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Sample records for core expansion reactivity

  1. Radial core expansion reactivity feedback in advanced LMRs: uncertainties and their effects on inherent safety

    International Nuclear Information System (INIS)

    Wigeland, R.A.; Moran, T.J.

    1988-01-01

    An analytical model for calculating radial core expansion, based on the thermal and elastic bowing of a single subassembly at the core periphery, is used to quantify the effect of uncertainties on this reactivity feedback mechanism. This model has been verified and validated with experimental and numerical results. The impact of these uncertainties on the safety margins in unprotected transients is investigated with SASSYS/SAS4A, which includes this model for calculating the reactivity feedback from radial core expansion. The magnitudes of these uncertainties are not sufficient to preclude the use of radial core expansion reactivity feedback in transient analysis

  2. A simple reactivity feedback model accounting for radial core expansion effects in the liquid metal fast reactor

    International Nuclear Information System (INIS)

    Kwon, Young Min; Lee, Yong Bum; Chang, Won Pyo; Haha, Do Hee

    2002-01-01

    The radial core expansion due to the structure temperature rise is one of major negative reactivity insertion mechanisms in metallic fueled reactor. Thermal expansion is a result of both the laws of nature and the particular core design and it causes negative reactivity feedback by the combination of increased core volume captures and increased core surface leakage. The simple radial core expansion reactivity feedback model developed for the SSC-K code was evaluated by the code-to-code comparison analysis. From the comparison results, it can be stated that the radial core expansion reactivity feedback model employed into the SSC-K code may be reasonably accurate in the UTOP analysis

  3. A simple reactivity feedback model accounting for radial core expansion effects in the liquid metal fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Min; Lee, Yong Bum; Chang, Won Pyo; Haha, Do Hee [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    The radial core expansion due to the structure temperature rise is one of major negative reactivity insertion mechanisms in metallic fueled reactor. Thermal expansion is a result of both the laws of nature and the particular core design and it causes negative reactivity feedback by the combination of increased core volume captures and increased core surface leakage. The simple radial core expansion reactivity feedback model developed for the SSC-K code was evaluated by the code-to-code comparison analysis. From the comparison results, it can be stated that the radial core expansion reactivity feedback model employed into the SSC-K code may be reasonably accurate in the UTOP analysis.

  4. Reactivity variations associated with the core expansion of the MARIA research reactor after modernisation

    International Nuclear Information System (INIS)

    Krzysztoszek, G.

    1997-01-01

    Polish high flux research reactor MARIA is a pool type reactor moderated with beryllium and water and cooled with water. The fuel is 80% enriched uranium, in the shape of multitube fuel elements, each tube made up of UAl x alloy in aluminium cladding. MARIA reactor has been operated in the years of 1977-85 and then it was modernised and again put into operation in December 1992. The modernisation as regarded the reactor core comprises a beryllium matrix expansion from 20-48 blocks. Within the frame of the power start-up and trial operation the reactor has been extended from 12 to 18 fuel channels. On that stage of reactor operation the power of mostly loaded fuel channels was constrained to 1,6 MW. Reactor has been operated within the 100-hrs campaign for an irradiation of target materials and for performing measurements at the horizontal channel outlets. In the previous time it has been noticed substantial differences in reactivity changes of the core in similar campaigns of reactor operation. It concerns the reactivity losses during poisoning period of the reactor within the first 30-40 hrs of operation as well as in the fuel burning up process. An analysis of the reactivity variations during the core extension will made possible the fuel management optimisation in further reactor operation system. (author)

  5. Comparison of the SASSYS/SAS4A radial core expansion reactivity feedback model and the empirical correlation for FFTF

    International Nuclear Information System (INIS)

    Wigeland, R.A.

    1987-01-01

    The present emphasis on inherent safety for LMR designs has resulted in a need to represent the various reactivity feedback mechanisms as accurately as possible. The dominant negative reactivity feedback has been found to result from radial expansion of the core for most postulated ATWS events. For this reason, a more detailed model for calculating the reactivity feedback from radial core expansion has been recently developed for use with the SASSYS/SAS4A Code System. The purpose of this summary is to present an extension to the model so that it is more suitable for handling a core restraint design as used in FFTF, and to compare the SASSYS/SAS4A results using this model to the empirical correlation presently being used to account for radial core expansion reactivity feedback to FFTF

  6. The reactivity meter and core reactivity

    International Nuclear Information System (INIS)

    Siltanen, P.

    1999-01-01

    This paper discussed in depth the point kinetic equations and the characteristics of the point kinetic reactivity meter, particularly for large negative reactivities. From a given input signal representing the neutron flux seen by a detector, the meter computes a value of reactivity in dollars (ρ/β), based on inverse point kinetics. The prompt jump point of view is emphasised. (Author)

  7. Critical experiment tests of bowing and expansion reactivity calculations for LMRS

    International Nuclear Information System (INIS)

    Schaefer, R.W.

    1988-01-01

    Experiments done in several LMR-type critical assemblies simulated core axial expansion, core radial expansion and bowing, coolant expansion, and control driveline expansion. For the most part new experimental techniques were developed to do these experiments. Calculations of the experiments basically used design-level methods, except when it was necessary to investigate complexities peculiar to the experiments. It was found that these feedback reactivities generally are overpredicted, but the predictions are within 30% of the experimental values. 14 refs., 2 figs., 4 tabs

  8. Techniques for computing reactivity changes caused by fuel axial expansion in LMR's

    International Nuclear Information System (INIS)

    Khalil, H.

    1988-01-01

    An evaluation is made of the accuracy of methods used to compute reactivity changes caused by axial fuel relocation in fast reactors. Results are presented to demonstrate the validity of assumptions commonly made such as linearity of reactivity with fuel elongation, additivity of local reactivity contributions, and the adequacy of standard perturbation techniques. Accurate prediction of the reactivity loss caused by axial swelling of metallic fuel is shown to require proper representation of the burnup dependence of the expansion reactivity. Some accuracy limitations in the methods used in transient analyses, which are based on the use of fuel worth tables, are identified, and efficient ways to improve accuracy are described. Implementation of these corrections produced expansion reactivity estimates within 5% of higher-order method for a metal-fueled FFTF core representation. 18 refs., 3 figs., 3 tabs

  9. Reactivity accident analysis in MTR cores

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  10. Reactivity analysis of core distortion effects in the FFTF

    International Nuclear Information System (INIS)

    Knutson, B.J.

    1982-01-01

    An improved technique for evaluating core distortion reactivity effects was developed using reactivity analyses of two core geometry models (R-Z and HEX). This technique is incorporated into a new processor code called CORDIS. The advantages of this technique over existing reactivity models are that is preserves core heterogeneity, provides a control rod insertion effect model, uses row-dependent axial shape functions, and provides a flexible and cost efficient core distortion reactivity analysis method

  11. Inherent safety that the reactivity effect of core bending in fast reactors brings about

    International Nuclear Information System (INIS)

    Nakagawa, Masatoshi; Yagawa, Genki.

    1994-01-01

    FBRs have the merit on safety by low operation pressure and the large heat capacity of coolant, in addition, due to the core temperature rise at the time of accidents and the thermal expansion of core structures, the negative feedback of reactivity can be expected. Recently, attention has been paid to the negative feedback of reactivity due to core bending. It can be expected also in the core of limited free bow type. Bending is caused by the difference of thermal expansion on six surfaces of hexagonal wrapper tubes. The bending changes core reactivity and exerts effects to fuel exchange force and operation, insertion of control rods and the structural soundness of fuel assemblies. for the purpose of limiting the effect that core bending exerts to core characteristics to allowable range, core constraint mechanism is installed. The behavior of core bending at the time of anticipated transient without scram is explained. The example of the analysis of PRISM reactor is shown. The experiment that confirmed the negative feedback of reactivity due to core bending under the condition of ULOF was that at the fast flux test facility. (K.I.)

  12. Reactivity worth of gas expansion modules (GEMs) in the fast flux test facility

    International Nuclear Information System (INIS)

    Campbell, L.R.; Nelson, J.V.; Burke, T.M.; Rawlins, J.A.; Daughtry, J.W.; Bennett, R.A.

    1986-01-01

    A new passive shutdown device called a gas expansion module (GEM) has been developed at Hanford Engineering Development Laboratory to insert negative reactivity during a primary system loss of flow in a liquid-metal reactor (LMR). A GEM is a hollow removable core component which is sealed at the top and open at the bottom. An argon gas bubble trapped inside the assembly expands when core inlet pressure decreases (caused by a flow reduction) and expels sodium from the assembly. The GEMs are designed so that the level of the liquid-sodium primary system coolant within a GEM is above the top of the core when the primary pumps are operating at full flow and is below the bottom of the core when the primary pumps are off. When a GEM is placed at the boundary of the core and radial reflector, the drop in sodium level increases core neutron leakage and inserts negative reactivity. The results of these measurements confirm the effectiveness of GEMs in adding negative reactivity in loss-of-flow situations. It follows, therefore, that the inherent safety of LMRs, comparable in size to the FFTF, can be enhanced by the use of GEMs

  13. Initial Comparison of Direct and Legacy Modeling Approaches for Radial Core Expansion Analysis

    International Nuclear Information System (INIS)

    Shemon, Emily R.

    2016-01-01

    Radial core expansion in sodium-cooled fast reactors provides an important reactivity feedback effect. As the reactor power increases due to normal start up conditions or accident scenarios, the core and surrounding materials heat up, causing both grid plate expansion and bowing of the assembly ducts. When the core restraint system is designed correctly, the resulting structural deformations introduce negative reactivity which decreases the reactor power. Historically, an indirect procedure has been used to estimate the reactivity feedback due to structural deformation which relies upon perturbation theory and coupling legacy physics codes with limited geometry capabilities. With advancements in modeling and simulation, radial core expansion phenomena can now be modeled directly, providing an assessment of the accuracy of the reactivity feedback coefficients generated by indirect legacy methods. Recently a new capability was added to the PROTEUS-SN unstructured geometry neutron transport solver to analyze deformed meshes quickly and directly. By supplying the deformed mesh in addition to the base configuration input files, PROTEUS-SN automatically processes material adjustments including calculation of region densities to conserve mass, calculation of isotopic densities according to material models (for example, sodium density as a function of temperature), and subsequent re-homogenization of materials. To verify the new capability of directly simulating deformed meshes, PROTEUS-SN was used to compute reactivity feedback for a series of contrived yet representative deformed configurations for the Advanced Burner Test Reactor design. The indirect legacy procedure was also performed to generate reactivity feedback coefficients for the same deformed configurations. Interestingly, the legacy procedure consistently overestimated reactivity feedbacks by 35% compared to direct simulations by PROTEUS-SN. This overestimation indicates that the legacy procedures are in fact

  14. Initial Comparison of Direct and Legacy Modeling Approaches for Radial Core Expansion Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shemon, Emily R. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-10-10

    Radial core expansion in sodium-cooled fast reactors provides an important reactivity feedback effect. As the reactor power increases due to normal start up conditions or accident scenarios, the core and surrounding materials heat up, causing both grid plate expansion and bowing of the assembly ducts. When the core restraint system is designed correctly, the resulting structural deformations introduce negative reactivity which decreases the reactor power. Historically, an indirect procedure has been used to estimate the reactivity feedback due to structural deformation which relies upon perturbation theory and coupling legacy physics codes with limited geometry capabilities. With advancements in modeling and simulation, radial core expansion phenomena can now be modeled directly, providing an assessment of the accuracy of the reactivity feedback coefficients generated by indirect legacy methods. Recently a new capability was added to the PROTEUS-SN unstructured geometry neutron transport solver to analyze deformed meshes quickly and directly. By supplying the deformed mesh in addition to the base configuration input files, PROTEUS-SN automatically processes material adjustments including calculation of region densities to conserve mass, calculation of isotopic densities according to material models (for example, sodium density as a function of temperature), and subsequent re-homogenization of materials. To verify the new capability of directly simulating deformed meshes, PROTEUS-SN was used to compute reactivity feedback for a series of contrived yet representative deformed configurations for the Advanced Burner Test Reactor design. The indirect legacy procedure was also performed to generate reactivity feedback coefficients for the same deformed configurations. Interestingly, the legacy procedure consistently overestimated reactivity feedbacks by 35% compared to direct simulations by PROTEUS-SN. This overestimation indicates that the legacy procedures are in fact

  15. Reactor-core-reactivity control device

    International Nuclear Information System (INIS)

    Miura, Teruo; Sakuranaga, Tomonobu.

    1983-01-01

    Purpose: To improve the reactor safety upon failures of control rod drives by adapting a control rod not to drop out accidentally from the reactor core but be inserted into the reactor core. Constitution: The control rod is entered or extracted as usual from the bottom of the pressure vessel. A space is provided above the reactor core within the pressure vessel, in which the moving scope of the control rod is set between the space above the reactor core and the reactor core. That is, the control rod is situated above the reactor core upon extraction thereof and, if an accident occurs to the control rod drive mechanisms to detach the control rod and the driving rod, the control rod falls gravitationally into the reactor core to improve the reactor safety. In addition, since the speed limiter is no more required to the control rod, the driving force can be decreased to reduce the size of the rod drive mechanisms. (Ikeda, J.)

  16. Investigation of Reactivity Feedback Mechanism of Axial and Radial Expansion Effect of Metal-Fueled Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Seong, Seung-Hwan; Choi, Chi-Woong; Jeong, Tae-Kyung; Ha, Gi-Seok

    2015-01-01

    The major inherent reactivity feedback models for a ceramic fuel used in a conventional light water reactor are Doppler feedback and moderator feedback. The metal fuel has these two reactivity feedback mechanisms previously mentioned. In addition, the metal fuel has two more reactivity feedback models related to the thermal expansion phenomena of the metal fuel. Since the metal fuel has a good capability to expand according to the temperature changes of the core, two more feedback mechanisms exist. These additional two feedback mechanism are important to the inherent safety of metal fuel and can make metal-fueled SFR safer than oxide-fueled SFR. These phenomena have already been applied to safety analysis on design extended condition. In this study, the effect of these characteristics on power control capability was examined through a simple load change operation. The axial expansion mechanism is induced from the change of the fuel temperature according to the change of the power level of PGSFR. When the power increases, the fuel temperatures in the metal fuel will increase and then the reactivity will decrease due to the axial elongation of the metal fuel. To evaluate the expansion effect, 2 cases were simulated with the same scenario by using MMS-LMR code developed at KAERI. The first simulation was to analyze the change of the reactor power according to the change of BOP power without the reactivity feedback model of the axial and radial expansion of the core during the power transient event. That is to say, the core had only two reactivity feedback mechanism of Doppler and coolant temperature

  17. Sensitivity of reactivity feedback due to core bowing in a metallic-fueled core

    International Nuclear Information System (INIS)

    Nakagawa, Masatoshi; Kawashima, Masatoshi; Endo, Hiroshi; Nishimura, Tomohiro

    1991-01-01

    A sensitivity study has been carried out on negative reactivity feedback caused by core bowing to assess the potential effectiveness of FBR passive safety features in regard to withstanding an anticipated transient without scram (ATWS). In the present study, an analysis has been carried to obtain the best material and geometrical conditions concerning the core restraint system out for several power to flow rates (P/F), up to 2.0 for a 300 MWe metallic-fueled core. From this study, it was clarified that the pad stiffness at an above core loading pads (ACLP) needs to be large enough to ensure negative reactivity feedback against ATWS. It was also clarified that there is an upper limit for the clearances between ducts at ACLP. A new concept, in regard to increasing the absolute value for negative reactivity feedback due to core bowing at ATWS, is proposed and discussed. (author)

  18. Analytic function expansion nodal method for nuclear reactor core design

    International Nuclear Information System (INIS)

    Noh, Hae Man

    1995-02-01

    In most advanced nodal methods the transverse integration is commonly used to reduce the multi-dimensional diffusion equation into equivalent one- dimensional diffusion equations when derving the nodal coupling equations. But the use of the transverse integration results in some limitations. The first limitation is that the transverse leakage term which appears in the transverse integration procedure must be appropriately approximated. The second limitation is that the one-dimensional flux shapes in each spatial direction resulted from the nodal calculation are not accurate enough to be directly used in reconstructing the pinwise flux distributions. Finally the transverse leakage defined for a non-rectangular node such as a hexagonal node or a triangular node is too complicated to be easily handled and may contain non-physical singular terms of step-function and delta-function types. In this thesis, the Analytic Function Expansion Nodal (AFEN) method and its two variations : the Polynomial Expansion Nodal (PEN) method and the hybrid of the AFEN and PEN methods, have been developed to overcome the limitations of the transverse integration procedure. All of the methods solve the multidimensional diffusion equation without the transverse integration. The AFEN method which we believe is the major contribution of this study to the reactor core analysis expands the homogeneous flux distributions within a node in non-separable analytic basis functions satisfying the neutron diffusion equations at any point of the node and expresses the coefficients of the flux expansion in terms of the nodal unknowns which comprise a node-average flux, node-interface fluxes, and corner-point fluxes. Then, the nodal coupling equations composed of the neutron balance equations, the interface current continuity equations, and the corner-point leakage balance equations are solved iteratively to determine all the nodal unknowns. Since the AFEN method does not use the transverse integration in

  19. Void coefficient of reactivity calculation for AP-600 core

    International Nuclear Information System (INIS)

    Suparlina, L.; Budiono, T.A.; Mardha, A.; Tukiran

    1998-01-01

    Void coefficient of reactivity as one of reactor kinetics parameters has been carried out. The calculation was done into two steps which is cell calculation using WIMSD/4 and core calculation using Batan-2DIFF code programs with the condition of beginning of cycle with all fresh fuels elements and all control rods withdrawn. The one dimension transport program in four neutron energy groups is used to calculate the cell generation of various core materials cell has been calculated in 1/4 fuel element with cluster model and square pitch arrange. Moderator density have been reduced until 20% for the void coefficient of reactivity calculation. Macroscopic cross-section as the out put of WIMSD/4 is being used as the input at the diffusion neutron program for core calculation. The void coefficient of reactivity of the AP-600 core can be determined with regular neutron flux and adjoint in four energy groups and X-Y geometry. The results is shown that the K eff calculation value is different 5.2% from the design data

  20. Moderator temperature effects on reactivity of HEU core of MNSR

    International Nuclear Information System (INIS)

    Ahmad, Siraj-ul-Islam; Sahibzada, Tasveer Muhammad

    2012-01-01

    Highlights: ► The MNSR core was analyzed to see the cross section effects on moderator temperature coefficient of reactivity. ► WIMS-D code was used for cell calculations. ► The 3D diffusion theory code PRIDE was first validated using IAEA benchmark problem and then used for analysis of MNSR. ► The differences among results for various libraries were discussed. -- Abstract: In this article we report on analyses that were performed to investigate the influence of cross section differences among libraries released by various centers on reactivity of Miniature Neutron Source Reactors. The 3D model of the core was developed with WIMS-D and PRIDE codes and six cross section libraries were used including JENDL-3.2, JEF-2.2, JEFF-3.3, ENDF/B-VI and ENDF/B-VII, and IAEA library. It was observed that all the libraries predict the reactivity within 10%, with IAEA library giving minimum reactivity worth, and JEF-2.2 data library resulted in highest worth.

  1. Development and Investigation of Reactivity Measurement Methods in Subcritical Cores

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Johanna

    2005-05-01

    Subcriticality measurements during core loading and in future accelerator driven systems have a clear safety relevance. In this thesis two subcriticality methods are treated: the Feynman-alpha and the source modulation method. The Feynman-alpha method is a technique to determine the reactivity from the relative variance of the detector counts during a measurement period. The period length is varied to get the full time dependence of the variance-to-mean. The corresponding theoretical formula was known only with stationary sources. In this thesis, due to its relevance for novel reactivity measurement methods, the Feynman-alpha formulae for pulsed sources for both the stochastic and the deterministic cases are treated. Formulae neglecting as well as including the delayed neutrons are derived. The formulae neglecting delayed neutrons are experimentally verified with quite good agreement. The second reactivity measurement technique investigated in this thesis is the so-called source modulation technique. The theory of the method was elaborated on the assumption of point kinetics, but in practice the method will be applied by using the signal from a single local neutron detector. Applicability of the method therefore assumes point kinetic behaviour of the core. Hence, first the conditions of the point kinetic behaviour of subcritical cores was investigated. After that the performance of the source modulation technique in the general case as well as and in the limit of exact point kinetic behaviour was examined. We obtained the unexpected result that the method has a finite, non-negligible error even in the limit of point kinetic behaviour, and a substantial error in the operation range of future accelerator driven subcritical reactors (ADS). In practice therefore the method needs to be calibrated by some other method for on-line applications.

  2. Development and Investigation of Reactivity Measurement Methods in Subcritical Cores

    International Nuclear Information System (INIS)

    Wright, Johanna

    2005-05-01

    Subcriticality measurements during core loading and in future accelerator driven systems have a clear safety relevance. In this thesis two subcriticality methods are treated: the Feynman-alpha and the source modulation method. The Feynman-alpha method is a technique to determine the reactivity from the relative variance of the detector counts during a measurement period. The period length is varied to get the full time dependence of the variance-to-mean. The corresponding theoretical formula was known only with stationary sources. In this thesis, due to its relevance for novel reactivity measurement methods, the Feynman-alpha formulae for pulsed sources for both the stochastic and the deterministic cases are treated. Formulae neglecting as well as including the delayed neutrons are derived. The formulae neglecting delayed neutrons are experimentally verified with quite good agreement. The second reactivity measurement technique investigated in this thesis is the so-called source modulation technique. The theory of the method was elaborated on the assumption of point kinetics, but in practice the method will be applied by using the signal from a single local neutron detector. Applicability of the method therefore assumes point kinetic behaviour of the core. Hence, first the conditions of the point kinetic behaviour of subcritical cores was investigated. After that the performance of the source modulation technique in the general case as well as and in the limit of exact point kinetic behaviour was examined. We obtained the unexpected result that the method has a finite, non-negligible error even in the limit of point kinetic behaviour, and a substantial error in the operation range of future accelerator driven subcritical reactors (ADS). In practice therefore the method needs to be calibrated by some other method for on-line applications

  3. Three-dimensional core analysis on a super fast reactor with negative local void reactivity

    International Nuclear Information System (INIS)

    Cao Liangzhi; Oka, Yoshiaki; Ishiwatari, Yuki; Ikejiri, Satoshi

    2009-01-01

    Keeping negative void reactivity throughout the cycle life is one of the most important requirements for the design of a supercritical water-cooled fast reactor (super fast reactor). Previous conceptual design has negative overall void reactivity. But the local void reactivity, which is defined as the reactivity change when the coolant of one fuel assembly disappears, also needs to be kept negative throughout the cycle life because the super fast reactor is designed with closed fuel assemblies. The mechanism of the local void reactivity is theoretically analyzed from the neutrons balance point of view. Three-dimensional neutronics/thermal-hydraulic coupling calculation is employed to analyze the characteristics of the super fast reactor including the local void reactivity. Some configurations of the core are optimized to decrease the local void reactivity. A reference core is successfully designed with keeping both overall and local void reactivity negative. The maximum local void reactivity is less than -30 pcm

  4. Core reactivity estimation in space reactors using recurrent dynamic networks

    Science.gov (United States)

    Parlos, Alexander G.; Tsai, Wei K.

    1991-01-01

    A recurrent multilayer perceptron network topology is used in the identification of nonlinear dynamic systems from only the input/output measurements. The identification is performed in the discrete time domain, with the learning algorithm being a modified form of the back propagation (BP) rule. The recurrent dynamic network (RDN) developed is applied for the total core reactivity prediction of a spacecraft reactor from only neutronic power level measurements. Results indicate that the RDN can reproduce the nonlinear response of the reactor while keeping the number of nodes roughly equal to the relative order of the system. As accuracy requirements are increased, the number of required nodes also increases, however, the order of the RDN necessary to obtain such results is still in the same order of magnitude as the order of the mathematical model of the system. It is believed that use of the recurrent MLP structure with a variety of different learning algorithms may prove useful in utilizing artificial neural networks for recognition, classification, and prediction of dynamic systems.

  5. Combination of AC Transmission Expansion Planning and Reactive Power Planning in the restructured power system

    International Nuclear Information System (INIS)

    Hooshmand, Rahmat-Allah; Hemmati, Reza; Parastegari, Moein

    2012-01-01

    Highlights: ► To overcome the disadvantages of DC model in Transmission Expansion Planning, AC model should be used. ► The Transmission Expansion Planning associated with Reactive Power Planning results in fewer new transmission lines. ► Electricity market concepts should be considered in Transmission Expansion Planning problem. ► Reliability aspects should be considered in Transmission Expansion Planning problem. ► Particle Swarm Optimization is a suitable optimization method to solve Transmission Expansion Planning problem. - Abstract: Transmission Expansion Planning (TEP) is an important issue in power system studies. It involves decisions on location and number of new transmission lines. Before deregulation of the power system, the goal of TEP problem was investment cost minimization. But in the restructured power system, nodal prices, congestion management, congestion surplus and so on, have been considered too. In this paper, an AC model of TEP problem (AC-TEP) associated with Reactive Power Planning (RPP) is presented. The goals of the proposed planning problem are to minimize investment cost and maximize social benefit at the same time. In the proposed planning problem, in order to improve the reliability of the system the Expected Energy Not Supplied (EENS) index of the system is limited by a constraint. For this purpose, Monte Carlo simulation method is used to determine the EENS. Particle Swarm Optimization (PSO) method is used to solve the proposed planning problem which is a nonlinear mixed integer optimization problem. Simulation results on Garver and RTS systems verify the effectiveness of the proposed planning problem for reduction of the total investment cost, EENS index and also increasing social welfare of the system.

  6. The effect of core configuration on temperature coefficient of reactivity in IRR-1

    Energy Technology Data Exchange (ETDEWEB)

    Bettan, M.; Silverman, I.; Shapira, M.; Nagler, A. [Soreq Nuclear Research Center, Yavne (Israel)

    1997-08-01

    Experiments designed to measure the effect of coolant moderator temperature on core reactivity in an HEU swimming pool type reactor were performed. The moderator temperature coefficient of reactivity ({alpha}{sub {omega}}) was obtained and found to be different in two core loadings. The measured {alpha}{sub {omega}} of one core loading was {minus}13 pcm/{degrees}C at the temperature range of 23-30{degrees}C. This value of {alpha}{sub {omega}} is comparable to the data published by the IAEA. The {alpha}{sub {omega}} measured in the second core loading was found to be {minus}8 pcm/{degrees}C at the same temperature range. Another phenomenon considered in this study is core behavior during reactivity insertion transient. The results were compared to a core simulation using the Dynamic Simulator for Nuclear Power Plants. It was found that in the second core loading factors other than the moderator temperature influence the core reactivity more than expected. These effects proved to be extremely dependent on core configuration and may in certain core loadings render the reactor`s reactivity coefficient undesirable.

  7. Core expansion in young star clusters in the Large Magellanic Cloud

    International Nuclear Information System (INIS)

    Elson, R.A.W.; Freeman, K.C.; Lauer, T.R.

    1989-01-01

    The core radii of 18 rich star clusters in the LMC with ages from 10 Myr to 1 Gyr. Data for an additional 17 clusters with ages from 1 Myr to 10 Gyr are available in the literature. The combined sample shows that the core radii increase from about 0 to about 5 pc between about 1 Myr and 1 Gyr, and then begin to decrease again. The expansion of the cores is probably driven by mass loss from evolving stars. Models of cluster evolution show that the rate of increase in core radius is sensitive to the slope of the initial mass function. The observed core radius-age relation for the LMC clusters favors an intial mass function with slope slightly flatter than the Salpeter value. 20 refs

  8. On-line reconstruction of in-core power distribution by harmonics expansion method

    International Nuclear Information System (INIS)

    Wang Changhui; Wu Hongchun; Cao Liangzhi; Yang Ping

    2011-01-01

    Highlights: → A harmonics expansion method for the on-line in-core power reconstruction is proposed. → A harmonics data library is pre-generated off-line and a code named COMS is developed. → Numerical results show that the maximum relative error of the reconstruction is less than 5.5%. → This method has a high computational speed compared to traditional methods. - Abstract: Fixed in-core detectors are most suitable in real-time response to in-core power distributions in pressurized water reactors (PWRs). In this paper, a harmonics expansion method is used to reconstruct the in-core power distribution of a PWR on-line. In this method, the in-core power distribution is expanded by the harmonics of one reference case. The expansion coefficients are calculated using signals provided by fixed in-core detectors. To conserve computing time and improve reconstruction precision, a harmonics data library containing the harmonics of different reference cases is constructed. Upon reconstruction of the in-core power distribution on-line, the two closest reference cases are searched from the harmonics data library to produce expanded harmonics by interpolation. The Unit 1 reactor of DayaBay Nuclear Power Plant (DayaBay NPP) in China is considered for verification. The maximum relative error between the measurement and reconstruction results is less than 5.5%, and the computing time is about 0.53 s for a single reconstruction, indicating that this method is suitable for the on-line monitoring of PWRs.

  9. Study on the reactivity behavior partially loaded reactor cores using SIMULATE-3

    International Nuclear Information System (INIS)

    Holzer, Robert; Zeitz, Andreas; Grimminger, Werner; Lubczyk, Tobias

    2009-01-01

    The reactor core design for the NPP Gundremmingen unit B and C is performed since several years using the validated 3D reactor core calculation program SIMULATE-3. The authors describe a special application of the program to study the reactivity for different partial core loadings. Based on the comparison with results of the program CASMO-4 the program SIMULATE-3 was validated for the calculation of partially loaded reactor cores. For the planned reactor operation in NPP Gundremmingen using new MOX fuel elements the reactivity behavior was studied with respect to the KTA-Code requirements.

  10. Higher order polynomial expansion nodal method for hexagonal core neutronics analysis

    International Nuclear Information System (INIS)

    Jin, Young Cho; Chang, Hyo Kim

    1998-01-01

    A higher-order polynomial expansion nodal(PEN) method is newly formulated as a means to improve the accuracy of the conventional PEN method solutions to multi-group diffusion equations in hexagonal core geometry. The new method is applied to solving various hexagonal core neutronics benchmark problems. The computational accuracy of the higher order PEN method is then compared with that of the conventional PEN method, the analytic function expansion nodal (AFEN) method, and the ANC-H method. It is demonstrated that the higher order PEN method improves the accuracy of the conventional PEN method and that it compares very well with the other nodal methods like the AFEN and ANC-H methods in accuracy

  11. Various reactivity effects value for assuring fast reactor core inherent safety

    International Nuclear Information System (INIS)

    Belov, S.B.; Vasilyev, B.A.

    1991-01-01

    The paper presents the results of temperature and power reactivity feedback components calculations for fast reactors with different core volume when using oxide, carbide, nitride and metal fuel. Reactor parameters change in loss of flow without scram and transient over power without scram accidents was evaluated. The importance of various reactivity feedback components in restricting the consequences of these accidents has been analyzed. (author)

  12. Adjusted neutron spectra of STEK cores for reactivity calculations

    International Nuclear Information System (INIS)

    Dekker, J.W.M.; Dragt, J.B.; Janssen, A.J.; Heijboer, R.J.; Klippel, H.Th.

    1978-02-01

    Neutron flux and adjoint flux spectra form a pre-requisite in the analysis of reactivity worth data measured in the STEK facility. First, a survey of all available information about these spectra is given. Next a special application of a general adjustment method is described. This method has been used to obtain adjusted STEK group flux and adjoint flux spectra, starting from calculated spectra. These theoretical spectra were adjusted to reactivity worths of natural boron (nat. B) and 235 U as well as a number of fission reaction rates. As a by-product in this adjustment calculation adjusted fission group cross sections of 235 U were obtained. The results, viz. group fluxes and adjoint fluxes and adjusted fission cross sections of 235 U are given. They have been used for the interpretation of fission product reactivity worth measurements made in STEK

  13. Evaluation of the oxide and silicide fuels reactivity in the RSG-GAS core

    International Nuclear Information System (INIS)

    S, Tukiran; M S, Tagor; S, Lily; Pinem, S.

    2000-01-01

    Fuel exchange of The RSG-GAS reactor core from uranium oxide to uranium silicide in the same loading, density, and enrichment, that is, 250 gr, 2.98 gr/cm 3 , and 19.75 % respectively, will be performed in-step wise. In every cycle of exchange with 5/l mode, it is needed to evaluate the parameter of reactor core operation. One of the important operation parameters is fuel reactivity that gives effect to the core reactivity. The experiment was performed at core no. 36, BOC, low power which exist 2 silicide fuels. The evaluation was done based on the RSG-GAS control rod calibration consisting of 40 fuels and 8 control rod.s. From 40 fuels in the core, there are 2 silicide fuels, RI-225/A-9 and RI-224/C-3. For inserting 2 silicide fuels, the reactivity effect to the core must be know. To know this effect , it was performed fuels reactivity experiment, which based on control rod calibration. But in this case the RSG-GAS has no other fresh oxide fuel so that configuration of the RSG-GAS core was rearranged by taking out the both silicide fuels and this configuration is used as reference core. Then silicide fuel RI-224 was inserted to position F-3 replacing the fresh oxide fuel RI-260 so the different reactivity of the fuels is obtained. The experiment result showed that the fuel reactivity change is in amount of 12.85 cent (0.098 % ) The experiment result was compared to the calculation result, using IAFUEL code which amount to 13.49 cent (0.103 %) The result showed that the reactivity change of oxide to silicide fuel is small so that the fuel exchange from uranium oxide to uranium silicide in the first step can be done without any significant change of the operation parameter

  14. Measurement and analysis of reactivity worth of 237Np sample in cores of TCA and FCA

    International Nuclear Information System (INIS)

    Sakurai, Takeshi; Mori, Takamasa; Okajima, Shigeaki; Tani, Kazuhiro; Suzaki, Takenori; Saito, Masaki

    2009-01-01

    The reactivity worth of 22.87 grams of 237 Np oxide sample was measured and analyzed in seven uranium cores in the Tank-Type Critical Assembly (TCA) and two uranium cores in the Fast Critical Assembly (FCA) at the Japan Atomic Energy Agency. The TCA cores provided a systematic variation in the neutron spectrum between the thermal and resonance energy regions. The FCA cores, XXI and XXV, provided a hard neutron spectrum of the fast reactor and a soft one of the resonance energy region, respectively. Analyses were carried out using the JENDL-3.3 nuclear data library with a Monte Carlo method for the TCA cores and a deterministic method for the FCA cores. The ratios of calculated to experimental (C/E) reactivity worth were between 0.97 and 0.91, and showed no apparent dependence on the neutron spectrum. (author)

  15. Core concept of fast power reactor with zero sodium void reactivity

    International Nuclear Information System (INIS)

    Matveev, V.I.; Chebeskov, A.N.; Krivitsky, I.Y.

    1991-01-01

    The paper presents a core concept of BN-800 - type fast power reactor with zero sodium void reactivity (SVR). Consideration is given to the layout-and some design features of such a core. Some considerations on the determination of the required SVR value as one of the fast reactor safety criteria in accidents with coolant boiling are presented. Some methodical considerations an the development of calculation models that give a correct description of the new core features are stated. The results of the integral SVR calculation studies are included. reactivity excursions under different scenarios of sodium boiling are estimated, some corrections into the calculated SVR value are discussed. (author)

  16. Fuel density effect on parameter of reactivity coefficient of the Innovative Research Reactor core

    International Nuclear Information System (INIS)

    Rokhmadi; Tukiran S

    2013-01-01

    The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research reactor in Indonesia right now is already 25 year old. Therefor, it is needed to design a new research reactor as a alternative called it innovative research reactor (IRR) and then as an exchanger for old research reactor. The aim of this research is to complete RRI core design data as a requirement for design license. Calculation done is to get the RRI core reactivity coefficients with 5 x 5 core configuration and 20 MW of power, has more than 40 days cycle of length. The RRI core reactivity coefficient calculation is done for new U-"9Mo-Al fuel with variation of densities. The calculation is done by using WIMSD-5B and BATAN-FUEL computer codes. The result of calculation for conceptual design showed that the equilibrium RRI core with 5 x 5 configuration, 450 g, 550 g and 700 g of fuel loadings have negative reactivity coefficients of fuel temperature, moderator temperature, void fraction and density of moderator but the values of the reactivities are very variation. This results has met the safety criteria for RRI core conceptual design. (author)

  17. Analysis Of Temperature Effects On Reactivity Of The Rsg-Gas Core Using Silicide Fuels

    International Nuclear Information System (INIS)

    Surbakti, Tukiran; Pinem, Surian

    2001-01-01

    RSG-GAS has been operating using new silicide fuels so that it is necessary to estimate and to measure the effect of temperature on reactivity of the core. The parameters to be determined due to temperature effect are reactivity coefficient of moderator temperature, temperature coefficient of fuel element and power reactivity coefficient. By doing a couple compensation method, determination of reactivity coefficient as well as the reactivity coefficient of moderator temperature can be obtained. Furthermore, coefficient of the reactivity was successfully estimated using the combination of WIMS-D4 and Batan-2DIFF. The cell calculation was done by using WIMS-D4 code to get macroscopic cross section and Batan-2DIFF code is used for core calculation. The calculation and experimental results of reactivity coefficient do not show any deviation from RSG-GAS safety margin. The results are -2,84 sen/ o C, -1,29 sen/MW and -0,64 sen/ o C for reactivity coefficients of temperature, power, fuel element and moderator temperature, respectively. All of 3 parameters are absolutely met with safety criteria

  18. Impact of correlations between core configurations for the evaluation of nuclear data uncertainty propagation for reactivity

    International Nuclear Information System (INIS)

    Frosio, T.; Bonaccorsi, T.; Blaise, P.

    2017-01-01

    The precise estimation of Pearson correlations, also called 'representativity' coefficients, between core configurations is a fundamental quantity for properly assessing the nuclear data (ND) uncertainties propagation on integral parameters such as k-eff, power distributions, or reactivity coefficients. In this paper, a traditional adjoint method is used to propagate ND uncertainty on reactivity and reactivity coefficients and estimate correlations between different states of the core. We show that neglecting those correlations induces a loss of information in the final uncertainty. We also show that using approximate values of Pearson does not lead to an important error of the model. This calculation is made for reactivity at the beginning of life and can be extended to other parameters during depletion calculations. (authors)

  19. Reactivity changes in hybrid thermal-fast reactor systems during fast core flooding

    International Nuclear Information System (INIS)

    Pesic, M.

    1994-09-01

    A new space-dependent kinetic model in adiabatic approximation with local feedback reactivity parameters for reactivity determination in the coupled systems is proposed in this thesis. It is applied in the accident calculation of the 'HERBE' fast-thermal reactor system and compared to usual point kinetics model with core-averaged parameters. Advantages of the new model - more realistic picture of the reactor kinetics and dynamics during local large reactivity perturbation, under the same heat transfer conditions, are underlined. Calculated reactivity parameters of the new model are verified in the experiments performed at the 'HERBE' coupled core. The model has shown that the 'HERBE' safety system can shutdown reactor safely and fast even in the case of highly set power trip and even under conditions of big partial failure of the reactor safety system (author)

  20. analysis of reactivity accidents in MTR for various protection system parameters and core condition

    International Nuclear Information System (INIS)

    Mohamed, F.M.

    2011-01-01

    Egypt Second Research Reactor (ETRR-2) core was modified to irradiate LEU (Low Enriched Uranium) plates in two irradiation boxes for fission 99 Mo production. The old core comprising 29 fuel elements and one Co Irradiation Device (CID) and the new core comprising 27 fuel elements, CID, and two 99 Mo production boxes. The in core irradiation has the advantage of no special cooling or irradiation loop is required. The purpose of the present work is the analysis of reactivity accidents (RIA) for ETRR-2 cores. The analysis was done to evaluate the accidents from different point of view:1- Analysis of the new core for various Reactor Protection System (RPS) parameters 2- Comparison between the two cores. 3- Analysis of the 99 Mo production boxes.PARET computer code was employed to compute various parameters. Initiating events in RIA involve various modes of reactivity insertion, namely, prompt critical condition (p=1$), accidental ejection of partial and complete CID uncontrolled withdrawal of a control rod accident, and sudden cooling of the reactor core. The time histories of reactor power, energy released, and the maximum fuel, clad and coolant temperatures of fuel elements and LEU plates were calculated for each of these accidents. The results show that the maximum clad temperatures remain well below the clad melting of both fuel and uranium plates during these accidents. It is concluded that for the new core, the RIA with scram will not result in fuel or uranium plate failure.

  1. Mineralogy, geochemistry and expansion testing of an alkali-reactive basalt from western Anatolia, Turkey

    International Nuclear Information System (INIS)

    Copuroglu, Oguzhan; Andic-Cakir, Ozge; Broekmans, Maarten A.T.M.; Kuehnel, Radko

    2009-01-01

    In this paper, the alkali-silica reaction performance of a basalt rock from western Anatolia, Turkey is reported. It is observed that the rock causes severe gel formation in the concrete microbar test. It appears that the main source of expansion is the reactive glassy phase of the basalt matrix having approximately 70% of SiO 2 . The study presents the microstructural characteristics of unreacted and reacted basalt aggregate by optical and electron microscopy and discusses the possible reaction mechanism. Microstructural analysis revealed that the dissolution of silica is overwhelming in the matrix of the basalt and it eventually generates four consequences: (1) Formation of alkali-silica reaction gel at the aggregate perimeter, (2) increased porosity and permeability of the basalt matrix, (3) reduction of mechanical properties of the aggregate and (4) additional gel formation within the aggregate. It is concluded that the basalt rock is highly prone to alkali-silica reaction. As an aggregate, this rock is not suitable for concrete production.

  2. Mineralogy, geochemistry and expansion testing of an alkali-reactive basalt from western Anatolia, Turkey

    Energy Technology Data Exchange (ETDEWEB)

    Copuroglu, Oguzhan, E-mail: O.Copuroglu@CiTG.TUDelft.NL [Delft University of Technology, Faculty of CiTG, Materials and Environment, Stevinweg 1, 2628CN, Delft (Netherlands); Andic-Cakir, Ozge [Ege University, Civil Engineering Dept., 35100 Bornova, Izmir (Turkey); Broekmans, Maarten A.T.M. [Geological Survey of Norway, Dept. of Mineral Characterization, N-7491 Trondheim (Norway); Kuehnel, Radko [Burgemeester Merkusstraat 5, 2645 NJ, Delfgauw (Netherlands)

    2009-07-15

    In this paper, the alkali-silica reaction performance of a basalt rock from western Anatolia, Turkey is reported. It is observed that the rock causes severe gel formation in the concrete microbar test. It appears that the main source of expansion is the reactive glassy phase of the basalt matrix having approximately 70% of SiO{sub 2}. The study presents the microstructural characteristics of unreacted and reacted basalt aggregate by optical and electron microscopy and discusses the possible reaction mechanism. Microstructural analysis revealed that the dissolution of silica is overwhelming in the matrix of the basalt and it eventually generates four consequences: (1) Formation of alkali-silica reaction gel at the aggregate perimeter, (2) increased porosity and permeability of the basalt matrix, (3) reduction of mechanical properties of the aggregate and (4) additional gel formation within the aggregate. It is concluded that the basalt rock is highly prone to alkali-silica reaction. As an aggregate, this rock is not suitable for concrete production.

  3. HELIOS/DRAGON/NESTLE codes' simulation of void reactivity in a CANDU core

    International Nuclear Information System (INIS)

    Sarsour, H.N.; Rahnema, F.; Mosher, S.; Turinsky, P.J.; Serghiuta, D.; Marleau, G.; Courau, T.

    2002-01-01

    This paper presents results of simulation of void reactivity in a CANDU core using the NESTLE core simulator, cross sections from the HELIOS lattice physics code in conjunction with incremental cross sections from the DRAGON lattice physics code. First, a sub-region of a CANDU6 core is modeled using the NESTLE core simulator and predictions are contrasted with predictions by the MCNP Monte Carlo simulation code utilizing a continuous energy model. In addition, whole core modeling results are presented using the NESTLE finite difference method (FDM), NESTLE nodal method (NM) without assembly discontinuity factors (ADF), and NESTLE NM with ADF. The work presented in this paper has been performed as part of a project sponsored by the Canadian Nuclear Safety Commission (CNSC). The purpose of the project was to gather information and assess the accuracy of best estimate methods using calculational methods and codes developed independently from the CANDU industry. (author)

  4. Effects of moderation level on core reactivity and. neutron fluxes in natural uranium fueled and heavy water moderated reactors

    International Nuclear Information System (INIS)

    Khan, M.J.; Aslam; Ahmad, N.; Ahmed, R.; Ahmad, S.I.

