temperature, mu, are becoming frequent; these events skew temperature anomaly (delta T) profiles towards extreme values. Although general extreme value (GEV) distributions have modeled precipitation data, their application to temperatures have met with limited success. This work presents a modified three-parameter (mu, sigma and tau (skew)) Exponential-Gaussian (eGd) model that hindcasts decadal NH land winter (DJF) and summer (JJA) delta Ts from 1951 to 2011, and forecasts profiles for a business-as-usual (BAU) scenario for 2061-2071. We accessed 12 numerical binned (0.05 °C/bin) z-scored NH decadal datasets (posted online until August 2015) from the publicly available website http://www.columbia.edu/ mhs119/PerceptionsAndDice/ mentioned in Hansen et al, PNAS 109 E2415-E2423 (2012) and stated to be in the public domain. No pre-processing was done. Parameters were calculated for the 12 NH datasets pasted into Microsoft Excel™ through the method of moments for 1-tail distributions and through the BEST deconvolution program described by Pommé and Marroyo, Applied Radiation and Isotopes 96 148-153 (2015) for 2-tail distributions. We used maximum likelihood estimation (MLE), residual sum of squares (RSS) and F-test to find optimal parameter values. Calculated 1st (= sigma + tau) and 2nd (= sigma2 + tau2) moments were found to be within 0.5% of observed values. Land delta Ts were recovered from the z-score values by multiplying the winter data by its SD (1.2 °C) and likewise the summer data by 0.6 °C. Results were all within 0.05 °C of 10-year averages from the GHCNv3 NH land dataset. Assuming BAU (increases from 2.1 to 2.6 ppm/y CO2) and using temperature rises of 0.27 °C and 0.35 °C per decade, for summer and winter, respectively, and forecasting to 2071, we obtain for the transient climate response for doubled CO2 (560 ppm CO2) mean

  20. Polynomial approximation approach to transient heat conduction ...

    African Journals Online (AJOL)

    This work reports polynomial approximation approach to transient heat conduction in a long slab, long cylinder and sphere with linear internal heat generation. It has been shown that the polynomial approximation method is able to calculate average temperature as a function of time for higher value of Biot numbers.

  1. Correlating activity incorporation with properties of oxide films formed on material samples exposed to BWR and PWR coolants in Finnish nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bojinov, M.; Kinnunen, P.; Laitinen, T.; Maekelae, K.; Saario, T.; Sirkiae, P. [VTT Industrial Systems, Espoo (Finland); Buddas, T.; Halin, M.; Kvarnstroem, R.; Tompuri, K. [Fortum Power and Heat Oy, Loviisa Power Plant, Loviisa (Finland); Helin, M.; Muttilainen, E.; Reinvall, A. [Teollisuuden Voima Oy, Olkiluoto (Finland)

    2002-07-01

    The extent of activity incorporation on primary circuit surfaces in nuclear power plants is connected to the chemical composition of the coolant, to the corrosion behaviour of the material surfaces and to the structure and properties of oxide films formed on circuit surfaces due to corrosion. Possible changes in operational conditions may induce changes in the structure of the oxide films and thus in the rate of activity incorporation. To predict these changes, experimental correlations between water chemistry, oxide films and activity incorporation, as well as mechanistic understanding of the related phenomena need to be established. In order to do this, flow-through cells with material samples and facilities for high-temperature water chemistry monitoring have been installed at Olkiluoto unit 1 (BWR) and Loviisa unit 1 (PWR) in spring 2000. The cells are being used for two major purposes: To observe the changes in the structure and activity levels of oxide films formed on material samples exposed to the primary coolant. Correlating these observations with the abundant chemical and radiochemical data on coolant composition, dose rates etc. collected routinely by the plant, as well as with high-temperature water chemistry monitoring data such as the corrosion potentials of relevant material samples, the redox potential and the high-temperature conductivity of the primary coolant. We describe in this paper the scope of the work, give examples of the observations made and summarize the results on oxide films that have been obtained during one full fuel cycle at both plants. (authors)

  2. Analytical model of transient thermal effect on convectional cooled ...

    Indian Academy of Sciences (India)

    Home; Journals; Pramana – Journal of Physics; Volume 81; Issue 4. Analytical model of transient thermal effect on convectional cooled end-pumped laser rod ... The transient analytical solutions of temperature distribution, stress, strain and optical path difference in convectional cooled end-pumped laser rod are derived.

  3. BWR water chemistry guidelines and PWR primary water chemistry guidelines in Japan – Purpose and technical background

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Hirotaka, E-mail: kawamuh@criepi.denken.or.jp [Central Research Institute of Electric Power Industry (Japan); Hirano, Hideo [Central Research Institute of Electric Power Industry (Japan); Katsumura, Yousuke [University of Tokyo (Japan); Uchida, Shunsuke [Tohoku University (Japan); Mizuno, Takayuki [Mie University (Japan); Kitajima, Hideaki; Tsuzuki, Yasuo [Japan Nuclear Safety Institute (Japan); Terachi, Takumi [Institute of Nuclear Safety System, Inc. (Japan); Nagase, Makoto; Usui, Naoshi [Hitachi-GE Nuclear Energy, Ltd. (Japan); Takagi, Junichi; Urata, Hidehiro [Toshiba Corporation (Japan); Shoda, Yasuhiko; Nishimura, Takao [Mitsubishi Heavy Industry, Ltd. (Japan)

    2016-12-01

    Highlights: • Framework of BWR/PWR water chemistry Guidelines in Japan are presented. • Guideline necessity, definitions, philosophy and technical background are mentioned. • Some guideline settings for control parameters and recommendations are explaines. • Chemistry strategy is also mentioned. - Abstract: After 40 years of light water reactor (LWR) operations in Japan, the sustainable development of water chemistry technologies has aimed to ensure the highest coolant system component integrity and fuel reliability performance for maintaining LWRs in the world; additionally, it aimed to achieve an excellent dose rate reduction. Although reasonable control and diagnostic parameters are utilized by each boiling water reactor (BWR) and pressurized water reactor (PWR) owner, it is recognized that specific values are not shared among everyone involved. To ensure the reliability of BWR and PWR operation and maintenance, relevant members of the Atomic Energy Society of Japan (AESJ) decided to establish guidelines for water chemistry. The Japanese BWR and PWR water chemistry guidelines provide strategies to improve material and fuel reliability performance as well as to reduce dosing rates. The guidelines also provide reasonable “control values”, “diagnostic values” and “action levels” for multiple parameters, and they stipulate responses when these levels are exceeded. Specifically, “conditioning parameters” are adopted in the Japanese PWR primary water chemistry guidelines. Good practices for operational conditions are also discussed with reference to long-term experience. This paper presents the purpose, technical background and framework of the preliminary water chemistry guidelines for Japanese BWRs and PWRs. It is expected that the guidelines will be helpful as an introduction to achieve safety and reliability during operations.

  4. Metallurgical and mechanical parameters controlling alloy 718 stress corrosion cracking resistance in PWR primary water; Facteurs metallurgiques et mecaniques controlant l'amorcage de defauts de corrosion sous contrainte dans l'alliage 718 en milieu primaire des reacteurs a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Deleume, J

    2007-11-15

    Improving the performance and reliability of the fuel assemblies of the pressurized water reactors requires having a perfect knowledge of the operating margins of both the components and the materials. The choice of alloy 718 as reference material for this study is justified by the industrial will to identify the first order parameters controlling the excellent resistance of this alloy to Stress Corrosion Cracking (SCC). For this purpose, a specific slow strain rate (SSR) crack initiation test using tensile specimen with a V-shaped hump in the middle of the gauge length was developed and modeled. The selectivity of such SSR tests in simulated PWR primary water at 350 C was clearly established by characterizing the SCC resistance of nine alloy 718 thin strip heats. Regardless of their origin and in spite of a similar thermo-mechanical history, they did not exhibit the same susceptibility to SCC crack initiation. All the characterized alloy 718 heats develop oxide scale of similar nature for various exposure times to PWR primary medium in the temperature range [320 C - 360 C]. {delta} phase precipitation has no impact on alloy 718 SCC initiation behavior when exposed to PWR primary water, contrary to interstitial contents and the triggering of plastic instabilities (PLC phenomenon). (author)

  5. Tribological study of hard coatings without cobalt intended to isolation components of PWR primary cooling system; Etude tribologique de revetements durs sans cobalt destines aux organes d`isolement du circuit primaire des REP

    Energy Technology Data Exchange (ETDEWEB)

    Cachon, L.

    1995-10-18

    The objective is to qualify coatings without cobalt to replace ``Stellites`` coatings in isolation valves of PWR primary cooling system, as Co is activated when passing in the reactor core and contaminated the cooling loop. Three families of coatings were tested: PVD thin films from 1 to 8 {mu}m monolayers of Cr/C{sub x} with x varying between 1.6 and 9.5 at% or multilayers of pure chromium and Cr/C{sub 1.6} at%, coatings with a thickness between 100 and 200 {mu}m of cermets NiCr{sub y} (y varying from 5 to 35 at%) matrix binding chromium or tungsten carbides, and thick coatings 2 mm thickness of cermets Nitronic 60 or Inconel 625 matrix binding 10, 20 or 30% titanium or niobium carbides. Stellite 6 (2 mm) is the reference coating for tribology. Coatings were qualified and selected by thermal shocks, corrosion and plane friction. The thin film and the thick families were disqualified by their destruction or by their high friction coefficient. Then coatings between 100 and 200 {mu}m were used in a valve mock-up working in PWR primary cooling system pressure and temperature conditions. Tests show that these coatings have better wear or tightness performances than stellite 6, except for a slightly higher friction coefficient. (A.B.).

  6. Transient cavitation in pipelines

    NARCIS (Netherlands)

    Kranenburg, C.

    1974-01-01

    The aim of the present study is to set up a one-dimensional mathematical model, which describes the transient flow in pipelines, taking into account the influence of cavitation and free gas. The flow will be conceived of as a three-phase flow of the liquid, its vapour and non-condensible gas. The

  7. Investigation on the use digital controls instead of PID analog controls in the level control of steam generators of nuclear power PWR; Investigacao sobre o uso de controladores digitais em substituicao aos controladores analogicos PID para o controle de nivel de geradores de vapor de centrais nucleares PWR

    Energy Technology Data Exchange (ETDEWEB)

    Alvarenga, Marco Antonio Bayout

    2012-07-01

    The aim of this study is to identify current alternatives for the implementation of digital controllers in the level control of steam generators of nuclear power PWR (Pressurized Water Reaetor). It is intended to identify the types of digital controls that are available from the theoretical and conceptual viewpoints for this purpose. We investigate the advantages and disadvantages of each controller model. From this assessment are pointed the most suitable models in hierarchical scale. This evaluation also serves to suggest possible types of control installation as a whole, where the level control of the steam generators becomes just one of many controls that are part of the plant. In this case, the use of digital controls allows the non-linear and multivariable treatment which is characteristic of complex systems, such as the nuclear power generation. The treatment of nonlinearities and multivariable aspects allows a more detailed study of the stability of these plants when they are subject to transients or several accidents, such as the case of losing external power of transients. In the specific case of steam generators, the instabilities result from the emergence of the shrink and swell phenomenas, depending on the load variations of thermonuclear plant. The application of several types and digital controllers, considering these inherent characteristics of the level control of steam generators, allows to infer which types of controllers are more appropriate to treat instabilities of this type and to make conjectures in its use for the cases of more complex instabilities, considering the integration of all nucleus-plant controls.

  8. Enhanced Severe Transient Analysis for Prevention Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    This document outlines the development of a high fidelity, best estimate nuclear power plant severe transient simulation capability that will complement or enhance the integral system codes historically used for licensing and analysis of severe accidents. As with other tools in the Risk Informed Safety Margin Characterization (RISMC) Toolkit, the ultimate user of Enhanced Severe Transient Analysis and Prevention (ESTAP) capability is the plant decision-maker; the deliverable to that customer is a modern, simulation-based safety analysis capability, applicable to a much broader class of safety issues than is traditional Light Water Reactor (LWR) licensing analysis. Currently, the RISMC pathway’s major emphasis is placed on developing RELAP-7, a next-generation safety analysis code, and on showing how to use RELAP-7 to analyze margin from a modern point of view: that is, by characterizing margin in terms of the probabilistic spectra of the “loads” applied to systems, structures, and components (SSCs), and the “capacity” of those SSCs to resist those loads without failing. The first objective of the ESTAP task, and the focus of one task of this effort, is to augment RELAP-7 analyses with user-selected multi-dimensional, multi-phase models of specific plant components to simulate complex phenomena that may lead to, or exacerbate, severe transients and core damage. Such phenomena include: coolant crossflow between PWR assemblies during a severe reactivity transient, stratified single or two-phase coolant flow in primary coolant piping, inhomogeneous mixing of emergency coolant water or boric acid with hot primary coolant, and water hammer. These are well-documented phenomena associated with plant transients but that are generally not captured in system codes. They are, however, generally limited to specific components, structures, and operating conditions. The second ESTAP task is to similarly augment a severe (post-core damage) accident integral analyses code

  9. Compressive Transient Imaging

    KAUST Repository

    Sun, Qilin

    2017-04-01

    High resolution transient/3D imaging technology is of high interest in both scientific research and commercial application. Nowadays, all of the transient imaging methods suffer from low resolution or time consuming mechanical scanning. We proposed a new method based on TCSPC and Compressive Sensing to achieve a high resolution transient imaging with a several seconds capturing process. Picosecond laser sends a serious of equal interval pulse while synchronized SPAD camera\\'s detecting gate window has a precise phase delay at each cycle. After capturing enough points, we are able to make up a whole signal. By inserting a DMD device into the system, we are able to modulate all the frames of data using binary random patterns to reconstruct a super resolution transient/3D image later. Because the low fill factor of SPAD sensor will make a compressive sensing scenario ill-conditioned, We designed and fabricated a diffractive microlens array. We proposed a new CS reconstruction algorithm which is able to denoise at the same time for the measurements suffering from Poisson noise. Instead of a single SPAD senor, we chose a SPAD array because it can drastically reduce the requirement for the number of measurements and its reconstruction time. Further more, it not easy to reconstruct a high resolution image with only one single sensor while for an array, it just needs to reconstruct small patches and a few measurements. In this thesis, we evaluated the reconstruction methods using both clean measurements and the version corrupted by Poisson noise. The results show how the integration over the layers influence the image quality and our algorithm works well while the measurements suffer from non-trival Poisson noise. It\\'s a breakthrough in the areas of both transient imaging and compressive sensing.

  10. PWR design for low doses in the United Kingdom: The present and the future

    Energy Technology Data Exchange (ETDEWEB)

    Zodiates, A.M.; Willcock, A. [PWR Project Group, Knutsford, England (United Kingdom)

    1995-03-01

    The Pressurizer Water Reactor (PWR) design chosen for adoption by Nuclear Electric plc was based on the Westinghouse Standard Nuclear Unit Power Plant System (SNUPPS). This design was developed to meet the United Kingdom (UK) requirements and those improvements are embodied in the Sizewell B plant. Nuclear Electric plc is now looking to the design of the future PWRs to be built in the UK. These PWRs will be based as replicas of the Sizewell B design, but attention will be given to reducing operator doses further. This paper details the approach in operator protection improvements incorporated at Sizewall B, presents the estimated annual collective dose, and identifies the approach being adopted to reduce further operator doses in future plants.

  11. Basic investigation of particle swarm optimization performance in a reduced scale PWR passive safety system design

    Energy Technology Data Exchange (ETDEWEB)

    Cunha, Joao J. da [Eletronuclear Eletrobras Termonuclear, Gerencia de Analise de Seguranca Nuclear, Rua da Candelaria, 65, 7o andar. Centro, Rio de Janeiro 20091-906 (Brazil); Lapa, Celso Marcelo F., E-mail: lapa@ien.gov.b [Instituto de Engenharia Nuclear, Divisao de Reatores/PPGIEN, P.O. Box 68550, Rua Helio de Almeida 75 Cidade Universitaria, Ilha do Fundao, Rio de Janeiro 21941-972 (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (Brazil); Alvim, Antonio Carlos M. [Universidade Federal do Rio de Janeiro, COPPE/Nuclear, P.O. Box 68509, Cidade Universitaria, Ilha do Fundao s/n, Rio de Janeiro 21945-970 (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (Brazil); Lima, Carlos A. Souza [Instituto de Engenharia Nuclear, Divisao de Reatores/PPGIEN, P.O. Box 68550, Rua Helio de Almeida 75 Cidade Universitaria, Ilha do Fundao, Rio de Janeiro 21941-972 (Brazil); Instituto Politecnico, Universidade do Estado do Rio de Janeiro, Pos-Graduacao em Modelagem Computacional, Rua Alberto Rangel, s/n, Vila Nova, Nova Friburgo 28630-050 (Brazil); Pereira, Claudio Marcio do N.A. [Instituto de Engenharia Nuclear, Divisao de Reatores/PPGIEN, P.O. Box 68550, Rua Helio de Almeida 75 Cidade Universitaria, Ilha do Fundao, Rio de Janeiro 21941-972 (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (Brazil)

    2010-03-15

    This work presents a methodology to investigate the viability of using particle swarm optimization technique to obtain the best combination of physical and operational parameters that lead to the best adjusted dimensionless groups, calculated by similarity laws, that are able to simulate the most relevant physical phenomena in single-phase flow under natural circulation and to offer an appropriate alternative reduced scale design for reactor primary loops with this flow characteristics. A PWR reactor core, under natural circulation, based on LOFT test facility, was used as the case study. The particle swarm optimization technique was applied to a problem with these thermo-hydraulics conditions and results demonstrated the viability and adequacy of the method to design similar systems with these characteristics.

  12. Representing Operational Knowledge of PWR Plant by Using Multilevel Flow Modelling

    DEFF Research Database (Denmark)

    Zhang, Xinxin; Lind, Morten; Jørgensen, Sten Bay

    2014-01-01

    situation and support operational decisions. This paper will provide a general MFM model of the primary side in a standard Westinghouse Pressurized Water Reactor ( PWR ) system including sub - systems of Reactor Coolant System, Rod Control System, Chemical and Volume Control System, emergency heat removal......The aim of this paper is to explore the capability of representing operational knowledge by using Multilevel Flow Modelling ( MFM ) methodology. The paper demonstrate s how the operational knowledge can be inserted into the MFM models and be used to evaluate the plant state, identify the current...... systems. And the sub - systems’ functions will be decomposed into sub - models according to different operational situations. An operational model will be developed based on the operating procedure by using MFM symbols and this model can be used to implement coordination rules for organize the utilizati...

  13. Effect of Weld Properties on the Crush Strength of the PWR Spacer Grid

    Directory of Open Access Journals (Sweden)

    Kee-nam Song

    2012-01-01

    Full Text Available Mechanical properties in a weld zone are different from those in the base material because of different microstructures. A spacer grid in PWR fuel is a structural component with an interconnected and welded array of slotted grid straps. Previous research on the strength analyses of the spacer grid was performed using base material properties owing to a lack of mechanical properties in the weld zone. In this study, based on the mechanical properties in the weld zone of the spacer grid recently obtained by an instrumented indentation technique, the strength analyses considering the mechanical properties in the weld zone were performed, and the analysis results were compared with the previous research.

  14. IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

    Energy Technology Data Exchange (ETDEWEB)

    Gilles Youinou; Andrea Alfonsi

    2012-03-01

    This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis, the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  15. Numerical modeling of in-vessel melt water interaction in large scale PWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Kolev, N.I. [Siemens AG, KWU NA-M, Erlangen (Germany)

    1998-01-01

    This paper presents a comparison between IVA4 simulations and FARO L14, L20 experiments. Both experiments were performed with the same geometry but under different initial pressures, 51 and 20 bar respectively. A pretest prediction for test L21 which is intended to be performed under an initial pressure of 5 bar is also presented. The strong effect of the volume expansion of the evaporating water at low pressure is demonstrated. An in-vessel simulation for a 1500 MW el. PWR is presented. The insight gained from this study is: that at no time are conditions for the feared large scale melt-water intermixing at low pressure in force, with this due to the limiting effect of the expansion process which accelerates the melt and the water into all available flow paths. (author)

  16. PENENTUAN KOEFISIEN DISPERSI ATMOSFERIK UNTUK ANALISIS KECELAKAAN REAKTOR PWR DI INDONESIA

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-03-01

    Full Text Available Atmosfer merupakan pathway penting pada perpindahan radionuklida yang lepas dari Pembangkit Listrik Tenaga Nuklir (PLTN ke lingkungan dan manusia. Penerimaan dosis pada lingkungan dan manusia dipengaruhi oleh sourceterm dan kondisi tapak PLTN. Untuk mengetahui penerimaan dosis lingkungan untuk PLTN di Indonesia, maka diperlukan nilai koefisien dispersi untuk tapak potensial yang dipilih. Model perhitungan dalam penelitian ini menggunakan model yang diterapkan pada paket program pada modul ATMOS dan CONCERN dari PC-Cosyma yaitu model perhitungan segmented plume model. Perhitungan dilakukan untuk PLTN tipe PWR kapasitas 1000 MWe berbahan bakar UO2, postulasi kejadian untuk kecelakaan DBA, kondisi tapak kasar, untuk 6 tapak contoh tapak Semenanjung Muria, Pesisir Banten, dan tapak yang didominasi oleh stabilitas cuaca C,D,E, dan F. Koefisien dispersi dihitung untuk 8 kelompok nuklida produk fisi yang lepas dari PLTN yaitu: kelompok gas mulia, lantanida, logam mulia, halogen, logam alkali, tellurium, cerium, dan kelompok stronsium & barium. Perhitungan input menggunakan paket program ORIGEN-2 dan Arc View untuk penyiapan input perhitungan. Hasil pemetaan untuk parameter dispersi maksimum rerata diperoleh pada jarak radius 800 m dari sumber lepasan untuk nuklida dari kelompok logam mulia, logam alkali dan kelompok nuklida cerium. Parameter dispersi untuk Tapak Muria maksimum 1,53E-04 s/m3, Tapak Serang adalah 1,40E-03 s/m3, tapak dengan stabilitas C: 1,72E-04 s/m3, stabilitas D: 1,40E-04 s/m3, Stabilitas E: 1,07E-04 s/m3, dan tapak dengan stabilitas F : 2,14E-05 s/m3. Kata kunci: koefisien dispersi, atmosferik, PWR, kecelakaan, Indonesia   The atmosphere is an important pathway in the migration of radionuclides transport from the Nuclear Power Plant (NPP to the environment and humans. The dose accepted in the environment and humans is influenced by the sourceterm and NPP siting condition. Distribution of radionuclides in the atmosphere is determined

  17. Probabilistic safety assessment of the French PWR 900 MWE series results and insights

    Energy Technology Data Exchange (ETDEWEB)

    Corenwinder, F.; Bertrand, V.; Dupuy, P.; Gomane, C.; Mattei, J.M.; Pichereau, F. [CEA Fontenay aux Roses, Institut de Radioprotection et de Surete Nucleaire, IRSN, 92 (France)

    2002-06-01

    The Institute for Nuclear Safety and Protection (IRSN) has been performing an updating of the Probabilistic Safety Assessment of the standard French 900 MWe PWR in order to include design and operation modifications set up since the first version of this study published by IRSN in 1990. Moreover the updating takes into account new data and new knowledge. For these reasons many sequences have been reassessed and new sequences have been identified. The updating has highlighted significant accident sequences (cold over pressurization, interfacing LOCA, loss of ventilation systems and inadvertent dilution), which were not included or not analyzed in details in the first PSA version. Moreover, the update has allowed the verification of the design and operational changes implemented to decrease the frequencies of the dominant accident sequences identified by the 1990 PSA study. The purpose of the paper is to describe the dominant accident sequences and to present the main significant points of the analysis, the main insights and the conclusions. (authors)

  18. Degraded core accidents for the Sizewell PWR A sensitivity analysis of the radiological consequences

    CERN Document Server

    Kelly, G N; Clarke, R H; Ferguson, L; Haywood, S M; Hemming, C R; Jones, J A

    1982-01-01

    The radiological impact of degraded core accidents postulated for the Sizewell PWR was assessed in an earlier study. In this report the sensitivity of the predicted consequences to variation in the values of a number of important parameters is investigated for one of the postulated accidental releases. The parameters subjected to sensitivity analyses are the dose-mortality relationship for bone marrow irradiation, the energy content of the release, the warning time before the release to the environment, and the dry deposition velocity for airborne material. These parameters were identified as among the more important in determining the uncertainty in the results obtained in the initial study. With a few exceptions the predicted consequences were found to be not very sensitive to the parameter values investigated, the range of variation in the consequences for the limiting values of each parameter rarely exceeded a factor of a few and in many cases was considerably less. The conclusions reached are, however, p...