    2005-01-01

    The neutron moderation level in a nuclear reactor has a strong influence on core multiplication, reactivity control, fuel burnup, neutron fluxes etc. In the study presented in this article, the effects of neutron moderation level on core reactivity and neutron fluxes in a typical heavy water moderated nuclear research reactor is explored and the results are discussed. (author)

  5. Analysis of reactivity accidents of the RSG-GAS core with silicide fuel

    International Nuclear Information System (INIS)

    Tukiran

    2002-01-01

    The fuels of RSG-GAS reactor is changed from uranium oxide to uranium silicide. For time being, the fuel of RSG-GAS core are mixed up between oxide and silicide fuels with 250 gr of loading and 2.96 g U/cm 3 of density, respectively. While, silicide fuel with 300 gr of loading is still under research. The advantages of silicide fuels are can be used in high density, so that, it can be stayed longer in the core at higher burn-up, therefore, the length of cycle is longer. The silicide fuel in RSG-GAS core is used in step-wise by using mixed up core. Firstly, it is used silicide fuel with 250 gr of loading and then, silicide fuel with 300 gr of loading (3.55 g U/cm 3 of density). In every step-wise of fuel loading must be analysed its safety margin. In this occasion, it is analysed the reactivity accident of RSG-GAS core with 300 gr of silicide fuel loading. The calculation was done by using POKDYN code which available at P2TRR. The calculation was done by reactivity insertion at start up and power rangers. From all cases which were have been done, the results of analysis showed that there is no anomaly and safety margin break at RSG-GAS core with 300 gr silicide fuel loading

  6. Non-linear triangle-based polynomial expansion nodal method for hexagonal core analysis

    International Nuclear Information System (INIS)

    Cho, Jin Young; Cho, Byung Oh; Joo, Han Gyu; Zee, Sung Qunn; Park, Sang Yong

    2000-09-01

    This report is for the implementation of triangle-based polynomial expansion nodal (TPEN) method to MASTER code in conjunction with the coarse mesh finite difference(CMFD) framework for hexagonal core design and analysis. The TPEN method is a variation of the higher order polynomial expansion nodal (HOPEN) method that solves the multi-group neutron diffusion equation in the hexagonal-z geometry. In contrast with the HOPEN method, only two-dimensional intranodal expansion is considered in the TPEN method for a triangular domain. The axial dependence of the intranodal flux is incorporated separately here and it is determined by the nodal expansion method (NEM) for a hexagonal node. For the consistency of node geometry of the MASTER code which is based on hexagon, TPEN solver is coded to solve one hexagonal node which is composed of 6 triangular nodes directly with Gauss elimination scheme. To solve the CMFD linear system efficiently, stabilized bi-conjugate gradient(BiCG) algorithm and Wielandt eigenvalue shift method are adopted. And for the construction of the efficient preconditioner of BiCG algorithm, the incomplete LU(ILU) factorization scheme which has been widely used in two-dimensional problems is used. To apply the ILU factorization scheme to three-dimensional problem, a symmetric Gauss-Seidel Factorization scheme is used. In order to examine the accuracy of the TPEN solution, several eigenvalue benchmark problems and two transient problems, i.e., a realistic VVER1000 and VVER440 rod ejection benchmark problems, were solved and compared with respective references. The results of eigenvalue benchmark problems indicate that non-linear TPEN method is very accurate showing less than 15 pcm of eigenvalue errors and 1% of maximum power errors, and fast enough to solve the three-dimensional VVER-440 problem within 5 seconds on 733MHz PENTIUM-III. In the case of the transient problems, the non-linear TPEN method also shows good results within a few minute of

  7. Analysis of excess reactivity of JOYO MK-III performance test core

    International Nuclear Information System (INIS)

    Maeda, Shigetaka; Yokoyama, Kenji

    2003-10-01

    JOYO is currently being upgraded to the high performance irradiation bed JOYO MK-III core'. The MK-III core is divided into two fuel regions with different plutonium contents. To obtain a higher neutron flux, the active core height was reduced from 55 cm to 50 cm. The reflector subassemblies were replaced by shielding subassemblies in the outer two rows. Twenty of the MK-III outer core fuel subassemblies in the performance test core were partially burned in the transition core. Four irradiation test rigs, which do not contain any fuel material, were loaded in the center of the performance test core. In order to evaluate the excess reactivity of MK-III performance test core accurately, we evaluated it by applying not only the JOYO MK-II core management code system MAGI, but also the MK-III core management code system HESTIA, the JUPITER standard analysis method and the Monte Carlo method with JFS-3-J3.2R content set. The excess reactivity evaluations obtained by the JUPITER standard analysis method were corrected to results based on transport theory with zero mesh-size in space and angle. A bias factor based on the MK-II 35th core, which sensitivity was similar to MK-III performance test core's, was also applied, except in the case where an adjusted nuclear cross-section library was used. Exact three-dimensional, pin-by-pin geometry and continuous-energy cross sections were used in the Monte Carlo calculation. The estimated error components associated with cross-sections, methods correction factors and the bias factor were combined based on Takeda's theory. Those independently calculated values agree well and range from 2.8 to 3.4%Δk/kk'. The calculation result of the MK-III core management code system HESTLA was 3.13% Δk/kk'. The estimated errors for bias method range from 0.1 to 0.2%Δk/kk'. The error in the case using adjusted cross-section was 0.3%Δk/kk'. (author)

  8. Synthesis and Characterization of Core-Shell Acrylate Based Latex and Study of Its Reactive Blends

    Directory of Open Access Journals (Sweden)

    Ying Nie

    2008-03-01

    Full Text Available Techniques in resin blending are simple and efficient method for improving the properties of polymers, and have been used widely in polymer modification field. However, polymer latex blends such as the combination of latexes, especially the latexes with water-soluble polymers, were rarely reported. Here, we report a core-shell composite latex synthesized using methyl methacrylate (MMA, butyl acrylate (BA, 2-ethylhexyl acrylate (EHA and glycidyl methacrylate (GMA as monomers and ammonium persulfate and sodium bisulfite redox system as the initiator. Two stages seeded semi-continuous emulsion polymerization were employed for constructing a core-shell structure with P(MMA-co-BA component as the core and P(EHA-co-GMA component as the shell. Results of Transmission Electron Microscopy (TEM and Dynamics Light Scattering (DLS tests confirmed that the particles obtained are indeed possessing a desired core-shell structural character. Stable reactive latex blends were prepared by adding the latex with waterborne melamine-formaldehyde resin (MF or urea-formaldehyde resin (UF. It was found that the glass transition temperature, the mechanical strength and the hygroscopic property of films cast from the latex blends present marked enhancements under higher thermal treatment temperature. It was revealed that the physical properties of chemically reactive latexes with core-shell structure could be altered via the change of crosslinking density both from the addition of crosslinkers and the thermal treatment.

  9. Core dynamics analysis for reactivity insertion and loss of coolant flow tests using the HTTR

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

    2007-01-01

    The High Temperature engineering Test Reactor (HTTR) is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950degC (SAITO, 1994). Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs) (TACHIBANA 2002) (NAKAGAWA 2004). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named ACCORD (TAKAMATSU 2006), was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We used a conventional method, namely, a one-dimensional flow channel model and reactor kinetics model with a single temperature coefficient, taking into account the temperature changes in the core. However, a slight difference between the analytical and experimental results was observed. Therefore, we have modified this code to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the loss of coolant flow tests by tripping one or two out of three gas circulators. Finally, the pre-analytical result of

  10. Reactivity And Neutron Flux At Silicide Fuel Element In The Core Of RSG-GAS

    International Nuclear Information System (INIS)

    Hamzah, Amir

    2000-01-01

    In order to 4.8 and 5.2 gr U/cm exp 3 loading of U 3 Si 2 --Al fuel plates characterization, he core reactivity change and neutron flux depression had been done. Control rod calibration method was used to reactivity change measurement and neutron flux distribution was measured using foil activation method. Measurement of insertion of A-type of testing fuel element with U-loading above cannot be done due to technical reason, so the measurement using full type silicide fuel element of 2.96 gr U/cm exp 3 loading. The reactivity change measurement result of insertion in A-9 and C-3 is + 2.67 cent. The flux depression at silicide fuel in A-9 is 1.69 times bigger than oxide and in C-3 is 0.68 times lower than oxide

  11. Alteration of alkali reactive aggregates autoclaved in different alkali solutions and application to alkali-aggregate reaction in concrete (II) expansion and microstructure of concrete microbar

    International Nuclear Information System (INIS)

    Lu Duyou; Mei Laibao; Xu Zhongzi; Tang Mingshu; Mo Xiangyin; Fournier, Benoit

    2006-01-01

    The effect of the type of alkalis on the expansion behavior of concrete microbars containing typical aggregate with alkali-silica reactivity and alkali-carbonate reactivity was studied. The results verified that: (1) at the same molar concentration, sodium has the strongest contribution to expansion due to both ASR and ACR, followed by potassium and lithium; (2) sufficient LiOH can completely suppress expansion due to ASR whereas it can induce expansion due to ACR. It is possible to use the duplex effect of LiOH on ASR and ACR to clarify the ACR contribution when ASR and ACR may coexist. It has been shown that a small amount of dolomite in the fine-grained siliceous Spratt limestone, which has always been used as a reference aggregate for high alkali-silica reactivity, might dedolomitize in alkaline environment and contribute to the expansion. That is to say, Spratt limestone may exhibit both alkali-silica and alkali-carbonate reactivity, although alkali-silica reactivity is predominant. Microstructural study suggested that the mechanism in which lithium controls ASR expansion is mainly due to the favorable formation of lithium-containing less-expansive product around aggregate particles and the protection of the reactive aggregate from further attack by alkalis by the lithium-containing product layer

  12. Effects of Radial Reflector Composition on Core Reactivity and Peak Power

    International Nuclear Information System (INIS)

    Park, Sang Yoon; Lee, Kyung Hoon; Song, Jae Seung

    2007-10-01

    The effects of radial SA-240 alloy shroud on core reactivity and peak power are evaluated. The existence of radial SA-240 alloy shroud makes reflector water volume decrease, so the thermal absorption cross section of radial reflector is lower than without SA-240 alloy shroud case. Finally, the cycle length is increased from 788 EFPD to 845 EFPD and the peak power is decreased from 1.66 to 1.49. In the case of without SA-240 alloy shroud, a new core loading pattern search has been performed. For the guarantee of the same equivalent cycle length of with SA-240 alloy shroud case, the enrichment of U-235 should be increased from 4.22 w/o to 4.68 w/o. The nuclear key safety parameters of new core loading pattern have been calculated and recorded for the future

  13. Effects of Radial Reflector Composition on Core Reactivity and Peak Power

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Yoon; Lee, Kyung Hoon; Song, Jae Seung

    2007-10-15

    The effects of radial SA-240 alloy shroud on core reactivity and peak power are evaluated. The existence of radial SA-240 alloy shroud makes reflector water volume decrease, so the thermal absorption cross section of radial reflector is lower than without SA-240 alloy shroud case. Finally, the cycle length is increased from 788 EFPD to 845 EFPD and the peak power is decreased from 1.66 to 1.49. In the case of without SA-240 alloy shroud, a new core loading pattern search has been performed. For the guarantee of the same equivalent cycle length of with SA-240 alloy shroud case, the enrichment of U-235 should be increased from 4.22 w/o to 4.68 w/o. The nuclear key safety parameters of new core loading pattern have been calculated and recorded for the future.

  14. Interactions of solitary waves and compression/expansion waves in core-annular flows

    Science.gov (United States)

    Maiden, Michelle; Anderson, Dalton; El, Gennady; Franco, Nevil; Hoefer, Mark

    2017-11-01

    The nonlinear hydrodynamics of an initial step leads to the formation of rarefaction waves and dispersive shock waves in dispersive media. Another hallmark of these media is the soliton, a localized traveling wave whose speed is amplitude dependent. Although compression/expansion waves and solitons have been well-studied individually, there has been no mathematical description of their interaction. In this talk, the interaction of solitons and shock/rarefaction waves for interfacial waves in viscous, miscible core-annular flows are modeled mathematically and explored experimentally. If the interior fluid is continuously injected, a deformable conduit forms whose interfacial dynamics are well-described by a scalar, dispersive nonlinear partial differential equation. The main focus is on interactions of solitons with dispersive shock waves and rarefaction waves. Theory predicts that a soliton can either be transmitted through or trapped by the extended hydrodynamic state. The notion of reciprocity is introduced whereby a soliton interacts with a shock wave in a reciprocal or dual fashion as with the rarefaction. Soliton reciprocity, trapping, and transmission are observed experimentally and are found to agree with the modulation theory and numerical simulations. This work was partially supported by NSF CAREER DMS-1255422 (M.A.H.) and NSF GRFP (M.D.M.).

  15. Study on Characteristic of Temperature Coefficient of Reactivity for Plutonium Core of Pebbled Bed Reactor

    Science.gov (United States)

    Zuhair; Suwoto; Setiadipura, T.; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    As a part of the solution searching for possibility to control the plutonium, a current effort is focused on mechanisms to maximize consumption of plutonium. Plutonium core solution is a unique case in the high temperature reactor which is intended to reduce the accumulation of plutonium. However, the safety performance of the plutonium core which tends to produce a positive temperature coefficient of reactivity should be examined. The pebble bed inherent safety features which are characterized by a negative temperature coefficient of reactivity must be maintained under any circumstances. The purpose of this study is to investigate the characteristic of temperature coefficient of reactivity for plutonium core of pebble bed reactor. A series of calculations with plutonium loading varied from 0.5 g to 1.5 g per fuel pebble were performed by the MCNPX code and ENDF/B-VII library. The calculation results show that the k eff curve of 0.5 g Pu/pebble declines sharply with the increase in fuel burnup while the greater Pu loading per pebble yields k eff curve declines slighter. The fuel with high Pu content per pebble may reach long burnup cycle. From the temperature coefficient point of view, it is concluded that the reactor containing 0.5 g-1.25 g Pu/pebble at high burnup has less favorable safety features if it is operated at high temperature. The use of fuel with Pu content of 1.5 g/pebble at high burnup should be considered carefully from core safety aspect because it could affect transient behavior into a fatal accident situation.

  16. Uncertainty Evaluation of Reactivity Coefficients for a large advanced SFR Core Design

    International Nuclear Information System (INIS)

    Khamakhem, Wassim; Rimpault, Gerald

    2008-01-01

    Sodium Cooled Fast Reactors are currently being reshaped in order to meet Generation IV goals on economics, safety and reliability, sustainability and proliferation resistance. Recent studies have led to large SFR cores for a 3600 MWth power plants, cores which exhibit interesting features. The designs have had to balance between competing aspects such as sustainability and safety characteristics. Sustainability in neutronic terms is translated into positive breeding gain and safety into rather low Na void reactivity effects. The studies have been done on two SFR concepts using oxide and carbide fuels. The use of the sensitivity theory in the ERANOS determinist code system has been used. Calculations have been performed with different sodium evaluations: JEF2.2, ERALIB-1 and the most recent JEFF3.1 and ENDF/B-VII in order to make a broad comparison. Values for the Na void reactivity effect exhibit differences as large as 14% when using the different sodium libraries. Uncertainties due to nuclear data on the reactivity coefficients were performed with BOLNA variances-covariances data, the Na Void Effect uncertainties are near to 12% at 1σ. Since, the uncertainties are far beyond the target accuracy for a design achieving high performance, two directions are envisaged: the first one is to perform new differential measurements or in a second attempt use integral experiments to improve effectively the nuclear data set and its uncertainties such as performed in the past with ERALIB1. (authors)

  17. Reactivity Accidents in CAREM-25 Core with and Without Safety Systems Actuation

    International Nuclear Information System (INIS)

    Gimenez, Marcelo; Vertullo, Alicia; Schlamp, Miguel

    2000-01-01

    A reactivity accident in CAREM core can be provoked by different initiating events, a cold water injection in pressure vessel, a secondary side steam line breakage and a failure in the absorbing rods drive system.The present work analyses inadverted control rod withdraws transients.Maximum worth control rod (2.5 $) at normal velocity (1 cm/s) is adopted for the simulations (Reactivity ramp of 0.018 $/s).Different scenarios considering actuation of first shutdown system (FSS), second shutdown system (SSS) and selflimiting conditions were modeled.Results of the accident with actuation of FSS show that safety margins are well above critical values (DNBR and CPR).In the cases with failure of the FSS and success of SSS or selflimited, safety margins are below critical values, however, the SSS provides a reduction of elapsed time under advised margins

  18. Verification for excess reactivity on beginning equilibrium core of RSG GAS

    International Nuclear Information System (INIS)

    Daddy Setyawan; Budi Rohman

    2011-01-01

    BAPETEN is an institution authorized to control the use of nuclear energy in Indonesia. Control for the use of nuclear energy is carried out through three pillars: regulation, licensing, and inspection. In order to assure the safety of the operating research reactors, the assessment unit of BAPETEN is carrying out independent assessment in order to verify safety related parameters in the SAR including neutronic aspect. The work includes verification to the Power Peaking Factor in the equilibrium silicide core of RSG GAS reactor by computational method using MCNP-ORIGEN. This verification calculation results for is 9.4 %. Meanwhile, the RSG-GAS safety analysis report shows that the excess reactivity on equilibrium core of RSG GAS is 9.7 %. The verification calculation results show a good agreement with the report. (author)

  19. Reactivity considerations for the on-line refuelling of a pebble bed modular reactor—Illustrating safety for the most reactive core fuel load

    International Nuclear Information System (INIS)

    Reitsma, Frederik

    2012-01-01

    In the multi-pass fuel management scheme employed for the pebble bed modular reactor the fuel pebbles are re-circulated until they reach the target burn-up. The rate at which fresh fuel is loaded and burned fuel is discharged is a result of the core neutronics cycle analysis but in practice (on the plant) this has to be controlled and managed by the fuel handling and storage system and use of the burnup measurement system. The excess reactivity is the additional reactivity available in the core during operating conditions that is the result of loading a fuel mixture in the core that is more reactive (less burned) than what is required to keep the reactor critical at full power operational conditions. The excess reactivity is balanced by the insertion of the control rods to keep the reactor critical. The excess reactivity allows flexibility in operations, for example to overcome the xenon build up when power is decreased as part of load follow. In order to limit reactivity excursions and to ensure safe shutdown the excess reactivity and thus the insertion depth of the control rods at normal operating conditions has to be managed. One way to do this is by operational procedures. The reactivity effect of long-term operation with the control rods inserted deeper than the design point is investigated and a control rod insertion limit is proposed that will not limit normal operations. The effects of other phenomena that can increase the power defect, such as higher-than-expected fuel temperatures, are also introduced. All of these cases are then evaluated by ensuring cold shutdown is still achievable and where appropriate by reactivity insertion accident analysis. These aspects are investigated on the PBMR 400 MW design.

  20. The effects of stainless steel radial reflector on core reactivity for small modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Jung Kil, E-mail: jkkang@email.kings.ac.kr; Hah, Chang Joo, E-mail: changhah@kings.ac.kr [KINGS, 658-91, Haemaji-ro, Seosaeng-myeon, Ulju-gun, Ulsan, 689-882 (Korea, Republic of); Cho, Sung Ju, E-mail: sungju@knfc.co.kr; Seong, Ki Bong, E-mail: kbseong@knfc.co.kr [KNFC, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2016-01-22

    Commercial PWR core is surrounded by a radial reflector, which consists of a baffle and water. Radial reflector is designed to reflect neutron back into the core region to improve the neutron efficiency of the reactor and to protect the reactor vessels from the embrittling effects caused by irradiation during power operation. Reflector also helps to flatten the neutron flux and power distributions in the reactor core. The conceptual nuclear design for boron-free small modular reactor (SMR) under development in Korea requires to have the cycle length of 4∼5 years, rated power of 180 MWth and enrichment less than 5 w/o. The aim of this paper is to analyze the effects of stainless steel radial reflector on the performance of the SMR using UO{sub 2} fuels. Three types of reflectors such as water, water/stainless steel 304 mixture and stainless steel 304 are selected to investigate the effect on core reactivity. Additionally, the thickness of stainless steel and double layer reflector type are also investigated. CASMO-4/SIMULATE-3 code system is used for this analysis. The results of analysis show that single layer stainless steel reflector is the most efficient reflector.

  1. Laboratory measurements of the coefficient of thermal expansion of Olkiluoto drill core samples

    International Nuclear Information System (INIS)

    Aakesson, U.

    2012-04-01

    The coefficient of thermal expansion and the wet density has been determined on 22 specimens from the ONKALO drillholes ONK-PP167, ONK-PP199, ONK-PP224, ONK-PP225 and ONK-PP226, Olkiluoto, Finland. The coefficient of thermal expansion has been determined in the temperature interval 20-60 deg C. The results indicated that the thermal expansion was almost linear, and the coefficient of thermal expansion for the investigated specimens range between 3.2 and 14.4 x 10 -6 mm/mm deg C, and the wet density between 2,610 and 2,820 kg/m 3 . The granite pegmatite has slightly lower coefficient of thermal expansion and wet density than gneissic rocks. (orig.)

  2. Calculation of Reactivity Build up in KANUPP core in Case of Large Break LOCA

    International Nuclear Information System (INIS)

    Arshad, M. W.

    2012-01-01

    Loss of Coolant Accident (LOCA) in a Pressurized Heavy Water Reactor (PHWR) leads to coolant expulsion in a primary heat transport system resulting in depressurization and possible core voiding. This results in deterioration of cooling conditions in reactor channels and increase in power before reactor shutdown, leading to higher fuel temperatures.The objective of this thesis is to couple Thermal Hydraulics Data for finding status of 2288 fuel bundles having unique coolant density along with continuous changing state of coolant. WIMCER and CITCER are used for the core calculation in case of LOCA and Thermal Hydraulic Data is obtained from the Thermal Hydraulic code TUF (two unequal flows). These codes are coupled with each other in C programming. Due to degradation of coolant in case of LOCA, the power and reactivity start increasing. Near to 5 mk of reactivity the moderator dump start and reactor goes shut down. The result obtained from these code is followed the same trend as shown in KFSAR. (author)

  3. Reactivity management and burn-up management on JRR-3 silicide-fuel-core

    International Nuclear Information System (INIS)

    Kato, Tomoaki; Araki, Masaaki; Izumo, Hironobu; Kinase, Masami; Torii, Yoshiya; Murayama, Yoji

    2007-08-01

    On the conversion from uranium-aluminum-dispersion-type fuel (aluminide fuel) to uranium-silicon-aluminum-dispersion-type fuel (silicide fuel), uranium density was increased from 2.2 to 4.8 g/cm 3 with keeping uranium-235 enrichment of 20%. So, burnable absorbers (cadmium wire) were introduced for decreasing excess reactivity caused by the increasing of uranium density. The burnable absorbers influence reactivity during reactor operation. So, the burning of the burnable absorbers was studied and the influence on reactor operation was made cleared. Furthermore, necessary excess reactivity on beginning of operation cycle and the time limit for restart after unplanned reactor shutdown was calculated. On the conversion, limit of fuel burn-up was increased from 50% to 60%. And the fuel exchange procedure was changed from the six-batch dispersion procedure to the fuel burn-up management procedure. The previous estimation of fuel burn-up was required for the planning of fuel exchange, so that the estimation was carried out by means of past operation data. Finally, a new fuel exchange procedure was proposed for effective use of fuel elements. On the procedure, burn-up of spent fuel was defined for each loading position. The average length of fuel's staying in the core can be increased by two percent on the procedure. (author)

  4. Core design of long life-cycle fast reactors operating without reactivity margin

    International Nuclear Information System (INIS)

    Aristova, E. N.; Baydin, D. F.; Gol'din, V. Y.; Pestryakova, G. A.; Stoynov, M. I.

    2012-01-01

    In this paper we consider a possibility of designing a fast reactor core that operates without reactivity margin for a long time. This study is based on the physical principle of fast reactor operating in a self-adjustable neutron-nuclear regime (SANNR-1) introduced by L.P. Feoktistov (1988-1993) and improved by V. Ya. Gol'din SANNR-2 (1995). The mathematical modeling of active zones of fast reactors in SANNR modes is held by authors since 1992. The numerical simulation is based on solving the neutron transport equation coupled with quasi-diffusion equations. The calculations have been performed using standard 26 energy groups. We use a hierarchy of spatial models of 1D, 1.5D, 2D, and 3D geometries. The spatial models of higher dimensionality are used for verification of results. The calculations showed that operation of the reactor in this mode increases its efficiency, safety and simplifies management. It is possible to achieve continuous work of the reactor in SANNR-2 during 7-10 years without fuel overloads by means of further optimization of the mode. Small reactivity margin is used only for the reactor start up. After first 10-15 days the reactor in SANNR-2 operates without reactivity margin. (authors)

  5. Removal of Reactive Anionic Dyes from Binary Solutions by Adsorption onto Quaternized Kenaf Core Fiber

    Directory of Open Access Journals (Sweden)

    Intidhar Jabir Idan

    2017-01-01

    Full Text Available The most challenging mission in wastewater treatment plants is the removal of anionic dyes, because they are water-soluble and produce very shining colours in the water. In this regard, kenaf core fiber (KCF was chemically modified by the quaternized agent (3-chloro-2-hydroxypropyltrimethylammonium chloride to increase surface area and change the surface properties in order to improve the removing reactive anionic dyes from binary aqueous solution. The influencing operating factors like dye concentration, pH, adsorbent dosage, and contact time were examined in a batch mode. The results indicate that the percentage of removal of Reactive Red-RB (RR-RB and Reactive Black-5 (RB-5 dyes from binary solution was increased with increasing dyes concentrations and the maximum percentage of removal reached up to 98.4% and 99.9% for RR-RB and RB-5, respectively. Studies on effect of pH showed that the adsorption was not significantly influenced by pH. The equilibrium analyses explain that, in spite of the extended Langmuir model failure to describe the data in the binary system, it is better than the Jain and Snoeyink model in describing the adsorption behavior of binary dyes onto QKCF. Also, the pseudo-second-order model was better to represent the adsorption kinetics for RR-RB and RB-5 dyes on QKCF.

  6. Track 5: safety in engineering, construction, operations, and maintenance. Reactor physics design, validation, and operating experience. 5. A Negative Reactivity Feedback Device for Actinide Burner Cores

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Hejzlar, P.

    2001-01-01

    Lead-bismuth eutectic (LBE) cooled reactors are of considerable interest because they may be useful for destruction of actinides in a cost-effective manner, particularly cores fueled predominantly with minor actinides, which gain reactivity with burnup. However, they also pose several design challenges: 1. a small (and perhaps even slightly positive) Doppler feedback; 2. small effective delayed neutron yield; 3. a small negative feedback from axial fuel expansion; 4. positive coolant void and temperature coefficients for conventional designs. This has motivated a search for palliative measures, leading to conceptualization of the reactivity feedback device (RFD). The RFD consists of an in-core flask containing helium gas, tungsten wool, and a small reservoir of LBE that communicates with vertical tubes housing neutron absorber floats. The upper part of these guide tubes contains helium gas that is vented into a separate, cooler ex-core helium gas plenum. The principle of operation is as follows: 1. The tungsten wool, hence the helium gas in the in-core plenum, is heated by gammas and loses heat to the walls by convection and conduction (radiation is feeble for monatomic gases and, in any event, intercepted by the tungsten wool). An energy balance determines the gas temperature, hence, pressure, which is 10 atm here. The energy loss rate can be adjusted by using xenon or a gas mixture in place of helium. The tungsten wool mass, which is 1 vol% wool here, can also be increased to increase gamma heating and further retard convection; alternatively, a Dewar flask could be used in place of the additional wool. 2. An increase in core power causes a virtually instantaneous increase in gamma flux, hence, gas heatup: The thermal time constant of the tungsten filaments and their surrounding gas film is ∼40 μs. 3. The increased gas temperature is associated with an increased gas pressure, which forces more liquid metal into the float guide tubes: LBE will rise ∼100 cm

  7. Reactive Desorption of CO Hydrogenation Products under Cold Pre-stellar Core Conditions

    Science.gov (United States)

    Chuang, K.-J.; Fedoseev, G.; Qasim, D.; Ioppolo, S.; van Dishoeck, E. F.; Linnartz, H.

    2018-02-01

    The astronomical gas-phase detection of simple species and small organic molecules in cold pre-stellar cores, with abundances as high as ∼10‑8–10‑9 n H, contradicts the generally accepted idea that at 10 K, such species should be fully frozen out on grain surfaces. A physical or chemical mechanism that results in a net transfer from solid-state species into the gas phase offers a possible explanation. Reactive desorption, i.e., desorption following the exothermic formation of a species, is one of the options that has been proposed. In astronomical models, the fraction of molecules desorbed through this process is handled as a free parameter, as experimental studies quantifying the impact of exothermicity on desorption efficiencies are largely lacking. In this work, we present a detailed laboratory study with the goal of deriving an upper limit for the reactive desorption efficiency of species involved in the CO–H2CO–CH3OH solid-state hydrogenation reaction chain. The limit for the overall reactive desorption fraction is derived by precisely investigating the solid-state elemental carbon budget, using reflection absorption infrared spectroscopy and the calibrated solid-state band-strength values for CO, H2CO and CH3OH. We find that for temperatures in the range of 10 to 14 K, an upper limit of 0.24 ± 0.02 for the overall elemental carbon loss upon CO conversion into CH3OH. This corresponds with an effective reaction desorption fraction of ≤0.07 per hydrogenation step, or ≤0.02 per H-atom induced reaction, assuming that H-atom addition and abstraction reactions equally contribute to the overall reactive desorption fraction along the hydrogenation sequence. The astronomical relevance of this finding is discussed.

  8. Removal of Reactive Orange 16 Dye from Aqueous Solution by Using Modified Kenaf Core Fiber

    Directory of Open Access Journals (Sweden)

    Maytham Kadhim Obaid

    2016-01-01

    Full Text Available Evaluated removal of reactive orange 16 (RO16 dye from aqueous solution was studied in batch mode by using kenaf core fiber as low-cost adsorbents. In this attempt, kenaf core fiber with size 0.25–1 mm was treated by using (3-chloro-2-hydroxypropyl trimethylammonium chloride (CHMAC as quaternization agent. Then effective parameters include adsorbent dose, pH, and contact time and initial dye concentration on adsorption by modified kenaf core fiber was investigated. In addition, isotherms and kinetics adsorption studies were estimated for determination of the equilibrium adsorption capacity and reactions dynamics, respectively. Results showed that the best dose of MKCF was 0.1 g/100 mL, the maximum removal of RO16 was 97.25 at 30°C, pH = 6.5, and agitation speed was 150 rpm. The results also showed that the equilibrium data were represented by Freundlich isotherm with correlation coefficients R2=0.9924, and the kinetic study followed the pseudo-second-order kinetic model with correlation coefficients R2=0.9997 for Co=100 mg/L. Furthermore, the maximum adsorption capacity was 416.86 mg/g. Adsorption through kenaf was found to be very effective for the removal of the RO16 dye.

  9. The influence of spatial effects on the measurement results of reactivity in 'fast disturbances' of core parameters

    International Nuclear Information System (INIS)

    Tsyganov, S.V.; Shishkov, L.K.

    2001-01-01

    The analysis of methods for the determination of reactivity revealed an essential influence of spatial effect on the measurement precision. Using of reverse point kinetic equation for reactivity meter is assumed that the average neutron flux weigh with the importance function is known at every moment of the transient. In fact, reactivity meter represent behaviour of the neutron flux only of the part of the core, so measured value of reactivity can differ from really reactivity. Three-dimensional dynamic model of the core allow to evaluate such difference. It is supposed to evaluate correction factor for the neutron flux measured at the place where ion chamber situated with the three-dimensional model NOSTRA of the WWER core. On the basis of such algorithm we propose to build module allowing the influence of spatial effects on the results of the reactivity meter to be eliminated at real time regime. This code will be incorporated into the core monitoring system 'BLOK' (SCORPIO type) which is being developed for the Kola and Rostov NPP. The report illustrates utilization of such algorithm (Authors)

  10. Phase transition and thermal expansion studies of alumina thin films prepared by reactive pulsed laser deposition.

    Science.gov (United States)

    Balakrishnan, G; Thirumurugesan, R; Mohandas, E; Sastikumar, D; Kuppusami, P; Songl, J I

    2014-10-01

    Aluminium oxide (Al2O3) thin films were deposited on Si (100) substrates at an optimized oxygen partial pressure of 3 x 10(-3) mbar at room temperature by pulsed laser deposition (PLD). The films were characterized by high temperature X-ray diffraction (HTXRD), field emission scanning electron microscopy (FESEM) and atomic force microscopy (AFM). The HTXRD pattern showed the cubic y-Al2O3 phase in the temperature range 300-973 K. At temperatures ≥ 1073 K, the δ and θ-phases of Al2O3 were observed. The mean linear thermal expansion coefficient and volume thermal expansion coefficient of γ-Al2O3 was found to be 12.66 x 10(-6) K(-1) and 38.87 x 10(-6) K(-1) in the temperature range 300 K-1073 K. The field emission scanning electron microscopy revealed a smooth and structureless morphology of the films deposited on Si (100). The atomic force microscopy study indicated the increased crystallinity and surface roughness of the films after annealing at high temperature.

  11. Using nodal expansion method in calculation of reactor core with square fuel assemblies

    International Nuclear Information System (INIS)

    Abdollahzadeh, M. Y.; Boroushaki, M.

    2009-01-01

    A polynomial nodal method is developed to solve few-group neutron diffusion equations in cartesian geometry. In this article, the effective multiplication factor, group flux and power distribution based on the nodal polynomial expansion procedure is presented. In addition, by comparison of the results the superiority of nodal expansion method on finite-difference and finite-element are fully demonstrated. The comparison of the results obtained by these method with those of the well known benchmark problems have shown that they are in very good agreement.

  12. An Interval Bound Algorithm of optimizing reactor core loading pattern by using reactivity interval schema

    International Nuclear Information System (INIS)

    Gong Zhaohu; Wang Kan; Yao Dong

    2011-01-01

    Highlights: → We present a new Loading Pattern Optimization method - Interval Bound Algorithm (IBA). → IBA directly uses the reactivity of fuel assemblies and burnable poison. → IBA can optimize fuel assembly orientation in a coupled way. → Numerical experiment shows that IBA outperforms genetic algorithm and engineers. → We devise DDWF technique to deal with multiple objectives and constraints. - Abstract: In order to optimize the core loading pattern in Nuclear Power Plants, the paper presents a new optimization method - Interval Bound Algorithm (IBA). Similar to the typical population based algorithms, e.g. genetic algorithm, IBA maintains a population of solutions and evolves them during the optimization process. IBA acquires the solution by statistical learning and sampling the control variable intervals of the population in each iteration. The control variables are the transforms of the reactivity of fuel assemblies or the worth of burnable poisons, which are the crucial heuristic information for loading pattern optimization problems. IBA can deal with the relationship between the dependent variables by defining the control variables. Based on the IBA algorithm, a parallel Loading Pattern Optimization code, named IBALPO, has been developed. To deal with multiple objectives and constraints, the Dynamic Discontinuous Weight Factors (DDWF) for the fitness function have been used in IBALPO. Finally, the code system has been used to solve a realistic reloading problem and a better pattern has been obtained compared with the ones searched by engineers and genetic algorithm, thus the performance of the code is proved.

  13. Analysis of Moderator Temperature Reactivity Coefficient of the PWR Core Using WIMS-ANL

    International Nuclear Information System (INIS)

    Tukiran; Rokhmadi

    2007-01-01

    The Moderator Temperature Reactivity Coefficient (MTRC) is an important parameter in design, control and safety, particularly in PWR reactor. It is then very important to validate any new processed library for an accurate prediction of this parameter. The objective of this work is to validate the newly WIMS library based on ENDF/B-VI nuclear data files, especially for the prediction of the MTRC parameter. For this purpose, it is used a set of light water moderated lattice experiments as the NORA experiment and R1-100H critical reactors, both of reactors using UO 2 fuel pellet. Analysis is used with WIMSD/4 lattice code with original cross section libraries and WIMS-ANL with ENDF/B-VI cross section libraries. The results showed that the moderator temperatures reactivity coefficients for the NORA reactor using original libraries is - 5.039E-04 %Δk/k/℃ but for ENDF/B-VI libraries is - 2.925E-03 %Δk/k/℃. Compared to the designed value of the reactor core, the difference is in the range of 1.8 - 3.8 % for ENDF/B-IV libraries. It can be concluded that for reactor safety and control analysis, it has to be used ENDF/B- VI libraries because the original libraries is not accurate any more. (author)

  14. SULEU NTP Core with Passive Reactivity Control and Enhanced Submersion Safety

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, Paolo; Kim, Yong Hee [KAIST, Daejeon (Korea, Republic of); Eades, Michael J [The Ohio State University, Ohio (United States)

    2016-05-15

    In this summary, SULEU has been adapted to implement some of the latest developments of LEUNTP design efforts. These include the implementation of a rapid depletion burnable absorber to flatten the reactivity profile during operation and the addition of a lower axial reflector to help minimize the reactivity increase during the full submersion criticality accident. The purpose of this study is to show the state of current LEU-NTP designs in terms of resolving key issues such as minimizing control drum usage and resolving the full submersion criticality accident. Future work will include integrating the rapid depletion poison with other passive reactivity control devices (such as hydrogen density in the tie-tubes) and developing additional systems for mitigating the full submersion criticality accident. It is widely acknowledged that nuclear thermal propulsion (NTP) is an enabling technology for manned missions to Mars and other locations beyond low-Earth orbit. Without nuclear thermal propulsion, manned space travel will be severely limited by the propellant requirements of chemical propulsion and significantly longer travel times of electric propulsion. While the performance superiority of NTP is clear, its implementation has been to date unsuccessful due to the significant costs of development, implementation, and regulations associated with the heritage NTP designs. These new systems take heritage designs and experimental results and adapt them to use LEU fuel with minimum impact on the heritage system. This is done in order to ensure their continued relevance with existing NTP research efforts and enable their rapid implementation into existing NASA efforts for human Mars mission planning. Of the current baseline NTP designs being studied, this paper concerns itself with the improvement of the Superb Use of Low Enriched (SULEU) core.

  15. Series expansion of two-dimensional fields produced by iron-core magnets

    International Nuclear Information System (INIS)

    Satoh, Kotaro.

    1997-02-01

    This paper discusses the validity of a series expansion of two-dimensional magnetic fields with harmonic functions, and suggests that the series may not converge outside of the pole gap. It also points out that this difficulty may appear due to a slow convergence of the series near to the pole edge, even within the convergent area. (author)

  16. Numerical analysis of the reactivity for the dry lattices above the water level of the critical fuel cores

    International Nuclear Information System (INIS)

    Nauchi, Yasushi; Kameyama, Takanori

    2003-01-01

    Criticality analysis has been performed for dozens of tank type cores in which fuel lattices are loaded vertically and partially immersed in light water. The reactivity effect of dry part of lattices stuck above the critical water level has been calculated using the continuous energy Monte Carlo method. The reactivity effect exceeds 0.8% both for MOX and UOX fuel lattices of large buckling (B z 2 > 0.0025 cm -2 ). It is evaluated that at least 20 cm length of fuel rods above the critical water level has significant reactivity effect. (author)

  17. The BWR core simulator COSIMA with 2 group nodal flux expansion and control rod history

    International Nuclear Information System (INIS)

    Hoejerup, C.F.

    1989-08-01

    The boiling water simulator NOTAM has been modified and improved in several aspects: - The ''1 1/2'' energy group TRILUX nodal flux solution method has been exchanged with a 2 group modal expansion method. - Control rod ''history'' has been introduced. - Precalculated instrument factors have been introduced. The paper describes these improvements, which were considered sufficiently large to justify a new name to the programme: COSIMA. (author)

  18. Benchscale Assessment of the Efficacy of a Reactive Core Mat to Isolate PAH-spiked Aquatic Sediments.

    Science.gov (United States)

    Meric, Dogus; Barbuto, Sara; Sheahan, Thomas C; Shine, James P; Alshawabkeh, Akram N

    2014-01-01

    This paper describes the results of a benchscale testing program to assess the efficacy of a reactive core mat (RCM) for short term isolation and partial remediation of contaminated, subaqueous sediments. The 1.25 cm thick RCM (with a core reactive material such as organoclay with filtering layers on top and bottom) is placed on the sediment, and approximately 7.5 - 10 cm of overlying soil is placed on the RCM for stability and protection. A set of experiments were conducted to measure the sorption characteristics of the mat core (organoclay) and sediment used in the experiments, and to determine the fate of semi-volatile organic contaminants and non-reactive tracers through the sediment and reactive mat. The experimental study was conducted on naphthalene-spiked Neponset River (Milton, MA) sediment. The results show nonlinear sorption behavior for organoclay, with sorption capacity increasing with increasing naphthalene concentration. Neponset River sediment showed a notably high sorption capacity, likely due to the relatively high organic carbon fraction (14%). The fate and transport experiments demonstrated the short term efficiency of the reactive mat to capture the contamination that is associated with the post-capping period during which the highest consolidation-induced advective flux occurs, driving solid particles, pore fluid and soluble contaminants toward the reactive mat. The goal of the mat placement is to provide a physical filtering and chemically reactive layer to isolate contamination from the overlying water column. An important finding is that because of the high sorption capacity of the Neponset River sediment, the physical filtering capability of the mat is as critical as its chemical reactive capacity.

  19. The effects of applying silicon carbide coating on core reactivity of pebble-bed HTR in water ingress accident

    Energy Technology Data Exchange (ETDEWEB)

    Zuhair, S.; Setiadipura, Topan [National Nuclear Energy Agency of Indonesia, Serpong Tagerang Selatan (Indonesia). Center for Nuclear Reactor Technology and Safety; Su' ud, Zaki [Bandung Institute of Technology (Indonesia). Dept. of Physics

    2017-03-15

    Graphite is used as the moderator, fuel barrier material, and core structure in High Temperature Reactors (HTRs). However, despite its good thermal and mechanical properties below the radiation and high temperatures, it cannot avoid corrosion as a consequence of an accident of water/air ingress. Degradation of graphite as a main HTR material and the formation of dangerous CO gas is a serious problem in HTR safety. One of the several steps that can be adopted to avoid or prevent the corrosion of graphite by the water/air ingress is the application of a thin layer of silicon carbide (SiC) on the surface of the fuel element. This study investigates the effect of applying SiC coating on the fuel surfaces of pebble-bed HTR in water ingress accident from the reactivity points of view. A series of reactivity calculations were done with the Monte Carlo transport code MCNPX and continuous energy nuclear data library ENDF/B-VII at temperature of 1200 K. Three options of UO{sub 2}, PuO{sub 2}, and ThO{sub 2}/UO{sub 2} fuel kernel were considered to obtain the inter comparison of the core reactivity of pebble-bed HTR in conditions of water/air ingress accident. The calculation results indicated that the UO{sub 2}-fueled pebble-bed HTR reactivity was slightly reduced and relatively more decreased when the thickness of the SiC coating increased. The reactivity characteristic of ThO{sub 2}/UO{sub 2}-fueled pebble-bed HTR showed a similar trend to that of UO{sub 2}, but did not show reactivity peak caused by water ingress. In contrast with UO{sub 2}- and ThO{sub 2}-fueled pebble-bed HTR, although the reactivity of PuO{sub 2}-fueled pebble-bed HTR was the lowest, its characteristics showed a very high reactivity peak (0.33 Δk/k) and this introduction of positive reactivity is difficult to control. SiC coating on the surface of the plutonium fuel pebble has no significant impact. From the comparison between reactivity characteristics of uranium, thorium and plutonium cores with 0

  20. Neutronic characterization of cylindrical core of minor excess reactivity in the nuclear reactor IPEN/MB-01 from the measure of neutron flux distribution and its reactivity ratio

    Energy Technology Data Exchange (ETDEWEB)

    Bitelli, Ulysses d' Utra; Aredes, Vitor O.G.; Mura, Luiz E.C.; Santos, Diogo F. dos; Silva, Alexandre P. da, E-mail: ubitelli@ipen.br, E-mail: vitoraredes@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    When compared to a rectangular parallelepiped configuration the cylindrical configuration of a nuclear reactor core has a better neutron economy because in this configuration the probability of the neutron leakage is smaller, causing an increase in overall reactivity in the system to the same amount of fuel used. In this work we obtained a critical cylindrical configuration with the control rods 89.50% withdraw from the active region of the IPEN/MB-01 core. This is the cylindrical configuration minimum possible excess of reactivity. Thus we obtained a cylindrical configuration with a diameter of only 28 fuel rods with lowest possible excess of reactivity. For this purpose, 112 peripheral fuel rods are removed from standard reactor core (rectangular parallelepiped of 28x28 fuel rods). In this configuration the excesses of reactivity is approximated 279 pcm. From there, we characterize the neutron field by measuring the spatial distribution of the thermal and epithermal neutron flux for the reactor operating power of 83 watts measured by neutron noise analysis technique and 92.08± 0.07 watts measured by activation technique [10]. The values of thermal and epithermal neutron flux in different directions, axial, radial north-south and radial east-west, are obtained in the asymptotic region of the reactor core, away from the disturbances caused by the reflector and control bar, by irradiating thin gold foils infinitely diluted (1% Au - 99% Al) with and without (bare) cadmium cover. In addition to the distribution of neutron flux, the moderator temperature coefficient, the void coefficient, calibration of the control rods were measured. (author)

  1. Localized reactive flow in carbonate rocks: Core-flood experiments and network simulations

    Science.gov (United States)

    Wang, Haoyue; Bernabé, Yves; Mok, Ulrich; Evans, Brian

    2016-11-01

    We conducted four core-flood experiments on samples of a micritic, reef limestone from Abu Dhabi under conditions of constant flow rate. The pore fluid was water in equilibrium with CO2, which, because of its lowered pH, is chemically reactive with the limestone. Flow rates were between 0.03 and 0.1 mL/min. The difference between up and downstream pore pressures dropped to final values ≪1 MPa over periods of 3-18 h. Scanning electron microscope and microtomography imaging of the starting material showed that the limestone is mostly calcite and lacks connected macroporosity and that the prevailing pores are few microns large. During each experiment, a wormhole formed by localized dissolution, an observation consistent with the decreases in pressure head between the up and downstream reservoirs. Moreover, we numerically modeled the changes in permeability during the experiments. We devised a network approach that separated the pore space into competing subnetworks of pipes. Thus, the problem was framed as a competition of flow of the reactive fluid among the adversary subnetworks. The precondition for localization within certain time is that the leading subnetwork rapidly becomes more transmissible than its competitors. This novel model successfully simulated features of the shape of the wormhole as it grew from few to about 100 µm, matched the pressure history patterns, and yielded the correct order of magnitude of the breakthrough time. Finally, we systematically studied the impact of changing the statistical parameters of the subnetworks. Larger mean radius and spatial correlation of the leading subnetwork led to faster localization.