  19. Advanced PWR Core Design with Siemens High-Plutonium-Content MOX Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Dieter Porsch; Gerhard Schlosser; Hans-Dieter Berger

    2000-06-04

    The Siemens experience with plutonium recycling dates back to the late 1960s. Over the years, extensive research and development programs were performed for the qualification of mixed-oxide (MOX) technology and design methods. Today's typical reload enrichments for uranium and MOX fuel assemblies and modern core designs have become more demanding with respect to accuracy and reliability of design codes. This paper presents the status of plutonium recycling in operating high-burnup pressurized water reactor (PWR) cores. Based on actual examples, it describes the validation status of the design methods and stresses current and future needs for fuel assembly and core design including those related to the disposition of weapons-grade plutonium.

  20. Evaluation of full MOX core capability for a 900 MWe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung-Kook; Kim, Young-Jin; Jung, Hyung-Guk; Kim, Young-Il; Sohn, Dong-Seong [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1996-09-01

    Full MOX capability of a PWR core with 900 MWe capacity has been evaluated in view of plutonium consumption and design feasibility as an effective means for spent fuel management. Three full MOX cores have been conceptually designed; for annual cycle, for 18-month cycle, and for 18-month cycle with high moderation lattice. Fissile and total plutonium quantities at discharge are significantly reduced to 60% and 70% respectively of initial value for standard full MOX cores. It is estimated that one full MOX core demands about 1 tonne of plutonium per year to be reloaded, which is equivalent to reprocessing of spent nuclear fuels discharged from five nuclear reactors operating with uranium fuels. With low-leakage loading scheme, a full MOX core with either annual or 18-month cycle can be designed satisfactorily by installing additional rod cluster control system and modifying soluble boron system. Overall high moderation lattice case promises better core nuclear characteristics. (author)

  1. Decontamination as a precursor to decommissioning. Status report Task 2: process evaluation. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Divine, J.R.; Woodruff, E.M.; McPartland, S.A.; Zima, G.E.

    1983-05-01

    As part of the US Nuclear Regulatory Commission's program to reduce occupational exposure and waste volumes, the Pacific Northwest Laboratory is studying decontamination as a precursor to decommissioning. Eleven processes or solvents were examined for their behavior in decontaminating BWR carbon steel samples. The solvents included NS-1, a proprietary solvent of Dow Chemical Corporation, designed for BWR use, and AP-Citrox, a well-known, two-step process designed for PWR stainless steel; it was used to provide a reference for later comparison to other systems and processes. The decontamination factors observed in the tests performed in a small laboratory scale recirculating loop ranged from about 1 (no effect) to 222 (about 99.6% of the initial activity removed. Coordinated corrosion measurements were made using twelve chemical solvents and eight metal alloys found in a range of reactor types.

  2. Risk analysis of highly combustible gas storage, supply, and distribution systems in PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Simion, G.P. [Science Applications International Corp., Albuquerque, NM (United States); VanHorn, R.L.; Smith, C.L.; Bickel, J.H.; Sattison, M.B. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Bulmahn, K.D. [SCIENTECH, Inc., Idaho Falls, ID (United States)

    1993-06-01

    This report presents the evaluation of the potential safety concerns for pressurized water reactors (PWRs) identified in Generic Safety Issue 106, Piping and the Use of Highly Combustible Gases in Vital Areas. A Westinghouse four-loop PWR plant was analyzed for the risk due to the use of combustible gases (predominantly hydrogen) within the plant. The analysis evaluated an actual hydrogen distribution configuration and conducted several sensitivity studies to determine the potential variability among PWRs. The sensitivity studies were based on hydrogen and safety-related equipment configurations observed at other PWRs within the United States. Several options for improving the hydrogen distribution system design were identified and evaluated for their effect on risk and core damage frequency. A cost/benefit analysis was performed to determine whether alternatives considered were justifiable based on the safety improvement and economics of each possible improvement.

  3. Quantitative uncertainty and sensitivity analysis of a PWR control rod ejection accident

    Energy Technology Data Exchange (ETDEWEB)

    Pasichnyk, I.; Perin, Y.; Velkov, K. [Gesellschaft flier Anlagen- und Reaktorsicherheit - GRS mbH, Boltzmannstasse 14, 85748 Garching bei Muenchen (Germany)

    2013-07-01

    The paper describes the results of the quantitative Uncertainty and Sensitivity (U/S) Analysis of a Rod Ejection Accident (REA) which is simulated by the coupled system code ATHLET-QUABOX/CUBBOX applying the GRS tool for U/S analysis SUSA/XSUSA. For the present study, a UOX/MOX mixed core loading based on a generic PWR is modeled. A control rod ejection is calculated for two reactor states: Hot Zero Power (HZP) and 30% of nominal power. The worst cases for the rod ejection are determined by steady-state neutronic simulations taking into account the maximum reactivity insertion in the system and the power peaking factor. For the U/S analysis 378 uncertain parameters are identified and quantified (thermal-hydraulic initial and boundary conditions, input parameters and variations of the two-group cross sections). Results for uncertainty and sensitivity analysis are presented for safety important global and local parameters. (authors)

  4. The radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR

    CERN Document Server

    Kelly, G N; Charles, D; Hemming, C R

    1983-01-01

    This report contains an assessment of the radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR. Three of the degraded core accident releases postulated by the CEGB are analysed. The consequences, conditional upon each release, are evaluated in terms of the health impact on the exposed population and the impact of countermeasures taken to limit the exposure. Consideration is given to the risk to the Greater London population as a whole and to individuals within it. The consequences are evaluated using the NRPB code MARC (Methodology for Assessing Radiological Consequences). The results presented in this report are all conditional upon the occurrence of each release. In assessing the significance of the results, due account must be taken of the frequency with which such releases may be predicted to occur.

  5. Time response of temperature sensors using neural networks; Utilizacao de redes neurais artificiais para determinar o tempo de resposta de sensores de temperatura do tipo RTD

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Roberto Carlos dos

    2010-07-01

    In a PWR nuclear power plant, the primary coolant temperature and feedwater temperature are measured using RTDs (Resistance Temperature Detectors). These RTDs typically feed the plant's control and safety systems and must, therefore, be very accurate and have good dynamic performance. The response time of RTDs is characterized by a single parameter called the Plunge Time Constant defined as the time it takes the sensor output to achieve 63.2 percent of its final value after a step change in temperature. Nuclear reactor service conditions are difficult to reproduce in the laboratory, and an in-situ test method called LCSR (Loop Current Step Response) test was developed to measure remotely the response time of RTDs. >From this test, the time constant of the sensor is identified by means of the LCSR transformation that involves the dynamic response modal time constants determination using a nodal heat-transfer model. This calculation is not simple and requires specialized personnel. For this reason an Artificial Neural Network has been developed to predict the time constant of RTD from LCSR test transient. It eliminates the transformations involved in the LCSR application. A series of LCSR tests on RTDs generates the response transients of the sensors, the input data of the networks. Plunge tests are used to determine the time constants of the RTDs, the desired output of the ANN, trained using these sets of input/output data. This methodology was firstly applied to theoretical data simulating 10 RTDs with different time constant values, resulting in an average error of about 0.74 %. Experimental data from three different RTDs was used to predict time constant resulting in a maximum error of 3,34 %. The time constants values predicted from ANN were compared with those obtained from traditional way resulting in an average error of about 18 % and that shows the network is able to predict accurately the sensor time constant. (author)

  6. Developing and analyzing long-term fuel management strategies for an advanced Small Modular PWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

    Highlights: • Comprehensive introduction and supplementary concepts as a review paper. • Developing an integrated long-term fuel management strategy for a SMR. • High reliable 3-D core modeling over fuel pins against the traditional LRM. • Verifying the expert rules of large PWRs for an advanced small PWR. • Investigating large numbers of safety parameters coherently. - Abstract: In this paper, long-term fuel management (FM) strategies are introduced and analyzed for a new advanced Pressurized Light Water Reactor (PWR) type of Small Modular Reactors (SMRs). The FM strategies are developed to be safe and practical for implementation as much as possible. Safety performances, economy of fuel, and Quality Assurance (QA) of periodic equilibrium conditions are chosen as the main goals. Flattening power density distribution over fuel pins is the major method to ensure safety performance; also maximum energy output or permissible discharging burn up indicates economy of fuel fabrication costs. Burn up effects from BOC to EOC have been traced, studied, and highly visualized in both of transport lattice cell calculations and diffusion core calculations. Long-term characteristics are searched to gain periodical equilibrium characteristics. They are fissile changes, neutron spectrum, refueling pattern, fuel cycle length, core excess reactivity, average, and maximum burn up of discharged fuels, radial Power Peaking Factors (PPF), total PPF, radial and axial power distributions, batch effects, and enrichment effects for fine regulations. Traditional linear reactivity model have been successfully simulated and adapted via fine core and burn up calculations. Effects of high burnable neutron poison and soluble boron are analyzed. Different numbers of batches via different refueling patterns have been studied and visualized. Expert rules for large type PWRs have been influenced and well tested throughout accurate equilibrium core calculations.

  7. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  8. The combined thermohydraulics-neutronics code TRAB-SMABRE for 3D plant transient and accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Jaakko Miettinen; Timo Vanttola; Hanna Raety; Antti Daavittila [VTT Processes, P.O.Box 1604, FIN-02044 VTT (Finland)

    2005-07-01

    Full text of publication follows: TRAB-3D models the PWR and BWR reactor core using the two-group diffusion equations in homogenized fuel assembly geometry with a sophisticated nodal method. Thermohydraulics is described using four-equation formulation. The stand-alone version of the code also describes thermohydraulics of the rest of the BWR circuit with one-dimensional components. The SMABRE code models the thermohydraulics of light water reactors. The five-equation formulation with the drift flux phase separation is modelling the two-phase behaviour. Conservation equations are solved for the phase mass, mixture momentum and phase energy. Additional equations are for the noncondensable in gas and boron in liquid. The TRAB-3D and SMABRE codes have been coupled earlier by using the parallel coupling principle, where in the core section the 3-dimensional TRAB core, and the parallel channel coarse SMABRE core are solved in parallel, but rest of the circulation system is solved with SMABRE. As a new development the internal coupling to meet new requirements for the PWR and BWR transient analyses is being realised. Both the circuit and core thermohydraulics are solved in SMABRE. The core thermohydraulics solution inside the core wide iterations is repeated to allow rapid power changes. These are the fast pressure changes, control rod ejection and ATWS. The numerical solution in SMABRE has been improved to allow full core simulation with separate flow channel for each fuel element of a BWR core. For the PWR plants the method is used as well by simulating the core by one-dimensional parallel channels. New development is needed for the open core calculation. In general questions could be raised, what advantages are seen with the new internal coupling in comparison with the earlier realised parallel coupling, and which advantages may be seen in building the realtor physical model on the basis of the old code, developed since 1970's. The internal coupling allows

  9. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design; Calculo de los terminos fuente de combustibles irradiados PWR para el diseno de contenedores de almacenamiento y transporte

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, J. L.; Lopez, J.

    1986-07-01

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs.

  10. Transient FDTD simulation validation

    OpenAIRE

    Jauregui Tellería, Ricardo; Riu Costa, Pere Joan; Silva Martínez, Fernando

    2010-01-01

    In computational electromagnetic simulations, most validation methods have been developed until now to be used in the frequency domain. However, the EMC analysis of the systems in the frequency domain many times is not enough to evaluate the immunity of current communication devices. Based on several studies, in this paper we propose an alternative method of validation of the transients in time domain allowing a rapid and objective quantification of the simulations results.

  11. Transient cerebral ischemia.

    OpenAIRE

    Cusimano, M D; Ameli, F M

    1989-01-01

    Stroke is a major cause of disability and death in North America. About 30% to 40% of patients with stroke have had transient ischemic attacks (TIAs). The recognition and treatment of TIAs and possibly of asymptomatic stenoses of the carotid arteries may be beneficial in preventing stroke. We review the epidemiologic features, natural history, pathogenetic features, clinical presentation, methods of investigation and management of patients with TIAs.

  12. Transient analysis in the 3D nodal kinetics and thermal-hydraulics ANDES/COBRA coupled system

    Energy Technology Data Exchange (ETDEWEB)

    Lozano, Juan Andres; Aragones, Jose Maria; Garcia-Herranz, Nuria [Universidad Politecnica de Madrid, Madrid (Spain)

    2008-07-01

    Neutron kinetics has been implemented in the 3D nodal solver ANDES, which has been coupled to the core thermal-hydraulics (TH) code COBRA-III for core transient analysis. The purpose of this work is, first, to discuss and test the ability of the kinetics solver ANDES to model transients; and second, by means of a systematic analysis, including alternate kinetics schemes, time step size, nodal size, neutron energy groups and spectrum, to serve as a basis for the development of more accurate and efficient neutronics/thermal-hydraulics tools for general transient simulations. The PWR MOX/UO{sub 2} transient benchmark provided by the OECD/NEA and US NRC was selected for these goals. The obtained ANDES/COBRA-III results were consistent with other solutions to the benchmark; the differences in the TH feedback led to slight differences in the core power evolution, whereas very good agreements were found in the other requested parameters. The performed systematic analysis highlighted the optimum kinetics iterative scheme, and showed that neutronics spatial discretization effects have stronger influence than time discretization effects, in the semi-implicit scheme adopted, on the numerical solution. On the other hand, the number of energy groups has an important influence on the transient evolution, whereas the assumption of using the prompt neutron spectrum for delayed neutrons is acceptable as it leads to small relative errors. (authors)

  13. Transient Astrophysics Probe

    Science.gov (United States)

    Camp, Jordan

    2017-08-01

    Transient Astrophysics Probe (TAP), selected by NASA for a funded Concept Study, is a wide-field high-energy transient mission proposed for flight starting in the late 2020s. TAP’s main science goals, called out as Frontier Discovery areas in the 2010 Decadal Survey, are time-domain astrophysics and counterparts of gravitational wave (GW) detections. The mission instruments include unique imaging soft X-ray optics that allow ~500 deg2 FoV in each of four separate modules; a high sensitivity, 1 deg2 FoV soft X-ray telescope based on single crystal silicon optics; a passively cooled, 1 deg2 FoV Infrared telescope with bandpass 0.6-3 micron; and a set of ~8 small NaI gamma-ray detectors. TAP will observe many events per year of X-ray transients related to compact objects, including tidal disruptions of stars, supernova shock breakouts, neutron star bursts and superbursts, and high redshift Gamma-Ray Bursts. Perhaps most exciting is TAP’s capability to observe X-ray and IR counterparts of GWs involving stellar mass black holes detected by LIGO/Virgo, and possibly X-ray counterparts of GWs from supermassive black holes, detected by LISA and Pulsar Timing Arrays.

  14. Qualitative analysis of the maintenance politics of the systems of a typical PWR by artificial neural networks; Analise qualitativa da politica de manutencoes dos sistemas de um PWR tipico por redes neurais artificiais

    Energy Technology Data Exchange (ETDEWEB)

    Lourenco, Victor Hugo Moreno

    2010-02-15

    Proceedings and techniques in order to maximize the reliability and the availability of industrial plants have been used along the last decades by specialists and professionals of maintenance. However, the modem industrial systems' sizing, and the increasing complexity and interdependence among its components have become this activity's planning a more and more difficult task. Considering this scenario, the objective of the present work is to provide a computational tool which is able to help about the taking decision's task, and about planning policies of maintenance practiced in thermonuclear plants. The tool developed is based on the artificial neural networks (ANN) for the recognition of standards and establishment of correlations among events occurred in the components of pressurized water reactor (PWR) typical systems. The ANN work as miners of database of failure events, and are able to identify connections and to establish imperceptible inferences even for the most experienced specialists in maintenance of nuclear systems. The results were attained from realistic data and are confronted against the maintenance's classic policies which are practiced nowadays on PWR thermonuclear plants. These results show the solidity of the technique in valuing and predicting failures in a real power plant, and is able to be used as a tool for supporting decisions about planning maintenance policies on a typical PWR. (author)

  15. PFM Analysis for Pre-Existing Cracks on Alloy 182 Weld in PWR Primary Water Environment using Monte Carlo Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Phil; Bahn, Chi Bum [Pusan National University, Busan (Korea, Republic of)

    2015-10-15

    Probabilistic Fracture Mechanics (PFM) analysis was generally used to consider the scatter and uncertainty of parameters in complex phenomenon. Weld defects could be present in weld regions of Pressurized Water Reactors (PWRs), which cannot be considered by the typical fracture mechanics analysis. It is necessary to evaluate the effects of the pre-existing cracks in welds for the integrity of the welds. In this paper, PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out using a Monte Carlo simulation. PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out. It was shown that inspection decreases the gradient of the failure probability. And failure probability caused by the pre-existing cracks was stabilized after 15 years of operation time in this input condition.

  16. Comparative analysis between measured and calculated concentrations of major actinides using destructive assay data from Ohi-2 PWR

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2015-09-01

    Full Text Available In the paper, we assess the accuracy of the Monte Carlo continuous energy burnup code (MCB in predicting final concentrations of major actinides in the spent nuclear fuel from commercial PWR. The Ohi-2 PWR irradiation experiment was chosen for the numerical reconstruction due to the availability of the final concentrations for eleven major actinides including five uranium isotopes (U-232, U-234, U-235, U-236, U-238 and six plutonium isotopes (Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242. The main results were presented as a calculated-to-experimental ratio (C/E for measured and calculated final actinide concentrations. The good agreement in the range of ±5% was obtained for 78% C/E factors (43 out of 55. The MCB modeling shows significant improvement compared with the results of previous studies conducted on the Ohi-2 experiment, which proves the reliability and accuracy of the developed methodology.

  17. Mechanism of hydrogen absorption during the exposure of alloy 600-like single-crystals to PWR primary simulated media

    Energy Technology Data Exchange (ETDEWEB)

    Jambon, F., E-mail: fanny.jambon@cea.fr [CEA, DEN, DPC, SCCME, Laboratoire d' Etude de la Corrosion Aqueuse, F-91191 Gif-Sur-Yvette Cedex (France); Marchetti, L. [CEA, DEN, DPC, SCCME, Laboratoire d' Etude de la Corrosion Aqueuse, F-91191 Gif-Sur-Yvette Cedex (France); Jomard, F. [GEMaC, CNRS Meudon, 1, Place Aristide Briand 92195 Meudon Cedex (France); Chene, J. [UMR CEA-CNRS no 8587 - CEA, DEN, DPC, SCCME, Laboratoire d' Etude de la Corrosion Aqueuse, F-91191 Gif-Sur-Yvette Cedex (France)

    2011-07-31

    The main purpose of this study is to investigate the mechanism responsible for the hydrogen absorption in alloy 600 exposed to pressurized water reactors primary water in order to evaluate the possible role of hydrogen in the stress corrosion cracking (SCC) of this alloy exposed to these environmental conditions. Alloy 600-like single-crystals were exposed to a simulated PWR medium where the hydrogen atoms of water or of the pressuring hydrogen gas were isotopically substituted with deuterium, used as a tracer. SIMS profiling of deuterium was used to characterize the deuterium absorption and localization in the passivated alloy. The results show that the hydrogen absorption during the exposure of the alloy to PWR primary water is associated with the water molecules dissociation during the built up of the passive film.

  18. Constraints on silicates formation in the Si-Al-Fe system: Application to hard deposits in steam generators of PWR nuclear reactors

    Science.gov (United States)

    Berger, Gilles; Million-Picallion, Lisa; Lefevre, Grégory; Delaunay, Sophie

    2015-04-01

    Introduction: The hydrothermal crystallization of silicates phases in the Si-Al-Fe system may lead to industrial constraints that can be encountered in the nuclear industry in at least two contexts: the geological repository for nuclear wastes and the formation of hard sludges in the steam generator of the PWR nuclear plants. In the first situation, the chemical reactions between the Fe-canister and the surrounding clays have been extensively studied in laboratory [1-7] and pilot experiments [8]. These studies demonstrated that the high reactivity of metallic iron leads to the formation of Fe-silicates, berthierine like, in a wide range of temperature. By contrast, the formation of deposits in the steam generators of PWR plants, called hard sludges, is a newer and less studied issue which can affect the reactor performance. Experiments: We present here a preliminary set of experiments reproducing the formation of hard sludges under conditions representative of the steam generator of PWR power plant: 275°C, diluted solutions maintained at low potential by hydrazine addition and at alkaline pH by low concentrations of amines and ammoniac. Magnetite, a corrosion by-product of the secondary circuit, is the source of iron while aqueous Si and Al, the major impurities in this system, are supplied either as trace elements in the circulating solution or by addition of amorphous silica and alumina when considering confined zones. The fluid chemistry is monitored by sampling aliquots of the solution. Eh and pH are continuously measured by hydrothermal Cormet© electrodes implanted in a titanium hydrothermal reactor. The transformation, or not, of the solid fraction was examined post-mortem. These experiments evidenced the role of Al colloids as precursor of cements composed of kaolinite and boehmite, and the passivation of amorphous silica (becoming unreactive) likely by sorption of aqueous iron. But no Fe-bearing was formed by contrast to many published studies on the Fe

  19. Characterization of highly transient EUV emitting discharges

    Energy Technology Data Exchange (ETDEWEB)

    Mullen, Joost van der; Kieft, Erik; Broks, Bart [Department of Applied Physics, Eindhoven University of Technology, PO Box 513, 5600 MB Eindhoven (Netherlands)

    2006-07-15

    The method of disturbed Bilateral Relations (dBR) is used to characterize highly transient plasmas that are used for the generation of Extreme Ultra Violet (EUV), i.e. radiation with a wavelength around 13.5 nm. This dBR method relates equilibrium disturbing to equilibrium restoring processes and follows the degree of equilibrium departure from the global down to the elementary plasma-level. The study gives global values of the electron density and electron temperature. Moreover, it gives a method to construct the atomic state distribution function (ASDF). This ASDF, which is responsible for the spectrum generated by the discharge, is found to be far from equilibrium. There are two reasons for this: first, systems with high charge numbers radiate strongly, second the highly transient behaviour makes that the distribution over the various ionization stages lags behind the temperature evolution.

  20. simulation of a SGTR severe PWR-W with MELCOR code; Simulacion de un SGTR severo en un PWR-W con el codigo Melcor

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, A. J.; Jimenez Varas, G.; Israelsson, L. C.

    2014-04-01

    Steam Generator tube Rupture (SGTR) is a small break loss of coolant accident. the issues related to this kind of transients makes them different from the classics LOCA studies. SGTR accidents in Pressurized Water Reactor are known to be one of the most demanding transients for the operating crew. It this accident is not managed in a proper way it could lead to steam generator overfill and a severe accident inside containment . To simulate this accident the MELCOR code was chosen, whose aim is the assessment of the progression of severe accidents in Light Water Reactors. (Author)

  1. The combined thermohydraulics-neutronics code TRAB-SMABRE for 3D plant transient and accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Jaakko, M.; Antti, D.; Hanna, R.; Timo, V. [VTT Processes, (Finland)

    2004-07-01

    TRAB-3D models the PWR and BWR reactor core using the two-group diffusion equations in homogenized fuel assembly geometry with a sophisticated nodal method. Thermohydraulics is described using four equation formulation. The stand-alone version of the code also describes thermohydraulics of the rest of the BWR circuit with one dimensional components. The SMABRE code models the thermohydraulics of light water reactors. The five equation formulation with the drift flux phase separation is modelling the two-phase behaviour. Conservation equations are solved for the phase mass, mixture momentum and phase energy. Additional equations are for the noncondensable part in gas and boron in liquid. The TRAB-3D and SMABRE codes have been coupled earlier by using the parallel coupling principle, where in the core section the 3-dimensional TRAB core, and the parallel channel coarse SMABRE core are solved in parallel, but the rest of the circulation system is solved with SMABRE. As a new development the internal coupling to meet new requirements for the PWR and BWR transient analyses is being realised. Both the circuit and core thermohydraulics are solved in SMABRE. The core thermohydraulics solution inside the core wide iterations is repeated to allow rapid power changes. These are the fast pressure changes, control rod ejection and ATWS. The numerical solution in SMABRE has been improved to allow full core simulation with separate flow channel for each fuel element of a BWR core. For the PWR plants the method is used as well by simulating the core by one-dimensional parallel channels. New development is needed for the open core calculation. (authors)

  2. SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

    Energy Technology Data Exchange (ETDEWEB)

    J.W. Davis

    1996-08-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

  3. IGENPRO knowledge-based digital system for process transient diagnostics and management

    Energy Technology Data Exchange (ETDEWEB)

    Morman, J.A.; Reifman, J.; Wei, T.Y.C. [Argonne National Lab., IL (United States)] [and others

    1997-12-31

    Verification and validation issues have been perceived as important factors in the large scale deployment of knowledge-based digital systems for plant transient diagnostics and management. Research and development (R&D) is being performed on the IGENPRO package to resolve knowledge base issues. The IGENPRO approach is to structure the knowledge bases on generic thermal-hydraulic (T-H) first principles and not use the conventional event-basis structure. This allows for generic comprehensive knowledge, relatively small knowledge bases and above all the possibility of T-H system/plant independence. To demonstrate concept feasibility the knowledge structure has been implemented in the diagnostic module PRODIAG. Promising laboratory testing results have been obtained using data from the full scope Braidwood PWR operator training simulator. This knowledge structure is now being implemented in the transient management module PROMANA to treat unanticipated events and the PROTREN module is being developed to process actual plant data. Achievement of the IGENPRO R&D goals should contribute to the acceptance of knowledge-based digital systems for transient diagnostics and management.