  2. Thermal hydraulic And RSG-Gas Core Reactivity Characteristics Due To Cold Water Insertion Accident

    International Nuclear Information System (INIS)

    Hastuti, Endiah Puji; Suparlina, Lily; Tukiran

    2000-01-01

    Under normal operating condition,the primary coolant is circulated by 2 out of the 3 primary coolant pumps. Unnecessary operation of the reserve pump would result in a temperatur decrease of the primary coolant by less than 5 o C. the corresponding increase of reactivity amounts to Δρ ≤0,1 %. The analysis was done using silicide core configuration data with 3.55 gU /cm 3 fuel loading. The calculation model was done with and without automatic control rod. The calculation results for the worst case condition, shows that reactor reached the maximum power 28.52 MW at 81.1 seconds, after the accident occurred. The maximal fuel element, cladding and outlet coolant temperatures are 148.3 o C,142.1 o C, and 75.7 o C, respectively. Safety margins for DNBR and flow instability reached 1.25 and 4.20, respectively. Comparing to the RSG-GAS safety margin at transient condition reguirement >1.48, RSG-GAS has enough safety margin if the power trip executed at 114% of 25 MW

  3. Measurements of the isothermal temperature reactivity coefficient of KUCA C-Core with a D{sub 2}O tank

    Energy Technology Data Exchange (ETDEWEB)

    Pyeon, Cheol Ho [Research Reactor Institute, Kyoto Univ., Osaka (Japan); Shim, Hyung Jin; Choi, Sung Hoon; Jeon, Byoung Kyu [Seoul National Univ., Seoul (Korea, Republic of); Ryu, Eun Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    The Kyoto University Critical Assembly (KUCA) is a multi-core type critical assembly consisting of three independent cores in the Kyoto University Research Reactor Institute. The light-water-moderated core (Ccore) is a tank type reactor, and the experiments of the isothermal temperature reactivity coefficient (ITRC) of C-core with a D{sub 2}O tank were carried out with the use of six 10 kW heaters and a radiator system in a dump tank, one 10 kW heater in a core tank, and one 5 kW heater in the D{sub 2}O tank. The ITRCs of the C-core with the D{sub 2}O tank immersed in the core tank are considered important to investigate the mechanism of moderation and reflection effects of H{sub 2}O and D{sub 2}O in the core on the evaluation by numerical simulations. The objectives of this paper are to report the ITRC measurements for C-core with D{sub 2}O tank ranging between 26.7 .deg. C and 58.5 .deg. C, and to examine the accuracy of the numerical simulations by the Seoul National University Monte Carlo code, McCARD, through the comparison between measured and calculated results.

  4. Assessment of Core Failure Limits for Light Water Reactor Fuel under Reactivity Initiated Accidents

    International Nuclear Information System (INIS)

    Jernkvist, Lars Olof; Massih, Ali R.

    2004-12-01

    Core failure limits for high-burnup light water reactor UO 2 fuel rods, subjected to postulated reactivity initiated accidents (RIAs), are here assessed by use of best-estimate computational methods. The considered RIAs are the hot zero power rod ejection accident (HZP REA) in pressurized water reactors and the cold zero power control rod drop accident (CZP CRDA) in boiling water reactors. Burnup dependent core failure limits for these events are established by calculating the fuel radial average enthalpy connected with incipient fuel pellet melting for fuel burnups in the range of 30 to 70 MWd/kgU. The postulated HZP REA and CZP CRDA result in lower enthalpies for pellet melting than RIAs that take place at rated power. Consequently, the enthalpy thresholds presented here are lower bounds to RIAs at rated power. The calculations are performed with best-estimate models, which are applied in the FRAPCON-3.2 and SCANAIR-3.2 computer codes. Based on the results of three-dimensional core kinetics analyses, the considered power transients are simulated by a Gaussian pulse shape, with a fixed width of either 25 ms (REA) or 45 ms (CRDA). Notwithstanding the differences in postulated accident scenarios between the REA and the CRDA, the calculated core failure limits for these two events are similar. The calculated enthalpy thresholds for fuel pellet melting decrease gradually with fuel burnup, from approximately 960 J/gUO 2 at 30 MWd/kgU to 810 J/gUO 2 at 70 MWd/kgU. The decline is due to depression of the UO 2 melting temperature with increasing burnup, in combination with burnup related changes to the radial power distribution within the fuel pellets. The presented fuel enthalpy thresholds for incipient UO 2 melting provide best-estimate core failure limits for low- and intermediate-burnup fuel. However, pulse reactor tests on high-burnup fuel rods indicate that the accumulation of gaseous fission products within the pellets may lead to fuel dispersal into the coolant at

  5. In-situ Density and Thermal Expansion Measurements of Fe and Fe-S Alloying Liquids Under Planetary Core Conditions

    Science.gov (United States)

    Jing, Z.; Chantel, J.; Yu, T.; Sakamaki, T.; Wang, Y.

    2015-12-01

    Liquid iron is likely the dominant constituent in the cores of terrestrial planets and icy satellites such as Earth, Mars, Mercury, the Moon, Ganymede, and Io. Suggested by geophysical and geochemical observations, light elements such as S, C, Si, etc., are likely present in planetary cores. These light elements can significantly reduce the density and melting temperature of the Fe cores, and hence their abundances are crucial to our understanding of the structure and thermal history of planetary cores, as well as the generation of intrinsic magnetic fields. Knowledge on the density of Fe-light element alloying liquids at high pressures is critical to place constraints on the composition of planetary cores. However, density data on liquid Fe-light element alloys at core pressures are very limited in pressure and composition and are sometimes controversial. In this study, we extend the density dataset for Fe-rich liquids by measuring the density of Fe, Fe-10wt%S, Fe-20wt%S, Fe-27wt%S, and FeS liquids using the X-ray absorption technique in a DIA-type multianvil apparatus up to 7 GPa and 2173 K. An ion chamber (1D-detector) and a CCD camera (2D-detector) were used to measure intensities of transmitted monochromatic X-rays through molten samples, with the photon energy optimized at 40 keV. The densities were then determined from the Beer-Lambert law using the mass absorption coefficients, calibrated by solid standards using X-ray diffraction. At each pressure, density measurements were conducted at a range of temperatures above the liquidus of the samples, enabling the determination of thermal expansion. Combined with our previous results on the sound velocity of Fe and Fe-S liquids at high pressures (Jing et al., 2014, Earth Planet. Sci. Lett. 396, 78-87), these data provide tight constraints on the equation of state and thermodynamic properties such as the adiabatic temperature gradient for Fe-S liquids. We will discuss these results with implications to planetary

  6. Insertion of reactivity (RIA) without scram in the reactor core IEA-R1 using code PARET

    International Nuclear Information System (INIS)

    Alves, Urias F.; Castrillo, Lazara S.; Lima, Fernando A.

    2013-01-01

    The modeling and analysis thermo hydraulics of a research reactor with MTR type fuel elements - Material Testing Reactor - was performed using the code PARET (Program for the Analysis of Reactor Transients) when in the system some external event is introduced that changed the reactivity in the reactor core. Transients of Reactivity Insertion of 0.5 , 1.5 and 2.0$/ 0.7s in the brazilian reactor IEA-R1 will be presented, and will be shown under what conditions it is possible to ensure the safe operation of its nucleus. (author)

  7. Expansion of CORE-SINEs in the genome of the Tasmanian devil.

    Science.gov (United States)

    Nilsson, Maria A; Janke, Axel; Murchison, Elizabeth P; Ning, Zemin; Hallström, Björn M

    2012-05-06

    The genome of the carnivorous marsupial, the Tasmanian devil (Sarcophilus harrisii, Order: Dasyuromorphia), was sequenced in the hopes of finding a cure for or gaining a better understanding of the contagious devil facial tumor disease that is threatening the species' survival. To better understand the Tasmanian devil genome, we screened it for transposable elements and investigated the dynamics of short interspersed element (SINE) retroposons. The temporal history of Tasmanian devil SINEs, elucidated using a transposition in transposition analysis, indicates that WSINE1, a CORE-SINE present in around 200,000 copies, is the most recently active element. Moreover, we discovered a new subtype of WSINE1 (WSINE1b) that comprises at least 90% of all Tasmanian devil WSINE1s. The frequencies of WSINE1 subtypes differ in the genomes of two of the other Australian marsupial orders. A co-segregation analysis indicated that at least 66 subfamilies of WSINE1 evolved during the evolution of Dasyuromorphia. Using a substitution rate derived from WSINE1 insertions, the ages of the subfamilies were estimated and correlated with a newly established phylogeny of Dasyuromorphia. Phylogenetic analyses and divergence time estimates of mitochondrial genome data indicate a rapid radiation of the Tasmanian devil and the closest relative the quolls (Dasyurus) around 14 million years ago. The radiation and abundance of CORE-SINEs in marsupial genomes indicates that they may be a major player in the evolution of marsupials. It is evident that the early phases of evolution of the carnivorous marsupial order Dasyuromorphia was characterized by a burst of SINE activity. A correlation between a speciation event and a major burst of retroposon activity is for the first time shown in a marsupial genome.

  8. Expansion of CORE-SINEs in the genome of the Tasmanian devil

    Directory of Open Access Journals (Sweden)

    Nilsson Maria A

    2012-05-01

    Full Text Available Abstract Background The genome of the carnivorous marsupial, the Tasmanian devil (Sarcophilus harrisii, Order: Dasyuromorphia, was sequenced in the hopes of finding a cure for or gaining a better understanding of the contagious devil facial tumor disease that is threatening the species’ survival. To better understand the Tasmanian devil genome, we screened it for transposable elements and investigated the dynamics of short interspersed element (SINE retroposons. Results The temporal history of Tasmanian devil SINEs, elucidated using a transposition in transposition analysis, indicates that WSINE1, a CORE-SINE present in around 200,000 copies, is the most recently active element. Moreover, we discovered a new subtype of WSINE1 (WSINE1b that comprises at least 90% of all Tasmanian devil WSINE1s. The frequencies of WSINE1 subtypes differ in the genomes of two of the other Australian marsupial orders. A co-segregation analysis indicated that at least 66 subfamilies of WSINE1 evolved during the evolution of Dasyuromorphia. Using a substitution rate derived from WSINE1 insertions, the ages of the subfamilies were estimated and correlated with a newly established phylogeny of Dasyuromorphia. Phylogenetic analyses and divergence time estimates of mitochondrial genome data indicate a rapid radiation of the Tasmanian devil and the closest relative the quolls (Dasyurus around 14 million years ago. Conclusions The radiation and abundance of CORE-SINEs in marsupial genomes indicates that they may be a major player in the evolution of marsupials. It is evident that the early phases of evolution of the carnivorous marsupial order Dasyuromorphia was characterized by a burst of SINE activity. A correlation between a speciation event and a major burst of retroposon activity is for the first time shown in a marsupial genome.

  9. The Accident Analysis Due to Reactivity Insertion of RSG GAS 3.55 g U/cc Silicide Core

    International Nuclear Information System (INIS)

    Endiah Puji-Hastuti; Surbakti, Tukiran

    2004-01-01

    The fuels of RSG-GAS reactor was changed from uranium oxide with 250 g U of loading or 2.96 g U/cc of fuel loading to uranium silicide with the same loading. The silicide fuels can be used in higher density, staying longer in the reactor core and hence having a longer cycle length. The silicide fuel in RSG-GAS core was made up in step-wise by using mixed up core Firstly, it was used silicide fuel with 250 g U of loading and then, silicide fuel with 300 g U of loading (3.55 g U/cc of fuel loading). In every step-wise of fuel loading, it must be analyzed its safety margin. In this occasion, the reactivity accident of RSG-GAS core with 300 g U of silicide fuel loading is analyzed. The calculation was done using EUREKA-2/RR code available at P2TRR. The calculation was done by reactivity insertion at start up and power rangers. The worst case accident is transient due to control rod with drawl failure at start up by means of lowest initial power (0.1 W), either in power range. From all cases which have been done, the results of analysis showed that there is no anomaly and safety margin break at RSG-GAS core with 300 g U silicide fuel loading. (author)

  10. Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Kuniyoshi, E-mail: takamatsu.kuniyoshi@jaea.go.jp; Yan, Xing L.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

    2014-05-01

    It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3 h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7 h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the

  11. Core2D. A code for non-isothermal water flow and reactive solute transport. Users manual version 2

    International Nuclear Information System (INIS)

    Samper, J.; Juncosa, R.; Delgado, J.; Montenegro, L.

    2000-01-01

    Understanding natural groundwater quality patterns, quantifying groundwater pollution and assessing the effects of waste disposal, require modeling tools accounting for water flow, and transport of heat and dissolved species as well as their complex interactions with solid and gases phases. This report contains the users manual of CORE ''2D Version V.2.0, a COde for modeling water flow (saturated and unsaturated), heat transport and multicomponent Reactive solute transport under both local chemical equilibrium and kinetic conditions. it is an updated and improved version of CORE-LE-2D V0 (Samper et al., 1988) which in turns is an extended version of TRANQUI, a previous reactive transport code (ENRESA, 1995). All these codes were developed within the context of Research Projects funded by ENRESA and the European Commission. (Author)

  12. Core 2D. A code for non-isothermal water flow and reactive solute transport. Users manual version 2

    Energy Technology Data Exchange (ETDEWEB)

    Samper, J; Juncosa, R; Delgado, J; Montenegro, L [Universidad de A Coruna (Spain)

    2000-07-01

    Understanding natural groundwater quality patterns, quantifying groundwater pollution and assessing the effects of waste disposal, require modeling tools accounting for water flow, and transport of heat and dissolved species as well as their complex interactions with solid and gases phases. This report contains the users manual of CORE ''2D Version V.2.0, a COde for modeling water flow (saturated and unsaturated), heat transport and multicomponent Reactive solute transport under both local chemical equilibrium and kinetic conditions. it is an updated and improved version of CORE-LE-2D V0 (Samper et al., 1988) which in turns is an extended version of TRANQUI, a previous reactive transport code (ENRESA, 1995). All these codes were developed within the context of Research Projects funded by ENRESA and the European Commission. (Author)

  13. Core 2D. A code for non-isothermal water flow and reactive solute transport. Users manual version 2

    Energy Technology Data Exchange (ETDEWEB)

    Samper, J.; Juncosa, R.; Delgado, J.; Montenegro, L. [Universidad de A Coruna (Spain)

    2000-07-01

    Understanding natural groundwater quality patterns, quantifying groundwater pollution and assessing the effects of waste disposal, require modeling tools accounting for water flow, and transport of heat and dissolved species as well as their complex interactions with solid and gases phases. This report contains the users manual of CORE ''2D Version V.2.0, a COde for modeling water flow (saturated and unsaturated), heat transport and multicomponent Reactive solute transport under both local chemical equilibrium and kinetic conditions. it is an updated and improved version of CORE-LE-2D V0 (Samper et al., 1988) which in turns is an extended version of TRANQUI, a previous reactive transport code (ENRESA, 1995). All these codes were developed within the context of Research Projects funded by ENRESA and the European Commission. (Author)

  14. HALO EXPANSION IN COSMOLOGICAL HYDRO SIMULATIONS: TOWARD A BARYONIC SOLUTION OF THE CUSP/CORE PROBLEM IN MASSIVE SPIRALS

    Energy Technology Data Exchange (ETDEWEB)

    Maccio, A. V.; Stinson, G. [Max-Planck-Institut fuer Astronomie, 69117 Heidelberg (Germany); Brook, C. B.; Gibson, B. K. [University of Central Lancashire, Jeremiah Horrocks Institute for Astrophysics and Supercomputing, Preston PR1 2HE (United Kingdom); Wadsley, J.; Couchman, H. M. P. [Department of Physics and Astronomy, McMaster University, Hamilton, Ontario, L8S 4M1 (Canada); Shen, S. [Department of Astronomy and Astrophysics, University of California Santa Cruz, Santa Cruz, CA 95064 (United States); Quinn, T., E-mail: maccio@mpia.de, E-mail: stinson@mpia.de [Astronomy Department, University of Washington, Seattle, WA 98195-1580 (United States)

    2012-01-15

    A clear prediction of the cold dark matter (CDM) model is the existence of cuspy dark matter halo density profiles on all mass scales. This is not in agreement with the observed rotation curves of spiral galaxies, challenging on small scales the otherwise successful CDM paradigm. In this work we employ high-resolution cosmological hydrodynamical simulations to study the effects of dissipative processes on the inner distribution of dark matter in Milky Way like objects (M Almost-Equal-To 10{sup 12} M{sub Sun }). Our simulations include supernova feedback, and the effects of the radiation pressure of massive stars before they explode as supernovae. The increased stellar feedback results in the expansion of the dark matter halo instead of contraction with respect to N-body simulations. Baryons are able to erase the dark matter cuspy distribution, creating a flat, cored, dark matter density profile in the central several kiloparsecs of a massive Milky-Way-like halo. The profile is well fit by a Burkert profile, with fitting parameters consistent with the observations. In addition, we obtain flat rotation curves as well as extended, exponential stellar disk profiles. While the stellar disk we obtain is still partially too thick to resemble the Milky Way thin disk, this pilot study shows that there is enough energy available in the baryonic component to alter the dark matter distribution even in massive disk galaxies, providing a possible solution to the long-standing problem of cusps versus cores.

  15. Dissecting cross-reactivity in hymenoptera venom allergy by circumvention of alpha-1,3-core fucosylation.

    Science.gov (United States)

    Seismann, Henning; Blank, Simon; Braren, Ingke; Greunke, Kerstin; Cifuentes, Liliana; Grunwald, Thomas; Bredehorst, Reinhard; Ollert, Markus; Spillner, Edzard

    2010-01-01

    Hymenoptera venom allergy is known to cause life-threatening and sometimes fatal IgE-mediated anaphylactic reactions in allergic individuals. About 30-50% of patients with insect venom allergy have IgE antibodies that react with both honeybee and yellow jacket venom. Apart from true double sensitisation, IgE against cross-reactive carbohydrate determinants (CCD) are the most frequent cause of multiple reactivities severely hampering the diagnosis and design of therapeutic strategies by clinically irrelevant test results. In this study we addressed allergenic cross-reactivity using a recombinant approach by employing cell lines with variant capacities of alpha-1,3-core fucosylation. The venom hyaluronidases, supposed major allergens implicated in cross-reactivity phenomena, from honeybee (Api m 2) and yellow jacket (Ves v 2a and its putative isoform Ves v 2b) as well as the human alpha-2HS-glycoprotein as control, were produced in different insect cell lines. In stark contrast to production in Trichoplusia ni (HighFive) cells, alpha-1,3-core fucosylation was absent or immunologically negligible after production in Spodoptera frugiperda (Sf9) cells. Consistently, co-expression of honeybee alpha-1,3-fucosyltransferase in Sf9 cells resulted in the reconstitution of CCD reactivity. Re-evaluation of differentially fucosylated hyaluronidases by screening of individual venom-sensitised sera emphasised the allergenic relevance of Api m 2 beyond its carbohydrate epitopes. In contrast, the vespid hyaluronidases, for which a predominance of Ves v 2b could be shown, exhibited pronounced and primary carbohydrate reactivity rendering their relevance in the context of allergy questionable. These findings show that the use of recombinant molecules devoid of CCDs represents a novel strategy with major implications for diagnostic and therapeutic approaches. Copyright 2010 Elsevier Ltd. All rights reserved.

  16. Experimental and calculating substantiation of reactivity balance and energy-release distribution in BN-600 core

    International Nuclear Information System (INIS)

    Moiseev, A.V.; Khomyakov, Yu.S.; Surov, S.V.

    2013-01-01

    This paper presents the results of experimental and theoretical work done in 2003-2010 years on substantiation of neutron-physical characteristics of the BN-600 core. 1. Transition to the new core 01M2 with high burnup 11.2% h.a. (the 4-th upgrade of the BN-600 core). Transfer was made without changing the constructive of the core almost by reducing conservatism of design decisions. 2. The end of BN-600 design life cycle and extending it to 10-15 years. Need for analysis and comprehension of the BN-600 experience. 3. Development and introduction of new methods of analysis (precision method of Monte Carlo). 4. In the experiments was a change of equipment and measurement techniques

  17. Expansion of donor-reactive host T cells in primary graft failure after allogeneic hematopoietic SCT following reduced-intensity conditioning.

    Science.gov (United States)

    Koyama, M; Hashimoto, D; Nagafuji, K; Eto, T; Ohno, Y; Aoyama, K; Iwasaki, H; Miyamoto, T; Hill, G R; Akashi, K; Teshima, T

    2014-01-01

    Graft rejection remains a major obstacle in allogeneic hematopoietic SCT following reduced-intensity conditioning (RIC-SCT), particularly after cord blood transplantation (CBT). In a murine MHC-mismatched model of RIC-SCT, primary graft rejection was associated with activation and expansion of donor-reactive host T cells in peripheral blood and BM early after SCT. Donor-derived dendritic cells are at least partly involved in host T-cell activation. We then evaluated if such an expansion of host T cells could be associated with graft rejection after RIC-CBT. Expansion of residual host lymphocytes was observed in 4/7 patients with graft rejection at 3 weeks after CBT, but in none of the 17 patients who achieved engraftment. These results suggest the crucial role of residual host T cells after RIC-SCT in graft rejection and expansion of host T cells could be a marker of graft rejection. Development of more efficient T cell-suppressive conditioning regimens may be necessary in the context of RIC-SCT.

  18. Reactive effects of core fermion excitations on the inertial mass of a vortex

    International Nuclear Information System (INIS)

    Simanek, E.

    1995-01-01

    The time-dependent Schroedinger equation for a fermion two-dimensional superfluid containing a moving vortex is solved using the adiabatic approximation. The expectation value of the linear momentum of the vortex is found dominated by core fermion excitations. The resulting inertial vortex mass, obtained in the adiabatic limit, is larger than the standard core mass by a factor of (k F ξ) 2 where ξ is the coherence length at T=0. Anamalous velocity dependence of the mass, associated with the breakdown of the adiabatic approximation, is predicted

  19. Expansion of mycobacterium-reactive gamma delta T cells by a subset of memory helper T cells.

    Science.gov (United States)

    Vila, L M; Haftel, H M; Park, H S; Lin, M S; Romzek, N C; Hanash, S M; Holoshitz, J

    1995-04-01

    Human gamma delta T cells expressing the V gamma 9/V delta 2 T-cell receptor have been previously found to proliferate in response to certain microorganisms and to expand throughout life, presumably because of extrathymic activation by foreign antigens. In vitro expansion of V gamma 9/V delta 2 cells by mycobacteria has been previously shown to be dependent on accessory cells. In order to gain an insight into the mechanisms involved in the expansion of these cells, we have undertaken to identify the peripheral blood subset of cells on which proliferation of V gamma 9/V delta 2 cells in response to mycobacteria is dependent. Contrary to their role in antigen presentation to alpha beta T cells, professional antigen-presenting cells, such as monocytes, B cells, and dendritic cells, were unable to provide the cellular support for the expansion of V gamma 9/V delta 2 cells. Selective depletion of T-cell subsets, as well as the use of highly purified T-cell populations, indicated that the only subset of peripheral blood cells that could expand V gamma 9/V delta 2 cells were CD4+ CD45RO+ CD7- alpha beta T cells. These cells underwent distinct intracellular signaling events after stimulation with the mycobacterial antigen. Expansion of V gamma 9/V delta 2 cells by alpha beta T cells was dependent on cell-cell contact. This is the first evidence that a small subset of the memory helper T-cell population is exclusively responsible for the peripheral expansion of V gamma 9/V delta 2 cells. These data illustrate a unique aspect of antigen recognition by gamma delta T cells and provide new means to study their immune defense role.

  20. (Gold core)/(titania shell) nanostructures for plasmon-enhanced photon harvesting and generation of reactive oxygen species

    KAUST Repository

    Fang, Caihong; Jia, Henglei; Chang, Shuai; Ruan, Qifeng; Wang, Peng; Chen, Tao; Wang, Jianfang

    2014-01-01

    Integration of gold and titania in a nanoscale core/shell architecture can offer large active metal/semiconductor interfacial areas and avoid aggregation and reshaping of the metal nanocrystal core. Such hybrid nanostructures are very useful for studying plasmon-enhanced/enabled processes and have great potential in light-harvesting applications. Herein we report on a facile route to (gold nanocrystal core)/(titania shell) nanostructures with their plasmon band synthetically variable from ∼700 nm to over 1000 nm. The coating method has also been applied to other mono- and bi-metallic Pd, Pt, Au nanocrystals. The gold/titania nanostructures have been employed as the scattering layer in dye-sensitized solar cells, with the resultant cells exhibiting a 13.3% increase in the power conversion efficiency and a 75% decrease in the scattering-layer thickness. Moreover, under resonant excitation, the gold/titania nanostructures can efficiently utilize low-energy photons to generate reactive oxygen species, including singlet oxygen and hydroxyl radicals.

  1. Analysis of reactivity characteristics of the MONJU initial core using JENDL-3.2

    Energy Technology Data Exchange (ETDEWEB)

    Sasaki, Kenji; Suzuki, Takayuki; Suzuki, Norimichi [Power Reactor and Nuclear Fuel Development Corp., Tsuruga, Fukui (Japan). Monju Construction Office; Itagaki, Yoshihiko

    1998-03-01

    This paper describes the evaluated results of criticality, absorber rod worth and coolant worth in the MONJU initial cores based on the JENDL-3.2 library compared with those of the JENDL-2 library. We confirm that the ratios of calculated and experimental (C/E) values using the JENDL-3.2 library are slightly better than those based on the JENDL-2 library. (author)

  2. Automatic determination of pressurized water reactor core loading patterns that maximize beginning-of-cycle reactivity within power-peaking and burnup constraints

    International Nuclear Information System (INIS)

    Hobson, G.H.; Turinsky, P.J.

    1986-01-01

    Computational capability has been developed to automatically determine a good estimate of the core loading pattern, which minimizes fuel cycle costs for a pressurized water reactor (PWR). Equating fuel cycle cost minimization with core reactivity maximization, the objective is to determine the loading pattern that maximizes core reactivity while satisfying power peaking, discharge burnup, and other constraints. The method utilizes a two-dimensional, coarse-mesh, finite difference scheme to evaluate core reactivity and fluxes for an initial reference loading pattern. First-order perturbation theory is applied to determine the effects of assembly shuffling on reactivity, power distribution, end-of-cycle burnup. Monte Carlo integer programming is then used to determine a near-optimal loading pattern within a range of loading patterns near the reference pattern. The process then repeats with the new loading pattern as the reference loading pattern and terminates when no better loading pattern can be determined. The process was applied with both reactivity maximization and radial power-peaking minimization as objectives. Results on a typical large PWR indicate that the cost of obtaining an 8% improvement in radial power-peaking margin is ≅2% in fuel cycle costs, for the reload core loaded without burnable poisons that was studied

  3. Analysis Influence of Mixing Gd2O3 in the Silicide Fuel Element to Core Excess Reactivity of RSG-GAS

    International Nuclear Information System (INIS)

    Susilo, Jati

    2004-01-01

    Gadolinium (Gd 2 O 3 ) is a burnable poison material mixed in the pin fuel element of the LWR core used to decrease core excess reactivity. In this research, analysis influence of mixing Gd 2 O 3 in the silicide fuel element to excess reactivity of the RSG-GAS core had been done. Equivalent cell of the equilibrium core developed by L.E.Strawbridge from Westing House Co. burn-up calculation has been done using SRAC-PIJ computer code achieve infinite multiplication factor (k x ). Value of Gd 2 O 3 concentration in the fuel element (pcm) showed by mass ratio of Gd 2 O 3 (gram) to that U 3 Si 2 (gram) times 10 5 , that is 0 pcm ∼ 100 pcm. From the calculation results analysis showed that Gd 2 O 3 concentration added should be considered. because a large number of Gd 2 O 3 will result in not achieving criticality at the Beginning Of Cycle. The maximum concentration of Gd 2 O 3 for RSG-GAS equilibrium fueled silicide 2.96 grU/cc is 80 pcm or 52.02 mgram/fuel plate. Maximum reduction of core excess reactivity due to mixing of Gd 2 O 3 in the RSG-GAS silicide fuels was around 1.502 %Δk/k, and hence not achieving the standard nominal excess reactivity for RSG-GAS core using high density of U 3 Si 2 -Al fuel. (author)

  4. The influence of the reactivity ramp on the course of the power transient in the MARK 1A core of the SNR 300

    International Nuclear Information System (INIS)

    Froehlich, R.; Schmuck, P.

    1976-01-01

    The course of a hypothetic transient overpower accident caused by the onset of a not further specified reactivity ramp accompanied by the simultaneous failure of both shutdown systems must be analyzed in the SNR 300 Mark 1A core licensing procedure. The present study is limited to the discussion of the starting and shutdown phases of such accidents for the fresh core. Depending on the operational state of the reactor, the core geometry is still intact during the starting phase. In the following shutdown phase (core disassembly phase), large-scale mass transfer leads to the nuclear shutdown of the reactor. (orig./AK) [de

  5. Visualized fuelling process and 3 dimensional reactivity device and core monitor

    International Nuclear Information System (INIS)

    Patterson, B.; Cusson, R.; Crowell, T.

    2006-01-01

    A new reactor fueling animated graphical display and a 3 dimensional view of the reactor core display are presented that are useful for the physics fuelling engineer, the Control Room Operators, the fuel handling operators and the fuel handling support engineers. Data is downloaded from the online fuelling computer to a data server that is network accessible. The fuelling display and 3Dview display can run on any network connected Computer. The animated graphical fuelling display offers a huge reduction in cognitive workload for all users. The authors recommend that animated graphical displays be developed and utilized wherever personnel have to visualize complex equipment operation. (author)

  6. Optimizing the Performance of Reactive Molecular Dynamics Simulations for Multi-core Architectures

    Energy Technology Data Exchange (ETDEWEB)

    Aktulga, Hasan Metin [Michigan State Univ., East Lansing, MI (United States); Coffman, Paul [Argonne National Lab. (ANL), Argonne, IL (United States); Shan, Tzu-Ray [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Knight, Chris [Argonne National Lab. (ANL), Argonne, IL (United States); Jiang, Wei [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-12-01

    Hybrid parallelism allows high performance computing applications to better leverage the increasing on-node parallelism of modern supercomputers. In this paper, we present a hybrid parallel implementation of the widely used LAMMPS/ReaxC package, where the construction of bonded and nonbonded lists and evaluation of complex ReaxFF interactions are implemented efficiently using OpenMP parallelism. Additionally, the performance of the QEq charge equilibration scheme is examined and a dual-solver is implemented. We present the performance of the resulting ReaxC-OMP package on a state-of-the-art multi-core architecture Mira, an IBM BlueGene/Q supercomputer. For system sizes ranging from 32 thousand to 16.6 million particles, speedups in the range of 1.5-4.5x are observed using the new ReaxC-OMP software. Sustained performance improvements have been observed for up to 262,144 cores (1,048,576 processes) of Mira with a weak scaling efficiency of 91.5% in larger simulations containing 16.6 million particles.

  7. Study of graphite reactivity worth on well-defined cores assembled on LR-0 reactor

    International Nuclear Information System (INIS)

    Košťál, Michal; Rypar, Vojtěch; Milčák, Ján; Juříček, Vlastimil; Losa, Evžen; Forget, Benoit; Harper, Sterling

    2016-01-01

    Highlights: • A light water critical facility for graphite reactivity worth measurements. • Comparison of calculated and measured k eff . • Effect of graphite description on k eff . - Abstract: Graphite is an often-used moderating material on the basis of its good moderating power and very low absorption cross section. This small absorption cross section permits the use of natural or low-enriched uranium in graphite moderated reactors. Graphite is now being considered as the moderator for Fluoride-salt-cooled High Temperature Reactors (FHR). The critical moderator level was measured for various graphite block configurations in an experimental dry assembly of the LR-0 reactor. Comparisons with experiments were performed between Monte Carlo simulation tools for which satisfactory agreement was obtained with the exception of some systematic discrepancies. The larger discrepancies were observed when using the ENDF/B-VII.0 library. To decrease the uncertainties, based on conservative assumptions, relative comparisons were done. The results provided by the different nuclear data libraries are within 3 sigma interval of experimental uncertainties. It has been determined that differences between the results of calculations are caused by variations in the (n,n), (n,n′), (n,g) reactions and also by various angular distributions, while the (n,g) cross section variations play only a minor role for these configurations.

  8. Surface zwitterionicalization of poly(vinylidene fluoride) membranes from the entrapped reactive core-shell silica nanoparticles.

    Science.gov (United States)

    Zhu, Li-Jing; Zhu, Li-Ping; Zhang, Pei-Bin; Zhu, Bao-Ku; Xu, You-Yi

    2016-04-15

    We demonstrate the preparation and properties of poly(vinylidene fluoride) (PVDF) filtration membranes modified via surface zwitterionicalization mediated by reactive core-shell silica nanoparticles (SiO2 NPs). The organic/inorganic hybrid SiO2 NPs grafted with poly(methyl meth acrylate)-block-poly(2-dimethylaminoethyl methacrylate) copolymer (PMMA-b-PDMAEMA) shell were prepared by surface-initiated reversible addition fragmentation chain transfer (SI-RAFT) polymerization and then used as a membrane-making additive of PVDF membranes. The PDMAEMA exposed on membrane surface and pore walls were quaternized into zwitterionic poly(sulfobetaine methacrylate) (PSBMA) using 1,3-propane sultone (1,3-PS) as the quaternization agent. The membrane surface chemistry and morphology were analyzed by attenuated total reflectance Fourier transform infrared spectroscopy (ATR-FTIR), X-ray photoelectron spectroscopy (XPS) and scanning electron microscopy (SEM), respectively. The hydrophilicity, permeability and antifouling ability of the investigated membranes were evaluated in detail. It was found that the PSBMA chains brought highly-hydrophilic and strong fouling resistant characteristics to PVDF membranes due to the powerful hydration of zwitterionic surface. The SiO2 cores and PMMA chains in the hybrid NPs play a role of anchors for the linking of PSBMA chains to membrane surface. Compared to the traditional strategies for membrane hydrophilic modification, the developed method in this work combined the advantages of both blending and surface reaction. Copyright © 2016 Elsevier Inc. All rights reserved.

  9. Nonlinear thermo-optical properties of two-layered spherical system of gold nanoparticle core and water vapor shell during initial stage of shell expansion

    Directory of Open Access Journals (Sweden)

    Astafyeva Liudmila

    2011-01-01

    Full Text Available Abstract Nonlinear thermo-optical properties of two-layered spherical system of gold nanoparticle core and water vapor shell, created under laser heating of nanoparticle in water, were theoretically investigated. Vapor shell expansion leads to decreasing up to one to two orders of magnitude in comparison with initial values of scattering and extinction of the radiation with wavelengths 532 and 633 nm by system while shell radius is increased up to value of about two radii of nanoparticle. Subsequent increasing of shell radius more than two radii of nanoparticle leads to rise of scattering and extinction properties of system over initial values. The significant decrease of radiation scattering and extinction by system of nanoparticle-vapor shell can be used for experimental detection of the energy threshold of vapor shell formation and investigation of the first stages of its expansion. PACS: 42.62.BE. 78.67. BF

  10. Internal core tightener

    International Nuclear Information System (INIS)

    Brynsvold, G.V.; Snyder, H.J. Jr.

    1976-01-01

    An internal core tightener is disclosed which is a linear actuated (vertical actuation motion) expanding device utilizing a minimum of moving parts to perform the lateral tightening function. The key features are: (1) large contact areas to transmit loads during reactor operation; (2) actuation cam surfaces loaded only during clamping and unclamping operation; (3) separation of the parts and internal operation involved in the holding function from those involved in the actuation function; and (4) preloaded pads with compliant travel at each face of the hexagonal assembly at the two clamping planes to accommodate thermal expansion and irradiation induced swelling. The latter feature enables use of a ''fixed'' outer core boundary, and thus eliminates the uncertainty in gross core dimensions, and potential for rapid core reactivity changes as a result of core dimensional change. 5 claims, 12 drawing figures

  11. Tolerance without clonal expansion: self-antigen-expressing B cells program self-reactive T cells for future deletion.

    Science.gov (United States)

    Frommer, Friederike; Heinen, Tobias J A J; Wunderlich, F Thomas; Yogev, Nir; Buch, Thorsten; Roers, Axel; Bettelli, Estelle; Müller, Werner; Anderton, Stephen M; Waisman, Ari

    2008-10-15

    B cells have been shown in various animal models to induce immunological tolerance leading to reduced immune responses and protection from autoimmunity. We show that interaction of B cells with naive T cells results in T cell triggering accompanied by the expression of negative costimulatory molecules such as PD-1, CTLA-4, B and T lymphocyte attenuator, and CD5. Following interaction with B cells, T cells were not induced to proliferate, in a process that was dependent on their expression of PD-1 and CTLA-4, but not CD5. In contrast, the T cells became sensitive to Ag-induced cell death. Our results demonstrate that B cells participate in the homeostasis of the immune system by ablation of conventional self-reactive T cells.

  12. Investigation of reactivity variations of the Isfahan MNSR reactor due to variations in the thickness of the core top beryllium layer using WIMSD and MCNP codes

    Directory of Open Access Journals (Sweden)

    A Shirani

    2010-12-01

    Full Text Available In this work, the Isfahan Miniature Neutron Source Reactor (MNSR is first simulated using the WIMSD code, and its fuel burn-up after 7 years of operation ( when the reactor was revived by adding a 1.5 mm thick beryllium shim plate to the top of its core and also after 14 years of operation (total operation time of the reactor is calculated. The reactor is then simulated using the MCNP code, and its reactivity variation due to adding a 1.5 mm thick beryllium shim plate to the top of the reactor core, after 7 years of operation, is calculated. The results show good agreement with the available data collected at the revival time. Exess reactivity of the reactor at present time (after 14 years of operation and after 7 years of the the reactor revival time is also determined both experimentally and by calculation, which show good agreement, and indicate that at the present time there is no need to add any further beryllium shim plate to the top of the reactor core. Furthermore, by adding more beryllium layers with various thicknesses to the top of the reactor core, in the input program of the MCNP program, reactivity value of these layers is calculated. From these results, one can predict the necessary beryllium thickness needed to reach a desired reactivity in the MNSR reactor.

  13. In-core power sharing and fuel requirement study for a decommissioning Boiling Water Reactor using the linear reactivity model

    International Nuclear Information System (INIS)

    Chen, Chung-Yuan; Tung, Wu-Hsiung; Yaur, Shung-Jung; Kuo, Weng-Sheng

    2014-01-01

    Highlights: • Linear reactivity model (LRM) was modified and applied to Boiling Water Reactor. • The power sharing and fuel requirement study of the last cycle and two cycles before decommissioning was implemented. • The loading pattern design concept for the cycles before decommissioning is carried out. - Abstract: A study of in-core power sharing and fuel requirement for a decommissioning BWR (Boiling Water Reactor) was carried out using the linear reactivity model (LRM). The power sharing of each fuel batch was taken as an independent variable, and the related parameters were set and modified to simulate actual cases. Optimizations of the last cycle and two cycles before decommissioning were both implemented; in the last-one-cycle optimization, a single cycle optimization was carried out with different upper limits of fuel batch power, whereas, in the two-cycle optimization, two cycles were optimized with different cycle lengths, along with two different optimization approaches which are the simultaneous optimization of two cycles (MO) and two successive single-cycle optimizations (SO). The results of the last-one-cycle optimization show that it is better to increase the fresh fuel power and decrease the thrice-burnt fuel power as much as possible. It also shows that relaxing the power limit is good to the fresh fuel requirement which will be reduced under lower power limit. On the other hand, the results of the last-two-cycle (cycle N-1 and N) optimization show that the MO is better than SO, and the power of fresh fuel batch should be decreased in cycle N-1 to save its energy for the next cycle. The results of the single-cycle optimization are found to be the same as that in cycle N of the multi-cycle optimization. Besides that, under the same total energy requirement of two cycles, a long-short distribution of cycle length design can save more fresh fuel

  14. Measurement of reactivity worths of burnable poison rods in enriched uranium graphite-moderated core simulated to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi; Takeuchi, Motoyoshi; Kitadate, Kenji; Yoshifuji, Hisashi; Kaneko, Yoshihiko

    1980-11-01

    As the core design for the Experimental Very High Temperature Gas Cooled Reactor progresses, evaluation of design precision has become increasingly important. For a high precision design, it is required to have adequate group constants based on accurate nuclear data, as well as calculation methods properly describing the physical behavior of neutrons. We, therefore, assembled a simulation core for VHTR, SHE-14, using a graphite-moderated 20%-enriched uranium Semi-Homogeneous Experimental Critical Facility (SHE), and obtained useful experimental data in evaluating the design precision. The VHTR is designed to accommodate burnable poison and control rods for reactivity compensation. Accordingly, the experimental burnable poison rods which are similar to those to be used in the experimental reactor were prepared, and their reactivity values were measured in the SHE-14 core. One to three rods of the above experimental burnable poison rods were inserted into the central column of the SHE-14 core, and the reactivity values were measured by the period and fuel rod substitution method. The results of the measurements have clearly shown that due to the self-shielding effect of B 4 C particles the reactivity value decreases with increasing particle diameter. For the particle diameter, the reactivity value is found to increase linearly with the logarithm of boron content. The measured values and those calculated are found to agree with each other within 5%. These results indicate that the reactivity of the burnable poison rod can be estimated fairly accurately by taking into account the self-shielding effect of B 4 C particles and the heterogeneity of the lattice cell. (author)

  15. Automatic determination of pressurized water reactor core loading patterns which maximize end-of-cycle reactivity within power peaking and burnup constraints

    International Nuclear Information System (INIS)

    Hobson, G.H.

    1985-01-01

    An automated procedure for determining the optimal core loading pattern for a pressurized water reactor which maximizes end-of-cycle k/sub eff/ while satisfying constraints on power peaking and discharge burnup has been developed. The optimization algorithm combines a two energy group, two-dimensional coarse-mesh finite difference diffusion theory neutronics model to simulate core conditions, a perturbation theory approach to determine reactivity, flux, power and burnup changes as a function of assembly shuffling, and Monte Carlo integer programming to select the optimal loading pattern solution. The core examined was a typical Cycle 2 reload with no burnable poisons. Results indicate that the core loading pattern that maximizes end-of-cycle k/sub eff/ results in a 5.4% decrease in fuel cycle costs compared with the core loading pattern that minimizes the maximum relative radial power peak

  16. PBDOWN: A computer code for simulation of core material discharge and expansion in the upper coolant plenum in a hypothetical unprotected loss of flow accident in a LMFBR

    International Nuclear Information System (INIS)

    Royl, P.

    1985-01-01

    The report gives a description of the code PBDOWN (Pool Blow Down), its equations, input specifications and subroutines and it lists the input and output for some samples. Besides that some analysis results for the SNR-300 are discussed, that were obtained with this code. PBDOWN is an integral blow-down and expansion code, which simulates core material discharge and expansion into a sodium filled upper coolant plenum after build-up of vapour pressures in an unprotected loss of flow accident. The model includes the effect of sodium entrainment into an expending bubble of fuel or steel vapour with various assumptions for the heat transfer and vaporization of the entrained sodium droplets. The expanding vapour bubble is connected to the discharging pool via an orifice of a given size through which a time dependent ejection is simulated using quasi-stationary blow down correlations. The model allows bounding analysis of the possible influence of sodium vapour as a secondary working fluid, that is activated outside the pool on the overall expansion energy and discharge

  17. Evaluation of CANDU6 PCR (power coefficient of reactivity) with a 3-D whole-core Monte Carlo Analysis

    International Nuclear Information System (INIS)

    Motalab, Mohammad Abdul; Kim, Woosong; Kim, Yonghee

    2015-01-01

    Highlights: • The PCR of the CANDU6 reactor is slightly negative at low power, e.g. <80% P. • Doppler broadening of scattering resonances improves noticeably the FTC and make the PCR more negative or less positive in CANDU6. • The elevated inlet coolant condition can worsen significantly the PCR of CANDU6. • Improved design tools are needed for the safety evaluation of CANDU6 reactor. - Abstract: The power coefficient of reactivity (PCR) is a very important parameter for inherent safety and stability of nuclear reactors. The combined effect of a relatively less negative fuel temperature coefficient and a positive coolant temperature coefficient make the CANDU6 (CANada Deuterium Uranium) PCR very close to zero. In the original CANDU6 design, the PCR was calculated to be clearly negative. However, the latest physics design tools predict that the PCR is slightly positive for a wide operational range of reactor power. It is upon this contradictory observation that the CANDU6 PCR is re-evaluated in this work. In our previous study, the CANDU6 PCR was evaluated through a standard lattice analysis at mid-burnup and was found to be negative at low power. In this paper, the study was extended to a detailed 3-D CANDU6 whole-core model using the Monte Carlo code Serpent2. The Doppler broadening rejection correction (DBRC) method was implemented in the Serpent2 code in order to take into account thermal motion of the heavy uranium nucleus in the neutron-U scattering reactions. Time-average equilibrium core was considered for the evaluation of the representative PCR of CANDU6. Two thermal hydraulic models were considered in this work: one at design condition and the other at operating condition. Bundle-wise distributions of the coolant properties are modeled and the bundle-wise fuel temperature is also considered in this study. The evaluated nuclear data library ENDF/B-VII.0 was used throughout this Serpent2 evaluation. In these Monte Carlo calculations, a large number

  18. Reactivity feedback models for SSC-K

    Energy Technology Data Exchange (ETDEWEB)

    Han, Do Hee; Kwon, Young Min; Kim, Kyung Du; Chang, Won Pyo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    Safety of KALIMER is assured by the inherent safety of the core and passive safety of the safety-related systems. For the safety analysis of a new reactor design such as KALIMER, analysis models, which are consistent with the design, have to be developed for a plant-wide transient and safety analysis code. Efforts for the development of reactivity feedback models for SSC-K, which is now being developed for the safety analysis of KALIMER, is described in this report. Models for Doppler, sodium density/void, fuel axial expansion, core radial expansion, and CRDL expansion have been developed. Test runs have been performed for the unprotected accident for the verification of the models. Use of KALIMER reactivity coefficients and future development of models for GEM and PSDRS would make it possible to analyze the response of KALIMER under TOP as well as LOF and LOHS accident conditions using SSC-K. (author). 5 refs., 64 figs., 2 tabs.