  4. Comparison of computational performance of GA and PSO optimization techniques when designing similar systems - Typical PWR core case

    Energy Technology Data Exchange (ETDEWEB)

    Souza Lima, Carlos A. [Instituto de Engenharia Nuclear - Divisao de Reatores/PPGIEN, Rua Helio de Almeida 75, Cidade Universitaria - Ilha do Fundao, P.O. Box: 68550 - Zip Code: 21941-972, Rio de Janeiro (Brazil); Instituto Politecnico, Universidade do Estado do Rio de Janeiro, Pos-Graduacao em Modelagem Computacional, Rua Alberto Rangel - s/n, Vila Nova, Nova Friburgo, Zip Code: 28630-050, Nova Friburgo (Brazil); Lapa, Celso Marcelo F.; Pereira, Claudio Marcio do N.A. [Instituto de Engenharia Nuclear - Divisao de Reatores/PPGIEN, Rua Helio de Almeida 75, Cidade Universitaria - Ilha do Fundao, P.O. Box: 68550 - Zip Code: 21941-972, Rio de Janeiro (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (INCT) (Brazil); Cunha, Joao J. da [Eletronuclear Eletrobras Termonuclear - Gerencia de Analise de Seguranca Nuclear, Rua da Candelaria, 65, 7 andar. Centro, Zip Code: 20091-906, Rio de Janeiro (Brazil); Alvim, Antonio Carlos M. [Universidade Federal do Rio de Janeiro, COPPE/Nuclear, Cidade Universitaria - Ilha do Fundao s/n, P.O.Box 68509 - Zip Code: 21945-970, Rio de Janeiro (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (INCT) (Brazil)

    2011-06-15

    Research highlights: > Performance of PSO and GA techniques applied to similar system design. > This work uses ANGRA1 (two loop PWR) core as a prototype. > Results indicate that PSO technique is more adequate than GA to solve this kind of problem. - Abstract: This paper compares the performance of two optimization techniques, particle swarm optimization (PSO) and genetic algorithm (GA) applied to the design a typical reduced scale two loop Pressurized Water Reactor (PWR) core, at full power in single phase forced circulation flow. This comparison aims at analyzing the performance in reaching the global optimum, considering that both heuristics are based on population search methods, that is, methods whose population (candidate solution set) evolve from one generation to the next using a combination of deterministic and probabilistic rules. The simulated PWR, similar to ANGRA 1 power plant, was used as a case example to compare the performance of PSO and GA. Results from simulations indicated that PSO is more adequate to solve this kind of problem.

  5. Evaluation of stress corrosion cracking of irradiated 304L stainless steel in PWR environment using heavy ion irradiation

    Science.gov (United States)

    Gupta, J.; Hure, J.; Tanguy, B.; Laffont, L.; Lafont, M.-C.; Andrieu, E.

    2016-08-01

    IASCC has been a major concern regarding the structural and functional integrity of core internals of PWR's, especially baffle-to-former bolts. Despite numerous studies over the past few decades, additional evaluation of the parameters influencing IASCC is still needed for an accurate understanding and modeling of this phenomenon. In this study, Fe irradiation at 450 °C was used to study the cracking susceptibility of 304 L austenitic stainless steel. After 10 MeV Fe irradiation to 5 dpa, irradiation-induced damage in the microstructure was characterized and quantified along with nano-hardness measurements. After 4% plastic strain in a PWR environment, quantitative information on the degree of strain localization, as determined by slip-line spacing, was obtained using SEM. Fe-irradiated material strained to 4% in a PWR environment exhibited crack initiation sites that were similar to those that occur in neutron- and proton-irradiated materials, which suggests that Fe irradiation may be a representative means for studying IASCC susceptibility. Fe-irradiated material subjected to 4% plastic strain in an inert argon environment did not exhibit any cracking, which suggests that localized deformation is not in itself sufficient for initiating cracking for the irradiation conditions used in this study.

  6. Comparative study of the contribution of various PWR spacer grid components to hydrodynamic and wall pressure characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharjee, Saptarshi, E-mail: saptarshi.bhattacharjee@outlook.com [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France); Ricciardi, Guillaume [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Viazzo, Stéphane [Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France)

    2017-06-15

    Highlights: • Complex geometry inside a PWR fuel assembly is simulated using simplified 3D models. • Structured meshes are generated as far as possible. • Fluctuating hydrodynamic and wall pressure field are analyzed using LES. • Comparative studies between square spacer grid, circular spacer grid and mixing vanes are presented. • Simulations are compared with experimental data. - Abstract: Flow-induced vibrations in a pressurized water reactor (PWR) core can cause fretting wear in fuel rods. These vibrations can compromise safety of a nuclear reactor. So, it is necessary to know the random fluctuating forces acting on the rods which cause these vibrations. In this paper, simplified 3D models like square spacer grid, circular spacer grid and symmetric mixing vanes have been used inside an annular pipe. Hydrodynamic and wall pressure characteristics are evaluated using large eddy simulations (LES). Structured meshes are generated as far as possible. Simulations are compared with an experiment. Results show that the grid and vanes have a combined effect: grid accelerates the flow whereas the vanes contribute to the swirl structures. Spectral analysis of the simulations illustrate vortex shedding phenomenon in the wake of spacer grids. This initial study opens up interesting perspectives towards improving the modeling strategy and understanding the complex phenomenon inside a PWR core.

  7. Transient thermohydraulic heat pipe modeling

    Science.gov (United States)

    Hall, Michael L.; Doster, Joseph M.

    Many space based reactor designs employ heat pipes as a means of conveying heat. In these designs, thermal radiation is the principle means for rejecting waste heat from the reactor system, making it desirable to operate at high temperatures. Lithium is generally the working fluid of choice as it undergoes a liquid-vapor transformation at the preferred operating temperature. The nature of remote startup, restart, and reaction to threats necessitates an accurate, detailed transient model of the heat pipe operation. A model is outlined of the vapor core region of the heat pipe which is part of a large model of the entire heat pipe thermal response. The vapor core is modeled using the area averaged Navier-Stokes equations in one dimension, which take into account the effects of mass, energy and momentum transfer. The core model is single phase (gaseous), but contains two components: lithium gas and a noncondensible vapor. The vapor core model consists of the continuity equations for the mixture and noncondensible, as well as mixture equations for internal energy and momentum.

  8. Familial Transient Global Amnesia

    Directory of Open Access Journals (Sweden)

    R.Rhys Davies

    2012-12-01

    Full Text Available Following an episode of typical transient global amnesia (TGA, a female patient reported similar clinical attacks in 2 maternal aunts. Prior reports of familial TGA are few, and no previous account of affected relatives more distant than siblings or parents was discovered in a literature survey. The aetiology of familial TGA is unknown. A pathophysiological mechanism akin to that in migraine attacks, comorbidity reported in a number of the examples of familial TGA, is one possibility. The study of familial TGA cases might facilitate the understanding of TGA aetiology.

  9. [Transient removable dentures].

    Science.gov (United States)

    Kouadio, A A; Jordana, F; N'Goran, J K; Le Bars, P

    2015-09-01

    Removable dentures are always transient current. The epidemiology and causes of tooth gaps demonstrate the need to master the different prosthetic treatment. This made whether to propose treatment plans that take into account psychological, physiological and technical support for this patient. Different situations may arise. A gradual transition may be considered or immediate passage to the total edentulous according to general criteria, local and desiderata of patients. After tooth extraction, the transitional prosthesis can control bone lysis thereby it is part of a complete treatment before prosthesis. It also facilitates a good psychological and physiological integration before the prosthesis use.

  10. Transient dimers of allergens.

    Directory of Open Access Journals (Sweden)

    Juha Rouvinen

    Full Text Available BACKGROUND: Allergen-mediated cross-linking of IgE antibodies bound to the FcepsilonRI receptors on the mast cell surface is the key feature of the type I allergy. If an allergen is a homodimer, its allergenicity is enhanced because it would only need one type of antibody, instead of two, for cross-linking. METHODOLOGY/PRINCIPAL FINDINGS: An analysis of 55 crystal structures of allergens showed that 80% of them exist in symmetric dimers or oligomers in crystals. The majority are transient dimers that are formed at high protein concentrations that are reached in cells by colocalization. Native mass spectrometric analysis showed that native allergens do indeed form transient dimers in solution, while hypoallergenic variants of them exist almost solely in the monomeric form. We created a monomeric Bos d 5 allergen and show that it has a reduced capability to induce histamine release. CONCLUSIONS/SIGNIFICANCE: The results suggest that dimerization would be a very common and essential feature for allergens. Thus, the preparation of purely monomeric variants of allergens could open up novel possibilities for specific immunotherapy.

  11. Transient regional osteoporosis

    Directory of Open Access Journals (Sweden)

    F. Trotta

    2011-09-01

    Full Text Available Transient osteoporosis of the hip and regional migratory osteoporosis are uncommon and probably underdiagnosed bone diseases characterized by pain and functional limitation mainly affecting weight-bearing joints of the lower limbs. These conditions are usually self-limiting and symptoms tend to abate within a few months without sequelae. Routine laboratory investigations are unremarkable. Middle aged men and women during the last months of pregnancy or in the immediate post-partum period are principally affected. Osteopenia with preservation of articular space and transitory edema of the bone marrow provided by magnetic resonance imaging are common to these two conditions, so they are also known by the term regional transitory osteoporosis. The appearance of bone marrow edema is not specific to regional transitory osteoporosis but can be observed in several diseases, i.e. trauma, reflex sympathetic dystrophy, avascular osteonecrosis, infections, tumors from which it must be differentiated. The etiology of this condition is unknown. Pathogenesis is still debated in particular the relationship with reflex sympathetic dystrophy, with which regional transitory osteoporosis is often identified. The purpose of the present review is to remark on the relationship between transient osteoporosis of the hip and regional migratory osteoporosis with particular attention to the bone marrow edema pattern and relative differential diagnosis.

  12. Transient Heat Transfer Model for Car Body Primer Curing

    OpenAIRE

    D. Zabala; N. Sánchez; J. Pinto

    2010-01-01

    A transient heat transfer mathematical model for the prediction of temperature distribution in the car body during primer baking has been developed by considering the thermal radiation and convection in the furnace chamber and transient heat conduction governing equations in the car framework. The car cockpit is considered like a structure with six flat plates, four vertical plates representing the car doors and the rear and front panels. The other two flat plates are the...

  13. 5 X 5 rod bundle flow field measurements downstream a PWR spacer grid

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Higor F.P.; Silva, Vitor V A.; Santos, André A.C.; Veloso, Maria A.F., E-mail: higorfabiano@gmail.com, E-mail: mdora@nuclear.ufmg.br, E-mail: vitors@cdtn.br, E-mail: aacs@cdtn.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil); Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The spacer grids are structures present in nuclear fuel assembly of Pressurized Water Reactors (PWR). They play an important structural role and also assist in heat removal through the assembly by promoting increased turbulence of the flow. Understanding the flow dynamics downstream the spacer grid is paramount for fuel element design and analysis. This paper presents water flow velocity profiles measurements downstream a spacer grid in a 5 x 5 rod bundle test rig with the objective of highlighting important fluid dynamic behavior near the grid and supplying data for CFD simulation validation. These velocity profiles were obtained at two different heights downstream the spacer grid using a LDV (Laser Doppler Velocimetry) through the top of test rig. The turbulence intensities and patterns of the swirl and cross flow were evaluated. The tests were conducted for Reynolds numbers ranging from 1.8 x 10{sup 4} to 5.4 x 10{sup 4}. This experimental research was carried out in thermo-hydraulics laboratory of Nuclear Technology Development Center – CDTN. Results show great repeatability and low uncertainties (< 1.24 %). Details of the flow field show how the mixture and turbulence induced by the spacer grid quickly decays downstream the spacer grid. It is shown that the developed methodology can supply high resolution low uncertainty results that can be used for validation of CFD simulations. (author)

  14. A stylized three dimensional PWR whole-core benchmark problem with Gadolinium

    Energy Technology Data Exchange (ETDEWEB)

    Douglass, Steven, E-mail: douglass.steven@gmail.co [Nuclear and Radiological Engineering/Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA 30332-0405 (United States); Rahnema, Farzad, E-mail: farzad@gatech.ed [Nuclear and Radiological Engineering/Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA 30332-0405 (United States); Margulies, Johann, E-mail: johann.margulies@me.gatech.ed [Nuclear and Radiological Engineering/Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA 30332-0405 (United States)

    2010-10-15

    Recent advancements in computer technology have led to an increased ability to solve complicated reactor problems, and as a result, there has been an interest in codes and methods that utilize whole-core transport in three-dimensional, large-scale, highly-heterogeneous configurations. Most current reactor benchmark problems are either small-scale or involve spatial homogenization (pin cell or full-assembly) for diffusion codes; therefore, new whole-core benchmark problems must be designed in order to validate whole-core transport methods. This study focuses on an PWR case, with a low leakage loading pattern and Gadolinium as a burnable absorber. In this paper, a detailed description of a new 3D whole-core numerical benchmark problem based on a 4-loop layout is presented with geometric, material and cross section specifications. A set of 2-group and 4-group region-wise neutron cross section libraries is generated using the lattice depletion code HELIOS, and an MCNP model is described for three configurations (all rods in, all rods out, power-shaping rods inserted). Eigenvalues are presented for the 2-group calculations of all three configurations, and detailed pin fission density results are presented for the power-shaping configuration.

  15. Assessment of Severe Accident Depressurization Valve Activation Strategy for Chinese Improved 1000 MWe PWR

    Directory of Open Access Journals (Sweden)

    Ge Shao

    2013-01-01

    Full Text Available To prevent HPME and DCH, SADV is proposed to be added to the pressurizer for Chinese improved 1000 MWe PWR NPP with the reference of EPR design. Rapid depressurization capability is assessed using the mechanical analytical code. Three typical severe accident sequences of TMLB’, SBLOCA, and LOFW are selected. It shows that with activation of the SADV the RCS pressure is low enough to prevent HPME and DCH. Natural circulation at upper RPV and hot leg is considered for the rapid depressurization capacity analysis. The result shows that natural circulation phenomenon results in heat transfer from the core to the pipes in RCS which may cause the creep rupture of pipes in RCS and delays the severe accident progression. Different SADV valve areas are investigated to the influence of depressurization of RCS. Analysis shows that the introduction of SADV with right valve area will delay progression of core degradation to RPV failure. Valve area is to be optimized since smaller SADV area will reduce its effect and too large valve area will lead to excessive loss of water inventory in RCS and makes core degradation progression to RPV failure faster without additional core cooling water sources.

  16. Verification of Cell-Homogenized Whole-Core Transport Calculations in Actual PWR Core Geometry

    Energy Technology Data Exchange (ETDEWEB)

    Akio Yamamoto; Masahiro Tatsumi; Tatsuya Kimoto; Shinya Kosaka; Etsuro Saji

    2000-11-12

    In this paper, a cell-homogenized whole-core transport calculation is applied to a pressurized water reactor (PWR) core and its prediction accuracy on reactivity and power distribution is verified in comparison with cell-heterogeneous whole-core transport calculation results obtained by the method of characteristics. Recent progress in common computer hardware promotes continuous improvements on core calculation methods. Among several different approaches, a three-dimensional pin-by-pin (fine-mesh) core calculation is considered as one of the candidates of advanced methods. In general, although the fine-mesh core calculation explicitly treats each individual pin, it still has three approximations: cell-homogenization, low-level transport theory (usually the SPN or the diffusion approximation) and few-group treatment. Therefore, the impact of these approximations on core characteristics should be evaluated to confirm the feasibility of the three-dimensional fine-mesh core calculation method. This paper focuses on two former approximations, and the impact of these approximations is evaluated.

  17. Development of a parametric containment event tree model of a severe PWR accident

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T. [OTO-Consulting Ay, Helsinki (Finland)

    1996-06-01

    The study supports the development project of STUK on `Living` PSA Level 2. The main work objective is to develop review tools for the Level 2 PSA studies underway at the utilities. The SPSA (STUK PSA) code is specifically designed for the purpose. In this work, SPSA is utilized as the Level 2 programming and calculation tool. A containment event tree (CET) model is built for analysis of severe accidents at the Loviisa pressurized water reactor (PWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to include new research results, and so it facilitates the Living PSA concept on Level 2 as well. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting at a low primary system pressure. Severe accident progression from five plant damage states (PDSs) is examined, however the integration with Level 1 is deferred to more definitive, integrated, safety assessments. (34 refs., 5 figs., 9 tabs.).

  18. Analysis of fission product transport under terminated LOCA conditions. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gieseke, J.A.; Baybutt, P.; Jordan, H.; Denning, R.S.; Wooton, R.O.

    1977-12-30

    An analytical model was developed to allow preditions of the source term to the containment as dependent on release from the fuel pins and deposition within the primary system. The model was developed into a flexible computer code adaptable to various geometrical arrangements and flow paths. The calculational framework was established in such a way to permit, in principle, the determination of particle and vapor transport and deposition in a general system of control volumes connected by fluid flow in an arbitrary way. This framework requires as input the rate of fission product release to the primary flow, as a function of time for each vapor and particulate species to be considered, and a complete, time dependent, thermal-hydraulic description of the system. TRAP-PWR, TRAP-BWR and TRACK computer codes are specializations, described in detail in the text of this report, of the general framework. The nature of these specializations depends strongly on the degree of detail with which the transporting fluid medium is modeled.

  19. LBB evaluation for a typical Japanese PWR primary loop by using the US NRC approved methods

    Energy Technology Data Exchange (ETDEWEB)

    Swamy, S.A.; Bhowmick, D.C.; Prager, D.E. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    1997-04-01

    The regulatory requirements for postulated pipe ruptures have changed significantly since the first nuclear plants were designed. The Leak-Before-Break (LBB) methodology is now accepted as a technically justifiable approach for eliminating postulation of double-ended guillotine breaks (DEGB) in high energy piping systems. The previous pipe rupture design requirements for nuclear power plant applications are responsible for all the numerous and massive pipe whip restraints and jet shields installed for each plant. This results in significant plant congestion, increased labor costs and radiation dosage for normal maintenance and inspection. Also the restraints increase the probability of interference between the piping and supporting structures during plant heatup, thereby potentially impacting overall plant reliability. The LBB approach to eliminate postulating ruptures in high energy piping systems is a significant improvement to former regulatory methodologies, and therefore, the LBB approach to design is gaining worldwide acceptance. However, the methods and criteria for LBB evaluation depend upon the policy of individual country and significant effort continues towards accomplishing uniformity on a global basis. In this paper the historical development of the U.S. LBB criteria will be traced and the results of an LBB evaluation for a typical Japanese PWR primary loop applying U.S. NRC approved methods will be presented. In addition, another approach using the Japanese LBB criteria will be shown and compared with the U.S. criteria. The comparison will be highlighted in this paper with detailed discussion.

  20. Regulatory Research of the PWR Severe Accident. Information Needs and Instrumentation for Hydrogen Control and Management

    Energy Technology Data Exchange (ETDEWEB)

    Park, Gun Chul; Suh, Kune Y.; Lee, Jin Yong; Lee, Seung Dong [Seoul Nat' l Univ., Seoul (Korea, Republic of)

    2001-03-15

    The current research is concerned with generation of basic engineering data needed in the process of developing hydrogen control guidelines as part of accident management strategies for domestic nuclear power plants and formulating pertinent regulatory requirements. Major focus is placed on identification of information needs and instrumentation methods for hydrogen control and management in the primary system and in the containment, development of decision-making trees for hydrogen management and their quantification, the instrument availability under severe accident conditions, critical review of relevant hydrogen generation model and phenomena In relation to hydrogen behavior, we analyzed the severe accident related hydrogen generation in the UCN 3{center_dot}4 PWR with modified hydrogen generation model. On the basis of the hydrogen mixing experiment and related GASFLOW calculation, the necessity of 3-dimensional analysis of the hydrogen mixing was investigated. We examined the hydrogen control models related to the PAR(Passive Autocatalytic Recombiner) and performed MAAP4 calculation in relation to the decision tree to estimate the capability and the role of the PAR during a severe accident.

  1. Determination of uncertainties of PWR spent fuel radionuclide inventory based on real operational history data

    Energy Technology Data Exchange (ETDEWEB)

    Fast, Ivan; Bosbach, Dirk [Institute of Energy- and Climate Research, Nuclear Waste Management and Reactor Safety Research, IEK-6, Forschungszentrum, Julich GmbH, (Germany); Aksyutina, Yuliya; Tietze-Jaensch, Holger [German Product Control Office for Radioactive Waste (PKS), Institute of Energy- and Climate Research, Nuclear Waste Management and Reactor Safety Research, IEK-6, Forschungszentrum Julich GmbH, (Germany)

    2015-07-01

    A requisite for the official approval of the safe final disposal of SNF is a comprehensive specification and declaration of the nuclear inventory in SNF by the waste supplier. In the verification process both the values of the radionuclide (RN) activities and their uncertainties are required. Burn-up (BU) calculations based on typical and generic reactor operational parameters do not encompass any possible uncertainties observed in real reactor operations. At the same time, the details of the irradiation history are often not well known, which complicates the assessment of declared RN inventories. Here, we have compiled a set of burnup calculations accounting for the operational history of 339 published or anonymized real PWR fuel assemblies (FA). These histories were used as a basis for a 'SRP analysis', to provide information about the range of the values of the associated secondary reactor parameters (SRP's). Hence, we can calculate the realistic variation or spectrum of RN inventories. SCALE 6.1 has been employed for the burn-up calculations. The results have been validated using experimental data from the online database - SFCOMPO-1 and -2. (authors)

  2. Validation of the scale system for PWR spent fuel isotopic composition analyses

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.; Bowman, S.M.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Laboratories, Las Vegas, NV (United States)

    1995-03-01

    The validity of the computation of pressurized-water-reactor (PWR) spent fuel isotopic composition by the SCALE system depletion analysis was assessed using data presented in the report. Radiochemical measurements and SCALE/SAS2H computations of depleted fuel isotopics were compared with 19 benchmark-problem samples from Calvert Cliffs Unit 1, H. B. Robinson Unit 2, and Obrigheim PWRs. Even though not exhaustive in scope, the validation included comparison of predicted and measured concentrations for 14 actinides and 37 fission and activation products. The basic method by which the SAS2H control module applies the neutron transport treatment and point-depletion methods of SCALE functional modules (XSDRNPM-S, NITAWL-II, BONAMI, and ORIGEN-S) is described in the report. Also, the reactor fuel design data, the operating histories, and the isotopic measurements for all cases are included in detail. The underlying radiochemical assays were conducted by the Materials Characterization. Center at Pacific Northwest Laboratory as part of the Approved Testing Material program and by four different laboratories in Europe on samples processed at the Karlsruhe Reprocessing Plant.

  3. HORIZONTAL LIFTING OF 5 DHLW/DOE LONG, 12-PWR LONG AND 24-BWR WASTE PACKAGES

    Energy Technology Data Exchange (ETDEWEB)

    V. de la Brosse

    2001-05-17

    The objective of this calculation was to determine the structural response of a 12-Pressurized Water Reactor (PWR) Long, a 24-Boiling Water Reactor (BWR) and a 5-Defense High Level Waste/Department of Energy (DHLW/DOE)--Long spent nuclear fuel waste packages lifted in a horizontal position. The scope of this calculation was limited to reporting the calculation results in terms of maximum stress intensities in the trunnion collar sleeves. In addition, the maximum stress intensities in the inner and outer shells of the waste packages were presented for illustrative purposes. The information provided by the sketches (Attachments I, II and III) is that of the potential design of the types of waste packages considered in this calculation, and all obtained results are valid for these designs only. This calculation is associated with the waste package design and was performed by the Waste Package Design Section in accordance with the ''Technical work plan for: Waste Package Design Description for LA'' (Ref. 7). AP-3.12Q, Calculations (Ref. 13), was used to perform the calculation and develop the document.

  4. Computer simulation of Angra-2 PWR nuclear reactor core using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Marcos P.C. de; Rebello, Wilson F., E-mail: eng.cavaliere@ime.eb.br, E-mail: rebello@ime.eb.br [Instituto Militar de Engenharia - Secao de Engenharia Nuclear, Rio de Janeiro, RJ (Brazil); Oliveira, Claudio L. [Universidade Gama Filho, Departamento de Matematica, Rio de Janeiro, RJ (Brazil); Vellozo, Sergio O., E-mail: vellozo@cbpf.br [Centro Tecnologico do Exercito. Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil); Silva, Ademir X. da, E-mail: ademir@nuclear.ufrj.br [Coordenacao dos Programas de Pos Gaduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    In this work the MCNPX (Monte Carlo N-Particle Transport Code) code was used to develop a computerized model of the core of Angra 2 PWR (Pressurized Water Reactor) nuclear reactor. The model was created without any kind of homogenization, but using real geometric information and material composition of that reactor, obtained from the FSAR (Final Safety Analysis Report). The model is still being improved and the version presented in this work is validated by comparing values calculated by MCNPX with results calculated by others means and presented on FSAR. This paper shows the results already obtained to K{sub eff} and K{infinity}, general parameters of the core, considering the reactor operating under stationary conditions of initial testing and operation. Other stationary operation conditions have been simulated and, in all tested cases, there was a close agreement between values calculated computationally through this model and data presented on the FSAR, which were obtained by other codes. This model is expected to become a valuable tool for many future applications. (author)

  5. Evaluation of radioactivity and gamma spectra in the secondary coolant system of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Zhaohuan, E-mail: lzhzuibang@sina.com [North China Electric Power University, Beijing 102206 (China); Chen, Yixue, E-mail: yxchen1972@126.com [North China Electric Power University, Beijing 102206 (China); Li, Lu; Song, Wen [North China Electric Power University, Beijing 102206 (China); Sun, Yeshuai [State Nuclear Power Technology Corporation, Beijing 100029 (China)

    2014-10-01

    Highlights: • The nuclear reaction data was extracted from ENDF/B-VII.1. • The benchmark was based on the data from ANSI/ANS-18.1-1999. • Mathematic models of the radionuclides generation and disappearance mechanism in the system were established. - Abstract: “Source Term” is the fundamental data used to evaluate the environmental impact of radioactive releases during normal operation. This paper presents a general investigation on the computational model of radiation source-term for the secondary coolant system of the pressurized water reactor (PWR). Research is carried out on the radionuclide migration, adsorption, retention, decay. Accordingly, mathematic models are described on the basis of the mechanism of radionuclides generation and disappearance in the system. Based on the implementation of these models, the corresponding function modules were developed and tested, which completes the source-term program previously developed. The nuclear reaction data of nuclides was extracted from the evaluated nuclear data library-ENDF/B-VII.1. The results obtained from preliminary verification between this work and ANSI/ANS-18, 1999 supported the models, indicating that the models could be used in the secondary coolant system for radiation shielding, accident prevention, environmental assessment and nuclear facility decommission.

  6. Wear performance and activity reduction effect of Co-free valves in PWR environment

    Energy Technology Data Exchange (ETDEWEB)

    Bahn, Chi Bum E-mail: bcb@snu.ac.kr; Han, Byoung Chan; Bum, Jin Sin; Hwang, Il Soon; Lee, Chan Bock

    2004-06-01

    Co-based hardfacing alloys exposed to PWR primary coolant may be replaced with Co-free alloys to lower occupational radiation exposure. To evaluate the viability of Co-free hardfacing alloys, we conducted hot-water tests for gate valves hardfaced with NOREM{sup TM} 02 (Fe-base), Deloro{sup TM} 50 (Ni-base), and Stellite{sup TM} 6 (Co-base). Using a high flow test loop, on-off cycling tests were conducted in 280 deg. C water. It was observed that NOREM 02 exhibited galling and excessive leak after 1000 cycle test whereas no leakage was developed with Deloro 50 after 2000 cycles. To estimate the activity reduction effect of Co-free hardfacing alloys, an existing activity transport model was modified. It is found that the main contributor of Co activity buildup is the corrosion of steam generator (SG) tubing. The Korean Next Generation Reactor (APR-1400) tubed with alloy 690 having a reduced cobalt impurity allowance is expected to have 73% lower Co activity on SG surface compared with the case of alloy 600 tubing. The complete replacement of Stellite 6 with Co-free hardfacing alloys is expected to cut additional 5% of activity which may be too small to justify the risk of galling and leakage development as revealed by the hot-water test.

  7. Spatial distribution of nanoparticles in PWR nanofluid coolant subjected to local nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Mirghaffari, Reza; Jahanfarnia, Gholamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering

    2016-12-15

    Nanofluids have shown to be promising as an alternative for a PWR reactor coolant or as a safety system coolant to cover the core in the event of a loss of coolant accident. The nanoparticles distribution and neutronic parameters are intensively affected by the local boiling of nanofluid coolant. The main goal of this study was the physical-mathematical modeling of the nanoparticles distribution in the nucleate boiling of nanofluids within the viscous sublayer. Nanoparticles concentration, especially near the heat transfer surfaces, plays a significant role in the enhancement of thermal conductivity of nanofluids and prediction of CHF, Hide Out and Return phenomena. By solving the equation of convection-diffusion for the liquid phase near the heating surface and the bulk stream, the effect of heat flux on the distribution of nanoparticles was studied. The steady state mass conservation equations for liquids, vapors and nanoparticles were written for the flow boiling within the viscous sublayer adjacent the fuel cladding surface. The derived differential equations were discretized by the finite difference method and were solved numerically. It was found out that by increasing the surface heat flux, the concentration of nanoparticles increased.

  8. Influence of Localized Plasticity on IASCC Sensitivity of Austenitic Stainless Steels under PWR Primary Water

    Science.gov (United States)

    Cissé, Sarata; Tanguy, Benoit; Laffont, Lydia; Lafont, Marie-Christine; Guerre, Catherine; Andrieu, Eric

    The sensibility of precipitation-strengthened A286 austenitic stainless steel to Stress Corrosion Cracking (SCC) is studied by means of Slow Strain Rate Tests (SSRT). First, alloy cold working by Low Cycle Fatigue (LCF) is investigated. Fatigue tests under plastic strain control are performed at different strain levels (Δ ɛp/2=0.2%, 0.5% and 0.8%) in order to establish correlation between stress softening and deformation microstructure resulting from LCF tests. Deformed microstructures have been identified through TEM investigations. Three states of cyclic behaviour for precipitation-strengthened A286 have been identified: hardening, cyclic softening and finally saturation of softening. It is shown that the A286 alloy cyclic softening is due to microstructural features such as defects — free deformation bands resulting from dislocations motion along family plans , that swept defects or γ' precipitates and lead to deformation localization. In order to quantify effects of plastic localized deformation on intergranular stress corrosion cracking (IGSCC) of the A286 alloy in PWR primary water, slow strain rate tests are conducted. For each cycling conditions, two specimens at a similar stress level are tested: the first containing free precipitate deformation bands, the other not significant of a localized deformation state. SSRT tests are still in progress.

  9. Very fast isotopic inventory and decay heat calculations in PWR spent nuclear fuel for industrial application

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, F.; Garzenne, C.; Cabrol, E. [EDF R D SINETICS, 92 - Clamart (France)

    2005-07-01

    We present a very fast method to calculate decay heat and isotopic inventory in PWR spent nuclear fuels for industrial application. This method was implemented in the new version of the STRAPONTIN code. Decay heat calculation is based on a mix of two classical approaches: isotopes inventory for longer cooling times and decay heat burst functions for shorter ones. STRAPONTIN gives accurate results for decay heat calculations for a wide range of cooling times, from 0.1 second to 3 million years after reactor shutdown. Mass and activity of 154 fission products and 26 actinides of interest for industrial applications are calculated in isotopic inventory. Comparisons between STRAPONTIN and the French reference codes system for the fuel cycle APOLLO2-DARWIN, developed and validated by the Cea, show a standard deviation of a few % for decay heat and inventory of major isotopes. It is coherent with the level of uncertainties associated to the qualification of DARWIN reference code. The computation time is at least 300 times faster with STRAPONTIN, for instance 0.5 second on a 1 GHz UltraSPARC IIIi processor, for a standard calculation of a 4.4% UO{sub x} with 32 cooling times ranging from 1 day to 2.5 million years. (authors)

  10. Risk-Informed External Hazards Analysis for Seismic and Flooding Phenomena for a Generic PWR

    Energy Technology Data Exchange (ETDEWEB)

    Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steve [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ma, Zhegang [Idaho National Lab. (INL), Idaho Falls, ID (United States); Spears, Bob [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kosbab, Ben [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-26

    This report describes the activities performed during the FY2017 for the US-DOE Light Water Reactor Sustainability Risk-Informed Safety Margin Characterization (LWRS-RISMC), Industry Application #2. The scope of Industry Application #2 is to deliver a risk-informed external hazards safety analysis for a representative nuclear power plant. Following the advancements occurred during the previous FYs (toolkits identification, models development), FY2017 focused on: increasing the level of realism of the analysis; improving the tools and the coupling methodologies. In particular the following objectives were achieved: calculation of buildings pounding and their effects on components seismic fragility; development of a SAPHIRE code PRA models for 3-loops Westinghouse PWR; set-up of a methodology for performing static-dynamic PRA coupling between SAPHIRE and EMRALD codes; coupling RELAP5-3D/RAVEN for performing Best-Estimate Plus Uncertainty analysis and automatic limit surface search; and execute sample calculations for demonstrating the capabilities of the toolkit in performing a risk-informed external hazards safety analyses.

  11. Fire protection of PWR-type nuclear power plants; Protection incendie des centrales nucleaires REP

    Energy Technology Data Exchange (ETDEWEB)

    Kaercher, M. [Electricite de France (EDF), 75 - Paris (France)

    2004-07-01

    This article presents the fire protection principles applied by Electricite de France (EdF) to PWR-type nuclear power plants, and stresses on the recent evolutions of the design and exploitation approach of fire protection which have contributed to the improvement of nuclear facilities safety: 1 - general presentation; 2 - objectives of the fire protection: integrity of safety systems, personnel and people protection, equipments protection and keeping up of availability; 3 - design requirements: fire prevention, fire confinement, fire mastery, checking approach, formalization of the design approach, experience feedback of the design approach; 4 - example of application, the Chooz B power plant, head of the N4 step series: common dispositions to the different buildings, protection of the reactor building, protection of electrical facilities, protection of the turbine building, protection of transformers, protection of the emergency power sets, protection of hydrogen hazard buildings; 5 - in-service protection: intervention of the operator, fire fighting organization, reactor operation during fire accident; 6 - conclusion. (J.S.)

  12. Inference of core barrel motion from neutron noise spectral density. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, J.C.; Shahrokhi, F.; Kryter, R.C.

    1977-03-15

    A method was developed for inference of core barrel motion from the following statistical descriptors: cross-power spectral density, autopower spectral density, and amplitude probability density. To quantify the core barrel motion in a typical pressurized water reactor (PWR), a scale factor was calculated in both one- and two-dimensional geometries using forward, variational, and perturbation methods of discrete ordinates neutron transport. A procedure for selection of the proper frequency band limits for the statistical descriptors was developed. It was found that although perturbation theory is adequate for the calculation of the scale factor, two-dimensional geometric effects are important enough to rule out the use of a one-dimensional approximation for all but the crudest calculations. It was also found that contributions of gamma rays can be ignored and that the results are relatively insensitive to the cross-section set employed. The proper frequency band for the statistical descriptors is conveniently determined from the coherence and phase information from two ex-core power range neutron monitors positioned diametrically across the reactor vessel. Core barrel motion can then be quantified from the integral of the band-limited cross-power spectral density of two diametrically opposed ex-core monitors or, if the coherence between the pair is greater than or equal to 0.7, from a properly band-limited amplitude probability density function. Wide-band amplitude probability density functions were demonstrated to yield erroneous estimates for the magnitude of core barrel motion.

  13. Low cycle fatigue of Alloy 690 and welds in a simulated PWR primary water environment

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jongdae; Cho, Pyungyeon; Jang, Changheui [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Cho, Pyungyeon [Khalifa Univ., Abu Dhabi (United Arab Emirates); Kim, Tae Soon; Lee, Yong Sung [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2013-05-15

    In this study, environmental fatigue tests for these materials were performed and the new prediction model of fatigue life of Alloy 690 and weld in primary water condition was proposed. To evaluate the fatigue life of Alloy 690 and 52M in a PWR environment, low cycle fatigue tests were performed and revised fatigue life prediction models and environmental factor were proposed. With the revised Fen model for Alloy 690 and 52M, the reliability of the fatigue life prediction has been improved. The reduction of low cycle fatigue life of metallic materials in the primary coolant water environments has been the subject of debate between the utility and regulator since 1980s. It became the significant licensing problem since the issue of RG-1.207 by U. S. NRC. The statistical model for the environmental factor, Fen, specified in RG-1.207 was based on the extensive test results accumulated by the ANL and Japanese national program. Of the materials, the limited fatigue life data of Ni-Cr-Fe alloys were used to develop the Fen for the alloys. Furthermore, test data for Alloy 690 and its weld are limited. Considering that Alloy 690 will be extensively used in the new nuclear power plants, additional effort to validate or improve current Fen model is required.

  14. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

    Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  15. Precursor evolution and SCC initiation of cold-worked alloy 690 in simulated PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Ziqing; Kruska, Karen; Toloczko, Mychailo B.; Bruemmer, Stephen M.

    2017-03-27

    Stress corrosion crack initiation of two thermally-treated, cold-worked (CW) alloy 690 materials was investigated in 360oC simulated PWR primary water using constant load tensile (CLT) tests and blunt notch compact tension (BNCT) tests equipped with direct current potential drop (DCPD) for in-situ detection of cracking. SCC initiation was not detected by DCPD for the 21% and 31%CW CLT specimens loaded at their yield stress after ~9,220 h, however intergranular (IG) precursor damage and isolated surface cracks were observed on the specimens. The two 31%CW BNCT specimens loaded at moderate stress intensity after several cyclic loading ramps showed DCPD-indicated crack initiation after 10,400h exposure at constant stress intensity, which resulted from significant growth of IG cracks. The 21%CW BNCT specimens only exhibited isolated small IG surface cracks and showed no apparent DCPD change throughout the test. Interestingly, post-test cross-section examinations revealed many grain boundary (GB) nano-cavities in the bulk of all the CLT and BNCT specimens particularly for the 31%CW materials. Cavities were also found along GBs extending to the surface suggesting an important role in crack nucleation. This paper provides an overview of the evolution of GB cavities and will discuss their effects on crack initiation in CW alloy 690.

  16. Transient Go: A Mobile App for Transient Astronomy Outreach

    Science.gov (United States)

    Crichton, D.; Mahabal, A.; Djorgovski, S. G.; Drake, A.; Early, J.; Ivezic, Z.; Jacoby, S.; Kanbur, S.

    2016-12-01

    Augmented Reality (AR) is set to revolutionize human interaction with the real world as demonstrated by the phenomenal success of `Pokemon Go'. That very technology can be used to rekindle the interest in science at the school level. We are in the process of developing a prototype app based on sky maps that will use AR to introduce different classes of astronomical transients to students as they are discovered i.e. in real-time. This will involve transient streams from surveys such as the Catalina Real-time Transient Survey (CRTS) today and the Large Synoptic Survey Telescope (LSST) in the near future. The transient streams will be combined with archival and latest image cut-outs and other auxiliary data as well as historical and statistical perspectives on each of the transient types being served. Such an app could easily be adapted to work with various NASA missions and NSF projects to enrich the student experience.

  17. Avaliação da suscetibilidade à corrosão sob tensão da ZAC do aço inoxidável AISI 316L em ambiente de reator nuclear PWR Stress corrosion cracking of stainless steel AISI 316L HAZ in PWR Nuclear reactor environment

    Directory of Open Access Journals (Sweden)

    Mônica Maria de Abreu Mendonça Schvartzman

    2009-09-01

    Full Text Available Aços carbono de baixa liga e aços inoxidáveis são amplamente utilizados nos circuitos primários de reatores nucleares do tipo PWR (Pressurized Water Reactor. Ligas de níquel são empregadas na soldagem destes materiais devido a características como elevadas resistências mecânica e à corrosão, coeficiente de expansão térmica adequado, etc. Nos últimos 30 anos, a corrosão sob tensão (CST tem sido observada principalmente nas regiões das soldas entre materiais dissimilares existentes nestes reatores. Este trabalho teve como objetivo avaliar, por comparação, a suscetibilidade à corrosão sob tensão da zona afetada pelo calor (ZAC do aço inoxidável austenítico AISI 316L quando submetida a um ambiente similar ao do circuito primário de um reator nuclear PWR nas temperaturas de 303ºC e 325ºC. Para esta avaliação empregou-se o ensaio de taxa de deformação lenta - SSRT (Slow Strain Rate Test. Os resultados indicaram que a CST é ativada termicamente e que a 325ºC pode-se observar a presença mais significativa de fratura frágil decorrente do processo de corrosão sob tensão.In pressurized water reactors (PWRs, low alloy carbon steels and stainless steel are widely used in the primary water circuits. In most cases, Ni alloys are used to joint these materials and form dissimilar welds. These alloys are known to accommodate the differences in composition and thermal expansion of the two materials. Stress corrosion cracking of metals and alloys is caused by synergistic effects of environment, material condition and stress. Over the last thirty years, CST has been observed in dissimilar metal welds. This study presents a comparative work between the CST in the HAZ (Heat Affected Zone of the AISI 316L in two different temperatures (303ºC and 325ºC. The susceptibility to stress corrosion cracking was assessed using the slow strain rate tensile (SSRT test. The results of the SSRT tests indicated that CST is a thermally

  18. The safety analysis and thermohydraulic methodologies for the power updating analyses in Spanish PWR plants; Methodologias de diseno termohidraulico y de analisis de seguridad en los aumentos de potencia de centrales PWR

    Energy Technology Data Exchange (ETDEWEB)

    Salesa, F.

    2014-02-01

    This article describes the Safety Analysis and Thermohydraulic methodologies used by ENUSA for the Power Updating analyses in Spanish PWR plants of Westinghouse design: Design tools have been developed over the first cycles resulting new correlations of DNB, fitted to the new fuel assemblies, new DNBR calculation methodology and other improvements in the design areas. Using these methodologies, the available margins between design and limit values are wider. These new margins have allowed to accomplish the design criteria under the new power updating operational conditions. (Author)

  19. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes; Calculo de fuerzas laterales hidraulicas en elementos combustibles tipo PWR con codigos de dinamica de fluidos coputacional

    Energy Technology Data Exchange (ETDEWEB)

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-08-01

    To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

  20. Study on the structural integrity of a PWR vessel according to the ductile cracking instability theory; Avaliacao da integridade estrutural de um vaso PWR segundo a teoria da instabilidade do rasgamento ductil

    Energy Technology Data Exchange (ETDEWEB)

    Tarpani, Jose Ricardo; Spinelli, Dirceu [Sao Paulo Univ., Sao Carlos, SP (Brazil). Escola de Engenharia. Dept. de Engenharia de Materiais

    1996-12-31

    Analytical predictions of PWR vessel instability have been made according to Linear Elastic criterion K{sub lC} and Elastic-Plastic Fracture Mechanics Ji, J{sub 50} and J {sub inst}. Special attention was given to the influence of crack depth and length on both approaches, as well as to the data extrapolation procedures on elasto-plasticity. Simple and didactic format has been supplied for the results evaluation in terms of internal pressure, wall-through strain gradient and ductile stable crack extension. (author) 4 refs., 4 figs., 1 tab.

  1. Application of mini-core of a PWR within the framework benchmark of the OECD Uncertainty Analysis in Modelling; Aplicacion del mininucleo de un PWR dentro del marco del benchmark de la OCDE Uncertainty Analysis in Modelling

    Energy Technology Data Exchange (ETDEWEB)

    Arenas Moreno, C.; Reventos Puigjaner, F.; Ivanov, K.

    2013-07-01

    This work studies the effect that produces the homogenization and condensation of the effective sections on jointly in the propagation of uncertainties of the effective sections the neutronic calculation of reactor. Discusses a mini nucleo PWR type according to specifications of the LWR UAM OECD benchmark. Applies the calculation sequence SCALE6.1/TSUNAMI-2D, based on the theory of generalized disturbance (GPT), on the mini nucleo complete, and compared with the result obtained using the two-step method, which combines the GPT with statistical random sampling of the effective sections at the level of the fuel element. Shown that uncertainty in keff increases.

  2. Frequency-Domain Transient Imaging.

    Science.gov (United States)

    Jingyu Lin; Yebin Liu; Jinli Suo; Qionghai Dai

    2017-05-01

    A transient image is the optical impulse response of a scene, which also visualizes the propagation of light during an ultra-short time interval. In contrast to the previous transient imaging which samples in the time domain using an ultra-fast imaging system, this paper proposes transient imaging in the frequency domain using a multi-frequency time-of-flight (ToF) camera. Our analysis reveals the Fourier relationship between transient images and the measurements of a multi-frequency ToF camera, and identifies the causes of the systematic error-non-sinusoidal and frequency-varying waveforms and limited frequency range of the modulation signal. Based on the analysis we propose a novel framework of frequency-domain transient imaging. By removing the systematic error and exploiting the harmonic components inside the measurements, we achieves high quality reconstruction results. Moreover, our technique significantly reduces the computational cost of ToF camera based transient image reconstruction, especially reduces the memory usage, such that it is feasible for the reconstruction of transient images at extremely small time steps. The effectiveness of frequency-domain transient imaging is tested on synthetic data, real data from the web, and real data acquired by our prototype camera.

  3. Pressure transients in pipeline systems

    DEFF Research Database (Denmark)

    Voigt, Kristian

    1998-01-01

    This text is to give an overview of the necessary background to do investigation of pressure transients via simulations. It will describe briefly the Method of Characteristics which is the defacto standard for simulating pressure transients. Much of the text has been adopted from the book Pressure...

  4. Development of a high sensitivity Cerenkov viewing device. Field test at the Ringhals 2 PWR facility, Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Trepte, O.; Hildingsson, L. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Chen, J.D.; Burton, G.R.; Young, G.J.; Attas, E.M. [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs

    1996-08-01

    Within the joint Swedish and Canadian Safeguards Support Program, a new concept for a Cerenkov viewing device (CVD), based on a scientific charge coupled device (SCCD), has been evaluated. A feasibility test has been performed using a commercially-available system. With this system, images of Cerenkov light originating from PWR fuel assemblies have been recorded at the Ringhals 2 facility in Vaeroebacka, Sweden. This test is a continuation of feasibility study complimenting earlier tests performed on BWR fuel. During the present tests, images were taken of Cerenkov light from PWR fuel assemblies of varying burnup and cooling time, including assemblies with stopper and insert structures. In addition, Cerenkov glow images from a high-density, and a skeleton non-fuel assembly were obtained. Pseudo-colour images were produced from the digitally recorded images to provide an additional means of assessing the distribution of light emitted over the assembly. Analysis of a series of images taken as the instrument was scanned over a spent fuel assembly has provided the first quantitative demonstration of the collimation effect. The variation of image quality as a function of frame rate was determined to assess the maximum achievable frame rate with the new concept system. A performance comparison of the new SCCD-based CVD with the Mark IV CVD was made. The new concept system shows superior image quality compared to the Mark IV CVD, and allows detection of new Cerenkov characteristics previously not detected on spent fuel. For example, the centre of the PWR fuel assembly is brighter than the edges of the assembly, and the high resolution, achieved with the new instrument, has permitted observation of the spacer grid structure below the top plate of the assembly. 5 refs.