  19. Control Rod Driveline Reactivity Feedback Model for Liquid Metal Reactors

    International Nuclear Information System (INIS)

    Kwon, Young-Min; Jeong, Hae-Yong; Chang, Won-Pyo; Cho, Chung-Ho; Lee, Yong-Bum

    2008-01-01

    The thermal expansion of the control rod drivelines (CRDL) is one important passive mitigator under all unprotected accident conditions in the metal and oxide cores. When the CRDL are washed by hot sodium in the coolant outlet plenum, the CRDL thermally expands and causes the control rods to be inserted further down into the active core region, providing a negative reactivity feedback. Since the control rods are attached to the top of the vessel head and the core attaches to the bottom of the reactor vessel (RV), the expansion of the vessel wall as it heats will either lower the core or raise the control rods supports. This contrary thermal expansion of the reactor vessel wall pulls the control rods out of the core somewhat, providing a positive reactivity feedback. However this is not a safety factor early in a transient because its time constant is relatively large. The total elongated length is calculated by subtracting the vessel expansion from the CRDL expansion to determine the net control rod expansion into the core. The system-wide safety analysis code SSC-K includes the CRDL/RV reactivity feedback model in which control rod and vessel expansions are calculated using single-nod temperatures for the vessel and CRDL masses. The KALIMER design has the upper internal structures (UIS) in which the CRDLs are positioned outside the structure where they are exposed to the mixed sodium temperature exiting the core. A new method to determine the CRDL expansion is suggested. Two dimensional hot pool thermal hydraulic model (HP2D) originally developed for the analysis of the stratification phenomena in the hot pool is utilized for a detailed heat transfer between the CRDL mass and the hot pool coolant. However, the reactor vessel wall temperature is still calculated by a simple lumped model

  20. Development of a standard data base for FBR core nuclear design (XIII). Analysis of small sample reactivity experiments at ZPPR-9

    International Nuclear Information System (INIS)

    Sato, Wakaei; Fukushima, Manabu; Ishikawa, Makoto

    2000-09-01

    A comprehensive study to evaluate and accumulate the abundant results of fast reactor physics is now in progress at O-arai Engineering Center to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as future commercial FBRs. The present report summarizes the analytical results of sample reactivity experiments at ZPPR-9 core, which has not been evaluated by the latest analytical method yet. The intention of the work is to extend and further generalize the standard data base for FBR core nuclear design. The analytical results of the sample reactivity experiments (samples: PU-30, U-6, DU-6, SS-1 and B-1) at ZPPR-9 core in JUPITER series, with the latest nuclear data library JENDL-3.2 and the analytical method which was established by the JUPITER analysis, can be concluded as follows: The region-averaged final C/E values generally agreed with unity within 5% differences at the inner core region. However, the C/E values of every sample showed the radial space-dependency increasing from center to core edge, especially the discrepancy of B-1 was the largest by 10%. Next, the influence of the present analytical results for the ZPPR-9 sample reactivity to the cross-section adjustment was evaluated. The reference case was a unified cross-section set ADJ98 based on the recent JUPITER analysis. As a conclusion, the present analytical results have sufficient physical consistency with other JUPITER data, and possess qualification as a part of the standard data base for FBR nuclear design. (author)

  1. Coupling of 3-D core computational codes and a reactor simulation software for the computation of PWR reactivity accidents induced by thermal-hydraulic transients

    International Nuclear Information System (INIS)

    Raymond, P.; Caruge, D.; Paik, H.J.

    1994-01-01

    The French CEA has recently developed a set of new computer codes for reactor physics computations called the Saphir system which includes CRONOS-2, a three-dimensional neutronic code, FLICA-4, a three-dimensional core thermal hydraulic code, and FLICA-S, a primary loops thermal-hydraulic transient computation code, which are coupled and applied to analyze a severe reactivity accident induced by a thermal hydraulic transient: the Steamline Break accident for a pressurized water reactor until soluble boron begins to accumulate in the core. The coupling of these codes has proved to be numerically stable. 15 figs., 7 refs

  2. Conceptual core design study for Japan sodium-cooled fast reactor: Review of sodium void reactivity worth evaluation

    International Nuclear Information System (INIS)

    Ohki, Shigeo

    2012-01-01

    The conceptual core design study for a large-scale Japan sodium-cooled fast reactor (JSFR) have been carried out in the framework of the FaCT project. The reference “High-internal conversion” core can satisfy the requirements for enhanced safety, as well as achieving economic competitiveness. In order to increase the design reliability, more rigorous uncertainty evaluation is important. Development of the verification and validation methodology of the core neutronic design method is currently underway. (author)

  3. Magnetic nuclear core restraint and control

    International Nuclear Information System (INIS)

    Cooper, M.H.

    1979-01-01

    A lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

  4. Magnetic nuclear core restraint and control

    International Nuclear Information System (INIS)

    Cooper, M.H.

    1979-01-01

    A lateral restraint and control systemm for a nuclear reactor core provides an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit is composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased by an amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

  5. Magnetic nuclear core restraint and control

    International Nuclear Information System (INIS)

    Cooper, M.H.

    1978-01-01

    Disclosed is a lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

  6. Investigation of space-energy effects in the reactivity measurement by neutron noise with ex-core detectors in a reflected LWR

    International Nuclear Information System (INIS)

    Lescano, V.H.; Behringer, K.

    1981-11-01

    Practical application of the zero-crossing correlation method for measuring slightly subcritical reactivities in a swimming pool reactor required the use of detector locations in the reflector zone near to the core boundary. Experimental investigations of neutron-noise cross-power spectra showed significant deviations from the point reactor model at higher frequencies (> 100 Hz). Nevertheless, the use of the point reactor model was found to be an useful approach in the analysis of the zero-crossing correlation method yielding results which agreed well with those obtained from the rod-drop method. The theoretical part of the work is concerned with a space-dependent model calculation in two-group diffusion theory to support the experimental findings. The model calculation can explain the trends observed in the neutron-noise spectra as well as the applicability of the point reactor model to the zero-crossing correlation method. To obtain better insight, the calculations have been extended to neutron-noise spectra when one or both detectors are located in the core zone. In the case of a large core and widely spaced detectors, with at least one detector in the core zone, a sink frequency appears in the spectra. This effect is well-known in coupled-core kinetics. (Auth.)

  7. Detection of antibodies to hepatitis B core antigen using the Abbott ARCHITECT anti-HBc assay: analysis of borderline reactive sera.

    Science.gov (United States)

    Ollier, Laurence; Laffont, Catherine; Kechkekian, Aurore; Doglio, Alain; Giordanengo, Valérie

    2008-12-01

    Routine use of the automated chemiluminescent microparticle immunoassay Abbott ARCHITECT anti-HBc for diagnosis of hepatitis B is limited in case of borderline reactive sera with low signal close to the cut-off index. In order to determine the significance of anti-HBc detection when borderline reactivity occurs using the ARCHITECT anti-HBc assay, a comparative study was designed. 3540 serum samples collected over a 2-month period in the hospital of Nice were examined for markers of HBV infection (HBsAg, anti-HBs and anti-HBc). One hundred seven samples with sufficient volume and with borderline reactivity by the ARCHITECT assay were tested by two other anti-HBc assays, a microparticle enzyme immunoassay (MEIA, AxSYM Core, Abbott Laboratories, IL, USA) and an enzyme linked fluorescent assay (ELFA, VIDAS Anti-HBc Total II, bioMérieux, Lyon, France). Only 46 samples were confirmed by the AxSYM and the VIDAS assays. Additional serological information linked to patient history showed that the remaining samples (61) were false positives (11), had low titer of anti-HBc antibodies (13), or were inconclusive (37). This comparative study highlighted the existence of a grey zone around the cut-off index. Confirmative results through a different immunoassay are needed to confirm the diagnosis of HBV on borderline reactive sera using the ARCHITECT anti-HBc assay.

  8. Contribution to the qualification of calculation methods of reactivity and of flux and power distributions in nuclear pressurized water reactor cores

    International Nuclear Information System (INIS)

    Abit, K.

    1984-01-01

    The last stage of the creation computer methods and calculations consists of verifying the running and qualifying the results obtained. The work of the present thesis consisted of improving a coupling method between radial and axial phenomena in a PWR core, refering to three-dimensional calculations, while ensuring a perfect coherence between the programmed physical models. The calculation results have been compared to measurements of reactivity and of flux distributions realized during start-up tests. Thus, the methods have been applied to the calculation of the evolution of a burnable poison (gadolinium) in view of operation in long campaign. 13 refs [fr

  9. Determining neutron multiplication factor in the infinite system by reactivity dependence on one dimension of the reactor core

    International Nuclear Information System (INIS)

    Pesic, M.

    1975-01-01

    The objective of this task was to apply Fermi age theory for determining τ and neutron multiplication factor in infinite medium by measuring reactivity coefficient of heavy water in heterogeneous mixed reactor lattice. Basis of experiment is the measurement of stable reactor period. Measurement of heavy water reactivity coefficient by measuring the stable reactor period is described for chosen overcritical heavy water levels. Calculated values of infinite multiplication factor for measured neutron age data are presented and they are compared to expected theoretical values

  10. Tradeoff of sodium void worth and burnup reactivity swing: Impacts on balance safety position in metallic-fueled cores

    International Nuclear Information System (INIS)

    Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

    1994-01-01

    A study has been conducted to investigate the effect of a lower sodium void worth on the consequences of severe accidents in metallic-fueled sodium-cooled reactors. Four 900 MWth designs were used for the study, where all of the reactor cores were designed based on the metallic fuel of the Integral Fast Reactor (IFR) concept. The four core designs each have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation was to determine the differences in severe accident response for the four core designs, in order to estimate the improvement in overall safety that could be achieved from a reduction in the sodium void worth for reactor cores which use a metallic fuel form

  11. Parametric study of postulated reactivity transients due to ingress of heavy water from the reflector tank into the converted core of APSARA reactor

    International Nuclear Information System (INIS)

    Sankaranarayanan, S.

    2004-01-01

    Research reactors in the power range 5-10 MW with useable neutron flux values >1.OE+14 n/sqcm/sec can be constructed using LEU fuel with light water for neutron moderation and fuel cooling. In order to obtain a large irradiation volume, a heavy water reflector is used where fairly high neutron flux levels can be obtained. A prototype LEU fuelled 5/10 MW reactor design has been developed in the Bhabha Atomic Research Centre in Trombay. Work is on hand to carry out technology simulation of this reactor design by converting the pool type reactor APSARA in BARC. Presently the Apsara reactor uses MTh type high enriched U-Al alloy plate type fuel loaded in a 7x7 grid with a square lattice pitch of 76.8 mm. The reactor has three control-scram-shut off rods and one regulating control rod. In the first phase of the simulation studies, it is proposed to use the existing high enriched uranium fuel in a modified core with 37 positions arranged with a square lattice pitch of 84.8 mm, surrounded by a 50 cm thick heavy water reflector. Subsequently the converted core will use plate-type low enriched uranium suicide fuel. One of the accident scenarios postulated for the safety evaluation of the modified APSARA reactor is the reactivity transient due to the ingress of heavy water into the core through a small sized rupture in the aluminium wall of the reflector tank. Parametric analyses were done for the safety evaluation of modified Apsara reactor, for postulated leak of heavy water into the core from the reflector tank. A simplified computer code REDYN, based on point model reactor kinetics with one effective group of delayed neutrons is used for the analyses. Results of several parametric cases used in the study show that it is possible to contain the consequences of this type of reactivity transient within acceptable fuel and coolant thermal safety limits

  12. Inadvertent pump start with gas expansion modules

    International Nuclear Information System (INIS)

    Campbell, L.R.; Harris, R.A.; Heard, F.J.; Dautel, W.A.

    1992-01-01

    Previous testing demonstrated the effectiveness of gas expansion modules (GEMs) in mitigating the consequences of a loss-of-flow-without-scram transient in Fast Flux Test Facility (FFTF)-sized sodium cooled cores. As a result, GEMs have been included in the advance liquid-metal reactor (PRISM) design project sponsored by the US Department of Energy. The PRISM design is under review at the US Nuclear Regulatory Commission for licensability. In the unlikely event that the reactor does not scram during a loss of low, the GEMs quickly insert sufficient negative reactivity to limit fuel and cladding temperatures to acceptable values. This is the positive benefit of the GEMs; however, the reverse situation must be considered. A primary pump could be inadvertently started from near-critical conditions resulting in a positive reactivity insertion and a power transient. One mitigating aspect of this event is that as the reactivity associated with the GEMs is inserted, the increasing flow increases core cooling. A test was conducted in the FFTF to demonstrate that the GEM and feedback reactivity are well predicted following pump start, and the reactivity transient is benign

  13. An amino-terminal segment of hantavirus nucleocapsid protein presented on hepatitis B virus core particles induces a strong and highly cross-reactive antibody response in mice

    International Nuclear Information System (INIS)

    Geldmacher, Astrid; Skrastina, Dace; Petrovskis, Ivars; Borisova, Galina; Berriman, John A.; Roseman, Alan M.; Crowther, R. Anthony; Fischer, Jan; Musema, Shamil; Gelderblom, Hans R.; Lundkvist, Aake; Renhofa, Regina; Ose, Velta; Krueger, Detlev H.; Pumpens, Paul; Ulrich, Rainer

    2004-01-01

    Previously, we have demonstrated that hepatitis B virus (HBV) core particles tolerate the insertion of the amino-terminal 120 amino acids (aa) of the Puumala hantavirus nucleocapsid (N) protein. Here, we demonstrate that the insertion of 120 amino-terminal aa of N proteins from highly virulent Dobrava and Hantaan hantaviruses allows the formation of chimeric core particles. These particles expose the inserted foreign protein segments, at least in part, on their surface. Analysis by electron cryomicroscopy of chimeric particles harbouring the Puumala virus (PUUV) N segment revealed 90% T = 3 and 10% T = 4 shells. A map computed from T = 3 shells shows additional density splaying out from the tips of the spikes producing the effect of an extra shell of density at an outer radius compared with wild-type shells. The inserted Puumala virus N protein segment is flexibly linked to the core spikes and only partially icosahedrally ordered. Immunisation of mice of two different haplotypes (BALB/c and C57BL/6) with chimeric core particles induces a high-titered and highly cross-reactive N-specific antibody response in both mice strains

  14. Magnitude and reactivity consequences of accidental moisture ingress into the Modular High-Temperature Gas-Cooled Reactor core

    International Nuclear Information System (INIS)

    Smith, O.L.

    1992-01-01

    Accidental admission of moisture into the primary system of a Modular High-Temperature Gas-Cooled Reactor (MHTGR) has been identified in US Department of Energy-sponsored studies as an important safety concern. The work described here develops an analytical methodology to quantify the pressure and reactivity consequences of steam-generator tube rupture and other moistureingress-related incidents. Important neutronic and thermohydraulic processes are coupled with reactivity feedback and safety and control system responses. Rate and magnitude of steam buildup are found to be dominated by major system features such as break size in comparison with safety valve capacity and reliability, while being less sensitive to factors such as heat transfer coefficients. The results indicate that ingress transients progress at a slower pace than previously predicted by bounding analyses, with milder power overshoots and more time for operator or automatic corrective actions

  15. Magnitude and reactivity consequences of moisture ingress into the modular High-Temperature Gas-Cooled Reactor core

    International Nuclear Information System (INIS)

    Smith, O.L.

    1992-12-01

    Inadvertent admission of moisture into the primary system of a modular high-temperature gas-cooled reactor has been identified in US Department of Energy-sponsored studies as an important safety concern. The work described here develops an analytical methodology to quantify the pressure and reactivity consequences of steam-generator tube rupture and other moisture-ingress-related incidents. Important neutronic and thermohydraulic processes are coupled with reactivity feedback and safety and control system responses. The rate and magnitude of steam buildup are found to be dominated by major system features such as break size compared with safety valve capacity and reliability and less sensitive to factors such as heat transfer coefficients. The results indicate that ingress transients progress at a slower pace than previously predicted by bounding analyses, with milder power overshoots and more time for operator or automatic corrective actions

  16. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  17. Multi-site reactivity: reduction of six equivalents of nitrite to give an Fe6(NO)6 cluster with a dramatically expanded octahedral core.

    Science.gov (United States)

    Harris, T David; Betley, Theodore A

    2011-09-07

    Reaction of NO(2)(-) with the octahedral cluster ((H)L)(2)Fe(6) in the presence of a proton source affords the hexanitrosyl cluster ((H)L)(2)Fe(6)(NO)(6). This species forms via a proton-induced reduction of six nitrite molecules per cluster, utilizing each site available on the polynuclear core. Formation of the hexanitrosyl cluster is accompanied by a near 2-fold expansion of the ((H)L)(2)Fe(6) core volume, where intracore Fe-Fe interactions are overcome by strong π-bonding between Fe centers and NO ligands. A core volume of this magnitude is rare in octahedral metal clusters not supported by interstitial atoms. Moreover, the structural flexibility afforded by the ((H)L)(2)Fe(6) platform highlights the potential for other reaction chemistry involving species with metal-ligand multiple bonds. Carrying out the reaction of the cluster [((H)L)(2)Fe(6)(NCMe)(6)](4+) with nitrite in the absence of a proton source serves to forestall the nitrite reduction and enables clean isolation of the intermediate hexanitro cluster [((H)L)(2)Fe(6)(NO(2))(6)](2-).

  18. Neutron flux distribution inside the cylindrical core of minor excess of reactivity in the IPEN/MB-01 reactor and comparison with citation code and MCNP- 5 code

    Energy Technology Data Exchange (ETDEWEB)

    Aredes, Vitor Ottoni; Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto C.; Santos, Diogo Feliciano dos; Lima, Ana Cecilia de Souza, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    This study aims to determine the distribution of thermal neutron flux in the IPEN/MB-01 nuclear reactor core assembled with cylindrical core configuration of minor excess of reactivity with 568 fuel rods (28 fuel rods in diameter). The thermal neutron flux at the positions of irradiation derive from the method of reaction rate using gold foils. The experiment consists in inserting gold activations foils with and without cadmium coverage (cadmium boxes with 0.0502 cm thickness) in several positions throughout the active core. After irradiation, activity induced by nuclear reaction rates over gold foils is assessed by gamma ray spectrometry using a high-purity germanium (HPGe) detector. Experimental results are compared to those derived from calculations performed using a three dimensional CITATION diffusion code and MCNP-5 code and a proper nuclear data library. While calculated neutron flux data shows good agreement with experimental values in regions with little disturbance in the neutron flux, also showing that in the region of the reflectors of neutrons and near the control rods, the diffusion theory is not very precise. The average value of thermal neutron flux obtained experimentally compared to the calculated value by CITATION code and MCNP-5 code respectively show a difference of 1.18% and 0.84% at a nuclear power level of 74.65 ± 3.28 % watts. The average measured value of thermal neutron flux is 4.10 10{sup 8} ± 5.25% n/cm{sup 2}s. (author)

  19. Advanced computational methods for the assessment of reactor core behaviour during reactivity initiated accidents. Final report; Fortschrittliche Rechenmethoden zum Kernverhalten bei Reaktivitaetsstoerfaellen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Pautz, A.; Perin, Y.; Pasichnyk, I.; Velkov, K.; Zwermann, W.; Seubert, A.; Klein, M.; Gallner, L.; Krzycacz-Hausmann, B.

    2012-05-15

    The document at hand serves as the final report for the reactor safety research project RS1183 ''Advanced Computational Methods for the Assessment of Reactor Core Behavior During Reactivity-Initiated Accidents''. The work performed in the framework of this project was dedicated to the development, validation and application of advanced computational methods for the simulation of transients and accidents of nuclear installations. These simulation tools describe in particular the behavior of the reactor core (with respect to neutronics, thermal-hydraulics and thermal mechanics) at a very high level of detail. The overall goal of this project was the deployment of a modern nuclear computational chain which provides, besides advanced 3D tools for coupled neutronics/ thermal-hydraulics full core calculations, also appropriate tools for the generation of multi-group cross sections and Monte Carlo models for the verification of the individual calculational steps. This computational chain shall primarily be deployed for light water reactors (LWR), but should beyond that also be applicable for innovative reactor concepts. Thus, validation on computational benchmarks and critical experiments was of paramount importance. Finally, appropriate methods for uncertainty and sensitivity analysis were to be integrated into the computational framework, in order to assess and quantify the uncertainties due to insufficient knowledge of data, as well as due to methodological aspects.

  20. Adsorbate reactivity and thermal mobility from simple modeling of high-resolution core-level spectra: application to O/Al(111)

    International Nuclear Information System (INIS)

    Schouborg, Jakob; Raarup, Merete K; Balling, Peter

    2009-01-01

    A high-resolution core-level spectroscopy investigation of the adsorption of oxygen on Al(111) at variable oxygen exposure demonstrates a low surface reactivity for an intensively cleaned surface. The threshold for oxide formation is as high as ∼200 L (langmuirs), at which point the coverage of the chemisorbed oxygen exceeds half a monolayer. A simple model is presented, using which it is possible to deduce the oxygen coverage from the core-level spectra and determine the initial sticking probability. For our data a value of 0.018 ± 0.004 is obtained. The changes in core-level spectra following low-temperature annealing of low-coverage O/Al(111) reflect the formation of gradually larger islands of oxygen atoms (Ostwald ripening). The island formation is consistent with a random-walk model from which the diffusion barrier can be deduced to be in the range of 0.80-0.90 eV.

  1. Neutron flux distribution inside the cylindrical core of minor excess of reactivity in the IPEN/MB-01 reactor and comparison with citation code and MCNP- 5 code

    International Nuclear Information System (INIS)

    Aredes, Vitor Ottoni; Bitelli, Ulysses d'Utra; Mura, Luiz Ernesto C.; Santos, Diogo Feliciano dos; Lima, Ana Cecilia de Souza

    2015-01-01

    This study aims to determine the distribution of thermal neutron flux in the IPEN/MB-01 nuclear reactor core assembled with cylindrical core configuration of minor excess of reactivity with 568 fuel rods (28 fuel rods in diameter). The thermal neutron flux at the positions of irradiation derive from the method of reaction rate using gold foils. The experiment consists in inserting gold activations foils with and without cadmium coverage (cadmium boxes with 0.0502 cm thickness) in several positions throughout the active core. After irradiation, activity induced by nuclear reaction rates over gold foils is assessed by gamma ray spectrometry using a high-purity germanium (HPGe) detector. Experimental results are compared to those derived from calculations performed using a three dimensional CITATION diffusion code and MCNP-5 code and a proper nuclear data library. While calculated neutron flux data shows good agreement with experimental values in regions with little disturbance in the neutron flux, also showing that in the region of the reflectors of neutrons and near the control rods, the diffusion theory is not very precise. The average value of thermal neutron flux obtained experimentally compared to the calculated value by CITATION code and MCNP-5 code respectively show a difference of 1.18% and 0.84% at a nuclear power level of 74.65 ± 3.28 % watts. The average measured value of thermal neutron flux is 4.10 10 8 ± 5.25% n/cm 2 s. (author)

  2. The determination of reactivity at a compact core with low absorption reflector by means of the pulsed neutron method

    International Nuclear Information System (INIS)

    Hecker, R.; Drueke, V.; Ahmad, S.K.; Kacmary, S.; Paul, N.

    1976-09-01

    Pulsed measurements at a compact critical facility with low absorption reflector are described. By a number of such measurements at different points of the system a proper value for the reactivity can be determined, which allows a meaningful comparison with values calculated by static computer codes in spite of the presence of kinetic distortion in the system. A modified area ratio method is developed and applied to the measurements; furthermore a method is shown, which allows to suppress in this systems the kinetic distortion in the fundamental mode by application of special detectors. (orig.) [de

  3. Study of diluting and absorber materials to control the reactivity during a postulated core meltdown accident in generation IV reactors

    International Nuclear Information System (INIS)

    Plevacova, Kamila

    2010-01-01

    In order to limit the consequences of a hypothetical core meltdown accident in Generation IV Sodium Fast Reactors, absorber materials in or near the core, such as boron carbide B 4 C, and diluting materials in the core catcher will be used to prevent recriticality within the mixture of molten oxide fuel and molten structures called corium. The aim of the PhD thesis was to select materials of both types and to understand their behaviour during their interaction with corium, from chemical and thermodynamic points of view. Concerning B 4 C, thermodynamic calculations and experiments agree with the formation of two immiscible phases at high temperature in the B 4 C - UO 2 system: one oxide and one boride. This separation of phases can reduce the efficiency of the neutrons absorption inside the molten fuel contained in the oxide phase. Moreover, volatilization of a part of the boron element can occur. According to these results, the necessary quantity of B 4 C to be introduced should be reconsidered for postulated severe accident sequence. Other solution could be the use of Eu 2 O 3 or HfO 2 as absorber material. These oxides form a solid solution with the oxide fuel. Concerning the diluting materials, mixed oxides Al 2 O 3 - HfO 2 and Al 2 O 3 - Eu 2 O 3 were preselected. These systems being completely unknown to date at high temperature in association with UO 2 , first points on the corresponding ternary phase diagrams were researched. Contrary to Al 2 O 3 - Eu 2 O 3 - UO 2 system, the Al 2 O 3 - HfO 2 - UO 2 mixture presents only one eutectic and thus only one solidification path which makes easier forecasting the behaviour of corium in the core catcher. (author)

  4. Study of diluting and absorber materials to control reactivity during a postulated core melt down accident in Generation IV reactors

    International Nuclear Information System (INIS)

    Plevacova, K.

    2010-01-01

    In order to limit the consequences of a hypothetical core meltdown accident in Generation IV Sodium Fast Reactors, absorber materials in or near the core, such as boron carbide B 4 C, and diluting materials in the core catcher will be used to prevent recriticality within the mixture of molten oxide fuel and molten structures called corium. The aim of the PhD thesis was to select materials of both types and to understand their behaviour during their interaction with corium, from chemical and thermodynamic point of view. Concerning B 4 C, thermodynamic calculations and experiments agree with the formation of two immiscible phases at high temperature in the B 4 C - UO 2 system: one oxide and one boride. This separation of phases can reduce the efficiency of the neutrons absorption inside the molten fuel contained in the oxide phase. Moreover, a volatilization of a part of the boron element can occur. According to these results, the necessary quantity of B 4 C to be introduced should be reconsidered for postulated severe accident sequence. Other solution could be the use of Eu 2 O 3 or HfO 2 as absorber material. These oxides form a solid solution with the oxide fuel. Concerning the diluting materials, mixed oxides Al 2 O 3 - HfO 2 and Al 2 O 3 - Eu 2 O 3 were preselected. These systems being completely unknown to date at high temperature in association with UO 2 , first points on the corresponding ternary phase diagrams were researched. Contrary to Al 2 O 3 - Eu 2 O 3 - UO 2 system, the Al 2 O 3 - HfO 2 - UO 2 mixture presents only one eutectic and thus only one solidification path which makes easier forecasting the behaviour of corium in the core catcher. (author) [fr

  5. Core Flooding Experiments and Reactive Transport Modeling of Seasonal Heat Storage in the Hot Deep Gassum Sandstone Formation

    DEFF Research Database (Denmark)

    Holmslykke, Hanne D.; Kjøller, Claus; Fabricius, Ida Lykke

    2017-01-01

    Seasonal storage of excess heat in hot deep aquifers is considered to optimize the usage of commonly available energy sources. The chemical effects of heating the Gassum Sandstone Formation to up to 150 degrees C is investigated by combining laboratory core flooding experiments with petrographic...... analysis and geochemical modeling. Synthetic formation water is injected into two sets of Gassum Formation samples at 25, 50 (reservoir temperature), 100, and 150 degrees C with a velocity of 0.05 and 0.1 PV/h, respectively. Results show a significant increase in the aqueous concentration of silicium...

  6. Core Flooding Experiments and Reactive Transport Modeling of Seasonal Heat Storage in the Hot Deep Gassum Sandstone Formation

    DEFF Research Database (Denmark)

    Holmslykke, Hanne D.; Kjøller, Claus; Fabricius, Ida Lykke

    2017-01-01

    Seasonal storage of excess heat in hot deep aquifers is considered to optimize the usage of commonly available energy sources. The chemical effects of heating the Gassum Sandstone Formation to up to 150 degrees C is investigated by combining laboratory core flooding experiments with petrographic ...... minor effects on the properties of the reservoir and that storage of excess heat in the Gassum Formation in the Stenlille area may be possible provided operational precautions are taken....... analysis and geochemical modeling. Synthetic formation water is injected into two sets of Gassum Formation samples at 25, 50 (reservoir temperature), 100, and 150 degrees C with a velocity of 0.05 and 0.1 PV/h, respectively. Results show a significant increase in the aqueous concentration of silicium...... and iron with increasing temperature due to dissolution of silica and siderite. Increasing the reservoir temperature from 50 to 100 degrees C enhanced the naturally occurring weathering of Na-rich feldspar to kaolinite. Dissolution of quartz increased sharply above 100 degrees C and was the dominating...

  7. A technique for computing bowing reactivity feedback in LMFBR's

    International Nuclear Information System (INIS)

    Finck, P.J.

    1987-01-01

    During normal or accidental transients occurring in a LMFBR core, the assemblies and their support structure are subjected to important thermal gradients which induce differential thermal expansions of the walls of the hexcans and differential displacement of the assembly support structure. These displacements, combined with the creep and swelling of structural materials, remain quite small, but the resulting reactivity changes constitute a significant component of the reactivity feedback coefficients used in safety analyses. It would be prohibitive to compute the reactivity changes due to all transients. Thus, the usual practice is to generate reactivity gradient tables. The purpose of the work presented here is twofold: develop and validate an efficient and accurate scheme for computing these reactivity tables; and to qualify this scheme

  8. Short-term vitamin E treatment impairs reactive oxygen species signaling required for adipose tissue expansion, resulting in fatty liver and insulin resistance in obese mice.

    Directory of Open Access Journals (Sweden)

    Martin Alcala

    Full Text Available The use of antioxidant therapy in the treatment of oxidative stress-related diseases such as cardiovascular disease, diabetes or obesity remains controversial. Our aim is to demonstrate that antioxidant supplementation may promote negative effects if used before the establishment of oxidative stress due to a reduced ROS generation under physiological levels, in a mice model of obesity.C57BL/6J mice were fed with a high-fat diet for 14 weeks, with (OE group or without (O group vitamin E supplementation.O mice developed a mild degree of obesity, which was not enough to induce metabolic alterations or oxidative stress. These animals exhibited a healthy expansion of retroperitoneal white adipose tissue (rpWAT and the liver showed no signs of lipotoxicity. Interestingly, despite achieving a similar body weight, OE mice were insulin resistant. In the rpWAT they presented a reduced generation of ROS, even below physiological levels (C: 1651.0 ± 212.0; O: 3113 ± 284.7; OE: 917.6 ±104.4 RFU/mg protein. C vs OE p< 0.01. ROS decay may impair their action as second messengers, which could account for the reduced adipocyte differentiation, lipid transport and adipogenesis compared to the O group. Together, these processes limited the expansion of this fat pad and as a consequence, lipid flux shifted towards the liver, causing steatosis and hepatomegaly, which may contribute to the marked insulin resistance.This study provides in vivo evidence for the role of ROS as second messengers in adipogenesis, lipid metabolism and insulin signaling. Reducing ROS generation below physiological levels when the oxidative process has not yet been established may be the cause of the controversial results obtained by antioxidant therapy.

  9. Global Structural Flexibility of Metalloproteins Regulates Reactivity of Transition Metal Ion in the Protein Core: An Experimental Study Using Thiol-subtilisin as a Model Protein.

    Science.gov (United States)

    Matsuo, Takashi; Kono, Takamasa; Shobu, Isamu; Ishida, Masaya; Gonda, Katsuya; Hirota, Shun

    2018-02-21

    The functions of metal-containing proteins (metalloproteins) are determined by the reactivities of transition metal ions at their active sites. Because protein macromolecular structures have several molecular degrees of freedom, global structural flexibility may also regulate the properties of metalloproteins. However, the influence of this factor has not been fully delineated in mechanistic studies of metalloproteins. Accordingly, we have investigated the relationship between global protein flexibility and the characteristics of a transition metal ion in the protein core using thiol-subtilisin (tSTL) with a Cys-coordinated Cu 2+ ion as a model system. Although tSTL has two Ca 2+ -binding sites, the Ca 2+ -binding status hardly affects its secondary structure. Nevertheless, guanidinium-induced denaturation and amide H/D exchange indicated the increase in the structural flexibility of tSTL by the removal of bound Ca 2+ ions. Electron paramagnetic resonance and absorption spectral changes have revealed that the protein flexibility determines the characteristics of a Cu 2+ ion in tSTL. Therefore, global protein flexibility should be recognized as an important factor that regulates the properties of metalloproteins. © 2018 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  10. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  11. Neutronic characterization of cylindrical core of minor excess reactivity in the nuclear reactor IPEN/MB-01 from the measure of spatial and energetic distribution of neutron flux distribution

    International Nuclear Information System (INIS)

    Aredes, Vitor Ottoni Garcia

    2014-01-01

    In this work was conducted the mapping of the thermal and epithermal neutrons flux and the energy spectrum of the neutrons in the reactor core IPEN/MB-01 for a cylindrical core configuration with minor excess reactivity, which is 28 x 28 fuel rods arranged in north-south and east-west directions. The calibration of control rods for this configuration determined their excess reactivity. The lower excess reactivity in the core decreased neutron flux disturbance caused by the neutron absorbing rods , given that the nuclear reactor was operated with the rods almost completely removed . Was used the 'Activation Analysis Technique' with the thin foil activation detectors ( infinitely diluted and hyper-pure), of different materials that work in different energy ranges, to calculate the saturation activity, used for determining the neutron flux and in the SANDBP code as input for the calculation of the neutrons energy spectrum. To discriminate thermal and epithermal flux , was used the 'Cadmium RatioTechnique' . The activation detectors were distributed in a total of 140 radial and axial positions in the reactor core and 16 irradiation, with bare and covered with cadmium activation foils. A model of this configuration was simulated by MCNP-5 code to determine the cadmium correction factor and comparison of the results obtained experimentally. The cylindrical configuration desired, with 17% less fuel than the standard rectangular configuration (28 x 26 fuel rods), reached criticality with the control rods approximately 90% removed, which decreased considerably the disturbance in neutron flux. Given the highest power density of the 28 x 28 cylindrical core, the neutron flux increased by over 50% in the central regions of the core compared to the values of the 28 x 26 standard rectangular core. (author)

  12. Thermal Expansivity Between 150 and 800°C of Hydrothermally Altered Conduit Dyke Samples from USDP-4 Drill Core (Mt Unzen, Shimabara, Japan)

    Science.gov (United States)

    Yilmaz, T. I.; Hess, K. U.; Vasseur, J.; Wadsworth, F. B.; Gilg, H. A.; Nakada, S.; Dingwell, D. B.

    2017-12-01

    When hot magma intrudes the crust, the surrounding rocks expand. Similarly, the cooling magma contracts. The expansion and contraction of these multiphase materials is not simple and often requires empirical constraint. Therefore, we constrained the thermal expansivity of Unzen dome and conduit samples using a NETZSCH® DIL 402C. Following experiments, those samples were scanned using a Phoenix v|tome|x m to observe the cracks that may have developed during the heating and cooling. The dome samples do not show petrological or chemical signs of alteration. However, the alteration of the conduit dykes is represented by the occurrence of the main secondary phases such as chlorite, sulfides, carbonates, R1 (Reichweite parameter) illite-smectite, and kaolinite. These alteration products indicate an (I) early weak to moderate argillic magmatic alteration, and a (II) second stage weak to moderate propylitic hydrothermal alteration. The linear thermal expansion coefficient aL of the dome material is K-1 between 150° and 800°C and shows a sharp peak of up to K-1 around the alpha-beta-quartz-transition ( 573°C). In contrast, aL of the hydrothermally altered conduit samples starts to increase around 180° and reaches K-1 at 400°C. We interpret this effect as being due to the water content of the kaolinite and the R1 illite-smectite, which induces larger expansions per degree temperature change. Furthermore, the altered conduit samples show a more pronounced increases of aL between 500 and 650°C of up to peaks at K-1, which is generated by the breakdown of chlorite, iron-rich dolomite solid solutions, calcite, and pyrite. We use a 1D conductive model of heat transfer to explore how the country rock around the Unzen conduit zone would heat up after intrusion. In turn, we convert these temperature profiles to thermal stress profiles, assuming the edifice is largely undeformable. We show that these high linear thermal expansion coefficients of the hydrothermally altered

  13. Influence of external source location in the reactivity calculation

    International Nuclear Information System (INIS)

    Silva, Adilson Costa da; Silva, Fernando Carvalho da; Martinez, Aquilino Senra

    2011-01-01

    We used the neutron diffusion equation with external neutron sources, in cartesian geometry and the two groups of energy, to verify the influence of external neutron source locations in the reactivity calculation. For this, a coarse mesh finite difference method was developed for the adjoint flux calculation and simplifies reactivity calculation in PWR type reactor, which uses the output of the nodal expansion method. The results were obtained for different locations on the two-dimensional plane, as well as for different types of fuel elements in the reactor core. (author)

  14. Influence of external source location in the reactivity calculation

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Adilson Costa da; Silva, Fernando Carvalho da; Martinez, Aquilino Senra, E-mail: asilva@con.ufrj.b, E-mail: fernando@con.ufrj.b, E-mail: Aquilino@lmp.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2011-07-01

    We used the neutron diffusion equation with external neutron sources, in cartesian geometry and the two groups of energy, to verify the influence of external neutron source locations in the reactivity calculation. For this, a coarse mesh finite difference method was developed for the adjoint flux calculation and simplifies reactivity calculation in PWR type reactor, which uses the output of the nodal expansion method. The results were obtained for different locations on the two-dimensional plane, as well as for different types of fuel elements in the reactor core. (author)

  15. Comparison of the behaviour of two core designs for ASTRID in case of severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Bertrand, F., E-mail: frederic.bertrand@cea.fr [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Marie, N.; Prulhière, G.; Lecerf, J. [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Seiler, J.M. [CEA, DEN, DTN, F-38054 Grenoble (France)

    2016-02-15

    Highlights: • Low void worth CFV and SFRv2 cores are compared for ASTRID pre-conceptual design. • Severe accident behaviour is assessed with a simplified calculation approach and tools. • Mitigation to limit reactivity inserted by core compaction is easier for CFV than for SFRv2 core. • When facing arbitrary reactivity ramps, CFV core would lead to lower energy release than SFRv2 core. • Time scale for core degradation is one order of magnitude larger for CFV than for SFRv2. - Abstract: The present paper is dedicated to the studies carried out during the first stage of the pre-conceptual design of the French demonstrator of fourth generation SFR reactors (ASTRID) in order to compare the behaviour of two envisaged core concepts under severe accident transients. Among the two studied core concepts, whose powers are 1500 MWth, the first one is a classical homogeneous core (called SFRv2) with large pin diameter whose the sodium overall voiding reactivity effect is 5 $. The second concept is an axially heterogeneous core (called CFV) whose global void reactivity effect is negative (−1.2 $ at the end of cycle at the equilibrium). The comparison of the cores relies on two typical accident families: a reactivity insertion (unprotected transient overpower, UTOP) and an overall loss of core cooling (unprotected loss of flow, ULOF). In the first part of the comparison, the primary phase of an UTOP is studied in order to assess typical features of the transient behaviour: power and reactivity evolutions, material heating and melting/vaporization and mechanical energy release due to fuel vapor expansion. The second part of the comparison deals with the calculation of the reactivity potential for degraded states (molten pools) representative of the secondary phase of a mild UTOP and of a strong UTOP (strong or mild qualifies the reactivity ramp inserted). According to the reactivity potential, the amount of fuel to extract from the core and the amount of absorber

  16. BN-600 MOX Core Benchmark Analysis. Results from Phases 4 and 6 of a Coordinated Research Project on Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects

    International Nuclear Information System (INIS)

    2013-12-01

    For those Member States that have or have had significant fast reactor development programmes, it is of utmost importance that they have validated up to date codes and methods for fast reactor physics analysis in support of R and D and core design activities in the area of actinide utilization and incineration. In particular, some Member States have recently focused on fast reactor systems for minor actinide transmutation and on cores optimized for consuming rather than breeding plutonium; the physics of the breeder reactor cycle having already been widely investigated. Plutonium burning systems may have an important role in managing plutonium stocks until the time when major programmes of self-sufficient fast breeder reactors are established. For assessing the safety of these systems, it is important to determine the prediction accuracy of transient simulations and their associated reactivity coefficients. In response to Member States' expressed interest, the IAEA sponsored a coordinated research project (CRP) on Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects. The CRP started in November 1999 and, at the first meeting, the members of the CRP endorsed a benchmark on the BN-600 hybrid core for consideration in its first studies. Benchmark analyses of the BN-600 hybrid core were performed during the first three phases of the CRP, investigating different nuclear data and levels of approximation in the calculation of safety related reactivity effects and their influence on uncertainties in transient analysis prediction. In an additional phase of the benchmark studies, experimental data were used for the verification and validation of nuclear data libraries and methods in support of the previous three phases. The results of phases 1, 2, 3 and 5 of the CRP are reported in IAEA-TECDOC-1623, BN-600 Hybrid Core Benchmark Analyses, Results from a Coordinated Research Project on Updated Codes and Methods to Reduce the

  17. Method of controlling reactivity

    International Nuclear Information System (INIS)

    Tochihara, Hiroshi.

    1982-01-01

    Purpose: To improve the reactivity controlling characteristics by artificially controlling the leakage of neutron from a reactor and providing a controller for controlling the reactivity. Method: A reactor core is divided into several water gaps to increase the leakage of neutron, its reactivity is reduced, a gas-filled control rod or a fuel assembly is inserted into the gap as required, the entire core is coupled in a system to reduce the leakage of the neutron, and the reactivity is increased. The reactor shutdown is conducted by the conventional control rod, and to maintain critical state, boron density varying system is used together. Futher, a control rod drive is used with that similar to the conventional one, thereby enabling fast reactivity variation, and the positive reactivity can be obtained by the insertion, thereby improving the reactivity controlling characteristics. (Yoshihara, H.)

  18. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  19. Thermal expansion

    International Nuclear Information System (INIS)

    Yun, Y.

    2015-01-01

    Thermal expansion of fuel pellet is an important property which limits the lifetime of the fuels in reactors, because it affects both the pellet and cladding mechanical interaction and the gap conductivity. By fitting a number of available measured data, recommended equations have been presented and successfully used to estimate thermal expansion coefficient of the nuclear fuel pellet. However, due to large scatter of the measured data, non-consensus data have been omitted in formulating the equations. Also, the equation is strongly governed by the lack of appropriate experimental data. For those reasons, it is important to develop theoretical methodologies to better describe thermal expansion behaviour of nuclear fuel. In particular, first-principles and molecular dynamics simulations have been certainly contributed to predict reliable thermal expansion without fitting the measured data. Furthermore, the two theoretical techniques have improved on understanding the change of fuel dimension by describing the atomic-scale processes associated with lattice expansion in the fuels. (author)

  20. Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

    International Nuclear Information System (INIS)

    Hong, S. G.; Kim, Y. H.; Venneri, F.

    2010-01-01

    In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

  1. Structure, Reactivity and Dynamics

    Indian Academy of Sciences (India)

    Understanding structure, reactivity and dynamics is the core issue in chemical ... functional theory (DFT) calculations, molecular dynamics (MD) simulations, light- ... between water and protein oxygen atoms, the superionic conductors which ...

  2. Effects of conversion ratio change on the core performances in medium to large TRU burning reactors

    International Nuclear Information System (INIS)

    Song, Hoon; Kim, Sang-Ji; Yoo, Jae-Woon; Kim, Yeong-Il

    2009-01-01

    Conceptual fast reactor core designs with sodium coolant are developed at 1,500, 3,000 and 4,500 MWt which are configured to transmute recycled transuranics (TRU) elements with external feeds consisting of LWR spent fuel. Even at each pre-determined power level, the performance parameters, reactivity coefficients and their implications on the safety analysis can be different when the target TRU conversion ratio changes. In order to address this aspect of design, a study on TRU conversion ratio change was performed. The results indicate that it is feasible to design a TRU burner core to accommodate a wide range of conversion ratios by employing different fuel cladding thicknesses. The TRU consumption rate is found to be proportional to the core power without any significant deterioration in the core performance at higher power levels. A low conversion ratio core has an increased TRU consumption rate and much faster burnup reactivity loss, which calls for appropriate means for reactivity compensation. As for the reactivity coefficients related with the conversion ratio change, the core with a low conversion ratio has a less negative Doppler coefficient, a more negative axial expansion coefficient, a more negative control rod worth per rod, a more negative radial expansion coefficient, a less positive sodium density coefficient and a less positive sodium void worth. A slight decrease in the delayed neutron fraction is also noted, reflecting the fertile U-238 fraction reduction. (author)

  3. Biosphere-atmosphere exchange of reactive nitrogen and greenhouse gases at the NitroEurope core flux measurement sites: Measurement strategy and first data sets

    DEFF Research Database (Denmark)

    Skiba, U.; Drewer, J.; Tang, Y.S.

    2009-01-01

    The NitroEurope project aims to improve understanding of the nitrogen (N) cycle at the continental scale and quantify the major fluxes of reactive N by a combination of reactive N measurements and modelling activities. As part of the overall measurement strategy, a network of 13 flux ‘super sites...

  4. Fast Flux Test Facility (FFTF) feedback reactivity components

    International Nuclear Information System (INIS)

    Nguyen, D.H.