  5. TREAT Transient Analysis Benchmarking for the HEU Core

    Energy Technology Data Exchange (ETDEWEB)

    Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-05-01

    This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to benchmark the transient calculations against temperature-limited transients performed in the final operating HEU TREAT core configuration. The MCNP code was used to evaluate steady-state neutronics behavior, and the point kinetics code TREKIN was used to determine core power and energy during transients. The first part of the benchmarking process was to calculate with MCNP all the neutronic parameters required by TREKIN to simulate the transients: the transient rod-bank worth, the prompt neutron generation lifetime, the temperature reactivity feedback as a function of total core energy, and the core-average temperature and peak temperature as a functions of total core energy. The results of these calculations were compared against measurements or against reported values as documented in the available TREAT reports. The heating of the fuel was simulated as an adiabatic process. The reported values were extracted from ANL reports, intra-laboratory memos and experiment logsheets and in some cases it was not clear if the values were based on measurements, on calculations or a combination of both. Therefore, it was decided to use the term “reported” values when referring to such data. The methods and results from the HEU core transient analyses will be used for the potential LEU core configurations to predict the converted (LEU) core’s performance.

  6. Counter-current flow limitation at the junction between the surge line and the pressurizer of a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Futatsugi, T.; Nariai, T.; Hayashi, K.; Tomiyama, A., E-mail: futatsugi@cfrg.scitec.kobe-u.ac.jp, E-mail: nariai@cfrg.scitec.kobe-u.ac.jp, E-mail: hayashi@mech.kobe-u.ac.jp, E-mail: tomiyama@mech.kobe-u.ac.jp [Kobe Univ., Graduate School of Engineering (Japan); Yanagi, C.; Murase, M., E-mail: yanagi.chihiro@inss.co.jp, E-mail: murase@inss.co.jp [Inst. of Nuclear Safety System, Inc. (Japan)

    2011-07-01

    An experimental study on counter-current flow limitation (CCFL) in vertical pipes is carried out to understand CCFL in the surge line in a PWR. Several upper tanks corresponding to the pressurizer and a lower tank are used to investigate effects of tank geometry and water levels in the tanks on CCFL characteristics. The experimental data clearly show that CCFL depends on the tank geometry and the water level, and therefore, these factors must be taken into account when modelling characteristics of CCFL in vertical pipes. (author)

  7. Burn-up credit in criticality safety of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mahmoud, Rowayda F., E-mail: Rowayda_mahmoud@yahoo.com [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Shaat, Mohamed K. [Nuclear Engineering, Reactors Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Nagy, M.E.; Agamy, S.A. [Professor of Nuclear Engineering, Nuclear and Radiation Department, Alexandria University (Egypt); Abdelrahman, Adel A. [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt)

    2014-12-15

    Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B{sub 4}C and Ag–In–Cd” in racks. • Sub-criticality was confirmed for fresh and burned fuel under specific cases. • Studies for BU credit recommend increasing fuel burn-up to 60.0 GWD/MTU. • Those studies require new core structure materials, fuel composition and cladding. - Abstract: The criticality safety calculations were performed for a proposed design of a wet spent fuel storage pool. This pool will be used for the storage of spent fuel discharged from a typical pressurized water reactor (PWR). The mathematical model based on the international validated codes, WIMS-5 and MCNP-5 were used for calculating the effective multiplication factor, k{sub eff}, for the spent fuel stored in the pool. The data library for the multi-group neutron microscopic cross-sections was used for the cell calculations. The k{sub eff} was calculated for several changes in water density, water level, assembly pitch and burn-up with different initial fuel enrichment and new types and amounts of fixed absorbers. Also, k{sub eff} was calculated for the conservative fresh fuel case. The results of the calculations confirmed that the effective multiplication factor for the spent fuel storage is sub-critical for all normal and abnormal states. The future strategy for the burn-up credit recommends increasing the fuel burn-up to a value >60.0 GWD/MTU, which requires new fuel composition and new fuel cladding material with the assessment of the effects of negative reactivity build up.

  8. On-line PWR RHR pump performance testing following motor and impeller replacement

    Energy Technology Data Exchange (ETDEWEB)

    DiMarzo, J.T.

    1996-12-01

    On-line maintenance and replacement of safety-related pumps requires the performance of an inservice test to determine and confirm the operational readiness of the pumps. In 1995, major maintenance was performed on two Pressurized Water Reactor (PWR) Residual Heat Removal (RHR) Pumps. A refurbished spare motor was overhauled with a new mechanical seal, new motor bearings and equipped with pump`s `B` impeller. The spare was installed into the `B` train. The motor had never been run in the system before. A pump performance test was developed to verify it`s operational readiness and determine the in-situ pump performance curve. Since the unit was operating, emphasis was placed on conducting a highly accurate pump performance test that would ensure that it satisfied the NSSS vendors accident analysis minimum acceptance curve. The design of the RHR System allowed testing of one train while the other was aligned for normal operation. A test flow path was established from the Refueling Water Storage Tank (RWST) through the pump (under test) and back to the RWST. This allowed staff to conduct a full flow range pump performance test. Each train was analyzed and an expression developed that included an error vector term for the TDH (ft), pressure (psig), and flow rate (gpm) using the variance error vector methodology. This method allowed the engineers to select a test instrumentation system that would yield accurate readings and minimal measurement errors, for data taken in the measurement of TDH (P,Q) versus Pump Flow Rate (Q). Test results for the `B` Train showed performance well in excess of the minimum required. The motor that was originally in the `B` train was similarly overhauled and equipped with `A` pump`s original impeller, re-installed in the `A` train, and tested. Analysis of the `A` train results indicate that the RHR pump`s performance was also well in excess of the vendors requirements.

  9. Applied hydraulic transients

    CERN Document Server

    Chaudhry, M Hanif

    2014-01-01

    This book covers hydraulic transients in a comprehensive and systematic manner from introduction to advanced level and presents various methods of analysis for computer solution. The field of application of the book is very broad and diverse and covers areas such as hydroelectric projects, pumped storage schemes, water-supply systems, cooling-water systems, oil pipelines and industrial piping systems. Strong emphasis is given to practical applications, including several case studies, problems of applied nature, and design criteria. This will help design engineers and introduce students to real-life projects. This book also: ·         Presents modern methods of analysis suitable for computer analysis, such as the method of characteristics, explicit and implicit finite-difference methods and matrix methods ·         Includes case studies of actual projects ·         Provides extensive and complete treatment of governed hydraulic turbines ·         Presents design charts, desi...

  10. Steady- and Transient-State Analyses of Fully Ceramic Microencapsulated Fuel with Randomly Dispersed Tristructural Isotropic Particles via Two-Temperature Homogenized Model—I: Theory and Method

    Directory of Open Access Journals (Sweden)

    Yoonhee Lee

    2016-06-01

    Full Text Available As a type of accident-tolerant fuel, fully ceramic microencapsulated (FCM fuel was proposed after the Fukushima accident in Japan. The FCM fuel consists of tristructural isotropic particles randomly dispersed in a silicon carbide (SiC matrix. For a fuel element with such high heterogeneity, we have proposed a two-temperature homogenized model using the particle transport Monte Carlo method for the heat conduction problem. This model distinguishes between fuel-kernel and SiC matrix temperatures. Moreover, the obtained temperature profiles are more realistic than those of other models. In Part I of the paper, homogenized parameters for the FCM fuel in which tristructural isotropic particles are randomly dispersed in the fine lattice stochastic structure are obtained by (1 matching steady-state analytic solutions of the model with the results of particle transport Monte Carlo method for heat conduction problems, and (2 preserving total enthalpies in fuel kernels and SiC matrix. The homogenized parameters have two desirable properties: (1 they are insensitive to boundary conditions such as coolant bulk temperatures and thickness of cladding, and (2 they are independent of operating power density. By performing the Monte Carlo calculations with the temperature-dependent thermal properties of the constituent materials of the FCM fuel, temperature-dependent homogenized parameters are obtained.

  11. [Transient global amnesia after general anaesthesia].

    Science.gov (United States)

    Galipienzo, J; Lablanca, M S; Zannin, I; Rosado, R; Zarza, B; Olarra, J

    2012-01-01

    Transient global amnesia is a neurological syndrome in which there is a sudden and brief inability to form new memories, as well as an intense retrograde amnesia. However, awareness, personal identity and attention remain intact. It is an uncommon condition seen after an anaesthetic procedure. There are several aetiopathogenic hypotheses (epileptic, migrainous or ischaemic origin) and triggering factors (pain, anxiety, temperature changes, exercise, Valsalva manoeuvres, diagnostic tests or certain drugs). We describe the case of a patient with a high level of pre-operative anxiety who suffered an episode of transient global amnesia after undergoing otolaryngology surgery. With an acute and continued amnesia after general anaesthesia, the first thing that must be done is to establish a suitable differencial diagnosis, which should include transient global amnesia, as this is mainly an exclusion diagnosis. Preoperative anxiety may be a triggering factor to take into account in this condition, with anxiolytic treatment prior to the surgery being important. Copyright © 2010 Sociedad Española de Anestesiología, Reanimación y Terapéutica del Dolor. Published by Elsevier España. All rights reserved.

  12. Electromagnetic transients in power cables

    CERN Document Server

    da Silva, Filipe Faria

    2013-01-01

    From the more basic concepts to the most advanced ones where long and laborious simulation models are required, Electromagnetic Transients in Power Cables provides a thorough insight into the study of electromagnetic transients and underground power cables. Explanations and demonstrations of different electromagnetic transient phenomena are provided, from simple lumped-parameter circuits to complex cable-based high voltage networks, as well as instructions on how to model the cables.Supported throughout by illustrations, circuit diagrams and simulation results, each chapter contains exercises,

  13. Machine Classification of Transient Images

    Science.gov (United States)

    Buisson, Lise Du; Sivanandam, Navin; Bassett, Bruce A.; Smith, Mathew

    2014-05-01

    Using transient imaging data from the 2nd and 3rd years of the SDSS supernova survey, we apply various machine learning techniques to the problem of classifying transients (e.g. SNe) from artefacts, one of the first steps in any transient detection pipeline, and one that is often still carried out by human scanners. Using features mostly obtained from PCA, we show that we can match human levels of classification success, and find that a K-nearest neighbours algorithm and SkyNet perform best, while the Naive Bayes, SVM and minimum error classifier have performances varying from slightly to significantly worse.

  14. Near interface traps in SiO{sub 2}/4H-SiC metal-oxide-semiconductor field effect transistors monitored by temperature dependent gate current transient measurements

    Energy Technology Data Exchange (ETDEWEB)

    Fiorenza, Patrick; La Magna, Antonino; Vivona, Marilena; Roccaforte, Fabrizio [Consiglio Nazionale delle Ricerche-Istituto per la Microelettronica e Microsistemi (CNR-IMM), Strada VIII 5, Zona Industriale 95121 Catania (Italy)

    2016-07-04

    This letter reports on the impact of gate oxide trapping states on the conduction mechanisms in SiO{sub 2}/4H-SiC metal-oxide-semiconductor field effect transistors (MOSFETs). The phenomena were studied by gate current transient measurements, performed on n-channel MOSFETs operated in “gate-controlled-diode” configuration. The measurements revealed an anomalous non-steady conduction under negative bias (V{sub G} > |20 V|) through the SiO{sub 2}/4H-SiC interface. The phenomenon was explained by the coexistence of a electron variable range hopping and a hole Fowler-Nordheim (FN) tunnelling. A semi-empirical modified FN model with a time-depended electric field is used to estimate the near interface traps in the gate oxide (N{sub trap} ∼ 2 × 10{sup 11} cm{sup −2}).

  15. TRANSIENT ELECTRONICS CATEGORIZATION

    Science.gov (United States)

    2017-08-24

    methods. Chemical reactions often require elevated temperatures and strong acids or alkaline solutions, which must be considered in developing transience...COPY AIR FORCE RESEARCH LABORATORY SENSORS DIRECTORATE WRIGHT-PATTERSON AIR FORCE BASE , OH 45433-7320 AIR FORCE MATERIEL COMMAND UNITED STATES AIR...patented invention that may relate to them. This report was cleared for public release by the USAF 88th Air Base Wing (88 ABW) Public Affairs Office

  16. Air ingression calculations for selected plant transients using MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Kmetyk, L.N.

    1994-01-01

    Two sets of MELCOR calculations have been completed studying the effects of air ingression on the consequences of various severe accident scenarios. One set of calculations analyzed a station blackout with surge line failure prior to vessel breach, starting from nominal operating conditions; the other set of calculations analyzed a station blackout occurring during shutdown (refueling) conditions. Both sets of analyses were for the Surry plant, a three-loop Westinghouse PWR. For both accident scenarios, a basecase calculation was done, and then repeated with air ingression from containment into the core region following core degradation and vessel failure. In addition to the two sets of analyses done for this program, a similar air-ingression sensitivity study was done as part of a low-power/shutdown PRA, with results summarized here; that PRA study also analyzed a station blackout occurring during shutdown (refueling) conditions, but for the Grand Gulf plant, a BWR/6 with Mark III containment. These studies help quantify the amount of air that would have to enter the core region to have a significant impact on the severe accident scenario, and demonstrate that one effect, of air ingression is substantial enhancement of ruthenium release. These calculations also show that, while the core clad temperatures rise more quickly due to oxidation with air rather than steam, the core also degrades and relocates more quickly, so that no sustained, enhanced core heatup is predicted to occur with air ingression.

  17. Fission product transport analysis: Task 2. Quarterly progress report, January--March 1976. [PWR; spent fuel casks

    Energy Technology Data Exchange (ETDEWEB)

    Gieseke, J.A.; Baybutt, P.; Denning, R.S.; Jordan, H.; Wooton, W.O.

    1976-06-09

    The escape of fission product activity from a water-cooled reactor under loss of coolant accident (LOCA) conditions is studied. It is the goal of the study to develop a primary system fission product transport and deposition model which is consistent with the emerging state of knowledge regarding fission product release from fuel rods and thermal-hydraulic analyses of emergency core cooling (ECC). The analyses to be developed will provide methods for more realistically calculating the rate and magnitude of fission product release to the containment. An analogous fission product transport and deposition model is also to be developed for spent fuel shipping casks under hypothetical LOCA conditions. Progress during this quarter was primarily in those areas associated with efforts to organize and initiate this study. Technical efforts were initiated (1) to develop the format to be used in the model analyzing fission product transport in pressurized water reactor (PWR) primary systems, (2) to define accidents for shipping casks which result in loss of coolant, (3) to derive from existing computer codes and additional calculations thermal-hydraulic conditions within PWR primary systems, and (4) to determine probable chemical forms for the various fission product species.

  18. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    Directory of Open Access Journals (Sweden)

    Andrej Prošek

    2013-01-01

    Full Text Available Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO. Long-term SBO in a pressurized water reactor (PWR leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs leaks assumed to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS. For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

  19. Criticality analysis for mixed thorium-uranium fuel in the Angra-2 PWR reactor using KENO-VI

    Energy Technology Data Exchange (ETDEWEB)

    Wichrowski, Caio C.; Gonçalves, Isadora C.; Oliveira, Claudio L.; Vellozo, Sergio O.; Baptista, Camila O., E-mail: wichrowski@ime.eb.br, E-mail: isadora.goncalves@ime.eb.br, E-mail: d7luiz@yahoo.com.br, E-mail: vellozo@ime.eb.br, E-mail: camila.oliv.baptista@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Seção de Engenharia Nuclear

    2017-07-01

    The increasing energy demand associated to the current sustainability challenges have given the thorium nuclear fuel cycle renewed interest in the scientific community. Studies have focused on energy production in different reactor designs through the fission of uranium 233, the product of thorium fertilization by neutrons. In order to make it possible for near future applications a strategy based on the adaptation of current nuclear reactors for the use of thorium fuels is being considered. In this work, bearing in mind these limitations, a code was used to evaluate the effect on criticality (k{sub inf}) of the mixing of thorium and uranium in different proportions in the fuel of a PWR, the German designed Angra-2 Brazilian reactor in order to scrutinise its behaviour and determine the feasibility of an adapted ThO{sub 2}-UO{sub 2} mixed fuel cycle using current PWR technology. The analysis is performed using the KENO-VI module in the SCALE 6.1 nuclear safety analysis simulation code and the information is taken from the Angra-2 FSAR (Final Security Analysis Report). (author)

  20. Analysis of a bending test on a full-scale PWR hot leg elbow containing a surface crack

    Energy Technology Data Exchange (ETDEWEB)

    Delliou, P. le [Electricite de France, EDF, 77 - Moret-sur-Loing (France). Dept. MTC; Julisch, P.; Hippelein, K. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Bezdikian, G. [Electricite de France, EDF, 92 - Paris la Defense (France). Direction Production Transport

    1998-11-01

    EDF, in co-operation with Framatome, has conducted a large research programme on the mechanical behaviour of thermally aged cast duplex stainless steel elbows, which are part of the main primary circuit of French PWR. One important task of this programme consisted of testing a full-scale PWR hot leg elbow. The elbow contained a semi-elliptical circumferential notch machined on the outer surface of the intrados as well as casting defects located on the flanks. To simulate the end-of-life condition of the component regarding material toughness, it had undergone a 2400 hours ageing heat treatment at 400 C. The test preparation and execution, as well as the material characterization programme, were committed to MPA. The test was conducted under constant internal pressure and in-plane bending (opening mode) at 200 C. For safety reasons, it took place on an open air-site: the Meppen military test ground. At the maximum applied moment (6000 kN.m), the notch did not initiate. This paper presents the experimental results and the fracture mechanics analysis of the test, based on finite element calculations. (orig.)

  1. New long-cycle small modular PWR cores using particle type burnable poisons for low boron operation

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Hoseong; Hwang, Dae Hee [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Hong, Ser Gi, E-mail: sergihong@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Shin, Ho Choel [Core and Fuel Analysis Group, Korea Hydro & Nuclear Power Central Research Institute (KHNP-CRI), Daejon 305-343 (Korea, Republic of)

    2017-04-01

    Highlights: • New advanced burnable poison rods (BPR) are suggested for low boron operation in PWR. • The new SMR cores have long cycle length of ∼4.5 EFPYs with low boron concentration. • The SMR core satisfies all the design targets and constraints. - Abstract: In this paper, new small long-cycle PWR (Pressurized Water Reactor) cores for low boron concentration operation are designed by employing advanced burnable poison rods (BPRs) in which the BISO (Bi-Isotropic) particles of burnable poison are distributed in a SiC matrix. The BPRs are designed by adjusting the kernel diameter, the kernel material and the packing fraction to effectively reduce the excess reactivity in order to reduce the boron concentration in the coolant and achieve a flat change in excess reactivity over a long operational cycle. In addition, axial zoning of the BPRs was suggested to improve the core performances, and it was shown that the suggested axial zoning of BPRs considerably extends the cycle length compared to a core with no BPR axial zoning. The results of the core physics analyses showed that the cores using BPRs with a B{sub 4}C kernel have long cycle lengths of ∼4.5 EFPYs (Effective Full Power Years), small maximum CBCs (Critical Boron Concentration) lower than 370 ppm, low power peaking factors, and large shutdown margins of control element assemblies.

  2. High Cooling Water Temperature Effects on Design and Operational Safety of NPPs in the Gulf Region

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byung Koo [Khalifa Univ., Abu Dhabi (United Arab Emirates); Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-12-15

    The Arabian Gulf region has one of the highest ocean temperatures, reaching above 35 degrees and ambient temperatures over 50 degrees in the summer. Two nuclear power plants (NPP) are being introduced in the region for the first time, one at Bushehr (1,000 MWe PWR plant from Russia), and a much larger one at Barakah (4Χ1,400 MWe PWR from Korea). Both plants take seawater from the Gulf for condenser cooling, having to modify the secondary/tertiary side cooling systems design by increasing the heat transfer surface area from the country of origin. This paper analyses the secondary side of a typical PWR plant operating under the Rankine cycle with a simplified thermal-hydraulic model. Parametric study of ocean cooling temperatures is conducted to estimate thermal efficiency variations and its associated design changes for the secondary side. Operational safety is reviewed to deliver rated power output with acceptable safety margins in line with technical specifications, mainly in the auxiliary systems together with the cooling water temperature. Impact on the Gulf seawater as the ultimate heat sink is considered negligible, affecting only the adjacent water near the NPP site, when compared to the solar radiation on the sea surface.

  3. HIGH COOLING WATER TEMPERATURE EFFECTS ON DESIGN AND OPERATIONAL SAFETY OF NPPS IN THE GULF REGION

    Directory of Open Access Journals (Sweden)

    BYUNG KOO KIM

    2013-12-01

    Full Text Available The Arabian Gulf region has one of the highest ocean temperatures, reaching above 35 degrees and ambient temperatures over 50 degrees in the summer. Two nuclear power plants (NPP are being introduced in the region for the first time, one at Bushehr (1,000 MWe PWR plant from Russia, and a much larger one at Barakah (4X1,400 MWe PWR from Korea. Both plants take seawater from the Gulf for condenser cooling, having to modify the secondary/tertiary side cooling systems design by increasing the heat transfer surface area from the country of origin. This paper analyses the secondary side of a typical PWR plant operating under the Rankine cycle with a simplified thermal-hydraulic model. Parametric study of ocean cooling temperatures is conducted to estimate thermal efficiency variations and its associated design changes for the secondary side. Operational safety is reviewed to deliver rated power output with acceptable safety margins in line with technical specifications, mainly in the auxiliary systems together with the cooling water temperature. Impact on the Gulf seawater as the ultimate heat sink is considered negligible, affecting only the adjacent water near the NPP site, when compared to the solar radiation on the sea surface.

  4. Electromagnetic Transients in Power Cables

    DEFF Research Database (Denmark)

    Silva, Filipe Faria Da; Bak, Claus Leth

    of electromagnetic phenomena associated to their operation, among them electromagnetic transients, increased as well. Transient phenomena have been studied since the beginning of power systems, at first using only analytical approaches, which limited studies to more basic phenomena; but as computational tools became...... more powerful, the analyses started to expand for the more complex phenomena. Being old phenomena, electromagnetic transients are covered in many publications, and classic books such as the 40-year-old Greenwood’s ‘‘Electric Transients in Power Systems’’ are still used to this day. However...... example.However, the book is not only intended for students . It can also be used by engineers who work in this area and need to understand the challenges/problems they are facing or who need to learn how to prepare their simulation models as well as their function. It also shows how to calculate...

  5. Transient heating of moving objects

    Directory of Open Access Journals (Sweden)

    E.I. Baida

    2014-06-01

    Full Text Available A mathematical model of transient and quasistatic heating of moving objects by various heat sources is considered. The mathematical formulation of the problem is described, examples of thermal calculation given.

  6. Transient thyrotoxicosis during nivolumab treatment

    NARCIS (Netherlands)

    van Kooten, M. J.; van den Berg, G.; Glaudemans, A. W. J. M.; Hiltermann, T. J. N.; Groen, H. J. M.; Rutgers, A.; Links, T. P.

    Two patients presented with transient thyrotoxicosis within 2-4 weeks after starting treatment with nivolumab. This thyrotoxicosis turned into hypothyroidism within 6-8 weeks. Temporary treatment with a beta blocker may be sufficient.

  7. Transient Tsunamis in Lakes

    Science.gov (United States)

    Couston, L.; Mei, C.; Alam, M.