    1988-04-01

    The static tests conducted during Cycle 8A (1986) of the FFTF have allowed, for the first time, the experimental determination of each of the feedback reactivities caused by the following mechanisms: fuel axial expansion, control rod repositioning, core radial expansion, and subassembly bowing. A semiempirical equation was obtained to describe each of these feedback components that depended only on the relevant reactor temperature (bowing was presented in a tabular form). The Doppler and sodium density reactivities were calculated using existing mechanistic methods. Although they could also be fitted with closed-form equations depending only on temperatures, these equations are not needed in transient analyses using whole core safety computer codes, which use mechanistic methods. The static feedback reactivity model was extended to obtain a dynamic model via the concept of ''time constants.'' Besides being used for transient analyses in the FFTF, these feedback equations constitute a database for the validation and/or calibration of mechanistic feedback reactivity models. 2 refs., 6 tabs

  5. TRACE analysis of Phenix core response to an increase of the core inlet sodium temperature

    Energy Technology Data Exchange (ETDEWEB)

    Chenu, A., E-mail: aurelia.chenu@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Ecole Polytechnique Federale (Switzerland); Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Adams, R., E-mail: robert.adams@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Eidgenossische Technische Hochschule, Zurich (Switzerland); Chawla, R., E-mail: rakesh.chawla@epfl.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Ecole Polytechnique Federale (Switzerland)

    2011-07-01

    This work presents the analysis, using the TRACE code, of the Phenix core response to an inlet sodium temperature increase. The considered experiment was performed in the frame of the Phenix End-Of-Life (EOL) test program of the CEA, prior to the final shutdown of the reactor. It corresponds to a transient following a 40°C increase of the core inlet temperature, which leads to a power decrease of 60%. This work focuses on the first phase of the transient, prior to the reactor scram and pump trip. First, the thermal-hydraulic TRACE model of the core developed for the present analysis is described. The kinetic parameters and feedback coefficients for the point kinetic model were first derived from a 3D static neutronic ERANOS model developed in a former study. The calculated kinetic parameters were then optimized, before use, on the basis of the experimental reactivity in order to minimize the error on the power calculation. The different reactivity feedbacks taken into account include various expansion mechanisms that have been specifically implemented in TRACE for analysis of fast-neutron spectrum systems. The point kinetic model has been used to study the sensitivity of the core response to the different feedback effects. The comparison of the calculated results with the experimental data reveals the need to accurately calculate the reactivity feedback coefficients. This is because the reactor response is very sensitive to small reactivity changes. This study has enabled us to study the sensitivity of the power change to the different reactivity feedbacks and define the most important parameters. As such, it furthers the validation of the FAST code system, which is being used to gain a more in-depth understanding of SFR core behavior during accidental transients. (author)

  6. Virial Expansion Bounds

    Science.gov (United States)

    Tate, Stephen James

    2013-10-01

    In the 1960s, the technique of using cluster expansion bounds in order to achieve bounds on the virial expansion was developed by Lebowitz and Penrose (J. Math. Phys. 5:841, 1964) and Ruelle (Statistical Mechanics: Rigorous Results. Benjamin, Elmsford, 1969). This technique is generalised to more recent cluster expansion bounds by Poghosyan and Ueltschi (J. Math. Phys. 50:053509, 2009), which are related to the work of Procacci (J. Stat. Phys. 129:171, 2007) and the tree-graph identity, detailed by Brydges (Phénomènes Critiques, Systèmes Aléatoires, Théories de Jauge. Les Houches 1984, pp. 129-183, 1986). The bounds achieved by Lebowitz and Penrose can also be sharpened by doing the actual optimisation and achieving expressions in terms of the Lambert W-function. The different bound from the cluster expansion shows some improvements for bounds on the convergence of the virial expansion in the case of positive potentials, which are allowed to have a hard core.

  7. Expansion dynamics

    International Nuclear Information System (INIS)

    Knoll, J.

    1985-10-01

    A quantum dynamical model is suggested which describes the expansion and disassembly phase of highly excited compounds formed in energetic heavy-ion collisions. First applications in two space and one time dimensional model world are discussed and qualitatively compared to standard freeze-out concepts. (orig.)

  8. expansion method

    Indian Academy of Sciences (India)

    of a system under investigation is to model the system in terms of some ... The organization of the paper is as follows: In §2, a brief account of the (G /G)- expansion ...... It is interesting to note that from the general results, one can easily recover.

  9. Development of a perturbation code, PERT-K, for hexagonal core geometry

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taek Kyum; Kim, Sang Ji; Song, Hoon; Kim, Young Il; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    A perturbation code for hexagonal core geometry has been developed based on Nodal Expansion Method. By using relevant output files of DIF3D code, it can calculate the reactivity changes caused by perturbation in composition or/and neutron cross section libraries. The accuracy of PERT-K code has been validated by calculating the reactivity changes due to fuel composition change, the sodium void coefficients, and the sample reactivity worths of BFS-73-1 critical experiments. In the case of 10% reduction in all fuel isotopics at a assembly located in the outer core, PERT-K computation agrees with the direct computation by DIF3D within 60 pcm. The sample reactivity worths of BFS-73-1 critical experiments are predicted with PERT-K code within the experimental error bounds. For 100% sodium void occurrence at the inner core, the maximum difference of reactivity changes between PERT-K and direct DIF3D computations is less than 40 pcm. On the other hand, the same sodium void condition at the outer core leads to a difference of reactivity change greater than 400 pcm. However, as sodium voiding becomes near zero value, the difference becomes less and rapidly falls within the acceptable bound, i.e. 40 pcm. (author). 11 refs., 9 figs., 6 tabs.

  10. Assessment of CRBR core disruptive accident energetics

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Bell, C.R.

    1984-03-01

    The results of an independent assessment of core disruptive accident energetics for the Clinch River Breeder Reactor are presented in this document. This assessment was performed for the Nuclear Regulatory Commission under the direction of the CRBR Program Office within the Office of Nuclear Reactor Regulation. It considered in detail the accident behavior for three accident initiators that are representative of three different classes of events; unprotected loss of flow, unprotected reactivity insertion, and protected loss of heat sink. The primary system's energetics accommodation capability was realistically, yet conservatively, determined in terms of core events. This accommodation capability was found to be equivalent to an isentropic work potential for expansion to one atmosphere of 2550 MJ or a ramp rate of about 200 $/s applied to a classical two-phase disassembly

  11. Prevalence and chemotherapy-induced reactivation of occult hepatitis B virus among hepatitis B surface antigen negative patients with diffuse large B-cell lymphoma: Significance of hepatitis B core antibodies screening

    International Nuclear Information System (INIS)

    Elbedewy, T.A.; Elashtokhy, H.A.; Rabee, E.S.; Kheder, G.E.

    2015-01-01

    Background: Occult hepatitis B infection (OBI) is characterized by negative hepatitis B surface antigen (HBsAg) and detectable hepatitis B virus (HBV)-DNA in the liver and/or serum, with or without hepatitis B core antibody (anti-HBc). Anti-HBc is the most sensitive marker of previous HBV. HBV reactivation in patients under immunosuppressive treatment is life-threatening, occurring in both overt and occult HBV especially in hematological malignancies. Aim of the work: To evaluate the prevalence and chemotherapy-induced reactivation of OBI among hepatitis B surface antigen negative patients with diffuse large B-cell lymphoma (DLBCL) patients and to determine the significance of anti-HBc screening among this group of patients before receiving chemotherapy. Patients and methods: This cross-sectional study included 72 DLBCL patients negative for HBsAg, HBsAb and hepatitis C virus antibodies (anti-HCV). Patients were subjected to investigations including anti-HBc. All patients underwent alanine transaminase (ALT) monitoring before each cycle of chemotherapy and monthly for 12 months after the end of chemotherapy. Patients with suspected OBI were tested for HBV-DNA using real-time polymerase chain reaction (PCR). Results: Anti-HBc was detected in 10 of 72 HBsAg negative sera (13.89%) (95% confidence interval 6.9-22.2%). Five of the 10 anti-HBc positive patients in this study had OBI reactivation. Conclusion: The study concluded that anti-HBc screening is mandatory before chemotherapy. HBsAg-negative/anti-HBc-positive patients should be closely observed for signs of HBV reactivation through the regular monitoring of ALT. Prophylaxis lamivudine is recommended for anti-HBc positive patients before chemotherapy.

  12. BN600 reactivity definition

    International Nuclear Information System (INIS)

    Zheltyshev, V.; Ivanov, A.

    2000-01-01

    Since 1980, the fast BN600 reactor with sodium coolant has been operated at Beloyarsk Nuclear Power Plant. The periodic monitoring of the reactivity modifications should be implemented in compliance with the standards and regulations applied in nuclear power engineering. The reactivity measurements are carried out in order to confirm the basic neutronic features of a BN600 reactor. The reactivity measurements are aimed to justify that nuclear safety is provided in course of the in-reactor installation of the experimental core components. Two reactivity meters are to be used on BN600 operation: 1. Digital on-line reactivity calculated under stationary reactor operation on power (approximation of the point-wise kinetics is applied). 2. Second reactivity meter used to define the reactor control rod operating components efficiency under reactor startup and take account of the changing efficiency of the sensor, however, this is more time-consumptive than the on-line reactivity meter. The application of two reactivity meters allows for the monitoring of the reactor reactivity under every operating mode. (authors)

  13. Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

    Energy Technology Data Exchange (ETDEWEB)

    Ammirabile, L.; Tsige-Tamirat, H. [European Commission, JRC, Inst. for Energy, Petten (Netherlands)

    2011-07-01

    In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

  14. Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

    International Nuclear Information System (INIS)

    Ammirabile, L.; Tsige-Tamirat, H.

    2011-01-01

    In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

  15. Analysis of fuel options for the breakeven core configuration of the Advanced Recycling Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Stauff, N.E.; Klim, T.K.; Taiwo, T.A. [Argonne National Laboratory, Argonne, IL (United States); Fiorina, C. [Politecnico di Milano, Milan (Italy); Franceschini, F. [Westinghouse Electric Company LLC., Cranberry Township, Pennsylvania (United States)

    2013-07-01

    A trade-off study is performed to determine the impacts of various fuel forms on the core design and core physics characteristics of the sodium-cooled Toshiba- Westinghouse Advanced Recycling Reactor (ARR). The fuel forms include oxide, nitride, and metallic forms of U and Th. The ARR core configuration is redesigned with driver and blanket regions in order to achieve breakeven fissile breeding performance with the various fuel types. State-of-the-art core physics tools are used for the analyses. In addition, a quasi-static reactivity balance approach is used for a preliminary comparison of the inherent safety performances of the various fuel options. Thorium-fueled cores exhibit lower breeding ratios and require larger blankets compared to the U-fueled cores, which is detrimental to core compactness and increases reprocessing and manufacturing requirements. The Th cores also exhibit higher reactivity swings through each cycle, which penalizes reactivity control and increases the number of control rods required. On the other hand, using Th leads to drastic reductions in void and coolant expansion coefficients of reactivity, with the potential for enhancing inherent core safety. Among the U-fueled ARR cores, metallic and nitride fuels result in higher breeding ratios due to their higher heavy metal densities. On the other hand, oxide fuels provide a softer spectrum, which increases the Doppler effect and reduces the positive sodium void worth. A lower fuel temperature is obtained with the metallic and nitride fuels due to their higher thermal conductivities and compatibility with sodium bonds. This is especially beneficial from an inherent safety point of view since it facilitates the reactor cool-down during loss of power removal transients. The advantages in terms of inherent safety of nitride and metallic fuels are maintained when using Th fuel. However, there is a lower relative increase in heavy metal density and in breeding ratio going from oxide to metallic

  16. Spring unit especially intended for a nuclear reactor core

    International Nuclear Information System (INIS)

    Brown, S.J.; Gorholt, Wilhelm.

    1977-01-01

    This invention relates to a spring unit or a group of springs bearing up a sprung mass against an unsprung mass. For instance, a gas cooled high temperature nuclear reactor includes a core of relatively complex structure supported inside a casing or vessel forming a shielded cavity enclosing the reactor core. This core can be assembled from a large number of graphite blocks of different sizes and shapes joined together to form a column. The blocks of each column can be fixed together so as to form together a loose side support. Under the effect of thermal expansion and contraction, shrinkage resulting from irradiation, the effects of pressure and the contraction and creep of the reactor vessel, it is not possible to confine all the columns of the reactor core in a cylindrical rigid structure. Further, the working of the nuclear reactor requires that the reactivity monitoring components may be inserted at any time in the reactor core. A standard process consists in mounting this loosely assembled reactor core in a floating manner by keeping it away from the vessel enclosure around it by means of a number of springs fitted between the lateral surfaces of the core unit and the reactor vessel. The core may be considered as a spring supported mass whereas, relatively, the reactor vessel is a mass that is not flexibly supported [fr

  17. Adiabatic supernova expansion into the circumstellar medium

    International Nuclear Information System (INIS)

    Band, D.L.; Liang, E.P.

    1987-01-01

    We perform one dimensional numerical simulations with a Lagrangian hydrodynamics code of the adiabatic expansion of a supernova into the surrounding medium. The early expansion follows Chevalier's analytic self-similar solution until the reverse shock reaches the ejecta core. We follow the expansion as it evolves towards the adiabatic blast wave phase. Some memory of the earlier phases of expansion is retained in the interior even when the outer regions expand as a blast wave. We find the results are sensitive to the initial configuration of the ejecta and to the placement of gridpoints. 6 refs., 2 figs

  18. A comparative design study of PB-BI cooled reactor cores with forced and natural convection cooling

    International Nuclear Information System (INIS)

    Mizuno, Tomoyasu; Enuma, Yasuhiro; Tanji, Mikio

    2003-01-01

    A comparative core design study is performed on Pb-Bi cooled reactors with forced and natural convection (FC and NC) cooling. Major interests of the study are core performance and core safety features. The designed core concepts with nitride fuel achieve reasonable breeding capability. The results of unprotected event analyses such as UTOP and ULOF show that both of concepts have possible features to withstand unprotected events due to negative reactivity feedback by Doppler effect, control rod drive line expansion, etc. These results lead to a conclusion that both of concepts have possible capability as one of future promising core concepts. A FC cooling core concept has more advantage if fuel recycle viewpoint is emphasized. (author)

  19. Reactive microencapsulation of carbon allotropes in polyamide shell-core structures and their transformation in hybrid composites with tailored electrical properties

    Directory of Open Access Journals (Sweden)

    F. Oliveira

    2016-02-01

    Full Text Available Polyamide 6 microcapsules (PAMC loaded with 2–10 wt% of different carbon allotropes: carbon black, multiwalled carbon nanotubes, carbon nanofibers and graphite were synthesized via activated anionic polymerization (AAROP of ε-caprolactam in solution performed in the presence of the respective micro- or nanosized loads. The forming high-molecular weight microporous PAMC showed typical diameters of 15–35 µm, the filler particles being entrapped in the core as proven by microscopy methods. The melt processing of the loaded microcapsules produced PA6/C-filler hybrid thermoplastic composites with homogeneous distribution of one or two C-fillers even at loads of up to 10% without any functionalization. The crystalline structure of all PAMC and molded composites was studied by thermal and X-ray diffraction methods focusing on possible structure modification during the transition from PAMC to molded plates. Mechanical tests in tension and electrical conductivity measurements showed that transforming loaded PAMC into composites by melt processing could be a facile and rapid method to fabricate polyamide composites with improved mechanical performance and tailored electrical and dielectric properties.

  20. High temperature chemical reactivity in the system (U, Zr,Fe, O). A contribution to the study of zirconia as a ''core catcher''

    International Nuclear Information System (INIS)

    Maurizi, A.; CEA Centre d'Etudes de Saclay, 91 - Gif-sur-Yvette

    1996-01-01

    Within the framework of the improvement of nuclear reactor safety, a device to recover corium is proposed to be installed under the reactor vessel to limit the consequences of a core melting. According to our bibliographic study, stabilised zirconia seems to be the best refractory material to play this role and to support the physicochemical, mechanical and thermal requirements imposed to the corium catcher. The nature of the chemical interactions between zirconia and iron of high temperature were established and experimental data on the (U, Fe, Zr, O) quaternary system which stands for the corium were determined. First of all, the Knudsen effusion mass-spectrometric method was used to establish the liquidus position for a (U, Zr, O) alloy representative of the corium (U/Zr = 1,5) at 2000 deg C. The oxygen solubility limit in a (U, Zr, O) liquid alloy is about 7 atomic %. In oxidising conditions, the reaction between zirconia and iron leads to the formation of a stabilised zirconia-iron oxide solid solution. Up to 10 atomic % of iron can be incorporated in the structure, leading to the stabilisation of cubic zirconia and a modification of lattice constants. The valence and localisation of those iron measured as a function of time and temperature from 1500 to 2400 deg C, after high frequency inductive heating, both on laboratory materials are commercial bricks. The reaction rate is governed by an activation energy of about 80 kJ/mol. Our results demonstrate that stabilised zirconia is able to efficiently absorb oxidised iron. (author)

  1. Reactivity costs in MARIA reactor

    International Nuclear Information System (INIS)

    Marcinkowska, Zuzanna E.; Pytel, Krzysztof M.; Frydrysiak, Andrzej

    2017-01-01

    Highlights: • The methodology for calculating consumed fuel cost of excess reactivity is proposed. • Correlation between time integral of the core excess reactivity and released energy. • Reactivity price gives number of fuel elements required for given excess reactivity. - Abstract: For the reactor operation at high power level and carrying out experiments and irradiations the major cost of reactor operation is the expense of nuclear fuel. In this paper the methodology for calculating consumed fuel cost-relatedness of excess reactivity is proposed. Reactivity costs have been determined on the basis of operating data. A number of examples of calculating the reactivity costs for processes such as: strong absorbing material irradiation, molybdenium-99 production, beryllium matrix poisoning and increased moderator temperature illustrates proposed method.

  2. Development and verification of an efficient spatial neutron kinetics method for reactivity-initiated event analyses

    International Nuclear Information System (INIS)

    Ikeda, Hideaki; Takeda, Toshikazu

    2001-01-01

    A space/time nodal diffusion code based on the nodal expansion method (NEM), EPISODE, was developed in order to evaluate transient neutron behavior in light water reactor cores. The present code employs the improved quasistatic (IQS) method for spatial neutron kinetics, and neutron flux distribution is numerically obtained by solving the neutron diffusion equation with the nonlinear iteration scheme to achieve fast computation. A predictor-corrector (PC) method developed in the present study enabled to apply a coarse time mesh to the transient spatial neutron calculation than that applicable in the conventional IQS model, which improved computational efficiency further. Its computational advantage was demonstrated by applying to the numerical benchmark problems that simulate reactivity-initiated events, showing reduction of computational times up to a factor of three than the conventional IQS. The thermohydraulics model was also incorporated in EPISODE, and the capability of realistic reactivity event analyses was verified using the SPERT-III/E-Core experimental data. (author)

  3. EXCURS: a computing programme for analysis of core transient behaviour in a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Saito, Shinzo

    1977-09-01

    In the code EXCURS developed for core transient behaviour calculation of a sodium-cooled fast reactor, a one-channel model is used to represent thermal behaviour of the reactor core. Calculations are made for three different channels; i.e. average, hot and hottest. In the average channel the power density and coolant velocity are equal to the mean values of the whole core. In the hot channel, a maximum power density of the core and a specific coolant velocity are introduced. In the hottest channel, engineering hot channel factors are considered to the hot channel. A one-point neutron kinetics equation with six delayed neutron groups is used to calculate the time-dependent power behaviour. Externally introduced reactivity effect and control rod movement in the case of a scram are taken into account. In the feedback effects evaluated on the basis of the average channel temperatures are considered Doppler effect, fuel axial expansion, cladding expansion, coolant expansion and structure expansion. The decay heat after reactor scram is also considered. Heat balance is taken in each cross section, neglecting the axial heat transfer except for the coolant region. Temperature dependence of the physical properties of materials is considered by second-order polynomials approximation, and also the fuel melting process. Each channel can be divided into a maximum of 20 regions in both radially and axially. The reactor core transient behaviour due to reactivity insertion or loss-of-coolant flow can be studied by EXCURS. The calculated results are plotted optionally by connected code EXPLOT. (auth.)

  4. Automated reactivity anomaly surveillance in the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Knutson, B.J.; Harris, R.A.; Honeyman, D.J.; Shook, A.T.; Krohn, C.N.

    1985-01-01

    The automated technique for monitoring core reactivity during power operation used at the Fast Flux Test Facility (FFTF) is described. This technique relies on comparing predicted to measured rod positions to detect any anomalous (or unpredicted) core reactivity changes. It is implemented on the Plant Data System (PDS) computer and, thus, provides rapid indication of any abnormal core conditions. The prediction algorithms use thermal-hydraulic, control rod position and neutron flux sensor information to predict the core reactivity state

  5. Fort St. Vrain core performance

    International Nuclear Information System (INIS)

    McEachern, D.W.; Brown, J.R.; Heller, R.A.; Franek, W.J.

    1977-07-01

    The Fort St. Vrain High Temperature Gas Cooled Reactor core performance has been evaluated during the startup testing phase of the reactor operation. The reactor is graphite moderated, helium cooled, and uses coated particle fuel and on-line flow control to each of the 37 refueling regions. Principal objectives of startup testing were to determine: core and control system reactivity, radial power distribution, flow control capability, and initial fission product release. Information from the core demonstrates that Technical Specifications are being met, performance of the core and fuel is as expected, flow and reactivity control are predictable and simple for the operator to carry out

  6. FBR type reactor core

    International Nuclear Information System (INIS)

    Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

    1995-01-01

    Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

  7. Reactor core and control rod assembly in FBR type reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

    1993-01-01

    Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

  8. Neutronic characterization of cylindrical core of minor excess reactivity in the nuclear reactor IPEN/MB-01 from the measure of spatial and energetic distribution of neutron flux distribution; Caracterizacao do nucleo cilindrico de menor excesso de reatividade do reator IPEN/MB-01, pela medida da distribuicao espacial e energetica do fluxo de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Aredes, Vitor Ottoni Garcia

    2014-07-01

    In this work was conducted the mapping of the thermal and epithermal neutrons flux and the energy spectrum of the neutrons in the reactor core IPEN/MB-01 for a cylindrical core configuration with minor excess reactivity, which is 28 x 28 fuel rods arranged in north-south and east-west directions. The calibration of control rods for this configuration determined their excess reactivity. The lower excess reactivity in the core decreased neutron flux disturbance caused by the neutron absorbing rods , given that the nuclear reactor was operated with the rods almost completely removed . Was used the 'Activation Analysis Technique' with the thin foil activation detectors ( infinitely diluted and hyper-pure), of different materials that work in different energy ranges, to calculate the saturation activity, used for determining the neutron flux and in the SANDBP code as input for the calculation of the neutrons energy spectrum. To discriminate thermal and epithermal flux , was used the 'Cadmium RatioTechnique' . The activation detectors were distributed in a total of 140 radial and axial positions in the reactor core and 16 irradiation, with bare and covered with cadmium activation foils. A model of this configuration was simulated by MCNP-5 code to determine the cadmium correction factor and comparison of the results obtained experimentally. The cylindrical configuration desired, with 17% less fuel than the standard rectangular configuration (28 x 26 fuel rods), reached criticality with the control rods approximately 90% removed, which decreased considerably the disturbance in neutron flux. Given the highest power density of the 28 x 28 cylindrical core, the neutron flux increased by over 50% in the central regions of the core compared to the values of the 28 x 26 standard rectangular core. (author)

  9. Plutonium cores of zenith

    Energy Technology Data Exchange (ETDEWEB)

    Barclay, F R; Cameron, I R; Drageset, A; Freemantle, R G; Wilson, D J

    1965-03-15

    The report describes a series of experiments carried out with plutonium fuel in the heated zero power reactor ZENITH, with the aim of testing current theoretical methods, with particular reference to excess reactivity, temperature coefficients, differential spectrum and reaction rate distributions. Two cores of widely different fissile/moderator atom ratios were loaded in order to test the theory under significantly varied spectrum conditions.

  10. Regarding KUR Reactivity Measurement System

    International Nuclear Information System (INIS)

    Nakamori, Akira; Hasegawa, Kei; Tsuchiyama, Tatsuo; Yamamoto, Toshihiro; Okumura, Ryo; Sano, Tadafumi

    2012-01-01

    This article reported: (1) the outline of the reactivity measurement system of Kyoto University Research Reactor (KUR), (2) the calibration data of control rod, (3) the problems and the countermeasures for range switching of linear output meter. For the laptop PC for the reactivity measurement system, there are four input signals: (1) linear output meter, (2) logarithmic output meter, (3) core temperature gauge, and (4) control rod position. The hardware of reactivity measurement system is controlled with Labview installed on the laptop. Output, reactivity, reactor period, and the change in reactivity due to temperature effect or Xenon effect are internally calculated and displayed in real-time with Labview based on the four signals above. Calculation results are recorded in the form of a spreadsheet. At KUR, the reactor core arrangement was changed, so the control rod was re-calibrated. At this time, calculated and experimental values of reactivity based on the reactivity measurement system were compared, and it was confirmed that the reactivity calculation by Labview was accurate. The range switching of linear output meter in the nuclear instrumentation should automatically change within the laptop, however sometimes this did not function properly in the early stage. It was speculated that undefined percent values during the transition of percent value were included in the calculation and caused calculation errors. The range switching started working properly after fixing this issue. (S.K.)

  11. Reactive Arthritis

    Directory of Open Access Journals (Sweden)

    Eren Erken

    2013-06-01

    Full Text Available Reactive arthritis is an acute, sterile, non-suppurative and inflammatory arthropaty which has occured as a result of an infectious processes, mostly after gastrointestinal and genitourinary tract infections. Reiter syndrome is a frequent type of reactive arthritis. Both reactive arthritis and Reiter syndrome belong to the group of seronegative spondyloarthropathies, associated with HLA-B27 positivity and characterized by ongoing inflammation after an infectious episode. The classical triad of Reiter syndrome is defined as arthritis, conjuctivitis and urethritis and is seen only in one third of patients with Reiter syndrome. Recently, seronegative asymmetric arthritis and typical extraarticular involvement are thought to be adequate for the diagnosis. However, there is no established criteria for the diagnosis of reactive arthritis and the number of randomized and controlled studies about the therapy is not enough. [Archives Medical Review Journal 2013; 22(3.000: 283-299

  12. Negative thermal expansion materials

    International Nuclear Information System (INIS)

    Evans, J.S.O.

    1997-01-01

    The recent discovery of negative thermal expansion over an unprecedented temperature range in ZrW 2 O 8 (which contracts continuously on warming from below 2 K to above 1000 K) has stimulated considerable interest in this unusual phenomenon. Negative and low thermal expansion materials have a number of important potential uses in ceramic, optical and electronic applications. We have now found negative thermal expansion in a large new family of materials with the general formula A 2 (MO 4 ) 3 . Chemical substitution dramatically influences the thermal expansion properties of these materials allowing the production of ceramics with negative, positive or zero coefficients of thermal expansion, with the potential to control other important materials properties such as refractive index and dielectric constant. The mechanism of negative thermal expansion and the phase transitions exhibited by this important new class of low-expansion materials will be discussed. (orig.)

  13. Proteomics Core

    Data.gov (United States)

    Federal Laboratory Consortium — Proteomics Core is the central resource for mass spectrometry based proteomics within the NHLBI. The Core staff help collaborators design proteomics experiments in a...

  14. Nuclear data propagation with burnup. Impact on SFR reactivity coefficients

    International Nuclear Information System (INIS)

    Buiron, Laurent; Plisson-Rieunier, Daniele

    2017-01-01

    thermal sodium expansion coefficient and the Doppler Effect. Using these sensitivities, a global evaluation of impact of the fuel depletion can be quantified for these reactivity effects at core scale for end of cycle state. An illustration is given for a GEN IV SFR industrial core design (SFR V2B). A first glance at preliminary uncertainty level is presented using current covariance matrices available at CEA. (author)

  15. Insertion material for controlling reactivity

    International Nuclear Information System (INIS)

    Baba, Iwao.

    1994-01-01

    Moderators and a group of suspended materials having substantially the same density as the moderator are sealed in a hollow rod vertically inserted to a fuel assembly. Specifically, the group of suspended materials is adapted to have a density changing stepwise from density of the moderator at the exit temperature of the reactor core to that at the inlet temperature of the reactor core. Reactivity is selectively controlled for a portion of high power and a portion of high reactivity by utilizing the density of the moderator and the distribution of the density. That is, if the power distribution is flat, the density of the moderators changes at a constant rate over the vertical direction of the reactor core and the suspended materials stay at a portion of the same density, to form a uniform distribution. Further, upon reactor shutdown, since the liquid temperature of the moderators is lowered and the density is increased, all of beads are collected at the upper portion to remove water at the upper portion of the reactor core of low burnup degree thereby selectively controlling the reactivity at a portion of high power and a portion of high reactivity. (N.H.)

  16. Spatial Linkage and Urban Expansion: AN Urban Agglomeration View

    Science.gov (United States)

    Jiao, L. M.; Tang, X.; Liu, X. P.

    2017-09-01

    Urban expansion displays different characteristics in each period. From the perspective of the urban agglomeration, studying the spatial and temporal characteristics of urban expansion plays an important role in understanding the complex relationship between urban expansion and network structure of urban agglomeration. We analyze urban expansion in the Yangtze River Delta Urban Agglomeration (YRD) through accessibility to and spatial interaction intensity from core cities as well as accessibility of road network. Results show that: (1) Correlation between urban expansion intensity and spatial indicators such as location and space syntax variables is remarkable and positive, while it decreases after rapid expansion. (2) Urban expansion velocity displays a positive correlation with spatial indicators mentioned above in the first (1980-1990) and second (1990-2000) period. However, it exhibits a negative relationship in the third period (2000-2010), i.e., cities located in the periphery of urban agglomeration developing more quickly. Consequently, the hypothesis of convergence of urban expansion in rapid expansion stage is put forward. (3) Results of Zipf's law and Gibrat's law show urban expansion in YRD displays a convergent trend in rapid expansion stage, small and medium-sized cities growing faster. This study shows that spatial linkage plays an important but evolving role in urban expansion within the urban agglomeration. In addition, it serves as a reference to the planning of Yangtze River Delta Urban Agglomeration and regulation of urban expansion of other urban agglomerations.

  17. Core-in-shell sorbent for hot coal gas desulfurization

    Science.gov (United States)

    Wheelock, Thomas D.; Akiti, Jr., Tetteh T.

    2004-02-10

    A core-in-shell sorbent is described herein. The core is reactive to the compounds of interest, and is preferably calcium-based, such as limestone for hot gas desulfurization. The shell is a porous protective layer, preferably inert, which allows the reactive core to remove the desired compounds while maintaining the desired physical characteristics to withstand the conditions of use.

  18. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  19. Expansion joints for LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Dzenus, M.; Hundhausen, W.; Jansing, W.

    1979-10-15

    This discourse recounts efforts put into the SNR-2 project; specifically the development of compensation devices. The various prototypes of these compensation devices are described and the state of development reviewed. The expansion joints were developed on the basis of specific design criteria whereby differentiation is made between expansion joints of small and large nominal diameter. Expansion joints for installation in the sodium-filled primary piping are equipped with safety bellows in addition to the actual working bellows.

  20. Reactivity monitoring during reactor-reloading operations

    International Nuclear Information System (INIS)

    Baumann, N.P.; Ahlfeld, C.F.; Ridgely, G.C.

    1983-01-01

    At the Savannah River Plant (SRP) reloading operations during shutdown present special considerations in reactivity monitoring and control. Large reactivity changes may occur during reloading operations because of the heterogeneous nature of some core designs. This paper describes an improved monitoring system

  1. Development of Regulatory Audit Core Safety Code : COREDAX

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chae Yong; Jo, Jong Chull; Roh, Byung Hwan [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Jun; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2005-07-01

    Korea Institute of Nuclear Safety (KINS) has developed a core neutronics simulator, COREDAX code, for verifying core safety of SMART-P reactor, which is technically supported by Korea Advanced Institute of Science and Technology (KAIST). The COREDAX code would be used for regulatory audit calculations of 3- dimendional core neutronics. The COREDAX code solves the steady-state and timedependent multi-group neutron diffusion equation in hexagonal geometry as well as rectangular geometry by analytic function expansion nodal (AFEN) method. AFEN method was developed at KAIST, and it was internationally verified that its accuracy is excellent. The COREDAX code is originally programmed based on the AFEN method. Accuracy of the code on the AFEN method was excellent for the hexagonal 2-dimensional problems, but there was a need for improvement for hexagonal-z 3-dimensional problems. Hence, several solution routines of the AFEN method are improved, and finally the advanced AFEN method is created. COREDAX code is based on the advanced AFEN method . The initial version of COREDAX code is to complete a basic framework, performing eigenvalue calculations and kinetics calculations with thermal-hydraulic feedbacks, for audit calculations of steady-state core design and reactivity-induced accidents of SMART-P reactor. This study describes the COREDAX code for hexagonal geometry.

  2. Reactive Systems

    DEFF Research Database (Denmark)

    Aceto, Luca; Ingolfsdottir, Anna; Larsen, Kim Guldstrand

    A reactive system comprises networks of computing components, achieving their goals through interaction among themselves and their environment. Thus even relatively small systems may exhibit unexpectedly complex behaviours. As moreover reactive systems are often used in safety critical systems......, the need for mathematically based formal methodology is increasingly important. There are many books that look at particular methodologies for such systems. This book offers a more balanced introduction for graduate students and describes the various approaches, their strengths and weaknesses, and when...... they are best used. Milner's CCS and its operational semantics are introduced, together with the notions of behavioural equivalences based on bisimulation techniques and with recursive extensions of Hennessy-Milner logic. In the second part of the book, the presented theories are extended to take timing issues...

  3. Convergence of mayer expansions

    International Nuclear Information System (INIS)

    Brydges, D.C.

    1986-01-01

    The tree graph bound of Battle and Federbush is extended and used to provide a simple criterion for the convergence of (iterated) Mayer expansions. As an application estimates on the radius of convergence of the Mayer expansion for the two-dimensional Yukawa gas (nonstable interaction) are obtained

  4. A study of temperature coefficients of reactivity for a Savannah River Site tritium-producing charge

    International Nuclear Information System (INIS)

    George, D.L.; Frost, R.L.

    1991-01-01

    Temperature coefficients of reactivity have been calculated for the Mark 22 assembly in the K-14 charge at the Savannah River Site. Temperature coefficients are the most important reactivity feedback mechanism in SRS reactors; they are used in all safety analyses performed in support of the Safety Analysis Report, and in operations to predict reactivity changes with control rod moves. The effects of the radial location of the assembly in the reactor, isotope depletion, and thermal expansion of the metal components on the temperature coefficients have also been investigated. With the exception of the dead space coefficient, all of the regional temperature coefficients were found to be negative or zero. All of the temperature coefficients become more negative with isotopic depletion over the fuel cycle. Coefficients also become more negative with increasing radial distance of the assembly from the center of the core; this is proven from first principles and confirmed by calculations. It was found that axial and radial thermal expansion effects on the metal fuel and target tubes counteract one another, indicating these effects do not need to be considered in future temperature coefficient calculations for the Mark 22 assembly. The moderator coefficient was found to be nonlinear with temperature; thus, the values derived for accidents involving increases in moderator temperature are significantly different than those for decreases in moderator temperature, although the moderator coefficient is always negative

  5. Neutronics aspects associated to irregular lattices in sodium fast reactors cores

    International Nuclear Information System (INIS)

    Gentili, Michele

    2015-01-01

    The fuel assemblies of SFR cores (sodium fast reactors) are normally arranged in hexagonal regular lattices, whose compactness is ensured in nominal operating conditions by thermal expansion of assemblies pads disposed on the six assembly wrapper faces. During the reactor operations, thermal expansion phenomena and irradiation creep phenomena occur and they cause the fuel assemblies to bow and to deform both radially and axially. The main goal of this PhD is the understanding of the neutronic aspects and phenomena occurring in case of core and lattice deformations, as much as the design and implementation of deterministic neutronic calculation schemes and methods in order to evaluate the consequences for the core design activities and the safety analysis. The first part of this work is focused on the development of an analytical model with the purpose to identify the neutronic phenomena that are the main contributors to the reactivity changes induced by lattice and core deformations. A first scheme based on the spatial mesh projection method has been conceived and implemented for the ERANOS codes (BISTRO, H3D and VARIANT) and to the SNATCH solver. The second calculation scheme propose is based on mesh deformation: the computing mesh is deformed as a function of the assembly displacement field. This methodology has been implemented for the solver SNATCH, which normally allows the Boltzmann equation to be solved for a regular mesh. Finally, an iterative method has been developed in order to fulfill an a-priori estimation of the maximal reactivity insertion as a function of the postulated mechanical energy provided to the core, as much as the deformation causing it. (author) [fr

  6. Low reactivity penalty burnable poison rods

    International Nuclear Information System (INIS)

    1978-01-01

    A nuclear reactor burnable poison rod is described which consists of an elongated tubular sheath enclosing a neutron absorbing material which, at least during reactor operation, also encloses a neutron moderating material. The excess reactivity existing at the beginning of core life is compensated for by the depletion of the burnable poison throughout the life of the core, so that the life of the core is extended. (UK)

  7. Determining neutron multiplication factor in the infinite system by reactivity dependence on one dimension of the reactor core; Odredjivanje faktora umnozavanja neutrona u beskonacnom sistemu pomocu zavisnosti reaktivnosti od jedne dimenzije reaktorskog jezgra, Diplomski rad

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1975-07-01

    The objective of this task was to apply Fermi age theory for determining {tau} and neutron multiplication factor in infinite medium by measuring reactivity coefficient of heavy water in heterogeneous mixed reactor lattice. Basis of experiment is the measurement of stable reactor period. Measurement of heavy water reactivity coefficient by measuring the stable reactor period is described for chosen overcritical heavy water levels. Calculated values of infinite multiplication factor for measured neutron age data are presented and they are compared to expected theoretical values.

  8. Stationary liquid fuel fast reactor SLFFR – Part I: Core design

    Energy Technology Data Exchange (ETDEWEB)

    Jing, T.; Yang, G.; Jung, Y.S.; Yang, W.S., E-mail: yang494@purdue.edu

    2016-12-15

    Highlights: • An innovative fast reactor concept SLFFR based on liquid metal fuel is proposed for TRU burning. • A compact core design of 1000 MWt SLFFR is developed to achieve a zero conversion ratio and passive safety. • The core size and the control requirement are significantly reduced compared to the conventional solid fuel reactor with same conversion ratio. - Abstract: For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named the stationary liquid fuel fast reactor (SLFFR) has been proposed based on a stationary molten metallic fuel. A compact core design of a 1000 MWt SLFFR has been developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches have been adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses have been performed to evaluate the steady-state performance characteristics. The analysis results indicate that the SLFFR of a zero TRU conversion ratio is feasible while satisfying the conservatively imposed thermal design constraints. A theoretical maximum TRU consumption rate of 1.01 kg/day is achieved with uranium-free fuel. Compared to the solid fuel reactors with the same TRU conversion ratio, the core size and the reactivity control requirement are reduced significantly. The primary and secondary control systems provide sufficient shutdown margins, and the calculated reactivity feedback coefficients show that the prompt fuel expansion coefficient is sufficiently negative.

  9. Controlled Thermal Expansion Alloys

    Data.gov (United States)

    National Aeronautics and Space Administration — There has always been a need for controlled thermal expansion alloys suitable for mounting optics and detectors in spacecraft applications.  These alloys help...

  10. Evaluation of In-Core Fuel Management for the Transition Cores of RSG-GAS Reactor to Full-Silicide Core

    International Nuclear Information System (INIS)

    S, Tukiran; MS, Tagor; P, Surian

    2003-01-01

    The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 gU/cc has been done. The core-of RSG-GAS reactor has been operated full core of silicide fuels which is started with the mixed core of oxide-silicide start from core 36. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 9 transition cores (core 36 - 44) to achieve a full-silicide core (core 45). The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters. Conversion core was achieved by transition cores mixed oxide-silicide fuels. Each transition core is calculated and measured core parameter such as, excess reactivity and shutdown margin. Calculation done by Batan-EQUIL-2D code and measurement of the core parameters was carried out using the method of compensation of couple control rods. The results of calculation and experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safely to a full-silicide core

  11. Fuel Thermal Expansion (FTHEXP)

    International Nuclear Information System (INIS)

    Reymann, G.A.

    1978-07-01

    A model is presented which deals with dimensional changes in LWR fuel pellets caused by changes in temperature. It is capable of dealing with any combination of UO 2 and PuO 2 in solid, liquid or mixed phase states, and includes expansion due to the solid-liquid phase change. The function FTHEXP models fuel thermal expansion as a function of temperature, fraction of PuO 2 , and the fraction of fuel which is molten

  12. Transport-diffusion comparisons for small core LMFBR disruptive accidents

    International Nuclear Information System (INIS)

    Tomlinson, E.T.

    1977-11-01

    A number of numerical experiments were performed to assess the validity of diffusion theory for calculating the reactivity state of various small core LMFBR disrupted geometries. The disrupted configurations correspond, in general, to various configurations predicted by SAS3A for transient undercooling (TUC) and transient overpower (TOP) accidents for homogeneous cores and to the ZPPR-7 configurations for heterogeneous core. In all TUC cases diffusion theory was shown to be inadequate for the calculation of reactivity changes during core disassembly

  13. Resonant state expansions

    International Nuclear Information System (INIS)

    Lind, P.

    1993-02-01

    The completeness properties of the discrete set of bound state, virtual states and resonances characterizing the system of a single nonrelativistic particle moving in a central cutoff potential is investigated. From a completeness relation in terms of these discrete states and complex scattering states one can derive several Resonant State Expansions (RSE). It is interesting to obtain purely discrete expansion which, if valid, would significantly simplify the treatment of the continuum. Such expansions can be derived using Mittag-Leffler (ML) theory for a cutoff potential and it would be nice to see if one can obtain the same expansions starting from an eigenfunction theory that is not restricted to a finite sphere. The RSE of Greens functions is especially important, e.g. in the continuum RPA (CRPA) method of treating giant resonances in nuclear physics. The convergence of RSE is studied in simple cases using square well wavefunctions in order to achieve high numerical accuracy. Several expansions can be derived from each other by using the theory of analytic functions and one can the see how to obtain a natural discretization of the continuum. Since the resonance wavefunctions are oscillating with an exponentially increasing amplitude, and therefore have to be interpreted through some regularization procedure, every statement made about quantities involving such states is checked by numerical calculations.Realistic nuclear wavefunctions, generated by a Wood-Saxon potential, are used to test also the usefulness of RSE in a realistic nuclear calculation. There are some fundamental differences between different symmetries of the integral contour that defines the continuum in RSE. One kind of symmetry is necessary to have an expansion of the unity operator that is idempotent. Another symmetry must be used if we want purely discrete expansions. These are found to be of the same form as given by ML. (29 refs.)

  14. Analysis on void reactivity of DCA lattice

    International Nuclear Information System (INIS)

    Min, B. J.; Noh, K. H.; Choi, H. B.; Yang, M. K.

    2001-01-01

    In case of loss of coolant accident, the void reactivity of CANDU fuel provides the positive reactivity and increases the reactor power rapidly. Therefore, it is required to secure credibility of the void reactivity for the design and analysis of reactor, which motivated a study to assess the measurement data of void reactivity. The assessment of lattice code was performed with the experimental data of void reactivity at 30, 70, 87 and 100% of void fractions. The infinite multiplication factors increased in four types of fuels as the void fractions of them grow. The infinite multiplication factors of uranium fuels are almost within 1%, but those of Pu fuels are over 10% by the results of WIMS-AECL and MCNP-4B codes. Moreover, coolant void reactivity of the core loaded with plutonium fuel is more negative compared with that with uranium fuel because of spectrum hardening resulting from large void fraction

  15. Analysis Of Core Management For The Transition Cores Of RSG-GAS Reactor To Full-Silicide Core

    International Nuclear Information System (INIS)

    Malem Sembiring, Tagor; Suparlina, Lily; Tukiran

    2001-01-01

    The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 g U/cc is still doing. At the end of 2000, the reactor has been operated for 3 transition cores which is the mixed core of oxide-silicide. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 10 transition cores to achieve a full-silicide core. The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters such as excess reactivity and shutdown margin. The measurement of the core parameters was carried out using the method of compensation of couple control rods. The experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safety to a full-silicide core

  16. Core calculational techniques and procedures

    International Nuclear Information System (INIS)

    Romano, J.J.

    1977-10-01

    Described are the procedures and techniques employed by B and W in core design analyses of power peaking, control rod worths, and reactivity coefficients. Major emphasis has been placed on current calculational tools and the most frequently performed calculations over the operating power range

  17. Expansion joints for LMFBR

    International Nuclear Information System (INIS)

    Dzenus, M.; Hundhausen, W.; Jansing, W.

    1980-01-01

    This discourse recounts efforts put into the SNR-2 project; specifically the development of compensation devices. The various prototypes of these compensation devices are described and the state of the development reviewed. Large Na (sodium)-heat transfer systems require a lot of valuable space if the component lay-out does not include compensation devices. So, in order to condense the spatial requirement as much as possible, expansion joints must be integrated into the pipe system. There are two basic types to suit the purpose: axial expansion joints and angular expansion joints. The expansion joints were developed on the basis of specific design criteria whereby differentiation is made between expansion joints of small and large nominal diameter. Expansion joints for installation in the sodium-filled primary piping are equipped with safety bellows in addition to the actual working bellows. Expansion joints must be designed and mounted in a manner to completely withstand seismic forces. The design must exclude any damage to the bellows during intermittent operations, that is, when sodium is drained the bellows' folds must be completely empty; otherwise residual solidified sodium could destroy the bellows when restarting. The expansion joints must be engineered on the basis of the following design data for the secondary system of the SNR project: working pressure: 16 bar; failure mode pressure: 5 events; failure mode: 5 sec., 28.5 bar, 520 deg. C; working temperature: 520 deg. C; temperature transients: 30 deg. C/sec.; service life: 200,000 h; number of load cycles: 10 4 ; material: 1.4948 or 1.4919; layer thickness of folds: 0.5 mm; angular deflection (DN 800): +3 deg. C or; axial expansion absorption (DN 600): ±80 mm; calculation: ASME class. The bellows' development work is not handled within this scope. The bellows are supplied by leading manufacturers, and warrant highest quality. Multiple bellows were selected on the basis of maximum elasticity - a property

  18. Accelerating the loop expansion

    International Nuclear Information System (INIS)

    Ingermanson, R.

    1986-01-01

    This thesis introduces a new non-perturbative technique into quantum field theory. To illustrate the method, I analyze the much-studied phi 4 theory in two dimensions. As a prelude, I first show that the Hartree approximation is easy to obtain from the calculation of the one-loop effective potential by a simple modification of the propagator that does not affect the perturbative renormalization procedure. A further modification then susggests itself, which has the same nice property, and which automatically yields a convex effective potential. I then show that both of these modifications extend naturally to higher orders in the derivative expansion of the effective action and to higher orders in the loop-expansion. The net effect is to re-sum the perturbation series for the effective action as a systematic ''accelerated'' non-perturbative expansion. Each term in the accelerated expansion corresponds to an infinite number of terms in the original series. Each term can be computed explicitly, albeit numerically. Many numerical graphs of the various approximations to the first two terms in the derivative expansion are given. I discuss the reliability of the results and the problem of spontaneous symmetry-breaking, as well as some potential applications to more interesting field theories. 40 refs

  19. Conformal expansions and renormalons

    Energy Technology Data Exchange (ETDEWEB)

    Rathsman, J.

    2000-02-07

    The coefficients in perturbative expansions in gauge theories are factorially increasing, predominantly due to renormalons. This type of factorial increase is not expected in conformal theories. In QCD conformal relations between observables can be defined in the presence of a perturbative infrared fixed-point. Using the Banks-Zaks expansion the authors study the effect of the large-order behavior of the perturbative series on the conformal coefficients. The authors find that in general these coefficients become factorially increasing. However, when the factorial behavior genuinely originates in a renormalon integral, as implied by a postulated skeleton expansion, it does not affect the conformal coefficients. As a consequence, the conformal coefficients will indeed be free of renormalon divergence, in accordance with previous observations concerning the smallness of these coefficients for specific observables. The authors further show that the correspondence of the BLM method with the skeleton expansion implies a unique scale-setting procedure. The BLM coefficients can be interpreted as the conformal coefficients in the series relating the fixed-point value of the observable with that of the skeleton effective charge. Through the skeleton expansion the relevance of renormalon-free conformal coefficients extends to real-world QCD.

  20. Transformer core

    NARCIS (Netherlands)

    Mehendale, A.; Hagedoorn, Wouter; Lötters, Joost Conrad

    2008-01-01

    A transformer core includes a stack of a plurality of planar core plates of a magnetically permeable material, which plates each consist of a first and a second sub-part that together enclose at least one opening. The sub-parts can be fitted together via contact faces that are located on either side

  1. Transformer core

    NARCIS (Netherlands)

    Mehendale, A.; Hagedoorn, Wouter; Lötters, Joost Conrad

    2010-01-01

    A transformer core includes a stack of a plurality of planar core plates of a magnetically permeable material, which plates each consist of a first and a second sub-part that together enclose at least one opening. The sub-parts can be fitted together via contact faces that are located on either side

  2. Thermal expansion data

    International Nuclear Information System (INIS)

    Taylor, D.

    1984-01-01

    This paper gives regression data for a modified second order polynomial fitted to the expansion data of, and percentage expansions for dioxides with (a) the fluorite and antifluorite structure: AmO 2 , BkO 2 , CeO 2 , CmO 2 , HfO 2 , Li 2 O, NpO 2 , PrO 2 , PuO 2 , ThO 2 , UO 2 , ZrO 2 , and (b) the rutile structure: CrO 2 , GeO 2 , IrO 2 , MnO 2 , NbO 2 , PbO 2 , SiO 2 , SnO 2 , TeO 2 , TiO 2 and VO 2 . Reduced expansion curves for the dioxides showed only partial grouping into iso-electronic series for the fluorite structures and showed that the 'law of corresponding states' did not apply to the rutile structures. (author)

  3. Safety characteristics of mid-sized MOX fueled liquid metal reactor core of high converter type in the initiating phase of unprotected loss of flow accident. Effect of low specific fuel power density on ULOF behavior brought by employment of large diameter fuel pins

    International Nuclear Information System (INIS)

    Ishida, Masayoshi; Kawada, Kenichi; Niwa, Hajime

    2003-07-01

    Safety characteristics in core disruptive accidents (CDAs) of mid-sized MOX fueled liquid metal reactor core of high converter type have been examined by using the CDA initiating phase analysis code SAS4A. The design concept of high converter type reactor core has been studied as one of options in the category of sodium-cooled reactor in Phase II of Feasibility Study on Commercialized Fast Reactor Cycle System. An unprotected loss-of-flow accident (ULOF) has been selected as a representative CDA initiator for this study. A core concept of high converter type, which employed a large diameter fuel pin of 11.1 mm with 1.2 m core height to get a large fuel volume fraction in the core to achieve high internal conversion ratio was proposed in JFY2001. Each fuel subassembly of the core (abbreviated here as UPL120)was provided with an upper sodium plenum directly above the core to reduce the sodium void reactivity worth. Because of the large fuel pin diameter, average specific fuel power density (31 kW/kg-MOX) of UPL120 is about one half of those of conventional large MOX cores. The reactivity worth of sodium voiding is 6$ in the whole core, and -1$ in the all upper plenums. Initiating phase of ULOF accident in UPL120 under the conditions of nominal design and best estimate analysis resulted in a slightly super-prompt critical power burst. The causes of the super-prompt criticality have been identified twofold: (a) the low specific fuel power density of core reduced the effectiveness of prompt negative reactivity feedback of Doppler and axial fuel expansion effects upon increase in reactor power, and (b) the longer core height compared with conventional 1m cores brought, together with the lower specific power density, a remarkable delay in insertion of negative fuel dispersion reactivity after the onset of fuel disruption in sodium voided subassembly due to the lower linear heat rating in the top portion of the core. During the delay, burst-type fuel failures in sodium un

  4. Uniform gradient expansions

    CERN Document Server

    Giovannini, Massimo

    2015-01-01

    Cosmological singularities are often discussed by means of a gradient expansion that can also describe, during a quasi-de Sitter phase, the progressive suppression of curvature inhomogeneities. While the inflationary event horizon is being formed the two mentioned regimes coexist and a uniform expansion can be conceived and applied to the evolution of spatial gradients across the protoinflationary boundary. It is argued that conventional arguments addressing the preinflationary initial conditions are necessary but generally not sufficient to guarantee a homogeneous onset of the conventional inflationary stage.

  5. IAEA sodium void reactivity benchmark calculations

    International Nuclear Information System (INIS)

    Hill, R.N.; Finck, P.J.

    1992-01-01

    In this paper, the IAEA-1 992 ''Benchmark Calculation of Sodium Void Reactivity Effect in Fast Reactor Core'' problem is evaluated. The proposed design is a large axially heterogeneous oxide-fueled fast reactor as described in Section 2; the core utilizes a sodium plenum above the core to enhance leakage effects. The calculation methods used in this benchmark evaluation are described in Section 3. In Section 4, the calculated core performance results for the benchmark reactor model are presented; and in Section 5, the influence of steel and interstitial sodium heterogeneity effects is estimated

  6. Reactivity anomalies in the FFTF [Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Knutson, B.J.; Harris, R.A.

    1987-04-01

    Experience using an automated core reactivity monitoring technique at the Fast Flux Test Facility (FFTF) through eight operating cycles is described. This technique relies on comparing predicted to measured rod positions to detect any anomalous (or unpredicted) core reactivity changes. Reactivity worth predictions of core state changes (e.g., temperature and irradiation changes) and compensating control rod movements are required for the rod position comparison. A substantial data base now exists to evaluate changes in temperature reactivity feedback effects operational in the FFTF, rod worth changes due to core loading, temperature and irradiation effects and burnup effects associated with transmutation of fuel materials. This report summarizes preliminary work of correlating zero power and at-power rod worth measurement data, calculated burnup rates and rod worths using the latest ENDF/B-V cross section set for each cycle to evaluate the prediction models and attempt to resolve observed reactivity anomalies. 2 figs., 2 tabs

  7. Low-temperature thermal expansion

    International Nuclear Information System (INIS)

    Collings, E.W.

    1986-01-01

    This chapter discusses the thermal expansion of insulators and metals. Harmonicity and anharmonicity in thermal expansion are examined. The electronic, magnetic, an other contributions to low temperature thermal expansion are analyzed. The thermodynamics of the Debye isotropic continuum, the lattice-dynamical approach, and the thermal expansion of metals are discussed. Relative linear expansion at low temperatures is reviewed and further calculations of the electronic thermal expansion coefficient are given. Thermal expansions are given for Cu, Al and Ti. Phenomenologic thermodynamic relationships are also discussed

  8. Core lifter

    Energy Technology Data Exchange (ETDEWEB)

    Pavlov, N G; Edel' man, Ya A

    1981-02-15

    A core lifter is suggested which contains a housing, core-clamping elements installed in the housing depressions in the form of semirings with projections on the outer surface restricting the rotation of the semirings in the housing depressions. In order to improve the strength and reliability of the core lifter, the semirings have a variable transverse section formed from the outside by the surface of the rotation body of the inner arc of the semiring aroung the rotation axis and from the inner a cylindrical surface which is concentric to the outer arc of the semiring. The core-clamping elements made in this manner have the possibility of freely rotating in the housing depressions under their own weight and from contact with the core sample. These semirings do not have weakened sections, have sufficient strength, are inserted into the limited ring section of the housing of the core lifter without reduction in its through opening and this improve the reliability of the core lifter in operation.

  9. Progress and problems in modelling HTR core dynamics

    International Nuclear Information System (INIS)

    Scherer, W.; Gerwin, H.

    1991-01-01

    In recent years greater effort has been made to establish theoretical models for HTR core dynamics. At KFA Juelich the TINTE (TIme dependent Neutronics and TEmperatures) code system has been developed, which is able to model the primary circuit of an HTR plant using modern numerical techniques and taking into account the mutual interference of the relevant physical variables. The HTR core is treated in 2-D R-Z geometry for both nucleonics and thermo-fluid-dynamics. 2-energy-group diffusion theory is used in the nuclear part including 6 groups of delayed neutron precursors and 14 groups of decay heat producers. Local and non-local heat sources are incorporated, thus simulating gamma ray transport. The thermo-fluid-dynamics module accounts for heterogeneity effects due to the pebble bed structure. Pipes and other components of the primary loop are modelled in 1-D geometry. Forced convection may be treated as well as natural convection in case of blower breakdown accidents. Validation of TINTE has started using the results of a comprehensive experimental program that has been performed at the Arbeitsgemeinschaft Versuchsreaktor GmbH (AVR) high temperature pebble bed reactor at Juelich. In the frame of this program power transients were initiated by varying the helium blower rotational speed or by moving the control rods. In most cases a good accordance between experiment and calculation was found. Problems in modelling the special AVR reactor geometry in two dimensions are described and suggestions for overcoming the uncertainties of experimentally determined control rod reactivities are given. The influence of different polynomial expansions of xenon cross sections to long term transients is discussed together with effects of burnup during that time. Up to now the TINTE code has proven its general applicability to operational core transients of HTR. The effects of water ingress on reactivity, fuel element corrosion and cooling gas properties are now being

  10. Lace expansion for dummies

    NARCIS (Netherlands)

    Bolthausen, Erwin; Van Der Hofstad, Remco; Kozma, Gady

    2018-01-01

    We show Green's function asymptotic upper bound for the two-point function of weakly self-Avoiding walk in d >4, revisiting a classic problem. Our proof relies on Banach algebras to analyse the lace-expansion fixed point equation and is simpler than previous approaches in that it avoids Fourier

  11. OPEC future capacity expansions

    International Nuclear Information System (INIS)

    Sandrea, I.

    2005-01-01

    This conference presentation examined OPEC future capacity expansions including highlights from 2000-2004 from the supply perspective and actions by OPEC; OPEC spare capacity in 2005/2006; medium-term capacity expansion and investments; long-term scenarios, challenges and opportunities; and upstream policies in member countries. Highlights from the supply perspective included worst than expected non-OPEC supply response; non-OPEC supply affected by a number of accidents and strikes; geopolitical tensions; and higher than expected demand for OPEC crude. OPEC's actions included closer relationship with other producers and consumers; capacity expansions in 2004 and 2005/2006; and OPEC kept the market well supplied with crude in 2004. The presentation also provided data using graphical charts on OPEC net capacity additions until 2005/2006; OPEC production versus spare capacity from 2003 to 2005; OPEC production and capacity to 2010; and change in required OPEC production from 2005-2020. Medium term expansion to 2010 includes over 60 projects. Medium-term risks such as project execution, financing, costs, demand, reserves, depletion, integration of Iraq, and geopolitical tensions were also discussed. The presentation concluded that in the long term, large uncertainties remain; the peak of world supply is not imminent; and continued and enhanced cooperation is essential to market stability. tabs., figs

  12. AUTO-EXPANSIVE FLOW

    Science.gov (United States)

    Physics suggests that the interplay of momentum, continuity, and geometry in outward radial flow must produce density and concomitant pressure reductions. In other words, this flow is intrinsically auto-expansive. It has been proposed that this process is the key to understanding...

  13. Core calculations of JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Yoshiharu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    In material testing reactors like the JMTR (Japan Material Testing Reactor) of 50 MW in Japan Atomic Energy Research Institute, the neutron flux and neutron energy spectra of irradiated samples show complex distributions. It is necessary to assess the neutron flux and neutron energy spectra of an irradiation field by carrying out the nuclear calculation of the core for every operation cycle. In order to advance core calculation, in the JMTR, the application of MCNP to the assessment of core reactivity and neutron flux and spectra has been investigated. In this study, in order to reduce the time for calculation and variance, the comparison of the results of the calculations by the use of K code and fixed source and the use of Weight Window were investigated. As to the calculation method, the modeling of the total JMTR core, the conditions for calculation and the adopted variance reduction technique are explained. The results of calculation are shown. Significant difference was not observed in the results of neutron flux calculations according to the difference of the modeling of fuel region in the calculations by K code and fixed source. The method of assessing the results of neutron flux calculation is described. (K.I.)

  14. Ice Cores

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Records of past temperature, precipitation, atmospheric trace gases, and other aspects of climate and environment derived from ice cores drilled on glaciers and ice...

  15. A core design study for 'zero-sodium-void-worth' cores

    International Nuclear Information System (INIS)

    Kawashima, Masatoshi; Suzuki, Masao; Hill, R.N.

    1992-01-01

    Recently, a number of low sodium-void-worth metal-fueled core design concepts have been proposed; to provide for flexibility in transuranic nuclide management strategy, core designs which exhibit a wide range of breeding characteristics have been developed. Two core concepts, a flat annular (transuranic burning) core and an absorber-type parfait (transuranic self-sufficient) core, are selected for this study. In this paper, the excess reactivity management schemes applied in the two designs are investigated in detail. In addition, the transient effect of reactivity insertions on the parfait core design is assessed. The upper and lower core regions in the parfait design are neutronically decoupled; however, the common coolant channel creates thermalhydraulic coupling. This combination of neutronic and thermalhydraulic characteristics leads to unique behavior in anticipated transient overpower events. (author)

  16. Expansion at Olympic Dam

    International Nuclear Information System (INIS)

    Lewis, C.

    1997-01-01

    The Olympic Dam orebody is the 6th largest copper and the single largest uranium orebody in the world. Mine production commenced in June 1988, at an annual production rate of around 45,000 tonnes of copper and 1,000 tonnes of uranium. Western Mining Corporation announced in 1996 a proposed $1.25 billion expansion of the Olympic Dam operation to raise the annual production capacity of the mine to 200,000 tonnes of copper, approximately 3,700 tonnes of uranium, 75,000 ounces of gold and 950,000 ounces of silver by 2001. Further optimisation work has identified a faster track expansion route, with an increase in the capital cost to $1.487 billion but improved investment outcome, a new target completion date of end 1999, and a new uranium output of 4,600 tonnes per annum from that date

  17. Financing electricity expansion

    International Nuclear Information System (INIS)

    Hyman, L.S.

    1994-01-01

    Expansion of electricity supply is associated with economic development. The installation and enlargement of power systems in developing countries entails a huge financial burden, however. Energy consumers in such countries must pay not only for supplies but for the cost of raising the capital for expansion on the international markets. Estimates are presented for the capital expenditure for electricity supply over the period 1990 to 2020 for the major world regions, using approximations for the cost of plant and capital and for the returns earned. These data lead to the conclusion that the five regions with the lowest per capita incomes are those which will need the major part of the capital expenditure and the highest percentage of external finance. (6 tables) (UK)

  18. Bigravity from gradient expansion

    International Nuclear Information System (INIS)

    Yamashita, Yasuho; Tanaka, Takahiro

    2016-01-01

    We discuss how the ghost-free bigravity coupled with a single scalar field can be derived from a braneworld setup. We consider DGP two-brane model without radion stabilization. The bulk configuration is solved for given boundary metrics, and it is substituted back into the action to obtain the effective four-dimensional action. In order to obtain the ghost-free bigravity, we consider the gradient expansion in which the brane separation is supposed to be sufficiently small so that two boundary metrics are almost identical. The obtained effective theory is shown to be ghost free as expected, however, the interaction between two gravitons takes the Fierz-Pauli form at the leading order of the gradient expansion, even though we do not use the approximation of linear perturbation. We also find that the radion remains as a scalar field in the four-dimensional effective theory, but its coupling to the metrics is non-trivial.

  19. Expansion of magnetic clouds

    International Nuclear Information System (INIS)

    Suess, S.T.

    1987-01-01

    Magnetic clouds are a carefully defined subclass of all interplanetary signatures of coronal mass ejections whose geometry is thought to be that of a cylinder embedded in a plane. It has been found that the total magnetic pressure inside the clouds is higher than the ion pressure outside, and that the clouds are expanding at 1 AU at about half the local Alfven speed. The geometry of the clouds is such that even though the magnetic pressure inside is larger than the total pressure outside, expansion will not occur because the pressure is balanced by magnetic tension - the pinch effect. The evidence for expansion of clouds at 1 AU is nevertheless quite strong so another reason for its existence must be found. It is demonstrated that the observations can be reproduced by taking into account the effects of geometrical distortion of the low plasma beta clouds as they move away from the Sun

  20. Evaluation of core distortion in FBR

    International Nuclear Information System (INIS)

    Ikarimoto, I.; Tanaka, M.; Okubo, Y.

    1984-01-01

    The analyses of FBR's core distortion are mainly performed in order to evaluate the following items: 1) Change of reactivity; 2) Force at pads on core assemblies; 3) Withdrawal force at refueling; 4) Loading, refueling and residual deviations of wrapper tubes (core assemblies) at the top; 5) Bowing modes of guide tubes for control rods. The analysis of core distortion are performed by using computer program for two-dimensional row deformation analysis or three-dimensional core deformation if necessary, considering these evaluated items which become design conditions. This report shows the relationship between core deformation analysis and component design, a point of view of choosing an analysis program for design considering core characteristics, and computing examples of core deformation of prototype class reactor by the above code. (author)

  1. IKEA's International Expansion

    OpenAIRE

    Harapiak, Clayton

    2013-01-01

    This case concerns a global retailing firm that is dealing with strategic management and marketing issues. Applying a scenario of international expansion, this case provides a thorough analysis of the current business environment for IKEA. Utilizing a variety of methods (e.g. SWOT, PESTLE, McKinsey Matrix) the overall objective is to provide students with the opportunity to apply their research skills and knowledge regarding a highly competitive industry to develop strategic marketing strateg...

  2. Symmetric eikonal expansion

    International Nuclear Information System (INIS)

    Matsuki, Takayuki

    1976-01-01

    Symmetric eikonal expansion for the scattering amplitude is formulated for nonrelativistic and relativistic potential scatterings and also for the quantum field theory. The first approximations coincide with those of Levy and Sucher. The obtained scattering amplitudes are time reversal invariant for all cases and are crossing symmetric for the quantum field theory in each order of approximation. The improved eikonal phase introduced by Levy and Sucher is also derived from the different approximation scheme from the above. (auth.)

  3. Series expansions without diagrams

    International Nuclear Information System (INIS)

    Bhanot, G.; Creutz, M.; Horvath, I.; Lacki, J.; Weckel, J.

    1994-01-01

    We discuss the use of recursive enumeration schemes to obtain low- and high-temperature series expansions for discrete statistical systems. Using linear combinations of generalized helical lattices, the method is competitive with diagrammatic approaches and is easily generalizable. We illustrate the approach using Ising and Potts models. We present low-temperature series results in up to five dimensions and high-temperature series in three dimensions. The method is general and can be applied to any discrete model

  4. JOYO MK-II core characteristics database

    International Nuclear Information System (INIS)

    Tabuchi, Shiro; Aoyama, Takafumi; Nagasaki, Hideaki; Kato, Yuichi

    1998-12-01

    The experimental fast reactor JOYO served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, extensive data were accumulated from the core characteristics tests conducted in thirty-one duty operations and thirteen special test operations. These core management data and core characteristics data were compiled into a database. The code system MAGI has been developed and used for core management of JOYO MK-II, and the core characteristics and the irradiation test conditions were calculated using MAGI on the basis of three dimensional diffusion theory with seven neutron energy groups. The core management data include extensive data, which were recorded on CD-ROM for user convenience. The data are specifications and configurations of the core, and for about 300 driver fuel subassemblies and about 60 uninstrumented irradiation subassemblies are core composition before and after irradiation, neutron flux, neutron fluences, fuel and control rod burn-up, and temperature and power distributions. MK-II core characteristics and test conditions were stored in the database for post analysis. Core characteristics data include excess reactivities, control rod worths, and reactivity coefficients, e.g., temperature, power and burn-up. Test conditions include both measured and calculated data for irradiation conditions. (author)

  5. Expansions for Coulomb wave functions

    NARCIS (Netherlands)

    Boersma, J.

    1969-01-01

    In this paper we derive a number of expansions for Whittaker functions, regular and irregular Coulomb wave functions. The main result consists of a new expansion for the irregular Coulomb wave functions of orders zero and one in terms of regular Coulomb wave functions. The latter expansions are

  6. Reactor core

    International Nuclear Information System (INIS)

    Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

    1990-01-01

    Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

  7. Radial expansion and multifragmentation

    International Nuclear Information System (INIS)

    Angelique, J.C.; Bizard, G.; Bougault, R.; Brou, R.; Buta, A.; Colin, J.; Cussol, D.; Durand, D.; Kerambrun, A.; Le Brun, C.; Lecolley, J.F.; Lopez, O.; Louvel, M.; Meslin, C.; Nakagawa, T.; Patry, J.P.; Peter, J.; Popescu, R.; Regimbart, R.; Steckmeyer, J.C.; Tamain, B.; Vient, E.; Yuasa-Nakagawa, K.; Wieloch, A.

    1998-01-01

    The light systems 36 Ar + 27 Al and 64 Zn + nat Ti were measured at several bombarding energies between ∼ 35 and 95 MeV/nucleon. It was found that the predominant part of the cross section is due to binary collisions. In this paper the focus is placed on the properties of the quasi-projectile nuclei. In the central collisions the excitation energies of the quasi-projectile reach values exceeding largely 10 MeV/nucleon. The slope of the high energy part of the distribution can give only an upper limit of the apparent temperature (the average temperature along the decay chain). The highly excited quasi-projectile may get rapidly fragmented rather than sequentially. The heavy fragments are excited and can emit light particles (n, p, d, t, 3 He, α,...) what perturbs additionally the spectrum of these particles. Concerning the expansion energy, one can determine the average kinetic energies of the product (in the quasi-projectile-framework) and compare with simulation values. To fit the experimental data an additional radial expansion energy is to be considered. The average expansion energy depends slightly on the impact parameter but it increases with E * / A, ranging from 0.4 to 1,2 MeV/nucleon for an excitation energy increasing from 7 to 10.5 MeV/nucleon. This collective radial energy seems to be independent of the fragment mass, what is possibly valid for the case of larger quasi-projectile masses. The origin of the expansion is to be determined. It may be due to a compression in the interaction zone at the initial stage of the collision, which propagates in the quasi-projectile and quasi-target, or else, may be due, simply, to the increase of thermal energy leading to a rapid fragment emission. The sequential de-excitation calculation overestimates light particle emission and consequently heavy residues, particularly, at higher excitation energies. This disagreement indicates that a sequential process can not account for the di-excitation of very hot nuclei

  8. Rethinking expansive learning

    DEFF Research Database (Denmark)

    Kolbæk, Ditte; Lundh Snis, Ulrika

    Abstract: This paper analyses an online community of master’s students taking a course in ICT and organisational learning. The students initiated and facilitated an educational design for organisational learning called Proactive Review in the organisation where they are employed. By using an online...... discussion forum on Google groups, they created new ways of reflecting and learning. We used netnography to select qualitative postings from the online community and expansive learning concepts for data analysis. The findings show how students changed practices of organisational learning...

  9. Load regulating expansion fixture

    International Nuclear Information System (INIS)

    Wagner, L.M.; Strum, M.J.

    1998-01-01

    A free standing self contained device for bonding ultra thin metallic films, such as 0.001 inch beryllium foils is disclosed. The device will regulate to a predetermined load for solid state bonding when heated to a bonding temperature. The device includes a load regulating feature, whereby the expansion stresses generated for bonding are regulated and self adjusting. The load regulator comprises a pair of friction isolators with a plurality of annealed copper members located there between. The device, with the load regulator, will adjust to and maintain a stress level needed to successfully and economically complete a leak tight bond without damaging thin foils or other delicate components. 1 fig

  10. SIMMER analysis of SRI high pressure bubble expansion experiments

    International Nuclear Information System (INIS)

    Rexroth, P.E.; Suo-Anttila, A.J.

    1979-01-01

    SIMMER-II was used to analyze the results of the SRI nitrogen bubble expansion experiments. Good agreement was found for all of the experiments analyzed as well as the theoretical isentropic limiting case. Scaling to a full size CRBR reactor reveals no significant scaling effects for the structureless core

  11. Thermal expansion of granite rocks

    International Nuclear Information System (INIS)

    Stephansson, O.

    1978-04-01

    The thermal expansion of rocks is strongly controlled by the thermal expansion of the minerals. The theoretical thermal expansion of the Stripa Granite is gound to be 21 . 10 -6 [deg C] -1 at 25 deg C and 38 . 10 -6 [deg C] -1 at 400 deg C. The difference in expansion for the rock forming minerals causes micro cracking at heating. The expansion due to micro cracks is found to be of the same order as the mineral expansion. Most of the micro cracks will close at pressures of the order of 10 - 20 MPa. The thermal expansion of a rock mass including the effect of joints is determined in the pilot heater test in the Stripa Mine

  12. Provincial hydro expansions

    Energy Technology Data Exchange (ETDEWEB)

    Froschauer, K J

    1993-01-01

    A study of the development of five provincial hydroelectric utilities in Canada indicates that power companies and the state invited manufacturers to use hydroelectricity and natural resources in order to diversify provincial economies. These hydro expansions also show that utilities and government designed hydro projects to serve continental requirements; serving both objectives became problematic. It is argued that when the Canadian state and firms such as utilities use hydro expansions to serve both continentalism and industrialization, then at best they foster dependent industrialization and staple processing. At worst, they overbuild the infrastructure to generate provincial surplus energy for continental, rather than national, integration. Hydro developments became subject to state intervention in Canada mainly through the failures of private utilities to provide power for the less-lucrative industrial markets within provincial subregions. Although the state and utilities invited foreign firms to manufacture hydro equipment within the provinces and others to use electricity to diversify production beyond resource processing, such a diversification did not occur. Since 1962, ca 80% of industrial energy was used to semi-process wood-derived products, chemicals, and metals. The idea for a national power network became undermined by interprovincial political-economic factors and since 1963, the federal national/continential power policy prevailed. 187 refs., 6 figs., 52 tabs.

  13. Measuring of tube expansion

    International Nuclear Information System (INIS)

    Vogeleer, J. P.

    1985-01-01

    The expansion of the primary tubes or sleeves of the steam generator of a nuclear reactor plant are measured while the tubes or sleeves are being expanded. A primary tube or sleeve is expanded by high pressure of water which flows through a channel in an expander body. The water is supplied through an elongated conductor and is introduced through a connector on the shank connected to the conductor at its outer end. A wire extends through the mandrel and through the conductor to the end of the connector. At its inner end the wire is connected to a tapered pin which is subject to counteracting forces produced by the pressure of the water. The force on the side where the wire is connected to the conductor is smaller than on the opposite side. The tapered pin is moved in the direction of the higher force and extrudes the wire outwardly of the conductor. The tapered surface of the tapered pin engages transverse captive plungers which are maintained in engagement with the expanding tube or sleeve as they are moved outwardly by the tapered pin. The wire and the connector extend out of the generator and, at its outer end, the wire is connected to an indicator which measures the extent to which the wire is moved by the tapered pin, thus measuring the expansion of the tube or sleeve as it progresses

  14. INDIAN POINT REACTOR REACTIVITY AND FLUX DISTRIBUTION MEASUREMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Batch, M. L.; Fischer, F. E.

    1963-11-15

    The reactivity of the Indian Point core was measured near zero reactivity at various shim and control rod patterns. Flux distribution measurements were also made, and the results are expressed in terms of power peaking factors and normalized detector response during rod withdrawal. (D.L.C.)

  15. Surface reactivity and layer analysis of chemisorbed reaction films in ...

    Indian Academy of Sciences (India)

    Administrator

    Surface reactivity and layer analysis of chemisorbed reaction films in ... in the nitrogen environment. Keywords. Surface reactivity ... sium (Na–K) compounds in the coating or core of the ..... Barkshire I R, Pruton M and Smith G C 1995 Appl. Sur.

  16. Reactive Kripke semantics

    CERN Document Server

    Gabbay, Dov M

    2013-01-01

    This text offers an extension to the traditional Kripke semantics for non-classical logics by adding the notion of reactivity. Reactive Kripke models change their accessibility relation as we progress in the evaluation process of formulas in the model. This feature makes the reactive Kripke semantics strictly stronger and more applicable than the traditional one. Here we investigate the properties and axiomatisations of this new and most effective semantics, and we offer a wide landscape of applications of the idea of reactivity. Applied topics include reactive automata, reactive grammars, rea

  17. Modelling The Effects of Aggregate Size on Alkali Aggregate Reaction Expansion

    Directory of Open Access Journals (Sweden)

    N. Z. Sekrane

    2014-06-01

    Full Text Available This work aims at developing models to predict the potential expansion of concrete containing alkali-reactive aggregates. The paper gives measurements in order to provide experimental data concerning the effect of particle size of an alkali-reactive siliceous limestone on mortar expansion. Results show that no expansion was measured on the mortars using small particles (0.5-1.0 mm while the particles (1.0–2.0 mm gave the largest expansions (0.217%. Two models are proposed, the first one studies the correlations between the measured expansions and the size of aggregates, the second one calculates the thickness of the porous zone necessary to take again all the volume of the gel created.

  18. Correlations among FBR core characteristics for various fuel compositions

    International Nuclear Information System (INIS)

    Maruyama, Shuhei; Ohki, Shigeo; Okubo, Tsutomu; Kawashima, Katsuyuki; Mizuno, Tomoyasu

    2012-01-01

    In the design of a fast breeder reactor (FBR) core for the light water reactor (LWR) to FBR transition stage, it is indispensable to grasp the effect of a wide range of fuel composition variations on the core characteristics. This study finds good correlations between burnup reactivity and safety parameters, such as the sodium void reactivity and Doppler coefficient, for various fuel compositions and determines the mechanisms behind these correlations with the aid of sensitivity analyses. It is clarified that the Doppler coefficient is actually correlated with the other core characteristics by considering the constraint imposed by the requirement of sustaining criticality on the fuel composition variations. These correlations make it easy to specify the various properties ranges for core reactivity control and core safety, which are important for core design in determining the core specifications and performance. They provide significant information for FBR core design for the transition stage. Moreover, as an application of the above-mentioned correlations, a simplified burnup reactivity index is developed for rapid and rational estimation of the core characteristic variations. With the use of this index and these correlations, the core characteristic variations can be estimated for various fuel compositions without repeating the core calculations. (author)

  19. The causal perturbation expansion revisited: Rescaling the interacting Dirac sea

    International Nuclear Information System (INIS)

    Finster, Felix; Grotz, Andreas

    2010-01-01

    The causal perturbation expansion defines the Dirac sea in the presence of a time-dependent external field. It yields an operator whose image generalizes the vacuum solutions of negative energy and thus gives a canonical splitting of the solution space into two subspaces. After giving a self-contained introduction to the ideas and techniques, we show that this operator is, in general, not idempotent. We modify the standard construction by a rescaling procedure giving a projector on the generalized negative-energy subspace. The resulting rescaled causal perturbation expansion uniquely defines the fermionic projector in terms of a series of distributional solutions of the Dirac equation. The technical core of the paper is to work out the combinatorics of the expansion in detail. It is also shown that the fermionic projector with interaction can be obtained from the free projector by a unitary transformation. We finally analyze the consequences of the rescaling procedure on the light-cone expansion.

  20. The causal perturbation expansion revisited: Rescaling the interacting Dirac sea

    Science.gov (United States)

    Finster, Felix; Grotz, Andreas

    2010-07-01

    The causal perturbation expansion defines the Dirac sea in the presence of a time-dependent external field. It yields an operator whose image generalizes the vacuum solutions of negative energy and thus gives a canonical splitting of the solution space into two subspaces. After giving a self-contained introduction to the ideas and techniques, we show that this operator is, in general, not idempotent. We modify the standard construction by a rescaling procedure giving a projector on the generalized negative-energy subspace. The resulting rescaled causal perturbation expansion uniquely defines the fermionic projector in terms of a series of distributional solutions of the Dirac equation. The technical core of the paper is to work out the combinatorics of the expansion in detail. It is also shown that the fermionic projector with interaction can be obtained from the free projector by a unitary transformation. We finally analyze the consequences of the rescaling procedure on the light-cone expansion.

  1. Thermal expansion of coking coals

    Energy Technology Data Exchange (ETDEWEB)

    Orlik, M.; Klimek, J. (Vyzkumny a Zkusebni Ustav Nova Hut, Ostrava (Czechoslovakia))

    1992-12-01

    Analyzes expansion of coal mixtures in coke ovens during coking. Methods for measuring coal expansion on both a laboratory and pilot plant scale are comparatively evaluated. The method, developed, tested and patented in Poland by the Institute for Chemical Coal Processing in Zabrze (Polish standard PN-73/G-04522), is discussed. A laboratory device developed by the Institute for measuring coal expansion is characterized. Expansion of black coal from 10 underground mines in the Ostrava-Karvina coal district and from 9 coal mines in the Upper Silesia basin in Poland is comparatively evaluated. Investigations show that coal expansion reaches a maximum for coal types with a volatile matter ranging from 20 to 25%. With increasing volatile matter in coal, its expansion decreases. Coal expansion increases with increasing swelling index. Coal expansion corresponds with coal dilatation. With increasing coal density its expansion increases. Coal mixtures should be selected in such a way that their expansion does not cause a pressure exceeding 40 MPa. 11 refs.

  2. Conceptual design study of LMFBR core with carbide fuel

    International Nuclear Information System (INIS)

    Tezuka, H.; Hojuyama, T.; Osada, H.; Ishii, T.; Hattori, S.; Nishimura, T.

    1987-01-01

    Carbide fuel is a hopeful candidate for demonstration FBR(DFBR) fuel from the plant cost reduction point of view. High thermal conductivity and high heavy metal content of carbide fuel lead to high linear heat rate and high breeding ratio. We have analyzed carbide fuel core characteristics and have clarified the concept of carbide fuel core. By survey calculation, we have obtained a correlation map between core parameters and core characteristics. From the map, we have selected a high efficiency core whose features are better than those of an oxide core, and have obtained reactivity coefficients. The core volume and the reactor fuel inventory are approximately 20% smaller, and the burn-up reactivity loss is 50% smaller compared with the oxide fuel core. These results will reduce the capital cost. The core reactivity coefficients are similar to the conventional oxide DFBR's. Therefore the carbide fuel core is regarded as safe as the oxide core. Except neutron fluence, the carbide fuel core has better nuclear features than the oxide core

  3. Identity Expansion and Transcendence

    Directory of Open Access Journals (Sweden)

    William Sims Bainbridge

    2014-05-01

    Full Text Available Emerging developments in communications and computing technology may transform the nature of human identity, in the process rendering obsolete the traditional philosophical and scientific frameworks for understanding the nature of individuals and groups.  Progress toward an evaluation of this possibility and an appropriate conceptual basis for analyzing it may be derived from two very different but ultimately connected social movements that promote this radical change. One is the governmentally supported exploration of Converging Technologies, based in the unification of nanoscience, biology, information science and cognitive science (NBIC. The other is the Transhumanist movement, which has been criticized as excessively radical yet is primarily conducted as a dignified intellectual discussion within a new school of philosophy about human enhancement.  Together, NBIC and Transhumanism suggest the immense transformative power of today’s technologies, through which individuals may explore multiple identities by means of online avatars, semi-autonomous intelligent agents, and other identity expansions.

  4. Three-dimensional static and dynamic reactor calculations by the nodal expansion method

    International Nuclear Information System (INIS)

    Christensen, B.

    1985-05-01

    This report reviews various method for the calculation of the neutron-flux- and power distribution in an nuclear reactor. The nodal expansion method (NEM) is especially described in much detail. The nodal expansion method solves the diffusion equation. In this method the reactor core is divided into nodes, typically 10 to 20 cm in each direction, and the average flux in each node is calculated. To obtain the coupling between the nodes the local flux inside each node is expressed by use of a polynomial expansion. The expansion is one-dimensional, so inside each node such three expansions occur. To calculate the expansion coefficients it is necessary that the polynomial expansion is a solution to the one-dimensional diffusion equation. When the one-dimensional diffusion equation is established a term with the transversal leakage occur, and this term is expanded after the same polynomials. The resulting equation system with the expansion coefficients as the unknowns is solved with weigthed residual technique. The nodal expansion method is built into a computer program (also called NEM), which is divided into two parts, one part for steady-state calculations and one part for dynamic calculations. It is possible to take advantage of symmetry properties of the reactor core. The program is very flexible with regard to the number of energy groups, the node size, the flux expansion order and the transverse leakage expansion order. The boundary of the core is described by albedos. The program and input to it are described. The program is tested on a number of examples extending from small theoretical one up to realistic reactor cores. Many calculations are done on the wellknown IAEA benchmark case. The calculations have tested the accuracy and the computing time for various node sizes and polynomial expansions. In the dynamic examples various strategies for variation of the time step-length have been tested. (author)

  5. Reactive perforating collagenosis

    Directory of Open Access Journals (Sweden)

    Yadav Mukesh

    2009-01-01

    Full Text Available Reactive perforating collagenosis is a rare cutaneous disorder of unknown etiology. We hereby describe a case of acquired reactive perforating collagenosis in a patient of diabetes and chronic renal failure.

  6. Evolution of density-dependent movement during experimental range expansions.

    Science.gov (United States)

    Fronhofer, E A; Gut, S; Altermatt, F

    2017-12-01

    Range expansions and biological invasions are prime examples of transient processes that are likely impacted by rapid evolutionary changes. As a spatial process, range expansions are driven by dispersal and movement behaviour. Although it is widely accepted that dispersal and movement may be context-dependent, for instance density-dependent, and best represented by reaction norms, the evolution of density-dependent movement during range expansions has received little experimental attention. We therefore tested current theory predicting the evolution of increased movement at low densities at range margins using highly replicated and controlled range expansion experiments across multiple genotypes of the protist model system Tetrahymena thermophila. Although rare, we found evolutionary changes during range expansions even in the absence of initial standing genetic variation. Range expansions led to the evolution of negatively density-dependent movement at range margins. In addition, we report the evolution of increased intrastrain competitive ability and concurrently decreased population growth rates in range cores. Our findings highlight the importance of understanding movement and dispersal as evolving reaction norms and plastic life-history traits of central relevance for range expansions, biological invasions and the dynamics of spatially structured systems in general. © 2017 European Society For Evolutionary Biology. Journal of Evolutionary Biology © 2017 European Society For Evolutionary Biology.

  7. Reactivity on the Web

    OpenAIRE

    Bailey, James; Bry, François; Eckert, Michael; Patrânjan, Paula Lavinia

    2005-01-01

    Reactivity, the ability to detect simple and composite events and respond in a timely manner, is an essential requirement in many present-day information systems. With the emergence of new, dynamic Web applications, reactivity on the Web is receiving increasing attention. Reactive Web-based systems need to detect and react not only to simple events but also to complex, real-life situations. This paper introduces XChange, a language for programming reactive behaviour on the Web,...

  8. Characteristic test of initial HTTR core

    International Nuclear Information System (INIS)

    Nojiri, Naoki; Shimakawa, Satoshi; Fujimoto, Nozomu; Goto, Minoru

    2004-01-01

    This paper describes the results of core physics test in start-up and power-up of the HTTR. The tests were conducted in order to ensure performance and safety of the high temperature gas cooled reactor, and was carried out to measure the critical approach, the excess reactivity, the shutdown margin, the control rod worth, the reactivity coefficient, the neutron flux distribution and the power distribution. The expected core performance and the required reactor safety characteristics were verified from the results of measurements and calculations

  9. Temperature coefficients of reactivity in the fourth loading of ZENITH

    International Nuclear Information System (INIS)

    Caro Manso, R.; Freemantle, R.G.; Rogers, J.D.

    1962-10-01

    Measurements have been made of the temperature coefficients of reactivity associated with the core plus end reflectors and the side reflector of the fourth core loading of ZENITH, which had a carbon/U235 atomic ratio of 7788 and no other absorber. (author)

  10. Temperature coefficients of reactivity in the fourth loading of ZENITH

    Energy Technology Data Exchange (ETDEWEB)

    Caro Manso, R; Freemantle, R G; Rogers, J D [Graphite Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1962-10-15

    Measurements have been made of the temperature coefficients of reactivity associated with the core plus end reflectors and the side reflector of the fourth core loading of ZENITH, which had a carbon/U235 atomic ratio of 7788 and no other absorber. (author)

  11. Core BPEL

    DEFF Research Database (Denmark)

    Hallwyl, Tim; Højsgaard, Espen

    The Web Services Business Process Execution Language (WS-BPEL) is a language for expressing business process behaviour based on web services. The language is intentionally not minimal but provides a rich set of constructs, allows omission of constructs by relying on defaults, and supports language......, does not allow omissions, and does not contain ignorable elements. We do so by identifying syntactic sugar, including default values, and ignorable elements in WS-BPEL. The analysis results in a translation from the full language to the core subset. Thus, we reduce the effort needed for working...

  12. Thermal expansion of beryllium oxide

    International Nuclear Information System (INIS)

    Solodukhin, A.V.; Kruzhalov, A.V.; Mazurenko, V.G.; Maslov, V.A.; Medvedev, V.A.; Polupanova, T.I.

    1987-01-01

    Precise measurements of temperature dependence of the coefficient of linear expansion in the 22-320 K temperature range on beryllium oxide monocrystals are conducted. A model of thermal expansion is suggested; the range of temperature dependence minimum of the coefficient of thermal expansion is well described within the frames of this model. The results of the experiment may be used for investigation of thermal stresses in crystals

  13. Monadic Functional Reactive Programming

    NARCIS (Netherlands)

    A.J. van der Ploeg (Atze); C Shan

    2013-01-01

    htmlabstractFunctional Reactive Programming (FRP) is a way to program reactive systems in functional style, eliminating many of the problems that arise from imperative techniques. In this paper, we present an alternative FRP formulation that is based on the notion of a reactive computation: a

  14. Electric Grid Expansion Planning with High Levels of Variable Generation

    Energy Technology Data Exchange (ETDEWEB)

    Hadley, Stanton W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); You, Shutang [Univ. of Tennessee, Knoxville, TN (United States); Shankar, Mallikarjun [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Liu, Yilu [Univ. of Tennessee, Knoxville, TN (United States); Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-02-01

    in the EI system. Incorporating more details of renewables in expansion planning will inevitably increase the computational burden. Therefore, high performance computing (HPC) techniques are urgently needed for power system operation and planning optimization. As a scoping study task, this project tested some preliminary parallel computation techniques such as breaking down the simulation task into several sub-tasks based on chronology splitting or sample splitting, and then assigning these sub-tasks to different cores. Testing results show significant time reduction when a simulation task is split into several sub-tasks for parallel execution.

  15. Full MOX core for PWRs

    International Nuclear Information System (INIS)

    Puill, A.; Aniel-Buchheit, S.

    1997-01-01

    Plutonium management is a major problem of the back end of the fuel cycle. Fabrication costs must be reduced and plant operation simplified. The design of a full MOX PWR core would enable the number of reactors devoted to plutonium recycling to be reduced and fuel zoning to be eliminated. This paper is a contribution to the feasibility studies for achieving such a core without fundamental modification of the current design. In view of the differences observed between uranium and plutonium characteristics it seems necessary to reconsider the safety of a MOX-fuelled PWR. Reduction of the control worth and modification of the moderator density coefficient are the main consequences of using MOX fuel in a PWR. The core reactivity change during a draining or a cooling is thus of prime interest. The study of core global draining leads to the following conclusion: only plutonium fuels of very poor quality (i.e. with low fissile content) cannot be used in a 900 MWe PWR because of a positive global voiding reactivity effect. During a cooling accident, like an spurious opening of a secondary-side valve, the hypothetical return to criticality of a 100% MOX core controlled by means of 57 control rod clusters (made of hafnium-clad B 4 C rods with a 90% 10 B content) depends on the isotopic plutonium composition. But safety criteria can be complied with for all isotopic compositions provided the 10 B content of the soluble boron is increased to a value of 40%. Core global draining and cooling accidents do not present any major obstacle to the feasibility of a 100% MOX PWR, only minor hardware modifications will be required. (author)

  16. Quadratic reactivity fuel cycle model

    International Nuclear Information System (INIS)

    Lewins, J.D.

    1985-01-01

    For educational purposes it is highly desirable to provide simple yet realistic models for fuel cycle and fuel economy. In particular, a lumped model without recourse to detailed spatial calculations would be very helpful in providing the student with a proper understanding of the purposes of fuel cycle calculations. A teaching model for fuel cycle studies based on a lumped model assuming the summability of partial reactivities with a linear dependence of reactivity usefully illustrates fuel utilization concepts. The linear burnup model does not satisfactorily represent natural enrichment reactors. A better model, showing the trend of initial plutonium production before subsequent fuel burnup and fission product generation, is a quadratic fit. The study of M-batch cycles, reloading 1/Mth of the core at end of cycle, is now complicated by nonlinear equations. A complete account of the asymptotic cycle for any order of M-batch refueling can be given and compared with the linear model. A complete account of the transient cycle can be obtained readily in the two-batch model and this exact solution would be useful in verifying numerical marching models. It is convenient to treat the parabolic fit rho = 1 - tau 2 as a special case of the general quadratic fit rho = 1 - C/sub tau/ - (1 - C)tau 2 in suitably normalized reactivity and cycle time units. The parabolic results are given in this paper

  17. Calculation of the Reactivity Equivalence of Control Rods in the Second Charge of the HBWR

    International Nuclear Information System (INIS)

    Weissglas, P.