    2013-12-01

    A large number of lakes are surrounded by steep and unstable mountains with slopes prone to failure. As a result, landslides are likely to occur and impact water sitting in closed reservoirs. These rare geological phenomena pose serious threats to dam reservoirs and nearshore facilities because they can generate unexpectedly large tsunami waves. In fact, the tallest wave experienced by contemporary humans occurred because of a landslide in the narrow bay of Lituya in 1958, and five years later, a deadly landslide tsunami overtopped Lake Vajont's dam, flooding and damaging villages along the lakefront and in the Piave valley. If unstable slopes and potential slides are detected ahead of time, inundation maps can be drawn to help people know the risks, and mitigate the destructive power of the ensuing waves. These maps give the maximum wave runup height along the lake's vertical and sloping boundaries, and can be obtained by numerical simulations. Keeping track of the moving shorelines along beaches is challenging in classical Eulerian formulations because the horizontal extent of the fluid domain can change over time. As a result, assuming a solid slide and nonbreaking waves, here we develop a nonlinear shallow-water model equation in the Lagrangian framework to address the problem of transient landslide-tsunamis. In this manner, the shorelines' three-dimensional motion is part of the solution. The model equation is hyperbolic and can be solved numerically by finite differences. Here, a 4th order Runge-Kutta method and a compact finite-difference scheme are implemented to integrate in time and spatially discretize the forced shallow-water equation in Lagrangian coordinates. The formulation is applied to different lake and slide geometries to better understand the effects of the lake's finite lengths and slide's forcing mechanism on the generated wavefield. Specifically, for a slide moving down a plane beach, we show that edge-waves trapped by the shoreline and free

  8. Transient Cooling of Waxy Crude Oil in a Floating Roof Tank

    OpenAIRE

    Jian Zhao; Yang Liu; LiXin Wei; Hang Dong

    2014-01-01

    The transient cooling of waxy crude oil stored in a floating roof tank located in alpine region is studied by means of numerical simulation, accomplished with a two-dimensional model in cylindrical coordinates with the finite volumes method. The typical evolution of transient natural convection and temperature distribution is investigated which can be divided into four stages. For the transient natural convection, it is concluded as the formation, expansion, degradation, and vanishing stage, ...

  9. PENGEMBANGAN SISTEM PEMANTAUAN KONDISI UNTUK KESELAMATAN ROTATING MACHINE DI PWR DENGAN MOTOR CURRENT SIGNATURE ANALYSIS

    Directory of Open Access Journals (Sweden)

    Syaiful Bakhri

    2015-03-01

    Full Text Available Pemantauan kondisi rotating machine sangat diperlukan untuk menjamin keselamatan operasi sekaligus untuk meningkatkan efisiensi operasi di PWR. Salah satu teknik pemantauan kondisi terbaik yang dewasa ini dipilih karena mudah, non-invasive dan murah dalam implementasinya adalah Motor Current Signature Analysis (MCSA. Namun sayangnya penelitian aplikasi teknik ini untuk perangkat keras yang compact, low cost, berkelas industri dan layak untuk aplikasi pembangkit daya bertenaga nuklir sangat terbatas. Penelitian ini bertujuan untuk mengembangkan metode pemantauan kondisi berbasis MCSA dengan perangkat keras berkelas industri yang kompak untuk pembangkit daya tenaga nuklir. Penelitian meliputi aspek pengembangan perangkat keras real-time berbasis FPGA-CompactRIO, pembuatan modul untuk penampil early warning, pengujian unjuk kerja algoritma perangkat kerasnya, analisis spektrum berbagai kerusakan komponen motor elektrik, serta pengujian kinerjanya dalam mendeteksi berbagai kerusakan. Sistem pemantauan mampu mengeksekusi dengan total waktu eksekusi berkisar 164 ms, berhasil mendeteksi spektrum frekuensi berbagai kerusakan di motor induksi seperti stator shorted turn berkisar 75%, rotor broken bar 95%, eccentricity 65%, dan mechanical misalignment 85%, termasuk gangguan catu daya voltage unbalance 100%. Berdasarkan unjuk kerja perangkatnya, sistem pemantauan kondisi rotating machine ini menjadi salah satu alternatif terbaik untuk sistem pemantauan berbagai perangkat pemantauan di reaktor nuklir. Kata kunci : Pemantauan kondisi, rotating machine, Motor Current Signature Analysis (MCSA, Field Programmable Gate Array (FPGA   Condition monitoring of rotating machine is essential to guarantee the safety operation as well as to improve the efficiency of nuclear power plants operations. One of the promising condition monitoring techniques which has been preferred currently since it is simple, non-invasive and inexpensive is Motor Stator Signature Analysis

  10. Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1998-09-01

    The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

  11. Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1997-04-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package

  12. A Comparative Physics Study on Low Boron Concentration Small PWR Core Designs using Different Reflectors

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Ho Seong; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2015-10-15

    In this work, we considered not only water reflector but also several different solid reflectors for this purpose. In this work, we performed the comparative physics study on the long cycle small PWR core with the several different reflector assemblies. A reference core having water reflector was designed to have low boron concentration by using new FCM burnable poison rods with B4C kernel BISO particles. Then, the comparative study was performed by replacing the water reflector with the candidate reflector compositions in the reference core. The SMRs have been designed to simplify the system components so as to improve passive safety by integrating the main system components into reactor vessel with the removal of valves and pipes. Korea also has supported a project for designing the light water cooled SMR which can be operated without soluble boron or low boron concentration. As a part of this project, we have designed an advanced light water cooled SMR core with new Fully Ceramic Micro-encapsulated(FCM) burnable poison rods (BPR) which can be operated over 4 EFPYs with low boron concentration (<350 ppm). The objective of this work is to analyze the effects of different radial reflector materials on the performances of our SMR core having FCM BPRs because it is considered interesting to know how long we can extend the cycle length of the SMR core by optimizing the reflector design. Also, to improve the accuracy of the reflector homogenized two-group cross sections, we performed two-dimensional whole core transport calculations with DeCART2D to obtain the detailed spectrum of the reflector and used it in producing the reflector homogenized two-group cross sections. The study showed that the reference water reflected core has the shortest cycle length of 3.5 EFPYs and the lowest boron concentration of 221ppm while the other solid reflected cores have significantly longer cycle lengths but higher boron concentrations. Of the cores, the graphite reflected core has

  13. Tribocorrosion in pressurized high temperature water: a mass flow model based on the third body approach

    Energy Technology Data Exchange (ETDEWEB)

    Guadalupe Maldonado, S.

    2014-07-01

    Pressurized water reactors (PWR) used for power generation are operated at elevated temperatures (280-300 °C) and under higher pressure (120-150 bar). In addition to these harsh environmental conditions some components of the PWR assemblies are subject to mechanical loading (sliding, vibration and impacts) leading to undesirable and hardly controllable material degradation phenomena. In such situations wear is determined by the complex interplay (tribocorrosion) between mechanical, material and physical-chemical phenomena. Tribocorrosion in PWR conditions is at present little understood and models need to be developed in order to predict component lifetime over several decades. The goal of this project, carried out in collaboration with the French company AREVA NP, is to develop a predictive model based on the mechanistic understanding of tribocorrosion of specific PWR components (stainless steel control assemblies, stellite grippers). The approach taken here is to describe degradation in terms of electro-chemical and mechanical material flows (third body concept of tribology) from the metal into the friction film (i.e. the oxidized film forming during rubbing on the metal surface) and from the friction film into the environment instead of simple mass loss considerations. The project involves the establishment of mechanistic models for describing the single flows based on ad-hoc tribocorrosion measurements operating at low temperature. The overall behaviour at high temperature and pressure in investigated using a dedicated tribometer (Aurore) including electrochemical control of the contact during rubbing. Physical laws describing the individual flows according to defined mechanisms and as a function of defined physical parameters were identified based on the obtained experimental results and from literature data. The physical laws were converted into mass flow rates and solved as differential equation system by considering the mass balance in compartments

  14. Transient Analysis of a Magnetic Heat Pump

    Science.gov (United States)

    Schroeder, E. A.

    1985-01-01

    An experimental heat pump that uses a rare earth element as the refrigerant is modeled using NASTRAN. The refrigerant is a ferromagnetic metal whose temperature rises when a magnetic field is applied and falls when the magnetic field is removed. The heat pump is used as a refrigerator to remove heat from a reservoir and discharge it through a heat exchanger. In the NASTRAN model the components modeled are represented by one-dimensional ROD elements. Heat flow in the solids and fluid are analyzed. The problem is mildly nonlinear since the heat capacity of the refrigerant is temperature-dependent. One simulation run consists of a series of transient analyses, each representing one stroke of the heat pump. An auxiliary program was written that uses the results of one NASTRAN analysis to generate data for the next NASTRAN analysis.

  15. Transient conduction-radiation analysis of an absolute active cavity radiometer using finite elements

    Science.gov (United States)

    Mahan, J. R.; Kowsary, F.; Tira, N.; Gardiner, B. D.

    1987-01-01

    A NASA-developed finite element-based model of a generic active cavity radiometer (ACR) has been developed in order to study the dependence on operating temperature of the closed-loop and open-loop transient response of the instrument. Transient conduction within the sensing element is explored, and the transient temperature distribution resulting from the application of a time-varying radiative boundary condition is calculated. The results verify the prediction that operation of an ACR at cryogenic temperatures results in large gains in frequency response.

  16. Steady-state and transient heat transfer through fins of complex geometry

    Directory of Open Access Journals (Sweden)

    Taler Dawid

    2014-06-01

    Full Text Available Various methods for steady-state and transient analysis of temperature distribution and efficiency of continuous-plate fins are presented. For a constant heat transfer coefficient over the fin surface, the plate fin can be divided into imaginary rectangular or hexangular fins. At first approximate methods for determining the steady-state fin efficiency like the method of equivalent circular fin and the sector method are discussed. When the fin geometry is complex, thus transient temperature distribution and fin efficiency can be determined using numerical methods. A numerical method for transient analysis of fins with complex geometry is developed. Transient temperature distributions in continuous fins attached to oval tubes is computed using the finite volume - finite element methods. The developed method can be used in the transient analysis of compact heat exchangers to calculate correctly the heat flow rate transferred from the finned tubes to the fluid.

  17. Analytical model of transient thermal effect on convectional cooled ...

    Indian Academy of Sciences (India)

    Abstract. The transient analytical solutions of temperature distribution, stress, strain and optical path difference in convectional cooled end-pumped laser rod are derived. The results are compared with other works and good agreements are found. The effects of increasing the edge cooling and face cooling are studied.

  18. Lumped thermal capacitance analysis of transient heat conduction ...

    African Journals Online (AJOL)

    The thermal energy transferred by unsteady flow of the coolant to the vessel was determined as internal energy change. Numerical algorithms for Matlab Code were implemented to generate data for transient analysis and simulation. The simulations indicated that the temperature variations and the the-rmal stresses were ...

  19. Potential for process tube burnout during transient conditions

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, P.A.; Jones, S.S.

    1960-04-22

    This report is an interpretation of data (flow, pressure, temperatures within a process tube during events affecting single tubes) as applied to the most severe (rapid) K-reactor transients which are credible. Analyses indicate that no fuel channel burnout will result from a BPA power loss to the process pumps.

  20. Application of transient analysis methodology to heat exchanger performance monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Rampall, I.; Soler, A.I.; Singh, K.P. [Holtec International, Cherry Hill, NJ (United States); Scott, B.H. [Baltimore Gas and Electric Co., Lusby, MD (United States)

    1994-12-31

    A transient testing technique is developed to evaluate the thermal performance of industrial scale heat exchangers. A Galerkin-based numerical method with a choice of spectral basis elements to account for spatial temperature variations in heat exchangers is developed to solve the transient heat exchanger model equations. Testing a heat exchanger in the transient state may be the only viable alternative where conventional steady state testing procedures are impossible or infeasible. For example, this methodology is particularly suited to the determination of fouling levels in component cooling water system heat exchangers in nuclear power plants. The heat load on these so-called component coolers under steady state conditions is too small to permit meaningful testing. An adequate heat load develops immediately after a reactor shutdown when the exchanger inlet temperatures are highly time-dependent. The application of the analysis methodology is illustrated herein with reference to an in-situ transient testing carried out at a nuclear power plant. The method, however, is applicable to any transient testing application.

  1. The transient thermal response of a tubular solar collector

    Science.gov (United States)

    Lansing, F. L.

    1976-01-01

    A special analytical solution is provided for the timewise response of the circulating fluid temperatures when a sudden step change of the input solar radiation is imposed and remains constant thereafter. An example which demonstrates the transient temperatures at the exit section of a single collector with two different flow patterns is presented. This study is used to supplement some numerical solutions to provide a fairly complete coverage for this type of solar collector.

  2. Modelling Phase Change in a 3D Thermal Transient Analysis

    OpenAIRE

    Haque, EEU; Hampson, PR

    2016-01-01

    A 3D thermal transient analysis of a gap profiling technique which utilises phase change material (plasticine) is conducted in ANSYS. Phase change is modelled by assigning enthalpy of fusion over a wide temperature range based on Differential Scanning Calorimetry (DSC) results. Temperature dependent convection is approximated using Nusselt number correlations. A parametric study is conducted on the thermal contact conductance value between the profiling device (polymer) and adjacent (metal) s...

  3. A numerical study of transient, thermally-conductive solar wind

    Science.gov (United States)

    Han, S. M.; Wu, S. T.; Dryer, M.

    1987-01-01

    A numerical analysis of transient solar wind starting at the solar surface and arriving at 1 AU is performed by an implicit numerical method. The model hydrodynamic equations include thermal conduction terms for both steady and unsteady simulations. Simulation results show significant influence of thermal conduction on both steady and time-dependent solar wind. Higher thermal conduction results in higher solar wind speed, higher temperature, but lower plasma density at 1 AU. Higher base temperature at the solar surface gives lower plasma speed, lower temperature, but higher density at 1 AU. Higher base density, on the other hand, gives lower velocity, lower temperature, but higher density at 1 AU.

  4. Structure analysis of a reactor pressure vessel by two- and three-dimensional models. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Sacher, H.; Mayr, M.

    1982-03-01

    This paper investigates the reactor pressure vessel of a 1300 MW pressurised water reactor. In order to determine the stresses and deformations of the vessel, two- and three-dimensional finite element models are used which represent the real structure with different degrees of accuracy. The results achieved by these different models are compared for the case of the transient called ''Start up of the nuclear power plant''. 5 refs.

  5. Recent development of transient electronics

    Directory of Open Access Journals (Sweden)

    Huanyu Cheng

    2016-01-01

    Full Text Available Transient electronics are an emerging class of electronics with the unique characteristic to completely dissolve within a programmed period of time. Since no harmful byproducts are released, these electronics can be used in the human body as a diagnostic tool, for instance, or they can be used as environmentally friendly alternatives to existing electronics which disintegrate when exposed to water. Thus, the most crucial aspect of transient electronics is their ability to disintegrate in a practical manner and a review of the literature on this topic is essential for understanding the current capabilities of transient electronics and areas of future research. In the past, only partial dissolution of transient electronics was possible, however, total dissolution has been achieved with a recent discovery that silicon nanomembrane undergoes hydrolysis. The use of single- and multi-layered structures has also been explored as a way to extend the lifetime of the electronics. Analytical models have been developed to study the dissolution of various functional materials as well as the devices constructed from this set of functional materials and these models prove to be useful in the design of the transient electronics.

  6. Analysis of the performance of fuel cells PWR with a single enrichment and radial distribution of enrichments; Analisis del desempeno de celdas combustibles PWR con un solo enriquecimiento y con distribucion radial de enriquecimientos

    Energy Technology Data Exchange (ETDEWEB)

    Vargas, S.; Gonzalez, J. A.; Alonso, G.; Del Valle, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D.F. 07738 (Mexico); Xolocostli M, J. V. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: nolosesamuel@prodigy.net.mx

    2008-07-01

    One of the main challenges in the design of fuel assemblies is the efficient use of uranium achieving burnt homogeneous of the fuel rods as well as the burnt maximum possible of the same ones to the unload. In the case of the assemblies type PWR has been decided actually for fuel assemblies with a single radial enrichment. The present work has like effect to show the because of this decision, reason why a comparison of the neutronic performance of two fuel cells takes place with the same enrichment average but one of them with radial distribution of enrichment and the other with a single enrichment equal to the average. The results shown in the present study of the behavior of the neutron flow as well as the power distribution through of assembly sustain the because of a single radial enrichment. (Author)

  7. Evaluation of the presence of a burnable absorber in an assembly 3x3 type PWR; Evaluacion de la presencia de un absorbedor quemable en un ensamble 3x3 tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez F, M. A.; Del Valle G, E.; Alonso V, G. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D. F. 07738 (Mexico)]. e-mail: mike_ipn_esfm@hotmail.com

    2008-07-01

    In the present work the effect is evaluated that causes the presence of a burnable absorber in an adjustment of rods of 3x3 of a fuel assembly type PWR using CASMO-4 code, when comparing the infinite multiplication factor and some average cross sections by means of codes MCNP-4A, CASMO-3 and HELIOS. For this evaluation two cases are evaluated: first consists of an adjustment of rods of 3x3 full completely of fuel and the second consists of a central rod full with a burnable absorber type wet annular burnable absorber (WABA) and the remaining full fuel rods. In both cases the enrichment of the fissile isotopes is varied, for two types of fuel, MOX degree armament and UO{sub 2}. (Author)

  8. Essays of leaching in cemented products containing simulated waste from evaporator concentrated of PWR reactor; Ensaios de lixiviacao em produtos cimentados contendo rejeito simulado de concentrado do evaporador de reator PWR

    Energy Technology Data Exchange (ETDEWEB)

    Haucz, Maria Judite A.; Calabria, Jaqueline A. Almeida; Tello, Cledola Cassia O.; Candido, Francisco Donizete; Seles, Sandro Rogerio Novaes, E-mail: hauczmj@cdtn.b, E-mail: jaalmeida@cdtn.b, E-mail: tellocc@cdtn.b, E-mail: fdc@cdtn.b, E-mail: seless@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-10-26

    This paper evaluated the results from leaching resistance essays of cemented products, prepared from three distinct formulations, containing simulated waste of concentrated from the PWR reactor evaporator. The leaching rate is a parameter of qualification of solidified products containing radioactive waste and is determined in accordance with regulation ISO 6961. This procedure evaluates the capacity of transfer organic and inorganic substances presents in the waste through dissolution in the extractor medium. For the case of radioactive waste it is reached the more retention of contaminants in the cemented product, i.e.the lesser value of lixiviation rate. Therefore, this work evaluated among the proposed formulation that one which attend the criterion established in the regulation CNEN-NN-6.09

  9. Time of Flight Transients in the Dipolar Glass Model

    OpenAIRE

    Novikov, S. V.; Tyutnev, A. P.; Schein, L. B.

    2013-01-01

    Using Monte Carlo simulation we investigated time of flight current transients predicted by the dipolar glass model for a random spatial distribution of hopping centers. Behavior of the carrier drift mobility was studied at room temperature over a broad range of electric field and sample thickness. A flat plateau followed by $j\\propto t^{-2}$ current decay is the most common feature of the simulated transients. Poole-Frenkel mobility field dependence was confirmed over 5 to 200 V/$\\mu$m as we...

  10. Explosive and Radio-Selected Transients: Transient Astronomy with ...

    Indian Academy of Sciences (India)

    In India, active research is going on in transient astronomy, especially in the fields of supernovae, gamma ray .... to ∼9 as of now. We currently understand them as catastrophic events generating a central engine which ..... bubbles blown by hot progenitor or with complex circumstellar environments set up by variable winds.

  11. Explosive and Radio-Selected Transients: Transient Astronomy with ...

    Indian Academy of Sciences (India)

    Therefore, multiwaveband observational efforts with wide fields of view will be the key to progress of transients astronomy from the middle 2020s offering unprecedented deep images and high spatial and spectral resolutions. Radio observations of Gamma Ray Bursts (GRBs) with SKA will uncover not only much fainter ...

  12. ANALISIS PERUBAHAN MASSA BAHAN FISIL DAN NON FISIL DALAM TERAS PWR 1000 MWe DENGAN ORIGEN-ARP 5.1

    Directory of Open Access Journals (Sweden)

    Anis Rohanda

    2015-03-01

    Full Text Available Teras reaktor merupakan tempat terjadinya reaksi pembelahan (fisi yang terkendali. Komponen reaktor seperti bahan bakar, kelongsong (cladding dan air pendingin memiliki peranan penting dalam keberlangsungan reaksi fisi. Reaksi fisi mengakibatkan terbentuknya sejumlah nuklida hasil fisi dan hasil aktivasi. Hasil fisi berasal dari reaksi tangkapan neutron termal dengan bahan fisil sedangkan hasil aktivasi berasal dari interaksi bahan non fisil seperti material kelongsong dan pendingin oleh neutron dan gamma. Pada setiap pengoperasian suatu reaktor, informasi perubahan massa bahan fisil dan non fisil sangat berguna untuk manajemen bahan bakar dalam teras, seperti pengaturan reaktivitas, optimasi dan pemuatan bahan bakar. Untuk itu perlu dilakukan penelitian mengenai perubahan bahan fisil dan non fisil tersebut dalam teras reaktor. Hal ini dapat dilakukan dengan mengamati perubahan massa dari material dalam teras reaktor. Penelitian ini memiliki tujuan untuk mengetahui perubahan massa unsur penyusun material dalam teras, seperti massa dari unsur penyusun elemen bahan bakar nuklir, kelongsong dan air pendingin setelah digunakan dalam teras. Dari perubahan massa tersebut dapat diketahui fraksi bakar atau tingkat konsumsi bahan bakar yang digunakan. Penelitian dilakukan pada basis reaktor PLTN tipe PWR buatan pabrikan asal Amerika Serikat berdaya 1000 MWe dengan menggunakan code penghitung inventori hasil fisi ORIGEN-ARP 5.1, yaitu versi terbaru dari ORIGEN dengan library khusus reaktor daya. Hasil analisis menunjukkan bahwa bahan fisil U-235 mengalami pengurangan massa hingga 58% atau lebih dari separuhnya dari massa U-235 awal untuk tiap kali siklus operasi. Bahan fertil U-238 hanya mengalami pengurangan massa sekitar 2% dari massa awalnya tiap kali siklus operasi. Lain halnya dengan bahan non fisil yang mengalami perubahan massa yang berbeda-beda untuk tiap kali siklus operasinya yang tergantung pada tampang lintang aktivasi serta laju peluruhan dan

  13. Simulation of German PKL refill/reflood experiment K9A using RELAP4/MOD7. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, M.T.; Davis, C.B.; Behling, S.R.

    1981-11-01

    This paper describes a RELAP4/MOD7 simulation of West Germany's Kraftwerk Union (KWU) Primary Coolant Loop (PKL) refill/reflood experiment K9A. RELAP4/MOD7, a best-estimate computer program for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This study was the first major simulation using RELAP4/MOD7 since its release by the Idaho National Engineering Laboratory (INEL). The PKL facility is a reduced scale (1:134) representation of a typical West German four-loop 1300 MW pressurized water reactor (PWR). A prototypical scale of the total volume to power ratio was maintained. The test facility was designed specifically for an experiment simulating the refill/reflood phase of a Loss-of-Coolant Accident (LOCA).

  14. The radiological consequences of degraded core accidents for the Sizewell PWR The impact of adopting revised frequencies of occurrence

    CERN Document Server

    Kelly, G N

    1983-01-01

    The radiological consequences of degraded core accidents postulated for the Sizewell PWR were assessed in an earlier study and the results published in NRPB-R137. Further analyses have since been made by the Central Electricity Generating Board (CEGB) of degraded core accidents which have led to a revision of their predicted frequencies of occurrence. The implications of these revised frequencies, in terms of the risk to the public from degraded core accidents, are evaluated in this report. Increases, by factors typically within the range of about 1.5 to 7, are predicted in the consequences, compared with those estimated in the earlier study. However, the predicted risk from degraded core accidents, despite these increases, remains exceedingly small.

  15. Analysis of the performance of the Westinghouse reactor vessel level indicating system for tests at semiscale. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hardy, J.E.; Miller, G.N.

    1982-10-01

    The Westinghouse Reactor Vessel Level Indicating System (RVLIS), a differential pressure level measurement system, was tested at SEMISCALE. This report contains the analyses of these tests and the conclusions of these analyses. The tests performed included small break and intermediate break tests. Also, frequency response and natural circulation tests were run and analyzed. The RVLIS always indicated a level less than the two phase froth level. The RVLIS output in early small break tests indicated a level 200 cm greater than actual collapsed liquid level. This discrepancy was caused by structural differences between SEMISCALE and a Westinghouse reactor. Once modifications were made so that SEMISCALE better simulated a Westinghouse PWR, the maximum difference between RVLIS and SEMISCALE instrumentation was 30 cm or 3% which is less than the stated uncertainty of the Westinghouse RVLIS.