    1960-11-01

    Full text: Using current methods the reactivity equivalence of 19 31 and 37 centrally located control rods in the second charge of the HBWR has been calculated. An estimate of the available excess reactivity with clean cold core has also been made. Insertion depth was taken as 0, l/3, 2/3 and 3/3 of the core length

  18. Calculation of the Reactivity Equivalence of Control Rods in the Second Charge of the HBWR.

    Energy Technology Data Exchange (ETDEWEB)

    Weissglas, P [The Swedish State Power Board, Stockholm (Sweden)

    1960-11-15

    Full text: Using current methods the reactivity equivalence of 19 31 and 37 centrally located control rods in the second charge of the HBWR has been calculated. An estimate of the available excess reactivity with clean cold core has also been made. Insertion depth was taken as 0, l/3, 2/3 and 3/3 of the core length.

  19. Digital reactivity meter

    International Nuclear Information System (INIS)

    Akkus, B.; Anac, H.; Alsan, S.; Erk, S.

    1991-01-01

    Nowadays, various digital methods making use of microcomputers for neutron detector signals and determining the reactivity by numerical calculations are used in reactor control systems in place of classical reactivity meters. In this work, a calculation based on the ''The Time Dependent Transport Equation'' has been developed for determining the reactivity numerically. The reactivity values have been obtained utilizing a computer-based data acquisition and control system and compared with the analog reactivity meter values as well as the values calculated from the ''Inhour Equation''

  20. Critical experiments of JMTRC MEU cores

    International Nuclear Information System (INIS)

    Nagaoka, Y.; Takeda, K.; Shimakawa, S.; Koike, S.; Oyamada, R.

    1984-01-01

    The JMTRC, the critical facility of the Japan Materials Testing Reactor (JMTR), went critical on August 29, 1983, with 14 medium enriched uranium (MEU, 45%) fuel elements. Experiments are now being carried out to measure the change in various reactor characteristics between the previous HEU core and the new MEU fueled core. This paper describes the results obtained thus far on critical mass, excess reactivity, control rod worths and flux distribution, including preliminary neutronics calculations for the experiments using the SRAC code. (author)

  1. Method of allowing for resonances in calculating reactivity values

    International Nuclear Information System (INIS)

    Kumpf, H.

    1985-01-01

    On the basis of the integral transport equation for the source density an expression has been derived for calculating reactivity values taking resonances in the core and in the sample into account. The model has been used for evaluating reactivities measured in the Rossendorf SEG IV configuration. It is shown that the influence of resonances in the core can be kept tolerable, if a sufficiently thick buffer zone of only slightly absorbing non-resonant material is arranged between the sample and the core. (author)

  2. Initial charge reactor core

    International Nuclear Information System (INIS)

    Kiyono, Takeshi

    1984-01-01

    Purpose: To effectivity burn fuels and improve the economical performance in an inital charge reactor core of BWR type reactors or the likes. Constitution: In a reactor core constituted with a plurality of fuel assemblies which are to be partially replaced upon fuel replacement, the density of the fissionable materials and the moderator - fuel ratio of a fuel assembly is set corresponding to the period till that fuel assembly is replaced, in which the density of the nuclear fissionable materials is lowered and the moderator - fuel ratio is increased for the fuel assembly with a shorter period from the fueling to the fuel exchange and, while on the other hand, the density of the fissionable materials is increased and the moderator - fuel ratio is decreased for the fuel assembly with a longer period from the fueling to the replacement. Accordingly, since the moderator - fuel ratio is increased for the fuel assembly to be replaced in a shorter period, the neutrons moderating effect is increased to increase the reactivity. (Horiuchi, T.)

  3. JOYO MK-II core characteristics database

    International Nuclear Information System (INIS)

    Ohkawachi, Yasushi; Maeda, Shigetaka; Sekine, Takashi; Aoyama, Takafumi

    2003-04-01

    The 'JOYO' MK-II core characteristics database was compiled and published in 1998. Comments and requests from many users led to the creation of a revised edition. The revisions include changes to the MAGI calculation code system to use the 70 group JFS-3-J3.2 constant set processed from the JENDL-3.2 library. Total control rod worth, reactor kinetic parameters and the MK-II core performance test results were included per user's requests. The core characteristics obtained from the 32 nd to 35 th operational cycles, which were conducted in the MK-III transition core, were newly added in this revised version. The MK-II core management data and core characteristics data were recorded to CD-ROM for user convenience. The Configuration Data' include the core arrangement and refueling record for each operational cycle. The 'Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of 362 driver fuel subassemblies and 69 irradiation test subassemblies. The 'Output Data' contain the calculated neutron flux, gamma flux, power density, linear heat rate, coolant and fuel temperature distribution of all the fuel subassemblies at the beginning and end of each operational cycle. The 'Core Characteristics Data' include the measured excess reactivity, control rod worth calibration curve, and reactivity coefficients of temperature, power and burn-up. (author)

  4. Renormalization group and Mayer expansions

    International Nuclear Information System (INIS)

    Mack, G.

    1984-02-01

    Mayer expansions promise to become a powerful tool in exact renormalization group calculations. Iterated Mayer expansions were sucessfully used in the rigorous analysis of 3-dimensional U(1) lattice gauge theory by Goepfert and the author, and it is hoped that they will also be useful in the 2-dimensional nonlinear sigma-model, and elsewhere. (orig.)

  5. Isotropic Negative Thermal Expansion Metamaterials.

    Science.gov (United States)

    Wu, Lingling; Li, Bo; Zhou, Ji

    2016-07-13

    Negative thermal expansion materials are important and desirable in science and engineering applications. However, natural materials with isotropic negative thermal expansion are rare and usually unsatisfied in performance. Here, we propose a novel method to achieve two- and three-dimensional negative thermal expansion metamaterials via antichiral structures. The two-dimensional metamaterial is constructed with unit cells that combine bimaterial strips and antichiral structures, while the three-dimensional metamaterial is fabricated by a multimaterial 3D printing process. Both experimental and simulation results display isotropic negative thermal expansion property of the samples. The effective coefficient of negative thermal expansion of the proposed models is demonstrated to be dependent on the difference between the thermal expansion coefficient of the component materials, as well as on the circular node radius and the ligament length in the antichiral structures. The measured value of the linear negative thermal expansion coefficient of the three-dimensional sample is among the largest achieved in experiments to date. Our findings provide an easy and practical approach to obtaining materials with tunable negative thermal expansion on any scale.

  6. Renormalization group and mayer expansions

    International Nuclear Information System (INIS)

    Mack, G.

    1984-01-01

    Mayer expansions promise to become a powerful tool in exact renormalization group calculations. Iterated Mayer expansions were sucessfully used in the rigorous analysis of 3-dimensional U (1) lattice gauge theory by Gopfert and the author, and it is hoped that they will also be useful in the 2-dimensional nonlinear σ-model, and elsewhere

  7. On summation of perturbation expansions

    International Nuclear Information System (INIS)

    Horzela, A.

    1985-04-01

    The problem of the restoration of physical quantities defined by divergent perturbation expansions is analysed. The Pad'e and Borel summability is proved for alternating perturbation expansions with factorially growing coefficients. The proof is based on the methods of the classical moments theory. 17 refs. (author)

  8. In-core Instrument Subcritical Verification (INCISV) - Core Design Verification Method - 358

    International Nuclear Information System (INIS)

    Prible, M.C.; Heibel, M.D.; Conner, S.L.; Sebastiani, P.J.; Kistler, D.P.

    2010-01-01

    According to the standard on reload startup physics testing, ANSI/ANS 19.6.1, a plant must verify that the constructed core behaves sufficiently close to the designed core to confirm that the various safety analyses bound the actual behavior of the plant. A large portion of this verification must occur before the reactor operates at power. The INCISV Core Design Verification Method uses the unique characteristics of a Westinghouse Electric Company fixed in-core self powered detector design to perform core design verification after a core reload before power operation. A Vanadium self powered detector that spans the length of the active fuel region is capable of confirming the required core characteristics prior to power ascension; reactivity balance, shutdown margin, temperature coefficient and power distribution. Using a detector element that spans the length of the active fuel region inside the core provides a signal of total integrated flux. Measuring the integrated flux distributions and changes at various rodded conditions and plant temperatures, and comparing them to predicted flux levels, validates all core necessary core design characteristics. INCISV eliminates the dependence on various corrections and assumptions between the ex-core detectors and the core for traditional physics testing programs. This program also eliminates the need for special rod maneuvers which are infrequently performed by plant operators during typical core design verification testing and allows for safer startup activities. (authors)

  9. Neutronic Design of KALIMER-600 Core with Moderator Rods

    International Nuclear Information System (INIS)

    Ser Gi Hong; Sang Ji Kim; Hoon Song; Yeong Il Kim

    2004-01-01

    Recently, the liquid-metal reactor research team of the Korea Atomic Energy Research Institute (KAERI) designed a 600 MWe sodium-cooled, metallic fueled fast reactor meeting the goals of Generation-IV, such as economics and proliferation resistance. In this paper, the core design analysis and its performance are reported. The core is designed to have a conversion ratio slightly larger than unity with no blanket assemblies in order not to produce an excess amount of high grade plutonium and to have no need for external feeds of fissile materials. To mitigate the sodium void reactivity of the fuel-self-sufficient core with no blanket assemblies, several design changes from a reference core are tried; reduction of the active core height, annular type cores with central dummy assemblies, and the use of moderator (BeO or ZrH 2 ) rods. As a result of the analysis, it is found that of the considered designs the use of moderator rods for the softening of the core neutron spectrum is the best choice for reducing the sodium void worth with the smallest changes from the reference fuel and assembly designs. The core analysis shows that the sodium void reactivity is reduced by ∼2$ in comparison with the reference core and the core has a much more negative fuel temperature reactivity feedback in comparison with the reference core. (authors)

  10. Core clamping device for a nuclear reactor

    International Nuclear Information System (INIS)

    Guenther, R.W.

    1974-01-01

    The core clamping device for a fast neutron reactor includes clamps to support the fuel zone against the pressure vessel. The clamps are arranged around the circumference of the core. They consist of torsion bars arranged parallel at some distance around the core with lever arms attached to the ends whose force is directed in the opposite direction, pressing against the wall of the pressure vessel. The lever arms and pressure plates also actuated by the ends of the torsion bars transfer the stress, the pressure plates acting upon the fuel elements or fuel assemblies. Coupling between the ends of the torsion bars and the pressure plates is achieved by end carrier plates directly attached to the torsion bars and radially movable. This clamping device follows the thermal expansions of the core, allows specific elements to be disengaged in sections and saves space between the core and the neutron reflectors. (DG) [de

  11. Comparison of the unitary pole and Adhikari-Sloan expansions in the three nucleon system

    International Nuclear Information System (INIS)

    Afnan, I.R.; Birrell, N.D.

    1977-01-01

    The binding energy of 3 H, percentage S-, S'- and D-state probability, and charge form factor of 3 He are calculated using the unitary pole and Adhikari-Sloan separable expansions to the Reid soft core potential. Comparison of the results for the two separable expansions show that the expansion of Adhikari and Sloan has the better convergence property, and the lowest rank expansion considered (equivalent to the unitary pole approximation) gives a good approximation to the binding energy of 3 H and the charge form factor of 3 He, even at large momentum transfer (K 2 -2 ). (Author)

  12. Digital reactivity meter construction based on PC

    International Nuclear Information System (INIS)

    Yusi-Eko-Yulianto; Kristedjo-Kurnianto

    2003-01-01

    The reactivitymeter is a core reactivity measuring equipment, which inform the reactor operator the neutron flux development in the core. This digital reactivitymeter is needed to replace analog reactivitymeter, whenever it fails in the future. The replacement of thus reactivitymeter can keep the continuation of reactor operation. The digital reactivitymeter is constructed by using the digital signal processing and computer. Thus real time signal processing is displayed on the monitor graphically. This reactivitymeter has been tested in RSG-GAS and perform a good work. This performance is worthy to use this digital reactivitymeter for RSG-GAS operation

  13. IAEA Technical Meeting on Innovative Fast Reactor Designs with Enhanced Negative Reactivity Feedback Features. Working Material

    International Nuclear Information System (INIS)

    2012-01-01

    The objective of the TM was to review and discuss the safety characteristics and the performances of the core of innovative fast reactor concepts, as well as to present the ongoing R&D activities in the area of core design and advanced simulation tools and methods for fast reactor core physics analysis. The focus was on fast spectrum cores optimized for actinide utilization and transmutation and, in particular, on core designs with enhanced negative reactivity feedback effects

  14. IAEA Technical Meeting on Innovative Fast Reactor Designs with Enhanced Negative Reactivity Feedback Features. Presentations

    International Nuclear Information System (INIS)

    2012-01-01

    The objective of the TM is to review and discuss the safety characteristics and the performances of the core of innovative fast reactor concepts, as well as to present the ongoing R&D activities in the area of core design and advanced simulation tools and methods for fast reactor core physics analysis. The focus is on fast spectrum cores optimized for actinide utilization and transmutation and, in particular, on core designs with enhanced negative reactivity feedback effects

  15. Reactive Programming in Java

    CERN Document Server

    CERN. Geneva

    2017-01-01

    Reactive Programming in gaining a lot of excitement. Many libraries, tools, and frameworks are beginning to make use of reactive libraries. Besides, applications dealing with big data or high frequency data can benefit from this programming paradigm. Come to this presentation to learn about what reactive programming is, what kind of problems it solves, how it solves them. We will take an example oriented approach to learning the programming model and the abstraction.

  16. Plasma expansion: fundamentals and applications

    International Nuclear Information System (INIS)

    Engeln, R; Mazouffre, S; Vankan, P; Bakker, I; Schram, D C

    2002-01-01

    The study of plasma expansion is interesting from a fundamental point of view as well as from a more applied point of view. We here give a short overview of the way properties like density, velocity and temperature behave in an expanding thermal plasma. Experimental data show that the basic phenomena of plasma expansion are to some extent similar to those of the expansion of a hot neutral gas. From the application point of view, we present first results on the use of an expanding thermal plasma in the plasma-activated catalysis of ammonia, from N 2 -H 2 mixtures

  17. Development of a computer code for neutronic calculations of a hexagonal lattice of nuclear reactor using the flux expansion nodal method

    Directory of Open Access Journals (Sweden)

    Mohammadnia Meysam

    2013-01-01

    Full Text Available The flux expansion nodal method is a suitable method for considering nodalization effects in node corners. In this paper we used this method to solve the intra-nodal flux analytically. Then, a computer code, named MA.CODE, was developed using the C# programming language. The code is capable of reactor core calculations for hexagonal geometries in two energy groups and three dimensions. The MA.CODE imports two group constants from the WIMS code and calculates the effective multiplication factor, thermal and fast neutron flux in three dimensions, power density, reactivity, and the power peaking factor of each fuel assembly. Some of the code's merits are low calculation time and a user friendly interface. MA.CODE results showed good agreement with IAEA benchmarks, i. e. AER-FCM-101 and AER-FCM-001.

  18. Exploring the conformational and reactive dynamics of biomolecules in solution using an extended version of the glycine reactive force field

    DEFF Research Database (Denmark)

    Monti, Susanna; Corozzi, Alessandro; Fristrup, Peter

    2013-01-01

    In order to describe possible reaction mechanisms involving amino acids, and the evolution of the protonation state of amino acid side chains in solution, a reactive force field (ReaxFF-based description) for peptide and protein simulations has been developed as an expansion of the previously rep...

  19. Relations among several nuclear and electronic density functional reactivity indexes

    Science.gov (United States)

    Torrent-Sucarrat, Miquel; Luis, Josep M.; Duran, Miquel; Toro-Labbé, Alejandro; Solà, Miquel

    2003-11-01

    An expansion of the energy functional in terms of the total number of electrons and the normal coordinates within the canonical ensemble is presented. A comparison of this expansion with the expansion of the energy in terms of the total number of electrons and the external potential leads to new relations among common density functional reactivity descriptors. The formulas obtained provide explicit links between important quantities related to the chemical reactivity of a system. In particular, the relation between the nuclear and the electronic Fukui functions is recovered. The connection between the derivatives of the electronic energy and the nuclear repulsion energy with respect to the external potential offers a proof for the "Quantum Chemical le Chatelier Principle." Finally, the nuclear linear response function is defined and the relation of this function with the electronic linear response function is given.

  20. Growth outside the core.

    Science.gov (United States)

    Zook, Chris; Allen, James

    2003-12-01

    Growth in an adjacent market is tougher than it looks; three-quarters of the time, the effort fails. But companies can change those odds dramatically. Results from a five-year study of corporate growth conducted by Bain & Company reveal that adjacency expansion succeeds only when built around strong core businesses that have the potential to become market leaders. And the best place to look for adjacency opportunities is inside a company's strongest customers. The study also found that the most successful companies were able to consistently, profitably outgrow their rivals by developing a formula for pushing out the boundaries of their core businesses in predictable, repeatable ways. Companies use their repeatability formulas to expand into any number of adjacencies. Some companies make repeated geographic moves, as Vodafone has done in expanding from one geographic market to another over the past 13 years, building revenues from $1 billion in 1990 to $48 billion in 2003. Others apply a superior business model to new segments. Dell, for example, has repeatedly adapted its direct-to-customer model to new customer segments and new product categories. In other cases, companies develop hybrid approaches. Nike executed a series of different types of adjacency moves: it expanded into adjacent customer segments, introduced new products, developed new distribution channels, and then moved into adjacent geographic markets. The successful repeaters in the study had two common characteristics. First, they were extraordinarily disciplined, applying rigorous screens before they made an adjacency move. This discipline paid off in the form of learning curve benefits, increased speed, and lower complexity. And second, in almost all cases, they developed their repeatable formulas by studying their customers and their customers' economics very, very carefully.

  1. Conceptual core model for the reactor core test

    International Nuclear Information System (INIS)

    Swenson, L.D.

    1970-01-01

    Several design options for the ZrH Flight System Reactor were investigated which involved tradeoffs of core excess reactivity, reactor control, coolant mixing and cladding thickness. A design point was selected which is to be the basis for more detailed evaluation in the preliminary design phase. The selected design utilizes 295 elements with 0.670 inch element-to-element pitch, 32 mil thick Incoloy cladding, 18.00 inches long fuel meat, hydrogen content of 6.3 x 10 22 atoms/cc fuel, 10.5 w/o uranium, and a spiraled fin configuration with alternate elements having fins with spiral to the right, spiral to the left, and no fin at all (R-L-N fin configuration). Fin height is 30 mils for the center region of the core and 15 mils for the outer region. (U.S.)

  2. Thermal expansion of epoxy-fiberglass composite specimens

    International Nuclear Information System (INIS)

    McElroy, D.L.; Weaver, F.J.; Bridgman, C.

    1986-01-01

    The thermal expansion behavior of three epoxy-fiberglass composite specimens was measured from 20 to 120 0 C (70 to 250 0 F) using a fused quartz push-rod dilatometer. Billets produced by vacuum impregnating layers of two types of fiberglass cloth with an epoxy resin were core-drilled to produce cylindrical specimens. These were used to study expansion perpendicular and parallel to the fiberglass layers. The dilatometer is held at a preselected temperature until steady-state is indicated by stable length and temperature data. Before testing the composite specimens, a reliability check of the dilatometer was performed using a copper secondary standard. This indicated thermal expansion coefficient (α) values within +-2% of expected values from 20 to 200 0 C

  3. Design factors affecting dynamic behaviour of fast reactor cores. UK review paper

    Energy Technology Data Exchange (ETDEWEB)

    Brindley, K W [National Nuclear Corporation Ltd., Risley, Warrington (United Kingdom); Perks, M A [United Kingdom Atomic Energy Authority, Risley, Warrington (United Kingdom)

    1982-01-01

    This paper summarises the consideration that has been given in the UK to the following factors that affect the dynamic behaviour of fast reactor cores: fuel design - Pu/u homogeneity, fuel expansion, fuel-clad gaps, uranium fraction. Structural response - CR supports, diagrid, sub-assembly bowing sodium expansion coefficients - low void cores including heterogenous cores. Calculational methods and models are outlined and some experimental results are discussed. (author)

  4. Warp drive with zero expansion

    Energy Technology Data Exchange (ETDEWEB)

    Natario, Jose [Department of Mathematics, Instituto Superior Tecnico (Portugal)

    2002-03-21

    It is commonly believed that Alcubierre's warp drive works by contracting space in front of the warp bubble and expanding the space behind it. We show that this contraction/expansion is but a marginal consequence of the choice made by Alcubierre and explicitly construct a similar spacetime where no contraction/expansion occurs. Global and optical properties of warp-drive spacetimes are also discussed.

  5. Expansion lyre-shaped tube

    International Nuclear Information System (INIS)

    Andro, Jean.

    1973-01-01

    The invention relates the expansion lyre-shaped tube portions formed in dudgeoned tubular bundles between two bottom plates. An expansion lyre comprises at least two sets of tubes of unequal lengths coplanar and symmetrical with respect to the main tube axis, with connecting portions between the tubes forming said sets. The invention applies to apparatus such as heat exchangers, heaters, superheaters or breeders [fr

  6. Estimates of expansion time scales

    International Nuclear Information System (INIS)

    Jones, E.M.

    1979-01-01

    Monte Carlo simulations of the expansion of a spacefaring civilization show that descendants of that civilization should be found near virtually every useful star in the Galaxy in a time much less than the current age of the Galaxy. Only extreme assumptions about local population growth rates, emigration rates, or ship ranges can slow or halt an expansion. The apparent absence of extraterrestrials from the solar system suggests that no such civilization has arisen in the Galaxy. 1 figure

  7. Strategic Complexity and Global Expansion

    DEFF Research Database (Denmark)

    Oladottir, Asta Dis; Hobdari, Bersant; Papanastassiou, Marina

    2012-01-01

    The purpose of this paper is to analyse the determinants of global expansion strategies of newcomer Multinational Corporations (MNCs) by focusing on Iceland, Israel and Ireland. We argue that newcomer MNCs from small open economies pursue complex global expansion strategies (CGES). We distinguish....... The empirical evidence suggests that newcomer MNCs move away from simplistic dualities in the formulation of their strategic choices towards more complex options as a means of maintaining and enhancing their global competitiveness....

  8. Range expansion of heterogeneous populations.

    Science.gov (United States)

    Reiter, Matthias; Rulands, Steffen; Frey, Erwin

    2014-04-11

    Risk spreading in bacterial populations is generally regarded as a strategy to maximize survival. Here, we study its role during range expansion of a genetically diverse population where growth and motility are two alternative traits. We find that during the initial expansion phase fast-growing cells do have a selective advantage. By contrast, asymptotically, generalists balancing motility and reproduction are evolutionarily most successful. These findings are rationalized by a set of coupled Fisher equations complemented by stochastic simulations.

  9. Uncertainty reduction requirements in cores designed for passive reactivity shutdown

    International Nuclear Information System (INIS)

    Wade, D.C.

    1988-01-01

    The first purpose of this paper is to describe the changed focus of neutronics accuracy requirements existing in the current US advanced LMR development program where passive shutdown is a major design goal. The second purpose is to provide the background and rationale which supports the selection of a formal data fitting methodology as the means for the application of critical experiment measurements to meet these accuracy needs. 6 refs., 1 fig., 2 tabs

  10. Neutronic Core Performance of CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Villarino, Eduardo; Hergenreder, Daniel; Matzkin, S

    2000-01-01

    The actual design state of core of CAREM-25 reactor is presented.It is shown that the core design complains with the safety and operation established requirements.It is analyzed the behavior of the reactor safety and control systems (single failure of the fast shut down system, single failure of the shut down system, single failure of the second shut down system, reactivity worth of the adjust and control system in normal operation and hot shut down, reactivity worth of the adjust and control system and the scheme of movement of the control rod during the operation cycle).It is shown the burnup profile of fuel elements with the proposed scheme of refueling and the burnup and power density distribution at different moments of the operation cycle.The power peaking factor of the equilibrium core is 2.56, the minimum DNBR is 1.90 and its average is 2.09 during the operation cycle

  11. Side core lifter

    Energy Technology Data Exchange (ETDEWEB)

    Edelman, Ya A

    1982-01-01

    A side core lifter is proposed which contains a housing with guide slits and a removable core lifter with side projections on the support section connected to the core receiver. In order to preserve the structure of the rock in the core sample by means of guaranteeing rectilinear movement of the core lifter in the rock, the support and core receiver sections are hinged. The device is equipped with a spring for angular shift in the core-reception part.

  12. Electrospinning of reactive mesogens

    NARCIS (Netherlands)

    Yao, J.; Picot, O.T.; Hughes-Brittain, N.F.; Bastiaansen, C.W.M.; Peijs, T.

    2016-01-01

    The reinforcement potential of reactive liquid crystals or reactive mesogens (RMs) in electrospun fibers was investigated through the blending of two types of RMs (RM257 and RM82) with two types of thermoplastics; polyamide 6 (PA6) and poly(methyl methacrylate) (PMMA). Polymer/RM blends were

  13. In core system mapping reactor power distribution

    International Nuclear Information System (INIS)

    Yoriyaz, H.; Moreira, J.M.L.

    1989-01-01

    Based on the signals of SPND'S (Self Powered Neutron Detectors) distributed inside of a core, the spatial power distribution is obtained using the MAP program, developed in this work. The methodology applied in MAP program uses a least mean square technique to calculate expansion coefficients that depend on the SPND'S signals. The final power or neutron flux distribution is obtained by a combination of certains functions or expansion modes that are provided from diffusion calculation with the CITATION code. The MAP program is written in PASCAL language and will be used in IEA-R1 reactor for assisting its operation. (author) [pt

  14. Study of the core compaction effects and its monitoring in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Zylbersztejn, F.

    2012-01-01

    Conclusions: • On calculation of reactivity impacts of core compaction/flowering: → Upper bound of the reactivity coefficients for each type of deformation; → Uniform compaction model: significant reactivity impact; Circular symmetric model: small reactivity impact. • On the visibility of these phenomena by the neutron detectors: → The direct monitoring of the core compaction by neutron detector in the BCC is not possible. (the identification that the reactivity perturbations observed are due to variation of the core geometry). Perspectives of solutions: → Improved core design: reducing the effects. → Physical improvements: Steel resistance to deformations (irradiation, flexion); Direct devices: core constraint (prevents deformations). → Additional calculations: Considering more localized deformations; Advanced monitoring with neutron noise (in progress)

  15. Density-functional expansion methods: Grand challenges.

    Science.gov (United States)

    Giese, Timothy J; York, Darrin M

    2012-03-01

    We discuss the source of errors in semiempirical density functional expansion (VE) methods. In particular, we show that VE methods are capable of well-reproducing their standard Kohn-Sham density functional method counterparts, but suffer from large errors upon using one or more of these approximations: the limited size of the atomic orbital basis, the Slater monopole auxiliary basis description of the response density, and the one- and two-body treatment of the core-Hamiltonian matrix elements. In the process of discussing these approximations and highlighting their symptoms, we introduce a new model that supplements the second-order density-functional tight-binding model with a self-consistent charge-dependent chemical potential equalization correction; we review our recently reported method for generalizing the auxiliary basis description of the atomic orbital response density; and we decompose the first-order potential into a summation of additive atomic components and many-body corrections, and from this examination, we provide new insights and preliminary results that motivate and inspire new approximate treatments of the core-Hamiltonian.

  16. Reactivity margins in heavy water moderated production reactors

    International Nuclear Information System (INIS)

    Benton, F.D.

    1981-11-01

    The design of the reactor core and components of the heavy water moderated reactors at the Savannah River Plant (SFP) can be varied to produce a number of isotopes. For the past decade, the predominant reactor core design has been the enriched-depleted lattice. In this lattice, fuel assemblies of highly enriched uranium and target assemblies of depleted uranium, which produce plutonium, occupy alternate lattice positions. This heterogeneous lattice arrangement and a nonuniform control rod distribution result in a reactor core that requires sophisticated calculational methods for accurate reactivity margin and power distribution predictions. For maximum accuracy, techniques must exist to provide a base of observed data for the calculations. Frequent enriched-depleted lattice design changes are required as product demands vary. These changes provided incentive for the development of techniques to combine the results of calculations and observed reactivity data to accurately and conveniently monitor reactivity margins during operation

  17. Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

    Energy Technology Data Exchange (ETDEWEB)

    Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics

    2017-05-15

    Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.

  18. Neutronics calculation of RTP core

    Science.gov (United States)

    Rabir, Mohamad Hairie B.; Zin, Muhammad Rawi B. Mohamed; Karim, Julia Bt. Abdul; Bayar, Abi Muttaqin B. Jalal; Usang, Mark Dennis Anak; Mustafa, Muhammad Khairul Ariff B.; Hamzah, Na'im Syauqi B.; Said, Norfarizan Bt. Mohd; Jalil, Muhammad Husamuddin B.

    2017-01-01

    Reactor calculation and simulation are significantly important to ensure safety and better utilization of a research reactor. The Malaysian's PUSPATI TRIGA Reactor (RTP) achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. Since early 90s, neutronics modelling were used as part of its routine in-core fuel management activities. The are several computer codes have been used in RTP since then, based on 1D neutron diffusion, 2D neutron diffusion and 3D Monte Carlo neutron transport method. This paper describes current progress and overview on neutronics modelling development in RTP. Several important parameters were analysed such as keff, reactivity, neutron flux, power distribution and fission product build-up for the latest core configuration. The developed core neutronics model was validated by means of comparison with experimental and measurement data. Along with the RTP core model, the calculation procedure also developed to establish better prediction capability of RTP's behaviour.

  19. Modeling the reactor core of MNSR to simulate its dynamic behavior using the code PARET

    International Nuclear Information System (INIS)

    Hainoun, A.; Alhabet, F.

    2004-02-01

    Using the computer code PARET the core of the MNSR reactor was modelled and the neutronics and thermal hydraulic behaviour of the reactor core for the steady state and selected transients, that deal with step change of reactivity including control rod withdraw starting from steady state at various low power level, were simulated. For this purpose a PARET input model for the core of MNSR reactor has been developed enabling the simulation of neutron kinetic and thermal hydraulic of reactor core including reactivity feedback effects. The neutron kinetic model depends on the point kinetic with 15 groups delayed neutrons including photo neutrons of beryllium reflector. In this regard the effect of photo neutron on the dynamic behaviour has been analysed through two additional calculation. In the first the yield of photo neutrons was neglected completely and in the second its share was added to the sixth group of delayed neutrons. In the thermal hydraulic model the fuel elements with their cooling channels were distributed to 4 different groups with various radial power factors. The pressure lose factors for friction, flow direction change, expansion and contraction were estimated using suitable approaches. The post calculations of the relative neutron flux change and core average temperature were found to be consistent with the experimental measurements. Furthermore, the simulation has indicated the influence of photo neutrons of the Beryllium reflector on the neutron flux behaviour. For the reliability of the results sensitivity analysis was carried out to consider the uncertainty in some important parameters like temperature feedback coefficient and flow velocity. On the other hand the application of PARET in simulation of void formation in the subcooled boiling regime were tested. The calculation indicates the capability of PARET in modelling this phenomenon. However, big discrepancy between calculation results and measurement of axial void distribution were observed

  20. Development of Core Design Technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeong Il; Hong, S. G.; Jang, J. W. (and others)

    2007-06-15

    This report describes the contents of core design technology and computer code system development performed during 2005 and 2006 on the objects of nuclear proliferation resistant core and nuclear fuel basic key technology development security. Also, it is including the future application plans for the results and the developed methodology, important information and the materials acquired in this period. Two core designs with single enrichment were considered for the KALIMER-600 during the first year : 1) the first core uses the non-fuel rods such as B4C, ZrH1.8, and dummy rods, 2) the core using different cladding thickness for each core region (inner, middle, and outer cores) without non-fuel rods to flatten the power distribution. In particular, the latter design was intended to simplify the fuel assembly design by eliminating the heterogeneity. It was found that the proposed design satisfy all of the Gen IV SFR design goals on the cycle length longer than 18 EFPM, fuel discharge burnup larger than 80GWd/t, sodium void worth, conversion ratio, reactivity burnup swing and so on. For this object reactor, the structure integrity outside of reactor is confirmed for the radiation exposure during the plant life according to the result of shielding design and evaluation. The transmutation capability and the core characteristics of sodium cooled fast reactor was also evaluated according to the change of MA amount. The reactivity coefficients for the BN-600 reactor with MA fueled are calculated and the results are compared and evaluated with other participants results. Even though the discrepancies between the results of participants are somewhat large but the K-CORE results are close to the average within a standard deviation. To have the capability of 3-dimensional core dynamic analysis such as analyzing power distribution and reactivity variations according to the asymmetric insertion/withdrawal of control rods, the calculation module for core dynamic parameters was

  1. On genus expansion of superpolynomials

    Energy Technology Data Exchange (ETDEWEB)

    Mironov, Andrei, E-mail: mironov@itep.ru [Lebedev Physics Institute, Moscow 119991 (Russian Federation); ITEP, Moscow 117218 (Russian Federation); National Research Nuclear University MEPhI, Moscow 115409 (Russian Federation); Morozov, Alexei, E-mail: morozov@itep.ru [ITEP, Moscow 117218 (Russian Federation); National Research Nuclear University MEPhI, Moscow 115409 (Russian Federation); Sleptsov, Alexei, E-mail: sleptsov@itep.ru [ITEP, Moscow 117218 (Russian Federation); Laboratory of Quantum Topology, Chelyabinsk State University, Chelyabinsk 454001 (Russian Federation); KdVI, University of Amsterdam (Netherlands); Smirnov, Andrey, E-mail: asmirnov@math.columbia.edu [ITEP, Moscow 117218 (Russian Federation); Columbia University, Department of Mathematics, New York (United States)

    2014-12-15

    Recently it was shown that the (Ooguri–Vafa) generating function of HOMFLY polynomials is the Hurwitz partition function, i.e. that the dependence of the HOMFLY polynomials on representation R is naturally captured by symmetric group characters (cut-and-join eigenvalues). The genus expansion and expansion through Vassiliev invariants explicitly demonstrate this phenomenon. In the present paper we claim that the superpolynomials are not functions of such a type: symmetric group characters do not provide an adequate linear basis for their expansions. Deformation to superpolynomials is, however, straightforward in the multiplicative basis: the Casimir operators are β-deformed to Hamiltonians of the Calogero–Moser–Sutherland system. Applying this trick to the genus and Vassiliev expansions, we observe that the deformation is fully straightforward only for the thin knots. Beyond the family of thin knots additional algebraically independent terms appear in the Vassiliev and genus expansions. This can suggest that the superpolynomials do in fact contain more information about knots than the colored HOMFLY and Kauffman polynomials. However, even for the thin knots the beta-deformation is non-innocent: already in the simplest examples it seems inconsistent with the positivity of colored superpolynomials in non-(anti)symmetric representations, which also happens in I. Cherednik's (DAHA-based) approach to the torus knots.

  2. Dependence of calculated void reactivity on film-boiling representation

    International Nuclear Information System (INIS)

    Whitlock, J.; Garland, W.

    1992-01-01

    Partial voiding of a fuel channel can lead to complicated neutronic analysis, because of highly nonuniform spatial distributions. An investigation of the distribution dependence of void reactivity in a Canada deuterium uranium (CANDU) lattice, specifically in the regime of film boiling, was done. Although the core is not expected to be critical at the time of sheath dryout, this study augments current knowledge of void reactivity in this type of lattice

  3. (Electronic structure and reactivities of transition metal clusters)

    Energy Technology Data Exchange (ETDEWEB)

    1992-01-01

    The following are reported: theoretical calculations (configuration interaction, relativistic effective core potentials, polyatomics, CASSCF); proposed theoretical studies (clusters of Cu, Ag, Au, Ni, Pt, Pd, Rh, Ir, Os, Ru; transition metal cluster ions; transition metal carbide clusters; bimetallic mixed transition metal clusters); reactivity studies on transition metal clusters (reactivity with H{sub 2}, C{sub 2}H{sub 4}, hydrocarbons; NO and CO chemisorption on surfaces). Computer facilities and codes to be used, are described. 192 refs, 13 figs.

  4. Device for protecting deformations of reactor cores

    International Nuclear Information System (INIS)

    Kato, Yasuyoshi; Urushihara, Hiroshi.

    1975-01-01

    Object: To provide a fluid pressure cylinder, which is operated according to change in temperature of coolant for a reactor to restrain or release a core, to simply and effectively protect deformation of the core. Structure: A closed fluid pressure cylinder interiorly filled with suitable fluid is disposed in peripherally equally spaced relation in an annular space between a core barrel of a reactor and a reactor vessel. A piston is mounted in fluid-tight fashion in a plurality of piston openings made in the cylinder, the piston being slidably moved according to expansion and contraction of the fluid filled in the cylinder. The piston has a movable frame mounted at the foremost end thereof, the movable frame being moved integral with the piston, and the surface opposite the mount thereof biasing the outermost peripheral surface of the core. (Kamimura, M.)

  5. Analysis of LOF/TOP excursions in the SNR-300 Zielkern core with a stress controlled rip extension after failure

    International Nuclear Information System (INIS)

    Royl, P.

    1987-07-01

    An unprotected loss of flow (ULOF) accident can result in energetic power excursions due to reactivity effects from in-pile fuel motion and compaction after a near midplane failure of fuel pins in unvoided regions of the core (loss of flow driven transient overpower or LOF/TOP excursions). These excursions are strongly mitigated by a propagation of the failure rip, which reduces the fuel compaction and enhances fuel sweep-out. A simple model has been introduced into the SAS3D code in which this propagation is restricted to the expansion of the void zone that is firmed in the coolant channel due to thermal interactions of fuel and sodium (FCI) after failure. The rip propagation is demonstrated in this report in all investigated cases to be a strong mitigator that will limit the fuel compaction effects in the region where they originate

  6. Low Thermal Expansion Glass Ceramics

    CERN Document Server

    Bach, Hans

    2005-01-01

    This book appears in the authoritative series reporting the international research and development activities conducted by the Schott group of companies. This series provides an overview of Schott's activities for scientists, engineers, and managers from all branches of industry worldwide in which glasses and glass ceramics are of interest. Each volume begins with a chapter providing a general idea of the current problems, results, and trends relating to the subjects treated. This new extended edition describes the fundamental principles, the manufacturing process, and applications of low thermal expansion glass ceramics. The composition, structure, and stability of polycrystalline materials having a low thermal expansion are described, and it is shown how low thermal expansion glass ceramics can be manufactured from appropriately chosen glass compositions. Examples illustrate the formation of this type of glass ceramic by utilizing normal production processes together with controlled crystallization. Thus g...

  7. Low thermal expansion glass ceramics

    CERN Document Server

    1995-01-01

    This book is one of a series reporting on international research and development activities conducted by the Schott group of companies With the series, Schott aims to provide an overview of its activities for scientists, engineers, and managers from all branches of industry worldwide where glasses and glass ceramics are of interest Each volume begins with a chapter providing a general idea of the current problems, results, and trends relating to the subjects treated This volume describes the fundamental principles, the manufacturing process, and applications of low thermal expansion glass ceramics The composition, structure, and stability of polycrystalline materials having a low thermal expansion are described, and it is shown how low thermal expansion glass ceramics can be manufactured from appropriately chosen glass compositions Examples illustrate the formation of this type of glass ceramic by utilizing normal production processes together with controlled crystallization Thus glass ceramics with thermal c...

  8. Regulation of gas infrastructure expansion

    International Nuclear Information System (INIS)

    De Joode, J.

    2012-01-01

    The topic of this dissertation is the regulation of gas infrastructure expansion in the European Union (EU). While the gas market has been liberalised, the gas infrastructure has largely remained in the regulated domain. However, not necessarily all gas infrastructure facilities - such as gas storage facilities, LNG import terminals and certain gas transmission pipelines - need to be regulated, as there may be scope for competition. In practice, the choice of regulation of gas infrastructure expansion varies among different types of gas infrastructure facilities and across EU Member States. Based on a review of economic literature and on a series of in-depth case studies, this study explains these differences in choices of regulation from differences in policy objectives, differences in local circumstances and differences in the intrinsic characteristics of the infrastructure projects. An important conclusion is that there is potential for a larger role for competition in gas infrastructure expansion.

  9. EBRPOCO - a program to calculate detailed contributions of power reactivity components of EBR-II

    International Nuclear Information System (INIS)

    Meneghetti, D.; Kucera, D.A.

    1981-01-01

    The EBRPOCO program has been developed to facilitate the calculations of the power coefficients of reactivity of EBR-II loadings. The program enables contributions of various components of the power coefficient to be delineated axially for every subassembly. The program computes the reactivity contributions of the power coefficients resulting from: density reduction of sodium coolant due to temperature; displacement of sodium coolant by thermal expansions of cladding, structural rods, subassembly cans, and lower and upper axial reflectors; density reductions of these steel components due to temperature; displacement of bond-sodium (if present) in gaps by differential thermal expansions of fuel and cladding; density reduction of bond-sodium (if present) in gaps due to temperature; free axial expansion of fuel if unrestricted by cladding or restricted axial expansion of fuel determined by axial expansion of cladding. Isotopic spatial contributions to the Doppler component my also be obtained. (orig.) [de

  10. Space dependence of reactivity parameters on reactor dynamic perturbation measurements

    International Nuclear Information System (INIS)

    Maletti, R.; Ziegenbein, D.

    1985-01-01

    Practical application of reactor-dynamic perturbation measurements for on-power determination of differential reactivity weight of control rods and power coefficients of reactivity has shown a significant dependence of parameters on the position of outcore detectors. The space dependence of neutron flux signal in the core of a VVER-440-type reactor was measured by means of 60 self-powered neutron detectors. The greatest neutron flux alterations are located close to moved control rods and in height of the perturbation position. By means of computations, detector positions can be found in the core in which the one-point model is almost valid. (author)

  11. BN-600 hybrid core benchmark analyses

    International Nuclear Information System (INIS)

    Kim, Y.I.; Stanculescu, A.; Finck, P.; Hill, R.N.; Grimm, K.N.

    2003-01-01

    Benchmark analyses for the hybrid BN-600 reactor that contains three uranium enrichment zones and one plutonium zone in the core, have been performed within the frame of an IAEA sponsored Coordinated Research Project. The results for several relevant reactivity parameters obtained by the participants with their own state-of-the-art basic data and codes, were compared in terms of calculational uncertainty, and their effects on the ULOF transient behavior of the hybrid BN-600 core were evaluated. The comparison of the diffusion and transport results obtained for the homogeneous representation generally shows good agreement for most parameters between the RZ and HEX-Z models. The burnup effect and the heterogeneity effect on most reactivity parameters also show good agreement for the HEX-Z diffusion and transport theory results. A large difference noticed for the sodium and steel density coefficients is mainly due to differences in the spatial coefficient predictions for non fuelled regions. The burnup reactivity loss was evaluated to be 0.025 (4.3 $) within ∼ 5.0% standard deviation. The heterogeneity effect on most reactivity coefficients was estimated to be small. The heterogeneity treatment reduced the control rod worth by 2.3%. The heterogeneity effect on the k-eff and control rod worth appeared to differ strongly depending on the heterogeneity treatment method. A substantial spread noticed for several reactivity coefficients did not give a significant impact on the transient behavior prediction. This result is attributable to compensating effects between several reactivity effects and the specific design of the partially MOX fuelled hybrid core. (author)

  12. Animal MRI Core

    Data.gov (United States)

    Federal Laboratory Consortium — The Animal Magnetic Resonance Imaging (MRI) Core develops and optimizes MRI methods for cardiovascular imaging of mice and rats. The Core provides imaging expertise,...

  13. Digital reactivity meter

    International Nuclear Information System (INIS)

    Jiang Zongbing

    1996-02-01

    The importance and the usual methods of reactivity measurement in a nuclear reactor are presented. Emphasis is put upon the calculation principle, software and hardware components, main specifications, application, as well as the features of the digital reactivity meter. The test results of operation in various reactors shown that the meter possess the following features: high accuracy, short response time, low output noise, high resolution, wide measuring range, simple and flexible to operate, high stability and reliability. In addition, the reactivity meter can save the measuring data automatically and have a perfect capability of self-verifying. It not only meet the requirement of the reactivity measurement in nuclear power plant, but also can be applied to various types of reactors. (1 tab.)

  14. Stress Reactivity in Insomnia.

    Science.gov (United States)

    Gehrman, Philip R; Hall, Martica; Barilla, Holly; Buysse, Daniel; Perlis, Michael; Gooneratne, Nalaka; Ross, Richard J

    2016-01-01

    This study examined whether individuals with primary insomnia (PI) are more reactive to stress than good sleepers (GS). PI and GS (n = 20 per group), matched on gender and age, completed three nights of polysomnography. On the stress night, participants received a mild electric shock and were told they could receive additional shocks during the night. Saliva samples were obtained for analysis of cortisol and alpha amylase along with self-report and visual analog scales (VAS). There was very little evidence of increased stress on the stress night, compared to the baseline night. There was also no evidence of greater stress reactivity in the PI group for any sleep or for salivary measures. In the GS group, stress reactivity measured by VAS scales was positively associated with an increase in sleep latency in the experimental night on exploratory analyses. Individuals with PI did not show greater stress reactivity compared to GS.

  15. Taskable Reactive Agent Communities

    National Research Council Canada - National Science Library

    Myers, Karen

    2002-01-01

    The focus of Taskable Reactive Agent Communities (TRAC) project was to develop mixed-initiative technology to enable humans to supervise and manage teams of agents as they perform tasks in dynamic environments...

  16. The loop expansion as a divergent-power-series expansion

    International Nuclear Information System (INIS)

    Murai, N.

    1981-01-01

    The loop expansion should be divergent, possibly an asymptotic one, in the Euclidean path integral formulation. This consideration is important in applications of the symmetric and mass-independent renormalization. The [1,1] Pade approximant is calculated in a PHI 4 model. Its classical vacua may be not truely stable for nonzero coupling constant. (author)

  17. Measurements and calculation of reactivity in the IEA-R1 nuclear reactor

    International Nuclear Information System (INIS)

    Ferreira, P.S.B.