  16. Uncertainty Propagation Analysis for PWR Burnup Pin-Cell Benchmark by Monte Carlo Code McCARD

    Directory of Open Access Journals (Sweden)

    Ho Jin Park

    2012-01-01

    Full Text Available In the Monte Carlo (MC burnup analyses, the uncertainty of a tally estimate at a burnup step may be induced from four sources: the statistical uncertainty caused by a finite number of simulations, the nuclear covariance data, uncertainties of number densities, and cross-correlations between the nuclear data and the number densities. In this paper, the uncertainties of kinf, reaction rates, and number densities for a PWR pin-cell benchmark problem are quantified by an uncertainty propagation formulation in the MC burnup calculations. The required sensitivities of tallied parameters to the microscopic cross-sections and the number densities are estimated by the MC differential operator sampling method accompanied by the fission source perturbation. The uncertainty propagation analyses are conducted with two nuclear covariance data—ENDF/B-VII.1 and SCALE6.1/COVA libraries—and the numerical results are compared with each other.

  17. Practical Application of the MFM Suite on a PWR System: Modelling and Reasoning on Causes and Consequences of Process Anomalies

    DEFF Research Database (Denmark)

    Zhang, Xinxin; Thunem, Harald P - J; Lind, Morten

    2014-01-01

    Multilevel Flow Modelling (MFM) is a functional modelling methodology which applies means - end and parts - whole decomposition and aggregation techniques to handle the complexity of engineering systems. It has been adopted in several case studies to model the process goal and functions of PWR......, with the aim to facilitate the plant operators to evaluate and understand plant situations. The theoretical aspects have been established for the cooperative development of an MFM software tool, namely MFM Suite, by the Halden Reactor Project (HRP) and the Technical University of Denmark (DTU). The MFM Suite...... is equipped with an MFM Model Editing Interface to facilitate the modelling process and MFM model analysis modules to run diag nosis and prognosis analyses based on developed models. New features of the MFM Suite also include making corresponding process diagram for the plant being modelled with MFM...

  18. PEMODELAN DAN ANALISIS SEBARAN RADIONUKLIDA DARI PWR PADA KONDISI ABNORMAL DI TAPAK BOJANEGARA-SERANG

    Directory of Open Access Journals (Sweden)

    Sri Kuntjoro

    2015-04-01

    Full Text Available Penambahan pembangkit listrik yang baru khususnya pembangkit listrik tenaga nuklir (PLTN berpotensi memberikan konsekuensi radiologis pada masyarakat dan lingkungan, karena adanya lepasan radioaktif dalam kondisi operasi normal maupun abnormal. Oleh karena itu maka pengelola reaktor nuklir harus bisa menyediakan data dan argumentasi yang kuat untuk menjelaskan tentang keselamatan PLTN terhadap lingkungan. Untuk itu perlu dilakukan analisis kondisi abnormal yang terjadi pada PLTN yang akan memberikan konsekuensi radiologis pada lingkungan. Analisis dilakukan dengan membuat pemodelan simulasi kondisi abnormal yang dipostulasikan pada PLTN tipe PWR 1000 MWe serta simulasi dan pemodelan pola potensi lingkungan sebagai daya dukung tapak terhadap penerimaan konsekuensi radiologis tersebut. Pemodelan fenomena transport radionuklida dari teras reaktor sampai ke luar dari sungkup reaktor dilakukan menggunakan perangkat lunak EMERALD dan pemodelan pola dispersi radioaktivitas ke lingkungan dari reaktor meliputi simulasi kondisi meteorologi, distribusi penduduk, produksi dan konsumsi masyarakat pada kondisi ekstrim di daerah studi, menggunakan perangkat lunak GIS, Arcview, Windrose, dan PC COSYMA. Pemodelan konsekuensi radiologis menggunakan tapak contoh daerah Bojanegara-Kramatwatu Pantai Serang-Banten. Dengan menggunakan data sourceterm, data meteorologi dan data dispersi (sebaran penduduk, produksi pertanian dan ternak dan modeling alur paparan (pathway, dihasilkan model sebaran radionuklida dan penerimaan paparan radiasi di lingkungan tapak Bojanegara-Serang, dengan penerimaan dosis radiasi di bawah batas yang diijinkan badan regulator BAPETEN. Kata kunci : PLTN, radioaktivitas, pola dispersi, keselamatan   Additional of electrical power especially Nuclear Power Plant will give radiological consequences to population and environment due to radioactive release in normal and abnormal condition. In consequence the management of nuclear power plant must

  19. Radiolytic stability of colloidal AgI suspensions and potential impact on PWR safety

    Energy Technology Data Exchange (ETDEWEB)

    Guentay, Salih; Bruchertseifer, Horst; Jaeckel, Bernd [Paul Scherrer Institut, PSI, CH-5232 Villigen (Switzerland)

    2006-07-01

    }-irradiation under otherwise identical conditions did not produce any decomposition. The {gamma}-irradiations of AgI colloidal suspension solutions containing chloride ions prevailed volatile iodine as in the {beta}-irradiations only under strong oxidizing conditions. The investigations and extensive modelling work showed that the AgI stability is a complex function of the containment sump composition and conditions and in contrast to previous conceptions can also be a function of coupled break-up of the AgI bond by ionization of AgI colloids by the absorbed dose and direct radiolytic oxidation of the AgI particle surface and followed by formation and release of iodine radicals. Presence of chloride ions under strong oxidizing conditions, introduces an additional oxidation process, which, independent of the irradiation source type, very effectively enhances the decomposition of the AgI. The paper introduces the results of the experimental investigations as well as provides arguments on its potential impact on the PWR safety during a severe accident.

  20. Development, verification and validation of an FPGA-based core heat removal protection system for a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Yichun, E-mail: ycwu@xmu.edu.cn [College of Energy, Xiamen University, Xiamen 361102 (China); Shui, Xuanxuan, E-mail: 807001564@qq.com [College of Energy, Xiamen University, Xiamen 361102 (China); Cai, Yuanfeng, E-mail: 1056303902@qq.com [College of Energy, Xiamen University, Xiamen 361102 (China); Zhou, Junyi, E-mail: 1032133755@qq.com [College of Energy, Xiamen University, Xiamen 361102 (China); Wu, Zhiqiang, E-mail: npic_wu@126.com [State Key Laboratory of Reactor System Design Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Zheng, Jianxiang, E-mail: zwu@xmu.edu.cn [College of Energy, Xiamen University, Xiamen 361102 (China)

    2016-05-15

    Highlights: • An example on life cycle development process and V&V on FPGA-based I&C is presented. • Software standards and guidelines are used in FPGA-based NPP I&C system logic V&V. • Diversified FPGA design and verification languages and tools are utilized. • An NPP operation principle simulator is used to simulate operation scenarios. - Abstract: To reach high confidence and ensure reliability of nuclear FPGA-based safety system, life cycle processes of discipline specification and implementation of design as well as regulations verification and validation (V&V) are needed. A specific example on how to conduct life cycle development process and V&V on FPGA-based core heat removal (CHR) protection system for CPR1000 pressure water reactor (PWR) is presented in this paper. Using the existing standards and guidelines for life cycle development and V&V, a simplified FPGA-based CHR protection system for PWR has been designed, implemented, verified and validated. Diversified verification and simulation languages and tools are used by the independent design team and the V&V team. In the system acceptance testing V&V phase, a CPR1000 NPP operation principle simulator (OPS) model is utilized to simulate normal and abnormal operation scenarios, and provide input data to the under-test FPGA-based CHR protection system and a verified C code CHR function module. The evaluation results are applied to validate the under-test FPGA-based CHR protection system. The OPS model operation outputs also provide reasonable references for the tests. Using an OPS model in the system acceptance testing V&V is cost-effective and high-efficient. A dedicated OPS, as a commercial-off-the-shelf (COTS) item, would contribute as an important tool in the V&V process of NPP I&C systems, including FPGA-based and microprocessor-based systems.

  1. Scientific-technical cooperation with Russia. Transient analyses for alternative types of water-cooled reactors. Final report; WTZ mit Russland. Transientenanalysen fuer wassergekuehlte Kernreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Rohde, Ulrich [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Kozmenkov, Yaroslav [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Institute of Physics and Power Engineering, Obninsk (Russian Federation); Pivovarov, Valeri; Matveev, Yurij [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    2010-12-15

    The recently developed multi-group version DYN3D-MG of the reactor dynamics code DYN3D has been qualified for applications to water-cooled reactor concepts different from industrial PWR and BWR. An extended DYN3D version was applied to the graphite-moderated pressure tube reactor EGP-6 (NPP Bilibino) and conceptual design studies of an advanced Boiling Water Reactor with reduced moderation (RMWR) as well as the RUTA-70 reactor for low temperature heat supply. Concerning the RUTA reactor, safe heat removal by natural circulation of the coolant at low pressure has to be shown. For the corresponding validation of thermo-hydraulic system codes like ATHLET and RELAP5, experiments on flashing-induced natural circulation instabilities performed at the CIRCUS test facility at the TU Delft were simulated using the RELAP5 code. For the application to alternative water-cooled reactors, DYN3D model extensions and modifications were implemented, in particular adaptations of heat conduction and heat transfer models. Performing code-to-code comparisons with the Russian fine-mesh neutron diffusion code ACADEM contributed to the verification of DYN3D-MG. Validation has been performed by calculating reactor dynamics experiments at the NPP Bilibino. For the reactors EGP-6, RMWR and RUTA, analyses of various protected and unprotected control rod withdrawal and ejection transients were performed. The beyond design basis accident (BDBA) scenario ''Coast-down of all main coolant pumps at nominal power without scram'' for the RUTA reactor was analyzed using the code complexes DYN3D/ATHLET and DYN3D/RELAP5. It was shown, that the reactor passes over to a save asymptotic state at reduced power with coolant natural circulation. Analyzing the BDBA ''Unprotected withdrawal of a control rod group'' for the RMWR, the safety against Departure from Nucleate Boiling (DNB) could not be shown with the necessary confidence. Finally, conclusions have been drawn

  2. Evaluation of nonequilibrium effects in bundle dispersed-flow film boiling. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Morris, D.G.; Mullins, C.B.; Yoder, G.L.

    1983-01-01

    The effects of thermodynamic nonequilibrium in dispersed flow film boiling heat transfer are examined. Steady-state and transient rod-bundle data are used to evaluate several empirical heat-transfer models commonly employed to predict post-CHF behavior. The models that account for thermodynamic nonequilibrium perform adequately, while those that ignore nonequilibrium effects incur errors in wall superheat as high as 190/sup 0/K. Nonequilibrium effects can also be treated by explicitly modeling the phenomena. The thermal-hydraulic code COBRA-TF employs this approach. Using bundle data, the models in the code are evaluated. Analysis suggests that the interfacial heat transfer is overpredicted.

  3. Transient filament stretching rheometer II

    DEFF Research Database (Denmark)

    Kolte, Mette Irene; Rasmussen, Henrik K.; Hassager, Ole

    1997-01-01

    The Lagrangian sspecification is used to simulate the transient stretching filament rheometer. Simulations are performed for dilute PIB-solutions modeled as a four mode Oldroyd-B fluid and a semidilute PIB-solution modeled as a non-linear single integral equation. The simulations are compared...

  4. Design and calibration of a novel transient radiative heat flux meter for a spacecraft thermal test.

    Science.gov (United States)

    Sheng, Chunchen; Hu, Peng; Cheng, Xiaofang

    2016-06-01

    Radiative heat flux measurement is significantly important for a spacecraft thermal test. To satisfy the requirements of both high accuracy and fast response, a novel transient radiative heat flux meter was developed. Its thermal receiver consists of a central thermal receiver and two thermal guarded annular plates, which ensure the temperature distribution of the central thermal receiver to be uniform enough for reasonably applying lumped heat capacity method in a transient radiative heat flux measurement. This novel transient radiative heat flux meter design can also take accurate measurements regardless of spacecraft surface temperature and incident radiation spectrum. The measurement principle was elaborated and the coefficients were calibrated. Experimental results from testing a blackbody furnace and an Xenon lamp show that this novel transient radiative heat flux meter can be used to measure transient radiative heat flux up to 1400 W/m(2) with high accuracy and the response time of less than 10 s.

  5. Transient thermal effects in Alpine permafrost

    Directory of Open Access Journals (Sweden)

    J. Noetzli

    2009-04-01

    Full Text Available In high mountain areas, permafrost is important because it influences the occurrence of natural hazards, because it has to be considered in construction practices, and because it is sensitive to climate change. The assessment of its distribution and evolution is challenging because of highly variable conditions at and below the surface, steep topography and varying climatic conditions. This paper presents a systematic investigation of effects of topography and climate variability that are important for subsurface temperatures in Alpine bedrock permafrost. We studied the effects of both, past and projected future ground surface temperature variations on the basis of numerical experimentation with simplified mountain topography in order to demonstrate the principal effects. The modeling approach applied combines a distributed surface energy balance model and a three-dimensional subsurface heat conduction scheme. Results show that the past climate variations that essentially influence present-day permafrost temperatures at depth of the idealized mountains are the last glacial period and the major fluctuations in the past millennium. Transient effects from projected future warming, however, are likely larger than those from past climate conditions because larger temperature changes at the surface occur in shorter time periods. We further demonstrate the accelerating influence of multi-lateral warming in steep and complex topography for a temperature signal entering the subsurface as compared to the situation in flat areas. The effects of varying and uncertain material properties (i.e., thermal properties, porosity, and freezing characteristics on the subsurface temperature field were examined in sensitivity studies. A considerable influence of latent heat due to water in low-porosity bedrock was only shown for simulations over time periods of decades to centuries. At the end, the model was applied to the topographic setting of the Matterhorn

  6. Transient optical gain in germanium quantum wells

    Energy Technology Data Exchange (ETDEWEB)

    Chatterjee, Sangam; Lange, Christoph; Koester, Niko S.; Schaefer, Martin; Kira, Mackillo; Koch, Stephan W. [Faculty of Physics and Materials Sciences Center, Philipps-Universitaet Marburg (Germany); Chrastina, Daniel; Isella, Giovanni; Kaenel, Hans von [CNISM, Como (Italy); L-NESS, Dipartimento di Fisica del Politecnico di Milano, Como (Italy); Sigg, Hans [Laboratory for Micro and Nanotecnology, Paul Scherrer Institut, Villigen PSI (Switzerland)

    2010-07-01

    One of today's most-sought goals in semiconductor technology is the monolithic integration of microelectronics and photonics on Si. Optical gain is, in general, not expected for Si and Ge or its alloys due to the indirect nature of the band gap in this material system. Here, we show that Ge/SiGe QWs show transient optical gain and may thus be used as an optically-pumped amplifier at room temperature. Further, the nonequilibrium effects which govern the relaxation dynamics of the optically injected carrier distributions in this material were observed and analyzed using a microscopic many-body theory. Strong non-equilibrium gain was obtained on a sub-100 fs time scale. Long-lived gain arising from {gamma}-point transitions is overcompensated by a process bearing the character of free carrier absorption.

  7. Modelling Phase Change in a 3D Thermal Transient Analysis

    Directory of Open Access Journals (Sweden)

    E Haque

    2016-09-01

    Full Text Available A 3D thermal transient analysis of a gap profiling technique which utilises phase change material (plasticine is conducted in ANSYS. Phase change is modelled by assigning enthalpy of fusion over a wide temperature range based on Differential Scanning Calorimetry (DSC results. Temperature dependent convection is approximated using Nusselt number correlations. A parametric study is conducted on the thermal contact conductance value between the profiling device (polymer and adjacent (metal surfaces. Initial temperatures are established using a liner extrapolation based on experimental data. Results yield good correlation with experimental data.

  8. Sensitivity analysis of the spectra of the core neutronic source in the calculation of radiation damage in internal of PWR reactor vessel. Internal; Analisis de sensibilidad a los espectros de la fuente neutronica del nucleo en el calculo del dano por irradiacion en los internos de la vasija de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barrerira Pereira, P.

    2012-07-01

    This study is to analyze the sensitivity to the expected differences in the energy spectra characterizing the neutron source that radiates the vessel internals of a commercial PWR reactor, in order to quantify their influence in the quantities that determine the damage in materials metal.

  9. Study of the linearity of CABRI experimental ionization chambers during RIA transients

    Science.gov (United States)

    Lecerf, J.; Garnier, Y.; Hudelot, JP.; Duc, B.; Pantera, L.

    2018-01-01

    CABRI is an experimental pulse reactor operated by CEA at the Cadarache research center and funded by the French Nuclear Safety and Radioprotection Institute (IRSN). For the purpose of the CABRI International Program (CIP), operated and managed by IRSN under an OECD/NEA framework it has been refurbished since 2003 to be able to provide experiments in prototypical PWR conditions (155 bar, 300 °C) in order to study the fuel behavior under Reactivity Initiated Accident (RIA) conditions. This paper first reminds the objectives of the power commissioning tests performed on the CABRI facility. The design and location of the neutron detectors monitoring the core power are also presented. Then it focuses on the different methodologies used to calibrate the detectors and check the consistency and co-linearity of the measurements. Finally, it presents the methods used to check the linearity of the neutron detectors up to the high power levels ( 20 GW) reached during power transients. Some results obtained during the power tests campaign are also presented.

  10. Features of Convective Turbulent Flows Behind the Mixing Spacer Grids of the TVS-Kvadrat of The PWR-Type Reactor

    Science.gov (United States)

    Dmitriev, S. M.; Varentsov, A. V.; Doronkov, D. V.; Legchanov, M. A.; Sorokin, V. D.; Khrobostov, A. E.

    2017-09-01

    This paper presents the results of experimental investigations of the influence of mixing spacer grids with different types of deflectors on the coolant flow in the TVS-Kvadrat fuel assembly of the PWR-type reactor. Investigations were conducted on an aerodynamic stand with the use of a multichannel pneumometric probe. Analysis of the spatial distribution of projections of the absolute flow velocity permitted detailing the coolant flow pattern behind the mixing spacer grid with various types of deflectors of mixing grids. The results obtained in the present paper are used to verify three-dimensional CFD programs, and in applied cell by cell codes they also serve as a database in calculations of thermotechnical reliability of the cores of PWR-type reactors with the TVS-Kvadrat.

  11. Study of a 900 MW PWR by a substructuring method - Spectral response to a seismic excitation and comparison with a beam model

    Energy Technology Data Exchange (ETDEWEB)

    Rousseau, G.; Bianchini-Burlot, B.; Bosselut, D.; Jacquart, G. [Electricite de France (EDF), 92 - Clamart (France); Viallet, E. [Electricite de France (EDF), 69 - Villeurbanne (France)

    1997-03-01

    This report presents a three dimensional Finite Element Model (FEM) of a 900 MW Pressurized Water Reactor (PWR) which is described at first: its modal behaviour is computed by a sub-structuring method based upon a Component Mode Synthesis (CMS) method. All the substructures taken into account in the model are described. One model with equivalent beams is also described. Then, different approaches to take into account the fluid/structure interaction in the different models are investigated. Results of the modal analysis of each model are compared to each other and with experimental measures. This modal analysis is then used to compute the non linear and linear response of the PWR due to a seismic excitation. (author) 4 refs.

  12. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

    2016-08-01

    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

  13. Preliminary Analysis on Reactivity Insertion Transient of Natural Circulation Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Son, Hyung Min [KAERI, Daejeon (Korea, Republic of); Suh, Kune Yull [Seoul National University, Seoul (Korea, Republic of)

    2015-05-15

    When a malfunction of the reactor control system occurs, there's a chance that the positive reactivity is inserted into the core, resulting in the increase of the core power. With the combination of the failure of the related safety features, this may raise the temperature of the core material beyond the design limit to break its integrity. For the fast nuclear reactors like FFTF and CRBRP, the overpower trip is initiated when the power reaches 115% of rated value to keep the fuel from melting. In this study, the system response to the reactivity insertion transient on a liquid metal cooled natural circulation reactor is analyzed utilizing an in-house code based on a momentum integral model. Utilizing an in-house system analysis code, a set of numerical simulation is carried out on the reactivity ramp insertion transients which showed that the role of reactivity feedback is significant in mitigating the time to failure, and the evolution of the natural circulation mass flow is rather slow to generate meaningful feedback effect. It is also observed that in terms of the peak cladding temperature, smaller reactivity insertion transient generated more severe outcome owing to increased accumulation of the thermal energy within fuel pins. Thus, it may require an extra attention to carefully monitor and capture the mild transients to avoid potential drastic results.

  14. Experiment data report for Semiscale Mod-1 Test S-29-3; integral test from reduced initial pressure. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Crapo, H.S.; Collins, B.L.; Sackett, K.E.

    1976-09-01

    Recorded test data are presented for Test S-29-3 of the Semiscale Mod-1 special heat transfer test series. This test is among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor system. Test S-29-3 was conducted from an initial cold leg fluid temperature of 544/sup 0/F and an initial pressure of 1,760 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient starting from a lower initial pressure than that usually associated with pressurized water reactor operation. System flow was set to achieve a full core fluid temperature differential of 66/sup 0/F at full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a peaked radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core cooling injection into both the intact and broken loops. Coolant injection was continued until testtermination at 200 seconds after initiation of blowdown. The purpose of the report is to make available the uninterpreted data from Test S-29-3 for future data analysis and test results reporting activites. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to assure that they are reasonable and consistent.

  15. On the role of sulfur on the dissolution of pressure vessel steels at the tip of a propagating crack in PWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Combrade, P.; Foucault, M. (UNIREC 42- Firminy (FR)); Marcus, P. (Ecole Nationale Superieure de Chimie 75 - Paris (FR)); Slama, G. (Societe Franco-Americaine de Constructions Atomiques (Framatome), 92 - Courbevoie (FR))

    1990-03-01

    Different aspects of the effect of sulfur on the dissolution and film repair on pressure vessel steel exposed to PWR environment at 300{sup 0}C were examined. A monolayer of sulfur adsorbed on a bare surface was shown to inhibit the nucleation of a magnetite film. The comparison of this result with dissolution measurements performed by using CERT under controlled potential lead to the assumption that mechanical rupture steps are involved in the environmental effect on the crack propagation rate. 27 refs.

  16. Nonlinear Diffusion and Transient Osmosis

    Science.gov (United States)

    Akira, Igarashi; Lamberto, Rondoni; Antonio, Botrugno; Marco, Pizzi

    2011-08-01

    We investigate both analytically and numerically the concentration dynamics of a solution in two containers connected by a narrow and short channel, in which diffusion obeys a porous medium equation. We also consider the variation of the pressure in the containers due to the flow of matter in the channel. In particular, we identify a phenomenon, which depends on the transport of matter across nano-porous membranes, which we call “transient osmosis". We find that nonlinear diffusion of the porous medium equation type allows numerous different osmotic-like phenomena, which are not present in the case of ordinary Fickian diffusion. Experimental results suggest one possible candidate for transiently osmotic processes.

  17. Evaluation of integral continuing experimental capability (CEC) concepts for light water reactor research: PWR scaling concepts

    Energy Technology Data Exchange (ETDEWEB)

    Condie, K G; Larson, T K; Davis, C B; McCreery, G E

    1987-02-01

    In this report reactor transients and thermal-hydraulic phenomena of importance (based on probabilistic risk assessment and the International Code Assessment Program) to reactor safety were examined and identified. Established scaling methodologies were used to develop potential concepts for integral thermal-hydraulic testing facilities. Advantages and disadvantages of each concept are evaluated. Analysis is conducted to examine the scaling of various phenomena in each of the selected concepts. Results generally suggest that a facility capable of operating at typical reactor operating conditions will scale most phenomena reasonably well. Although many phenomena in facilities using Freon or water at nontypical pressure will scale reasonably well, those phenomena that are heavily dependent on quality (heat transfer or critical flow for example) can be distorted. Furthermore, relation of data produced in facilities operating with nontypical fluids or at nontypical pressures to large plants will be a difficult and time consuming process.

  18. Acceptance test for 900 MWe PWR unit replacement steam generators; Essai de reception des generateurs de vapeur de remplacement des tranches REP 900

    Energy Technology Data Exchange (ETDEWEB)

    Gourguechon, B.

    1993-12-31

    During the first half of 1994, the Gravelines 1 steam generators will be replaced (SG replacement procedure). The new SG`s differ from the former components notably by the alloy used for the tube bundle, in this case, the high chromium content Inconel 690. So, from this standpoint, they are to be considered as PWR 900 replacement SG first models and their thermal efficiency has consequently to be assessed. This will provide an opportunity of ensuring that the performance of the components delivered is in compliance with requirements and of making the necessary provisions if significant deviations are observed. The EFMT branch, which has been in charge of the instrumentation and acceptance of the different SG first models since the first PWR plants were commissioned, will be responsible for the acceptance tests and the ultimate validation of a performance assessment procedure applicable to the future replacement steam generators. The methods and tests proposed for SG expert appraisal are based on consideration of the importance of primary measurement quality for satisfactory SG assessment and of the new test facilities with which the 900 and 1 300 PWR plants are gradually being equipped. These facilities provide an on-site computer environment for tests compatible with the tools (PATTERN, etc.) used at EFMT and in other departments. This test is the first of this kind performed by EFMT and the test facility of a nuclear power plant. (author). 6 figs.

  19. A new correlation for convective heat transfer coefficient of water–alumina nanofluid in a square array subchannel under PWR condition

    Energy Technology Data Exchange (ETDEWEB)

    Shamim, Jubair A. [Department of Nuclear Engineering, Seoul National University, Seoul 08826 (Korea, Republic of); Department of Mechanical Engineering, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Bhowmik, Palash K. [Department of Nuclear Engineering, Seoul National University, Seoul 08826 (Korea, Republic of); Department of Nuclear Engineering, Missouri University of Science and Technology, 1201 N. State St., Rolla, MO 65409 (United States); Xiangyi, Chen [Department of Nuclear Engineering, Seoul National University, Seoul 08826 (Korea, Republic of); Suh, Kune Y., E-mail: kysuh@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, Seoul 08826 (Korea, Republic of)

    2016-11-15

    Highlights: • Thermo-hydrodynamic properties of water–Al{sub 2}O{sub 3} nanofluid at PWR condition is analyzed. • Details of CFD simulation and validation procedure is outlined. • Augmented heat transfer capacity of nanofluid is governed by larger pumping power. • A new correlation for nanofluid Nusselt number in subchannel geometry is proposed. - Abstract: The computational fluid dynamic (CFD) simulation is performed to determine on the thermo- and hydrodynamic performance of the water–alumina (Al{sub 2}O{sub 3}) nanofluid in a square array subchannel featuring pitch-to-diameter ratios of 1.25 and 1.35. Two fundamental aspects of thermal hydraulics, viz. heat transfer and pressure drop, are assessed under typical pressurized water reactor (PWR) conditions at various flow rates (3 × 10{sup 5} ⩽ Re ⩽ 6 × 10{sup 5}) using pure water and differing concentrations of water–alumina nanofluid (0.5–3.0 vol.%) as coolant. Numerical results are compared against predictions made by conventional single-phase convective heat transfer and pressure loss correlations for fully developed turbulent flow. It is observed that addition of tiny nanoparticles in PWR coolant can give rise to the convective heat transfer coefficient at the expense of larger pressure drop. Nevertheless, a modified correlation as a function of nanoparticle volume fraction is proposed to estimate nanofluid Nusselt number more precisely in square array subchannel.

  20. Simulation of Transient Viscoelastic Flow

    DEFF Research Database (Denmark)

    Rasmussen, Henrik Koblitz; Hassager, Ole

    1993-01-01

    The Lagrangian kinematic description is used to develop a numerical method for simulation of time-dependent flow of viscoelastic fluids described by integral models. The method is shown to converge to first order in the time step and at least second order in the spatial discretization. The method...... is tested on the established sphere in a cylinder benchmark problem, and an extension of the problem to transient flow is proposed....

  1. Simulation Model of a Transient

    DEFF Research Database (Denmark)

    Jauch, Clemens; Sørensen, Poul; Bak-Jensen, Birgitte

    2005-01-01

    This paper describes the simulation model of a controller that enables an active-stall wind turbine to ride through transient faults. The simulated wind turbine is connected to a simple model of a power system. Certain fault scenarios are specified and the turbine shall be able to sustain operation...... in case of such faults. The design of the controller is described and its performance assessed by simulations. The control strategies are explained and the behaviour of the turbine discussed....

  2. Transient virulence of emerging pathogens.

    Science.gov (United States)

    Bolker, Benjamin M; Nanda, Arjun; Shah, Dharmini

    2010-05-06

    Should emerging pathogens be unusually virulent? If so, why? Existing theories of virulence evolution based on a tradeoff between high transmission rates and long infectious periods imply that epidemic growth conditions will select for higher virulence, possibly leading to a transient peak in virulence near the beginning of an epidemic. This transient selection could lead to high virulence in emerging pathogens. Using a simple model of the epidemiological and evolutionary dynamics of emerging pathogens, along with rough estimates of parameters for pathogens such as severe acute respiratory syndrome, West Nile virus and myxomatosis, we estimated the potential magnitude and timing of such transient virulence peaks. Pathogens that are moderately evolvable, highly transmissible, and highly virulent at equilibrium could briefly double their virulence during an epidemic; thus, epidemic-phase selection could contribute significantly to the virulence of emerging pathogens. In order to further assess the potential significance of this mechanism, we bring together data from the literature for the shapes of tradeoff curves for several pathogens (myxomatosis, HIV, and a parasite of Daphnia) and the level of genetic variation for virulence for one (myxomatosis). We discuss the need for better data on tradeoff curves and genetic variance in order to evaluate the plausibility of various scenarios of virulence evolution.

  3. Cortical computations via transient attractors.

    Directory of Open Access Journals (Sweden)

    Oliver L C Rourke

    Full Text Available The ability of sensory networks to transiently store information on the scale of seconds can confer many advantages in processing time-varying stimuli. How a network could store information on such intermediate time scales, between typical neurophysiological time scales and those of long-term memory, is typically attributed to persistent neural activity. An alternative mechanism which might allow for such information storage is through temporary modifications to the neural connectivity which decay on the same second-long time scale as the underlying memories. Earlier work that has explored this method has done so by emphasizing one attractor from a limited, pre-defined set. Here, we describe an alternative, a Transient Attractor network, which can learn any pattern presented to it, store several simultaneously, and robustly recall them on demand using targeted probes in a manner reminiscent of Hopfield networks. We hypothesize that such functionality could be usefully embedded within sensory cortex, and allow for a flexibly-gated short-term memory, as well as conferring the ability of the network to perform automatic de-noising, and separation of input signals into distinct perceptual objects. We demonstrate that the stored information can be refreshed to extend storage time, is not sensitive to noise in the system, and can be turned on or off by simple neuromodulation. The diverse capabilities of transient attractors, as well as their resemblance to many features observed in sensory cortex, suggest the possibility that their actions might underlie neural processing in many sensory areas.

  4. Igniter heater EMI transient test

    Science.gov (United States)

    Cook, M.

    1989-01-01

    Testing to evaluate Redesigned Solid Rocket Motor igniter heater electromagnetic interference (EMI) effects on the Safe and Arm (S and A) device was completed. It was suspected that EMI generated by the igniter heater and it's associated electromechanical relay could cause a premature firing of the NASA Standard Initiators (NSIs) inside the S and A. The maximum voltage induced into the NSI fire lines was 1/4 of the NASA specified no-fire limit of one volt (SKB 26100066). As a result, the igniter heaters are not expected to have any adverse EMI effects on the NSIs. The results did show, however, that power switching causes occasional high transients within the igniter heater power cable. These transients could affect the sensitive equipment inside the forward skirt. It is therefore recommended that the electromechanical igniter heater relays be replaced with zero crossing solid state relays. If the solid state relays are installed, it is also recommended that they be tested for EMI transient effects.

  5. The effects of temperature and aeration on the corrosion of A508III low alloy steel in boric acid solutions at 25–95 °C

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, Qian [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steel, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Lu, Zhanpeng, E-mail: zplu@t.shu.edu.cn [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steel, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Shanghai Key Laboratory of Advanced Ferrometallurgy, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Chen, Junjie [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steel, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Yao, Meiyi [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steel, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Shanghai Key Laboratory of Advanced Ferrometallurgy, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Chen, Zhen; Ejaz, Ahsan [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China)

    2016-11-15

    The effects of temperature, solution composition and dissolved oxygen on the corrosion rate and electrochemical behavior of an A508III low alloy steel in boric acid solution with lithium hydroxide at 25–95 °C are investigated. In aerated solutions, increasing the boric acid concentration increases the corrosion rate and the anodic current density. The corrosion rate in deaerated solutions increases with increasing temperature. A corrosion rate peak value is found at approximately 75 °C in aerated solutions. Increasing temperature increases the oxygen diffusion coefficient, decreases the dissolved oxygen concentration, accelerates the hydrogen evolution reaction, and accelerates both the active dissolution and the film forming reactions. Increasing dissolved oxygen concentration does not significantly affect the corrosion rate at 50 and 60 °C, increases the corrosion rate at 70 and 80 °C, and decreases the corrosion rate at 87.5 and 95 °C in a high concentration boric acid solution with lithium hydroxide. - Highlights: • Effects of temperature on the corrosion of LAS in simulated PWR water are studied. • Increase of corrosion rate with increasing temperature in deaerated solution is observed. • Corrosion rate peaks at approximately 75 °C in aerated solution. • Deaeration decreases the corrosion rate in concentrated PWR water at 70–80 °C. • Deaeration increases the corrosion rate in concentrated PWR water at 87.5–95 °C.

  6. Thermal-hydraulically corrected neutron cross-sections for PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Santiago, Daniela M.N.; Alvim, Antonio C.M.; Silva, Fernando C., E-mail: dsantiago@con.ufrj.b, E-mail: alvim@con.ufrj.b, E-mail: fernando@con.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear

    2011-07-01

    Reactor core simulation codes ought to have a thermal-hydraulics feedback module. This module calculates, among other effects, the fuel temperature thermal-hydraulics feedback, that corrects neutron cross sections. In the nodal code developed at PEN/COPPE/UFRJ, the fuel temperature is obtained through an empirical model. This work presents a physical model to calculate this temperature. A finite volume technique was used to discretize the equation for temperature distribution, while the moderator coefficient of heat transfer was calculated using ASME routines, appended to the developed code. This model allows calculation of an average radial temperature per node, since the thermal-hydraulic feedback must follow the conditions imposed by the nodal code. The results obtained were compared with the ones obtained by the empirical model. The results show that, for fuel elements near core periphery, the empirical model overestimates the temperature in the fuel, as compared to our model, which may indicate that the physical model is more appropriate to calculate the thermal-hydraulic feedback temperatures. (author)

  7. Transient combustion in hybrid rockets

    Science.gov (United States)

    Karabeyoglu, Mustafa Arif

    1998-09-01

    Hybrid rockets regained interest recently as an alternative chemical propulsion system due to their advantages over the solid and liquid systems that are currently in use. Development efforts on hybrids revealed two important problem areas: (1) low frequency instabilities and (2) slow transient response. Both of these are closely related to the transient behavior which is a poorly understood aspect of hybrid operation. This thesis is mainly involved with a theoretical study of transient combustion in hybrid rockets. We follow the methodology of identifying and modeling the subsystems of the motor such as the thermal lags in the solid, boundary layer combustion and chamber gasdynamics from a dynamic point of view. We begin with the thermal lag in the solid which yield the regression rate for any given wall heat flux variation. Interesting phenomena such as overshooting during throttling and the amplification and phase lead regions in the frequency domain are discovered. Later we develop a quasi-steady transient hybrid combustion model supported with time delays for the boundary layer processes. This is integrated with the thermal lag system to obtain the thermal combustion (TC) coupled response. The TC coupled system with positive delays generated low frequency instabilities. The scaling of the instabilities are in good agreement with actual motor test data. Finally, we formulate a gasdynamic model for the hybrid chamber which successfully resolves the filling/emptying and longitudinal acoustic behavior of the motor. The TC coupled system is later integrated to the gasdynamic model to obtain the overall response (TCG coupled system) of gaseous oxidizer motors with stiff feed systems. Low frequency instabilities were also encountered for the TCG coupled system. Apart from the transient investigations, the regression rate behavior of liquefying hybrid propellants such as solid cryogenic materials are also studied. The theory is based on the possibility of enhancement

  8. Transient noise suppression algorithm in speech system

    Science.gov (United States)

    Cao, Keyu; Wang, Mingjiang

    2017-08-01

    In this paper, I mainly introduce the algorithm of transient noise suppression in speech system. Firstly, it divides into impulsive noise and other types of transient noise according to the characteristics of transient noise. In the impulse noise suppression algorithm, I mainly use the averaging energy threshold method to detect the impulse noise, and then I use the amplitude threshold method to reduce the impulse noise which was detected. In the other types of transient noise suppression algorithm, I mainly use the Optimally Modified-Log Spectral Amplitude estimation (OM-LSA) algorithm and the Minimum Control Recursive Average (MCRA) algorithm to suppress the transient noise.

  9. Analyses of PWR spent fuel composition using SCALE and SWAT code systems to find correction factors for criticality safety applications adopting burnup credit

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung; Suyama, Kenya; Mochizuki, Hiroki; Okuno, Hiroshi; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-01-01

    The isotopic composition calculations were performed for 26 spent fuel samples from the Obrigheim PWR reactor and 55 spent fuel samples from 7 PWR reactors using the SAS2H module of the SCALE4.4 code system with 27, 44 and 238 group cross-section libraries and the SWAT code system with the 107 group cross-section library. For the analyses of samples from the Obrigheim PWR reactor, geometrical models were constructed for each of SCALE4.4/SAS2H and SWAT. For the analyses of samples from 7 PWR reactors, the geometrical model already adopted in the SCALE/SAS2H was directly converted to the model of SWAT. The four kinds of calculation results were compared with the measured data. For convenience, the ratio of the measured to calculated values was used as a parameter. When the ratio is less than unity, the calculation overestimates the measurement, and the ratio becomes closer to unity, they have a better agreement. For many important nuclides for burnup credit criticality safety evaluation, the four methods applied in this study showed good coincidence with measurements in general. More precise observations showed, however: (1) Less unity ratios were found for Pu-239 and -241 for selected 16 samples out of the 26 samples from the Obrigheim reactor (10 samples were deselected because their burnups were measured with Cs-137 non-destructive method, less reliable than Nd-148 method the rest 16 samples were measured with); (2) Larger than unity ratios were found for Am-241 and Cm-242 for both the 16 and 55 samples; (3) Larger than unity ratios were found for Sm-149 for the 55 samples; (4) SWAT was generally accompanied by larger ratios than those of SAS2H with some exceptions. Based on the measured-to-calculated ratios for 71 samples of a combined set in which 16 selected samples and 55 samples were included, the correction factors that should be multiplied to the calculated isotopic compositions were generated for a conservative estimate of the neutron multiplication factor

  10. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant; Avaliacao da integridade de varetas combustiveis em reatores PWR a partir da analise espectrometrica da agua do primario

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Iara Arraes

    1999-02-15

    The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

  11. Modeling of transient electrical characteristics for granular semiconductors

    Science.gov (United States)

    Varpula, Aapo

    2010-08-01

    Modeling of the electrical large-signal response of granular n-type semiconductors is carried out at following three different levels: (i) simple fully analytical model, (ii) semianalytical numerical model, and (iii) numerical device simulation. The electrical transients induced by both voltage and temperature changes are calculated. The analysis is based on the dynamic electrical model of the grain-boundary (GB) region, the drift-diffusion theory, and electronic trapping in the acceptor-type electronic interface states at the GBs. The electronic trapping is described using the standard rate equation. The models are verified by performing numerical device simulations using SILVACO ATLAS. The agreement between the proposed semianalytical model and ATLAS results is excellent during the whole transient and up to rather high electric fields. Compared to ATLAS, the calculations performed with the present semianalytical model are four orders of magnitude faster on a standard PC computer. The approximative analytical formulas describing the response are valid when the voltage and temperature changes are small. The semianalytical model is also fitted to reported experimental data obtained from dc and transient measurements of ZnO powder samples. The semianalytical model fits to the data well. The current in the GB region has following three components: potential-barrier limited current, charging and discharging current, and capacitive current. The results show that the large-signal transient responses of granular semiconductors are complex, as they vary highly in both duration and magnitude. During a transient the current can change many orders of magnitude. This is mainly caused by the change in the GB trap occupancy.

  12. Analytical model of transient temperature and thermal stress in ...

    Indian Academy of Sciences (India)

    The result of this work is compared with a ... Keywords. Integral transform method; double end-pumped; laser rod; thermal stress. .... 2.2 Stress analysis. Although different methods have been used to predict failure stress, the maximum tensile hoop stress is proved to model the fracture in laser rod very well [9]. The hotter ...

  13. Chaotic Simulated Annealing by A Neural Network Model with Transient Chaos

    CERN Document Server

    Chen, L; Chen, Luonan; Aihara, Kazuyuki

    1997-01-01

    We propose a neural network model with transient chaos, or a transiently chaotic neural network (TCNN) as an approximation method for combinatorial optimization problem, by introducing transiently chaotic dynamics into neural networks. Unlike conventional neural networks only with point attractors, the proposed neural network has richer and more flexible dynamics, so that it can be expected to have higher ability of searching for globally optimal or near-optimal solutions. A significant property of this model is that the chaotic neurodynamics is temporarily generated for searching and self-organizing, and eventually vanishes with autonomous decreasing of a bifurcation parameter corresponding to the "temperature" in usual annealing process. Therefore, the neural network gradually approaches, through the transient chaos, to dynamical structure similar to such conventional models as the Hopfield neural network which converges to a stable equilibrium point. Since the optimization process of the transiently chaoti...

  14. Effects of RCP trip when recovering HPSI during LOCA in a Westinghouse PWR

    Energy Technology Data Exchange (ETDEWEB)

    Montero-Mayorga, Javier, E-mail: fj.montero@alumnos.upm.es; Queral, César; Rivas-Lewicky, Julio; González-Cadelo, Juan

    2014-12-15

    Highlights: • If HPSI is recovered during SBLOCA and RCPs are tripped core damage can be reached. • If the RCPs are tripped once the accumulators have injected the damage can be avoided. • If only 2 out of 3 RCPs are tripped the damage can be also avoided. • Improvements are proposed to the EOPs in order to avoid possible damage. - Abstract: Current Westinghouse Emergency Operating Procedures (EOPs) indicate initially that the operator must keep the reactor coolant pumps (RCPs) running during a Small Break Loss of Coolant Accident (SBLOCA) if there is unavailability of high pressure safety injection (HPSI) system in order to cool the core by forced convection. However, the crew must follow different EOPs along the transient depending on its evolution. In these EOPs there are several conditions which indicate the necessity of tripping one or more RCPs when HPSI is recovered. In this paper the occurrence of a SBLOCA with unavailability of HPSI has been analyzed with a model of Almaraz Nuclear Power Plant (Westinghouse 3 Loop) for TRACE code V5.0 patch 1. Two different approaches have been considered: the first one, taking into account Optimal Recovery Guidelines (ORGs) and in the second approach, the transition to Function Restoration Guidelines (FRGs) due to inadequate core cooling (ICC) conditions is considered. Results of this paper lead to the implementation of an improvement in current EOPs regarding how many RCPs should be tripped during SBLOCA sequences.

  15. Irradiation effects test series, test IE-5. Test results report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Croucher, D. W.; Yackle, T. R.; Allison, C. M.; Ploger, S. A.

    1978-01-01

    Test IE-5, conducted in the Power Burst Facility at the Idaho National Engineering Laboratory, employed three 0.97-m long pressurized water reactor type fuel rods, fabricated from previously irradiated zircaloy-4 cladding and one similar rod fabricated from unirradiated cladding. The objectives of the test were to evaluate the influence of simulated fission products, cladding irradiation damage, and fuel rod internal pressure on pellet-cladding interaction during a power ramp and on fuel rod behavior during film boiling operation. The four rods were subjected to a preconditioning period, a power ramp to an average fuel rod peak power of 65 kW/m, and steady state operation for one hour at a coolant mass flux of 4880 kg/s-m/sup 2/ for each rod. After a flow reduction to 1800 kg/s-m/sup 2/, film boiling occurred on one rod. Additional flow reductions to 970 kg/s-m/sup 2/ produced film boiling on the three remaining fuel rods. Maximum time in film boiling was 80s. The rod having the highest initial internal pressure (8.3 MPa) failed 10s after the onset of film boiling. A second rod failed about 90s after reactor shutdown. The report contains a description of the experiment, the test conduct, test results, and results from the preliminary postirradiation examination. Calculations using a transient fuel rod behavior code are compared with the test results.

  16. Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J. [Tractebel Energy Engineering, Brussels (Belgium)

    2001-07-01

    A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

  17. Thermal–stress analysis on the crack formation of tungsten during fusion relevant transient heat loads

    Directory of Open Access Journals (Sweden)

    Changjun Li

    2017-12-01

    Full Text Available In the future fusion devices, ELMs-induced transient heat flux may lead to the surface cracking of tungsten (W based plasma-facing materials (PFMs. In theory, the cracking is related to the material fracture toughness and the thermal stress-strain caused by transient heat flux. In this paper, a finite element model was successfully built to realize a theoretical semi infinite space. The temperature and stress-strain distribution as well as evolution of W during a single heating-cooling cycle of transient heat flux were simulated and analyzed. It showed that the generation of plastic deformation during the brittle temperature range between room temperature and DBTT (ductile to brittle transition temperature, ∼400 °C caused the cracking of W during the cooling phase. The cracking threshold for W under transient heat flux was successfully obtained by finite element analysis, to some extent, in consistent with the similar experimental results. Both the heat flux factors (FHF = P·t0.5 and the maximum surface temperatures at cracking thresholds were almost invariant for the transient heat fluxes with different pulse widths and temporal distributions. This method not only identified the theoretical conclusion but also obtained the detail values for W with actual temperature-dependent properties.

  18. Transient or permanent fisheye views

    DEFF Research Database (Denmark)

    Jakobsen, Mikkel Rønne; Hornbæk, Kasper

    2012-01-01

    programming environment. Fourteen participants performed varied tasks involving navigation and understanding of source code. Participants used the three interfaces for between four and six hours in all. Time and accuracy measures were inconclusive, but subjective data showed a preference for the permanent......, about the benefits and limitations of transient visualizations. We describe an experiment that compares the usability of a fisheye view that participants could call up temporarily, a permanent fisheye view, and a linear view: all interfaces gave access to source code in the editor of a widespread...

  19. Design, Construction and Testing of an In-Pile Loop for PWR (Pressurized Water Reactor) Simulation.

    Science.gov (United States)

    1987-06-01

    concentrations. It was found that raising coolant pH would decrease visual crud concentrations, but steam generator plenum and primary coolant pipe wall...built, the Loop duplicates the core and Steam Generator fluid surface film differential temperatures, bulk fluid temperatures, and wall fluid shear...2.4.2 Loop Plenum Design 39 2.4.3 Neutron Absorber Plug 44 2.4.4 Plenum Section Instrumentation 45 2.5 Steam Generator /Heat Rejection Section 45 5 2.5.1

  20. A prediction model of IASCC initiation stress for bolts in PWR core internals

    Energy Technology Data Exchange (ETDEWEB)

    Fukuya, Koji, E-mail: fukuya@inss.co.j [Institute of Nuclear Safety System, 64 Sata, Mihama, Mikata, Fukui 919-1205 (Japan); Fujii, Katsuhiko; Nishioka, Hiromasa [Institute of Nuclear Safety System, 64 Sata, Mihama, Mikata, Fukui 919-1205 (Japan); Takakura, Kenichi; Nakata, Kiyotomo [Japan Nuclear Energy Safety Organization, 3-17-1 Toranomon, Minato, Tokyo 105-0001 (Japan)

    2010-03-15

    A model of IASCC initiation stress for bolts of core internals in pressurized water reactors was developed considering differences in material property changes due to irradiation and material conditions. Assuming that IASCC initiation was controlled by grain boundary composition and yield strength, these values for each specimen of post-irradiation IASCC initiation tests were calculated by physical kinetic models considering dose rate, temperature, material composition and surface hardening. Then, correlations of grain boundary composition and yield strength with IASCC initiation stress were determined. The model predicted that the IASCC initiation stress became lower with dose and was lower for higher temperature, lower flux and higher surface hardening level.