    1988-01-01

    Techniques and experimentals procedures utilized in the measurement of some nuclear parameters related to reactivity are presented. Measurements of reactivity coefficients, such as void, temperature and power, and control rod worth were made in the IEA-R1 Research Reactor. The techniques used to perform the measurements were: i) stable period (control rod calibration), ii) inverse kinetics (digital reactivity meter), iii) aluminium slab insertion in the fuel element coolant channels (void reactivity), iv) nuclear reactor core temperature changes by means of the changes in the coolant systems of reactor core (isothermal reactivity coefficient) and v) by making perturbation in the core through the control rod motions (power reactivity coefficient and control rod calibration). By using the computer codes HAMMER, HAMMER-TECHNION and CITATION, the experiments realized in the IEA-R1 reactor were simulated. From this simulation, the theoretical reactivity parameters were estimated and compared with the respective experimental results. Furthermore, in the second fuel load of Angra-1 Nuclear Power Station, the IPEN-CNEN/SP digital reactivity - meter were used in the lower power test with the aim to assess the equipment performance. Among several tests, the reacticity-meter were used in parallel with a Westinghouse analogic reativimeter-meter) to measure the heat additiona point, critical boron concentration, control rod calibration, isothermal and moderator reactivity coefficient. These tests, and the results obtained by the digital reactivity-meter are described. The results were compared with those obtained by Westinghouse analogic reactivity meter, showing excellent agreement. (author) [pt

  18. Cosmological expansion and local physics

    International Nuclear Information System (INIS)

    Faraoni, Valerio; Jacques, Audrey

    2007-01-01

    The interplay between cosmological expansion and local attraction in a gravitationally bound system is revisited in various regimes. First, weakly gravitating Newtonian systems are considered, followed by various exact solutions describing a relativistic central object embedded in a Friedmann universe. It is shown that the 'all or nothing' behavior recently discovered (i.e., weakly coupled systems are comoving while strongly coupled ones resist the cosmic expansion) is limited to the de Sitter background. New exact solutions are presented which describe black holes perfectly comoving with a generic Friedmann universe. The possibility of violating cosmic censorship for a black hole approaching the big rip is also discussed

  19. Temperature expansions for magnetic systems

    International Nuclear Information System (INIS)

    Cangemi, D.; Dunne, G.

    1996-01-01

    We derive finite temperature expansions for relativistic fermion systems in the presence of background magnetic fields, and with nonzero chemical potential. We use the imaginary-time formalism for the finite temperature effects, the proper-time method for the background field effects, and zeta function regularization for developing the expansions. We emphasize the essential difference between even and odd dimensions, focusing on 2+1 and 3+1 dimensions. We concentrate on the high temperature limit, but we also discuss the T=0 limit with nonzero chemical potential. Copyright copyright 1996 Academic Press, Inc

  20. Coupled core criticality calculations with control rods located in the central reflector region

    Energy Technology Data Exchange (ETDEWEB)

    Sobhy, M [Reactor depatrment, nuclear research center, Inshaas (Egypt)

    1995-10-01

    The reactivity of a coupled core is controlled by a set of control rods distributed in the central reflector region. The reactor contains two compact cores cooled and moderated by light water. Control rods are designed to have reactivity worths sufficient to start, control and shutdown the coupled system. Each core in a coupled system is in subcritical conditions without any absorber then each core needs to the other core to fulfill nuclear chain reaction and to approach the criticality. In this case, each core is considered clean which is suitable for research reactor with low flux disturbance and better neutron economy, in addition to the advantage of disappearing the cut corner fuel baskets. This facilitate the in core fuel management with identical fuel baskets. Hot spots will disappear. This leads to a good heat transfer process. the excess reactivity and the shutdown margin are calculated for some of reflector as coupling region gives sufficient area for coupled core are calculated cost. The fluctuations of reactivity for coupled core are calculated by noise analysis technique and compared with that for rode core. The results show low reactivity perturbation associated with coupled core.

  1. Sensitivity of BWR shutdown margin tests to local reactivity anomalies

    International Nuclear Information System (INIS)

    Cokinos, D.M.; Carew, J.F.

    1987-01-01

    Successful shutdown margin (SDM) demonstration is a required procedure in the startup of a newly configured boiling water reactor (BWR) core. In its most reactive condition throughout a cycle, a BWR core must be capable of being made subcritical by a specified margin with the highest worth control rod fully withdrawn and all other rods at their fully inserted positions. Two different methods are used to demonstrate SDM: (a) the adjacent-rod test and (b) the in-sequence test. In the adjacent-rod test, the strongest rod is fully withdrawn and an adjacent rod is withdrawn to reach criticality. In the in-sequence test, control rods spread throughout the core are withdrawn in a predetermined sequence of withdrawals. Larger than expected core k/sub eff/ values have been observed during the performance of BWR SDM tests. The purpose of the work summarized in this paper has been to investigated and quantify the sensitivity of both the adjacent-rod and in-sequence SDM tests to local reactivity anomalies. This was accomplished by introducing reactivity perturbations at selected four-bundle cell locations and by evaluating their effect on core reactivity in each of the two tests

  2. Reactive sputter deposition

    CERN Document Server

    Mahieu, Stijn

    2008-01-01

    In this valuable work, all aspects of the reactive magnetron sputtering process, from the discharge up to the resulting thin film growth, are described in detail, allowing the reader to understand the complete process. Hence, this book gives necessary information for those who want to start with reactive magnetron sputtering, understand and investigate the technique, control their sputtering process and tune their existing process, obtaining the desired thin films.

  3. Bearing-Mounting Concept Accommodates Thermal Expansion

    Science.gov (United States)

    Nespodzany, Robert; Davis, Toren S.

    1995-01-01

    Pins or splines allow radial expansion without slippage. Design concept for mounting rotary bearing accommodates differential thermal expansion between bearing and any structure(s) to which bearing connected. Prevents buildup of thermal stresses by allowing thermal expansion to occur freely but accommodating expansion in such way not to introduce looseness. Pin-in-slot configuration also maintains concentricity.

  4. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    2013-01-01

    JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO 2 and MOX fuel rods, (3) analysis of isotopic composition data for UO 2 and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

  5. Validation study of core analysis methods for full MOX BWR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO{sub 2} and MOX fuel rods, (3) analysis of isotopic composition data for UO{sub 2} and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

  6. DENSE MOLECULAR CORES BEING EXTERNALLY HEATED

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Gwanjeong; Lee, Chang Won; Kim, Mi-Ryang [Radio Astronomy division, Korea Astronomy and Space Science Institute, 776 Daedeokdae-ro, Yuseong-gu, Daejeon, 34055 (Korea, Republic of); Gopinathan, Maheswar [Aryabhatta Research Institute of Observational Sciences, Manora Peak, Nainital 263129 (India); Jeong, Woong-Seob, E-mail: archer81@kasi.re.kr [Department of Astronomy and Space Science, University of Science and Technology, 217 Gajungro, Yuseong-gu, Daejeon, 34113 (Korea, Republic of)

    2016-06-20

    We present results of our study of eight dense cores, previously classified as starless, using infrared (3–160 μ m) imaging observations with the AKARI telescope and molecular line (HCN and N{sub 2}H{sup +}) mapping observations with the KVN telescope. Combining our results with the archival IR to millimeter continuum data, we examined the starless nature of these eight cores. Two of the eight cores are found to harbor faint protostars having luminosities of ∼0.3–4.4 L {sub ⊙}. The other six cores are found to remain starless and probably are in a dynamically transitional state. The temperature maps produced using multi-wavelength images show an enhancement of about 3–6 K toward the outer boundary of these cores, suggesting that they are most likely being heated externally by nearby stars and/or interstellar radiation fields. Large virial parameters and an overdominance of red asymmetric line profiles over the cores may indicate that the cores are set into either an expansion or an oscillatory motion, probably due to the external heating. Most of the starless cores show a coreshine effect due to the scattering of light by the micron-sized dust grains. This may imply that the age of the cores is of the order of ∼10{sup 5} years, which is consistent with the timescale required for the cores to evolve into an oscillatory stage due to external perturbation. Our observational results support the idea that the external feedback from nearby stars and/or interstellar radiation fields may play an important role in the dynamical evolution of the cores.

  7. Energy expansion planning by considering electrical and thermal expansion simultaneously

    International Nuclear Information System (INIS)

    Abbasi, Ali Reza; Seifi, Ali Reza

    2014-01-01

    Highlights: • This paper focused on the expansion planning optimization of energy systems. • Employing two form of energy: the expansion of electrical and thermal energies. • The main objective is to minimize the costs. • A new Modified Honey Bee Mating Optimization (MHBMO) algorithm is applied. - Abstract: This study focused on the expansion planning optimization of energy systems employing two forms of energy: the expansion of electrical and thermal energies simultaneously. The main objective of this investigation is confirming network adequacy by adding new equipment to the network, over a given planning horizon. The main objective of the energy expansion planning (EEP) is to minimize the real energy loss, voltage deviation and the total cost of installation equipments. Since the objectives are different and incommensurable, it is difficult to solve the problem by the conventional approaches that may optimize a single objective. So, the meta-heuristic algorithm is applied to this problem. Here, Honey Bee Mating Optimization algorithm (HBMO) as a new evolutionary optimization algorithm is utilized. In order to improve the total ability of HBMO for the global search and exploration, a new modification process is suggested such a way that the algorithm will search the total search space globally. Also, regarding the uncertainties of the new complicated energy systems, in this paper for the first time, the EEP problem is investigated in a stochastic environment by the use of probabilistic load flow technique based on Point Estimate Method (PEM). In order to evaluate the feasibility and effectiveness of the proposed algorithm, two modified test systems are used as case studies

  8. Evidence for coral range expansion accompanied by reduced diversity of Symbiodinium genotypes

    KAUST Repository

    Grupstra, Carsten G. B.

    2017-05-15

    Zooxanthellate corals are threatened by climate change but may be able to escape increasing temperatures by colonizing higher latitudes. To determine the effect of host range expansion on symbiont genetic diversity, we examined genetic variation among populations of Symbiodinium psygmophilum associated with Oculina patagonica, a range-expanding coral that acquires its symbionts through horizontal transmission. We optimized five microsatellite primer pairs for S. psygmophilum and tested them on Oculina spp. samples from the western North Atlantic and the Mediterranean. We then used them to compare symbiont genotype diversity between an Iberian core and an expansion front population of O. patagonica. Only one multilocus S. psygmophilum genotype was identified at the expansion front, and it was shared with the core population, which harbored seven multilocus genotypes. This pattern suggests that O. patagonica range expansion is accompanied by reduced symbiont genetic diversity, possibly due to limited dispersal of symbionts or local selection.

  9. The Thermal Expansion Of Feldspars

    Science.gov (United States)

    Hovis, G. L.; Medford, A.; Conlon, M.

    2009-12-01

    Hovis and others (1) investigated the thermal expansion of natural and synthetic AlSi3 feldspars and demonstrated that the coefficient of thermal expansion (α) decreases significantly, and linearly, with increasing room-temperature volume (VRT). In all such feldspars, therefore, chemical expansion limits thermal expansion. The scope of this work now has been broadened to include plagioclase and Ba-K feldspar crystalline solutions. X-ray powder diffraction data have been collected between room temperature and 925 °C on six plagioclase specimens ranging in composition from anorthite to oligoclase. When combined with thermal expansion data for albite (2,3,4) a steep linear trend of α as a function of VRT emerges, reflecting how small changes in composition dramatically affect expansion behavior. The thermal expansion data for five synthetic Ba-K feldspars ranging in composition from 20 to 100 mole percent celsian, combined with data for pure K-feldspar (3,4), show α-VRT relationships similar in nature to the plagioclase series, but with a slope and intercept different from the latter. Taken as a group all Al2Si2 feldspars, including anorthite and celsian from the present study along with Sr- (5) and Pb-feldspar (6) from other workers, show very limited thermal expansion that, unlike AlSi3 feldspars, has little dependence on the divalent-ion (or M-) site occupant. This apparently is due to the necessitated alternation of Al and Si in the tetrahedral sites of these minerals (7), which in turn locks the tetrahedral framework and makes the M-site occupant nearly irrelevant to expansion behavior. Indeed, in feldspar series with coupled chemical substitution it is the change away from a 1:1 Al:Si ratio that gives feldspars greater freedom to expand. Overall, the relationships among α, chemical composition, and room-temperature volume provide useful predictive tools for estimating feldspar thermal expansion and give insight into the controls of expansion behavior in

  10. Performance testing of a mixed TRIGA core

    Energy Technology Data Exchange (ETDEWEB)

    Schumacher, R F; Godsey, T A; Feltz, D E; Randall, J D [Texas A and M University (United States)

    1974-07-01

    The major operational problem experienced by the Nuclear Science Center Reactor at Texas A and M University is full burnup. With two shift operation caused by the high utilization of the facility, the reactor is operated more than 100 megawatt days per year. The solution chosen for this problem was conversion to FLIP fuel. Since funds were not available to load an entire FLIP core, a mixed core comprised of approximately one third FLIP fuel located in the central region was designed. The design core was loaded and went critical on July 1, 1973. The results of the following measurements on the mixed core are presented: Determination of Rod worths; Measurement of Reactivity Effects; Determination of Flux values; Measurement of Fuel temperatures; Preliminary Fuel Burnup Rate; Pulsing Calibration. (author)

  11. An alternative approach to the management of reactive metals: tolerant cementitious systems

    International Nuclear Information System (INIS)

    Swift, P.; Cox, J.; Wise, M.; McKinney, J.; Rhodes, C.

    2015-01-01

    In recent years research has focused on preventing or minimising corrosion of reactive metals to ensure long-term waste package integrity. An alternative approach to the encapsulation of reactive metals is being explored. The approach will identify a cementitious-based encapsulating material that will allow corrosion of reactive metals to occur in a controlled and predictable manner, rather than seeking to limit or prevent the corrosion, whilst retaining waste package integrity. A low strength grout will be developed that will be 'tolerant' to the expansive forces generated by the corrosion products of reactive metals. Novel cementitious systems (e.g. foamed cements, rubber composite cements, cenosphere composite cements, lime mortars, bentonite cements etc.) that may be tolerant to potentially expansive waste products, such as reactive metals will be considered and assessed in a series of small-scale preliminary trials (compressive strength, porosity, permeability, pore solution pH, etc.)

  12. Crude oil pipeline expansion summary

    International Nuclear Information System (INIS)

    2005-02-01

    The Canadian Association of Petroleum Producers has been working with producers to address issues associated with the development of new pipeline capacity from western Canada. This document presents an assessment of the need for additional oil pipeline capacity given the changing mix of crude oil types and forecasted supply growth. It is of particular interest to crude oil producers and contributes to current available information for market participants. While detailed, the underlying analysis does not account for all the factors that may come into play when individual market participants make choices about which expansions they may support. The key focus is on the importance of timely expansion. It was emphasized that if pipeline expansions lags the crude supply growth, then the consequences would be both significant and unacceptable. Obstacles to timely expansion are also discussed. The report reviews the production and supply forecasts, the existing crude oil pipeline infrastructure, opportunities for new market development, requirements for new pipeline capacity and tolling options for pipeline development. tabs., figs., 1 appendix

  13. Asymptotic Expansions - Methods and Applications

    International Nuclear Information System (INIS)

    Harlander, R.

    1999-01-01

    Different viewpoints on the asymptotic expansion of Feynman diagrams are reviewed. The relations between the field theoretic and diagrammatic approaches are sketched. The focus is on problems with large masses or large external momenta. Several recent applications also for other limiting cases are touched upon. Finally, the pros and cons of the different approaches are briefly discussed. (author)

  14. Model of clinker capacity expansion

    CSIR Research Space (South Africa)

    Stylianides, T

    1998-10-01

    Full Text Available This paper describes a model which has been applied in practice to determine an optimal plan for clinker capacity expansion. The problem was formulated as an integer linear program aiming to determine the optimal number, size and location of kilns...

  15. The bootstrap and edgeworth expansion

    CERN Document Server

    Hall, Peter

    1992-01-01

    This monograph addresses two quite different topics, in the belief that each can shed light on the other. Firstly, it lays the foundation for a particular view of the bootstrap. Secondly, it gives an account of Edgeworth expansion. Chapter 1 is about the bootstrap, witih almost no mention of Edgeworth expansion; Chapter 2 is about Edgeworth expansion, with scarcely a word about the bootstrap; and Chapters 3 and 4 bring these two themes together, using Edgeworth expansion to explore and develop the properites of the bootstrap. The book is aimed a a graduate level audience who has some exposure to the methods of theoretical statistics. However, technical details are delayed until the last chapter (entitled "Details of Mathematical Rogour"), and so a mathematically able reader without knowledge of the rigorous theory of probability will have no trouble understanding the first four-fifths of the book. The book simultaneously fills two gaps in the literature; it provides a very readable graduate level account of t...

  16. On Fourier re-expansions

    OpenAIRE

    Liflyand, E.

    2012-01-01

    We study an extension to Fourier transforms of the old problem on absolute convergence of the re-expansion in the sine (cosine) Fourier series of an absolutely convergent cosine (sine) Fourier series. The results are obtained by revealing certain relations between the Fourier transforms and their Hilbert transforms.

  17. On persistently positively expansive maps

    Directory of Open Access Journals (Sweden)

    Alexander Arbieto

    2010-06-01

    Full Text Available In this paper, we prove that any C¹-persistently positively expansive map is expanding. This improves a result due to Sakai (Sakai 2004.Neste artigo, mostramos que todo mapa C¹-persistentemente positivamente expansivo e expansor. Isto melhora um resultado devido a Sakai (Sakai 2004.

  18. Bypass Flow and Hot Spot Analysis for PMR200 Block-Core Design with Core Restraint Mechanism

    International Nuclear Information System (INIS)

    Lim, Hong Sik; Kim, Min Hwan

    2009-01-01

    The accurate prediction of local hot spot during normal operation is important to ensure core thermal margin in a very high temperature gas-cooled reactor because of production of its high temperature output. The active cooling of the reactor core determining local hot spot is strongly affected by core bypass flows through the inter-column gaps between graphite blocks and the cross gaps between two stacked fuel blocks. The bypass gap sizes vary during core life cycle by the thermal expansion at the elevated temperature and the shrinkage/swelling by fast neutron irradiation. This study is to investigate the impacts of the variation of bypass gaps during core life cycle as well as core restraint mechanism on the amount of bypass flow and thus maximum fuel temperature. The core thermo fluid analysis is performed using the GAMMA+ code for the PMR200 block-core design. For the analysis not only are some modeling features, developed for solid conduction and bypass flow, are implemented into the GAMMA+ code but also non-uniform bypass gap distribution taken from a tool calculating the thermal expansion and the shrinkage/swell of graphite during core life cycle under the design options with and without core restraint mechanism is used

  19. Alteration of helical vortex core without change in flow topology

    DEFF Research Database (Denmark)

    Velte, Clara Marika; Okulov, Valery; Hansen, Martin Otto Laver

    2011-01-01

    topology. The helical symmetry as such is preserved, although the characteristic parameters of helical symmetry of the vortex core transfer from a smooth linear variation to a different trend under the influence of a non-uniform pressure gradient, causing an increase in helical pitch without changing its......The abrupt expansion of the slender vortex core with changes in flow topology is commonly known as vortex breakdown. We present new experimental observations of an alteration of the helical vortex core in wall bounded turbulent flow with abrupt growth in core size, but without change in flow...

  20. Monitoring of the temperature reactivity coefficient at the PWR nuclear plant

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1996-01-01

    For monitoring temperature coefficient of reactivity of pressurized water reactor a method based on the correction of fluctuation in signals of i-core neutron detectors and core-exit thermocouples and neural network paradigm is used it is shown that the moderator temperature coefficient of relativity can be predicted with the aid of the back propagation neural network technique by measuring the frequency response function between the in-core neutron flux and the core-exit coolant temperature

  1. The expansion of electric power systems considering the operation reliability; Expansion de sistemas electricos considerando la securidad de servicio

    Energy Technology Data Exchange (ETDEWEB)

    Cortes, Marcelo A. [Universidad de Antofagasta (Chile). Dept. de Ingenieria Electrica; Moya A, Oscar [Chile Univ., Santiago (Chile). Dept. de Ingenieria Civil

    1997-12-31

    This publication describes a procedure based in an linear programming for the planning of the expansion of mediating term in power systems. The implicit alternatives for network expansion, whether power plants or transmission lines, are classified according to indexes of reinforcement. The indexes are gotten resulting shadow pricing upon resolving a problem of minimum cost of operation and outage. The network is subjected to the most probable contingencies and from these outputs the expected reinforcement results are obtained. The problem of the reactive power accounts for restrictions in line capacity and prior evaluation of A C power flux. Results obtained are used for methodology validation. (author) 12 refs., 1 fig., 3 tabs.; e-mail: mcortes at apolo.cc.uantof.cl.; osmoya at tamarugo.cec.uchile.cl

  2. Feasibility Study of Core Design with a Monte Carlo Code for APR1400 Initial core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jinsun; Chang, Do Ik; Seong, Kibong [KEPCO NF, Daejeon (Korea, Republic of)

    2014-10-15

    The Monte Carlo calculation becomes more popular and useful nowadays due to the rapid progress in computing power and parallel calculation techniques. There have been many attempts to analyze a commercial core by Monte Carlo transport code using the enhanced computer capability, recently. In this paper, Monte Carlo calculation of APR1400 initial core has been performed and the results are compared with the calculation results of conventional deterministic code to find out the feasibility of core design using Monte Carlo code. SERPENT, a 3D continuous-energy Monte Carlo reactor physics burnup calculation code is used for this purpose and the KARMA-ASTRA code system, which is used for a deterministic code of comparison. The preliminary investigation for the feasibility of commercial core design with Monte Carlo code was performed in this study. Simplified core geometry modeling was performed for the reactor core surroundings and reactor coolant model is based on two region model. The reactivity difference at HZP ARO condition between Monte Carlo code and the deterministic code is consistent with each other and the reactivity difference during the depletion could be reduced by adopting the realistic moderator temperature. The reactivity difference calculated at HFP, BOC, ARO equilibrium condition was 180 ±9 pcm, with axial moderator temperature of a deterministic code. The computing time will be a significant burden at this time for the application of Monte Carlo code to the commercial core design even with the application of parallel computing because numerous core simulations are required for actual loading pattern search. One of the remedy will be a combination of Monte Carlo code and the deterministic code to generate the physics data. The comparison of physics parameters with sophisticated moderator temperature modeling and depletion will be performed for a further study.

  3. Design comparisons of TRU burner cores with similar sodium void worth

    International Nuclear Information System (INIS)

    Sang Ji, Kim; Young Il, Kim; Young Jin, Kim; Nam Zin, Cho

    2001-01-01

    This study summarizes the neutronic performance and fuel cycle behavior of five geometrically-different transuranic (TRU) burner cores with similar low sodium void reactivity. The conceptual cores encompass core geometries for annular, two-region homogeneous, dual pin type, pan-shaped and H-shaped cores. They have been designed with the same assembly specifications and managed to have similar end-of-cycle sodium void reactivities and beginning-of-cycle peak power densities through the changes in the core size and configuration. The requirement of low sodium void reactivity is shown to lead each design concept to characteristic neutronics performance and fuel cycle behavior. The H-/pan-shaped cores allow the core compaction as well as higher rate of TRU burning. (author)

  4. Reactive power compensator

    Science.gov (United States)

    El-Sharkawi, Mohamed A.; Venkata, Subrahmanyam S.; Chen, Mingliang; Andexler, George; Huang, Tony

    1992-01-01

    A system and method for determining and providing reactive power compensation for an inductive load. A reactive power compensator (50,50') monitors the voltage and current flowing through each of three distribution lines (52a, 52b, 52c), which are supplying three-phase power to one or more inductive loads. Using signals indicative of the current on each of these lines when the voltage waveform on the line crosses zero, the reactive power compensator determines a reactive power compensator capacitance that must be connected to the lines to maintain a desired VAR level, power factor, or line voltage. Alternatively, an operator can manually select a specific capacitance for connection to each line, or the capacitance can be selected based on a time schedule. The reactive power compensator produces control signals, which are coupled through optical fibers (102/106) to a switch driver (110, 110') to select specific compensation capacitors (112) for connections to each line. The switch driver develops triggering signals that are supplied to a plurality of series-connected solid state switches (350), which control charge current in one direction in respect to ground for each compensation capacitor. During each cycle, current flows from ground to charge the capacitors as the voltage on the line begins to go negative from its positive peak value. The triggering signals are applied to gate the solid state switches into a conducting state when the potential on the lines and on the capacitors reaches a negative peak value, thereby minimizing both the potential difference and across the charge current through the switches when they begin to conduct. Any harmonic distortion on the potential and current carried by the lines is filtered out from the current and potential signals used by the reactive power compensator so that it does not affect the determination of the required reactive compensation.

  5. Reactive power compensator

    Energy Technology Data Exchange (ETDEWEB)

    El-Sharkawi, Mohamed A. (Renton, WA); Venkata, Subrahmanyam S. (Woodinville, WA); Chen, Mingliang (Kirkland, WA); Andexler, George (Everett, WA); Huang, Tony (Seattle, WA)

    1992-01-01

    A system and method for determining and providing reactive power compensation for an inductive load. A reactive power compensator (50,50') monitors the voltage and current flowing through each of three distribution lines (52a, 52b, 52c), which are supplying three-phase power to one or more inductive loads. Using signals indicative of the current on each of these lines when the voltage waveform on the line crosses zero, the reactive power compensator determines a reactive power compensator capacitance that must be connected to the lines to maintain a desired VAR level, power factor, or line voltage. Alternatively, an operator can manually select a specific capacitance for connection to each line, or the capacitance can be selected based on a time schedule. The reactive power compensator produces control signals, which are coupled through optical fibers (102/106) to a switch driver (110, 110') to select specific compensation capacitors (112) for connections to each line. The switch driver develops triggering signals that are supplied to a plurality of series-connected solid state switches (350), which control charge current in one direction in respect to ground for each compensation capacitor. During each cycle, current flows from ground to charge the capacitors as the voltage on the line begins to go negative from its positive peak value. The triggering signals are applied to gate the solid state switches into a conducting state when the potential on the lines and on the capacitors reaches a negative peak value, thereby minimizing both the potential difference and across the charge current through the switches when they begin to conduct. Any harmonic distortion on the potential and current carried by the lines is filtered out from the current and potential signals used by the reactive power compensator so that it does not affect the determination of the required reactive compensation.

  6. Stand-alone core sensitivity and uncertainty analysis of ALFRED from Monte Carlo simulations

    International Nuclear Information System (INIS)

    Pérez-Valseca, A.-D.; Espinosa-Paredes, G.; François, J.L.; Vázquez Rodríguez, A.; Martín-del-Campo, C.

    2017-01-01

    Highlights: • Methodology based on Monte Carlo simulation. • Sensitivity analysis of Lead Fast Reactor (LFR). • Uncertainty and regression analysis of LFR. • 10% change in the core inlet flow, the response in thermal power change is 0.58%. • 2.5% change in the inlet lead temperature the response is 1.87% in power. - Abstract: The aim of this paper is the sensitivity and uncertainty analysis of a Lead-Cooled Fast Reactor (LFR) based on Monte Carlo simulation of sizes up to 2000. The methodology developed in this work considers the uncertainty of sensitivities and uncertainty of output variables due to a single-input-variable variation. The Advanced Lead fast Reactor European Demonstrator (ALFRED) is analyzed to determine the behavior of the essential parameters due to effects of mass flow and temperature of liquid lead. The ALFRED core mathematical model developed in this work is fully transient, which takes into account the heat transfer in an annular fuel pellet design, the thermo-fluid in the core, and the neutronic processes, which are modeled with point kinetic with feedback fuel temperature and expansion effects. The sensitivity evaluated in terms of the relative standard deviation (RSD) showed that for 10% change in the core inlet flow, the response in thermal power change is 0.58%, and for 2.5% change in the inlet lead temperature is 1.87%. The regression analysis with mass flow rate as the predictor variable showed statistically valid cubic correlations for neutron flux and linear relationship neutron flux as a function of the lead temperature. No statistically valid correlation was observed for the reactivity as a function of the mass flow rate and for the lead temperature. These correlations are useful for the study, analysis, and design of any LFR.

  7. Considerations on the influence of fission products in whole core accidents

    International Nuclear Information System (INIS)

    Meyer Heine, A.; Pattoret, A.; Schmitz, F.

    1977-01-01

    If the hypothetical Whole Core Accidents which are taken into account in reactor safety analysis can change from one country to another, there is nevertheless a general agreement over their description and main phases. Furthermore the important parameters have also been identified by every laboratory. During the development of such core accidents the role of the fission products in essential. It is not the purpose of this paper to give an exhaustive description of the phases which can be influenced by the fission products, we will try however to focus this study on the most important ones. In a second step we will discuss the equation of state of irradiated fuels; here again one principal preoccupation being to quantify the influence of fission products on reactor accidents. It is not our purpose to enter on the fundamental aspects of the equation of state. The studies and the experimental program launched at the CEA will then be described. Special attention will be directed towards the eventual role of fission products in molten fuel-coolant interactions (MFCls) or the events leading to the initiation of whole core accidents. This paper will be limited to oxide fuels. Whether the whole core accident is initiated by a reactivity defect or a coolant coast-down, one has to deal with four great categories of phenomena. Loss of flow: the power is around the nominal value, while the coolant flow has been reduced by a factor of 5 to 10. This induces boiling and clad weakening. Will the plenum pressure lead to a clad rupture? In case of a rupture, what will be the effect on the voiding of the channel? Transient over power: influence of gases from gaseous and volatile fission products on the fuel movements? MFCIs: Influence of the fission products in the mode of contact between fuel and coolant? Influence on the fuel characteristics. Sodium vapour bubble expansion: influence of the fission products on the heat transfer and eventual condensation of the bubble?

  8. Whole core burnup calculations using 'MCNP'

    International Nuclear Information System (INIS)

    Haran, O.; Shaham, Y.

    1996-01-01

    Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors)

  9. Whole core burnup calculations using `MCNP`

    Energy Technology Data Exchange (ETDEWEB)

    Haran, O; Shaham, Y [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev

    1996-12-01

    Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors).

  10. Some concept for the TRIGA core design

    International Nuclear Information System (INIS)

    Aizawa, Otohiko

    1994-01-01

    There is the research reactor called TRIGA Mark-2 of 100 kW in Atomic Energy Research Laboratory, Musashi Institute of Technology. Recently, while the various calculations on the core were carried out, the author became aware of that this TRIGA core was designed at that time with excellent consideration. The reason for that is, although fuel is arranged in simple concentric circular state at a glance, it was known that in reality, this is the modification of the hexagonal core of triangular lattice. In the examination of square lattice fuel arrangement, the reactivity was calculated by using the gap between fuel rods as the parameter and by using ENDF/B-4 library and Monte Carlo code Keno-5. It is known that the design of the lattice with maximum reactivity cannot be done by the square lattice. The similar examination was carried out on triangular lattice, and it was found that the gap between fuel rods of 4 mm is the optimal design. The average neutron energy spectra in the fuel rods of the TRIGA Mark-2 core agreed considerably well with the energy spectra at 4.16 cm fuel rod pitch in triangular hexagonal core. In the reactor of about 100 kW, even if the gap between fuel rods is less than 4 mm, heat removal is sufficiently possible. (K.I.)

  11. An axially and radially two-zoned large liquid-metal fast breeder reactor core concept

    International Nuclear Information System (INIS)

    Kamei, T.; Arie, K.; Moriki, Y.; Suzuki, M.; Yamaoka, M.

    1985-01-01

    A new core concept that has advantages over conventional homogeneous cores in neutronics characteristics such as power peaking factor, burnup reactivity loss, and reactivity response to the movement of control rods in earthquakes has been evolved. Two options of the new core concept are feasible. One is the so-called axially heterogeneous core, with the internal blanket placed at the lower part of the core. The other concept is similar to the conventional homogeneous core, but has two different plutonium-enriched zones in the axial as well as in the radial direction, so it is a hybrid type of the conventional homogeneous core and the axially heterogeneous core. The new design concept is described and the way that the core characteristics are improved by the chosen key parameters is shown

  12. Feasibility study of SMART core with soluble boron

    International Nuclear Information System (INIS)

    Kim, Kang Seog; Lee, Chung Chan; Zee, Sung Quun

    2000-11-01

    The excess reactivity of SMART core without soluble boron is effectively controlled by 49 CEDM. We suggest another method to control the core excess reactivity using both the checkerboard type of 25 CEDM and soluble boron and perform a feasibility calculation. The soluble boron operation is categorized into the on-line and the off-line mechanisms. The former is to successively control the boron concentration according to the excess reactivity during operation and the latter is to add and change some soluble boron during refueling and repairing. Since the on-line soluble boron control system of SMART is conceptually identical to that of the commercial pressurized water reactor, we did not perform the analysis. Since the soluble boron in the complete off-line system increases the moderator temperature coefficient, the reactivity defect between hot and cold moderator temperature is decreased. However, the decrease of the reactivity is not big to satisfy the core reactivity limits. When using 25 CEDM, the possible mechanism is to control the excess reactivity by both control rod and on-line boron control mechanism between cold and hot zero power and by only control rod at hot full power. We selected the loading pattern satisfying the requirement in the view of nuclear design

  13. A core management system for JRR-3

    International Nuclear Information System (INIS)

    Soyama, Kazuhiko; Tsuruta, Harumichi; Ichikawa, Hiroki; Nemoto, Hiroyuki.

    1991-05-01

    Japan Research Reactor No.3 (JRR-3) was upgraded to the thermal output with 20 MW by replacing the core, cooling system and utilization facilities. It is a water moderated and cooled, pool type reactor using 20% enriched U · Alx fuel. A core management system for JRR-3 has been made. This code system can manage of reactivity, power distribution and burn up in consideration of the position of control rod, fuel arrangement and operation pattern. This report is the user's manual of this code system. (author)

  14. Digital reactivity meter

    International Nuclear Information System (INIS)

    Copie, M.; Valantic, B.

    1978-01-01

    Digital reactivity meters (DRM) are mostly used as measuring instruments, e.g. for calibration of control rods, and there are only a few cases of their incorporation into the control systems of the reactors. To move in this direction there is more development work needed. First of all, fast algorithms are needed for inverse kinetics equations to relieve the computer for more important tasks of reactor model solving in real time. The next problem, currently under investigation, is the incorporation of the reactor thermal-hydraulic model into the DRM so that it can be used in the power range. Such an extension of DHM allows presentation not only of the instantaneous reactivity of the system, but also the inserted reactivity can be estimated from the temperature reactivity feed-backs. One of the applications of this concept is the anomalous digital reactivity monitor (ADRN) as part of the reactor protection system. As a solution of the first problem, a fast algorithm for solving the inverse kinetics equations has been implemented in the off-line program RODCAL on CDC 1700 computer and tested for its accuracy by performing different control rod calibrations on the reactor TRIGA

  15. CP ESFR: Collaborative Project for a European Sodium Fast Reactor Core studies

    International Nuclear Information System (INIS)

    Buiron, L.; Vasile, A.; Sunderland, R.

    2013-01-01

    • Significant progress has been made in optimizing both the oxide and carbide ESFR cores; • For the oxide core the optimisation process concentrated on the reduction of the sodium void reactivity effect and on the evaluation of MA burning performances. The CONF2 axial configuration has provided a significant overall reduction of the sodium void reactivity effect. • The carbide core had a significantly higher reactivity loss over the fuel cycle compared to the oxide one. By increasing slightly the fuel pin diameter, whilst still retaining the advantages of lower fuel temperatures of carbide fuel, and making changes in the core layout, the reactivity loss over the cycle has been reduced to a level similar to that of the oxide core. By adopting the CONF2 axial configuration initially developed for the oxide core, the sodium void reactivity of the carbide core has also been reduced appreciably. • The MA transmutation performances of the optimized ESFR oxide core have been investigated with respect to two boundary configurations. The HET2 configuration shows a low MA transmutation rate sufficient to burn the MA produced by the ESFR core without affecting the safety parameters. The HOM4 configuration (where 4%wt. MA are loaded homogeneously in each core SA) is the most challenging configuration due to its impact on safety coefficients but it shows an high MA burning rate suitable for burning also MA accumulated by a thermal reactor fleet

  16. Analytical estimation of control rod shadowing effect for excess reactivity measurement of HTTR

    International Nuclear Information System (INIS)

    Nakano, Masaaki; Fujimoto, Nozomu; Yamashita, Kiyonobu

    1999-01-01

    The fuel addition method is generally used for the excess reactivity measurement of the initial core. The control rod shadowing effect for the excess reactivity measurement has been estimated analytically for High Temperature Engineering Test Reactor (HTTR). 3-dimensional whole core analyses were carried out. The movements of control rods in measurements were simulated in the calculation. It was made clear that the value of excess reactivity strongly depend on combinations of measuring control rods and compensating control rods. The differences in excess reactivity between combinations come from the control rod shadowing effect. The shadowing effect is reduced by the use of plural number of measuring and compensating control rods to prevent deep insertion of them into the core. The measured excess reactivity in the experiments is, however, smaller than the estimated value with shadowing effect. (author)

  17. Determination of the design excess reactivity for the TREAT Upgrade reactor

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.; Hanan, N.A.

    1983-01-01

    The excess reactivity designed to be built into a reactor core is a primary determinant of the fissile loadings of the fuel rods in the core. For the TREAT Upgrade (TU) reactor the considerations that enter into the determination of the excess reactivity are different from those of conventional power reactors. The reactor is designed to operate in an adiabatic transient mode for reactor safety in-pile test programs. The primary constituent of the excess reactivity is the calculated reactivity required to perform the most demanding transient experiments. Because of the unavailability of supporting critical experiments for the core design, the uncertainty terms that add on to this basic constituent are rather large. The burnup effects in TU are negligible and no refueling is planned. In this paper the determination of the design excess reactivity of the TREAT Upgrade reactor is discussed

  18. Core Hunter 3: flexible core subset selection.

    Science.gov (United States)

    De Beukelaer, Herman; Davenport, Guy F; Fack, Veerle

    2018-05-31

    Core collections provide genebank curators and plant breeders a way to reduce size of their collections and populations, while minimizing impact on genetic diversity and allele frequency. Many methods have been proposed to generate core collections, often using distance metrics to quantify the similarity of two accessions, based on genetic marker data or phenotypic traits. Core Hunter is a multi-purpose core subset selection tool that uses local search algorithms to generate subsets relying on one or more metrics, including several distance metrics and allelic richness. In version 3 of Core Hunter (CH3) we have incorporated two new, improved methods for summarizing distances to quantify diversity or representativeness of the core collection. A comparison of CH3 and Core Hunter 2 (CH2) showed that these new metrics can be effectively optimized with less complex algorithms, as compared to those used in CH2. CH3 is more effective at maximizing the improved diversity metric than CH2, still ensures a high average and minimum distance, and is faster for large datasets. Using CH3, a simple stochastic hill-climber is able to find highly diverse core collections, and the more advanced parallel tempering algorithm further increases the quality of the core and further reduces variability across independent samples. We also evaluate the ability of CH3 to simultaneously maximize diversity, and either representativeness or allelic richness, and compare the results with those of the GDOpt and SimEli methods. CH3 can sample equally representative cores as GDOpt, which was specifically designed for this purpose, and is able to construct cores that are simultaneously more diverse, and either are more representative or have higher allelic richness, than those obtained by SimEli. In version 3, Core Hunter has been updated to include two new core subset selection metrics that construct cores for representativeness or diversity, with improved performance. It combines and outperforms the

  19. Experimental Investigation of Reynolds Number Effects on Test Quality in a Hypersonic Expansion Tube

    Science.gov (United States)

    Rossmann, Tobias; Devin, Alyssa; Shi, Wen; Verhoog, Charles

    2017-11-01

    Reynolds number effects on test time and the temporal and spatial flow quality in a hypersonic expansion tube are explored using high-speed pressure, infrared optical, and Schlieren imaging measurements. Boundary layer models for shock tube flows are fairly well established to assist in the determination of test time and flow dimensions at typical high enthalpy test conditions. However, the application of these models needs to be more fully explored due to the unsteady expansion of turbulent boundary layers and contact regions separating dissimilar gasses present in expansion tube flows. Additionally, expansion tubes rely on the development of a steady jet with a large enough core-flow region at the exit of the acceleration tube to create a constant velocity region inside of the test section. High-speed measurements of pressure and Mach number at several locations within the expansion tube allow for the determination of an experimental x-t diagram. The comparison of the experimentally determined x-t diagram to theoretical highlights the Reynolds number dependent effects on expansion tube. Additionally, spatially resolved measurements of the Reynolds number dependent, steady core-flow in the expansion tube viewing section are shown. NSF MRI CBET #1531475, Lafayette College, McCutcheon Foundation.

  20. Core design for use with precision composite reflectors

    Science.gov (United States)

    Porter, Christopher C. (Inventor); Jacoy, Paul J. (Inventor); Schmitigal, Wesley P. (Inventor)

    1992-01-01

    A uniformly flexible core, and method for manufacturing the same, is disclosed for use between the face plates of a sandwich structure. The core is made of a plurality of thin corrugated strips, the corrugations being defined by a plurality of peaks and valleys connected to one another by a plurality of diagonal risers. The corrugated strips are orthogonally criss-crossed to form the core. The core is particularly suitable for use with high accuracy spherically curved sandwich structures because undesirable stresses in the curved face plates are minimized due to the uniform flexibility characteristics of the core in both the X and Y directions. The core is self venting because of the open geometry of the corrugations. The core can be made from any suitable composite, metal, or polymer. Thermal expansion problems in sandwich structures may be minimized by making the core from the same composite materials that are selected in the manufacture of the curved face plates because of their low coefficients of thermal expansion. Where the strips are made of a composite material, the core may be constructed by first cutting an already cured corrugated sheet into a plurality of corrugated strips and then secondarily bonding the strips to one another or, alternatively, by lying a plurality of uncured strips orthogonally over one another in a suitable jig and then curing and bonding the entire plurality of strips to one another in a single operation.

  1. Reactivity anomaly surveillance in the Fast Flux Test Facility through cycle 3

    International Nuclear Information System (INIS)

    Knutson, B.J.; Harris, R.A.

    1984-08-01

    The technique for monitoring core reactivity during power operation used at the Fast Flux Test Facility (FFTF) is described. This technique relies on comparing predicted to measured rod positions to detect any anomalous (or unpredicted) core reactivity changes. It is implemented on the Plant Data System (PDS) computer and thus provides rapid indication of any abnormal core conditions. The prediction algorithms use thermal-hydraulic, control rod position and neutron flux sensor information to predict the core reactivity state. Initial results of using this technique based mainly on theoretical formulations is presented. The results show that the reactivity changes due to increasing reactor power (power defect) and burnup of the fuel were within approx. 16% of predicted values. To increase the sensitivity and accuracy of this technique, the prediction algorithms were calibrated to actual operating data. The work of calibrating this technique and the results of using the calibrated technique up through the third full operating cycle are summarized

  2. A report on core 5 of ZENITH

    International Nuclear Information System (INIS)

    Barclay, F.R.; Cameron, I.R.; Freemantle, R.G.; Reed, D.L.; Wilson, D.J.

    1964-06-01

    The determination of the excess reactivity, control rod worths, flux fine structure, temperature coefficients, differential spectra and reaction rates of various nuclides for the fifth loading of the heated zero energy reactor ZENITH is described. The core contained 32.6 Kgm of U235, giving a carbon/U235 atomic ratio of 1072, and formed the least moderated of the range studied. Comparisons of the experimental results with calculations using multigroup diffusion codes are presented. (author)

  3. Measurement of xenon reactivity in the reactor of the nuclear ship 'MUTSU'

    International Nuclear Information System (INIS)

    Itagaki, Masafumi; Miyoshi, Yoshinori; Gakuhari, Kazuhiko; Okada, Noboru.

    1993-01-01

    This report deals with the measurement of reactivity changes caused by the increase and decrease of xenon concentration in the reactor core of the nuclear ship 'MUTSU' after a change from long-term operation at 70 % to zero power. The change in xenon reactivity was compensated by control-rod movements and the compensated reactivity was measured using a digital reactivity meter. The xenon override peak was recognized five and half hours after the start of power reduction. The equilibrium and peak reactivities of xenon were estimated by reading the initial and peak values of a theoretical curve which was fitted to the measured variation in xenon reactivity. The xenon reactivity results obtained by the present method can be considered to be accurate since no control-rod worth data were used and the measured quantity was the reactivity itself. (author)

  4. k-core covers and the core

    NARCIS (Netherlands)

    Sanchez-Rodriguez, E.; Borm, Peter; Estevez-Fernandez, A.; Fiestras-Janeiro, G.; Mosquera, M.A.

    This paper extends the notion of individual minimal rights for a transferable utility game (TU-game) to coalitional minimal rights using minimal balanced families of a specific type, thus defining a corresponding minimal rights game. It is shown that the core of a TU-game coincides with the core of

  5. Development of a Two-dimensional Thermohydraulic Hot Pool Model and ITS Effects on Reactivity Feedback during a UTOP in Liquid Metal Reactors

    International Nuclear Information System (INIS)

    Lee, Yong Bum; Jeong, Hae Yong; Cho, Chung Ho; Kwon, Young Min; Ha, Kwi Seok; Chang, Won Pyo; Suk, Soo Dong; Hahn, Do Hee

    2009-01-01

    The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect

  6. Exponential Expansion in Evolutionary Economics

    DEFF Research Database (Denmark)

    Frederiksen, Peter; Jagtfelt, Tue

    2013-01-01

    This article attempts to solve current problems of conceptual fragmentation within the field of evolutionary economics. One of the problems, as noted by a number of observers, is that the field suffers from an assemblage of fragmented and scattered concepts (Boschma and Martin 2010). A solution...... to this problem is proposed in the form of a model of exponential expansion. The model outlines the overall structure and function of the economy as exponential expansion. The pictographic model describes four axiomatic concepts and their exponential nature. The interactive, directional, emerging and expanding...... concepts are described in detail. Taken together it provides the rudimentary aspects of an economic system within an analytical perspective. It is argued that the main dynamic processes of the evolutionary perspective can be reduced to these four concepts. The model and concepts are evaluated in the light...

  7. Production expansion continues to accelerate

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This paper reports that Saudi Arabian Oil Co. (Saudi Aramco) is continuing its accelerated Crude Oil Expansion Program initiated in 1989 that aims at achieving a 10 million bpd productive capacity by 1995. In addition to major engineering, construction and renovation work related to production expansion, Saudi Aramco drilling and workover operations have been markedly expanded. Since January 1991, rig activity has doubled. As an indication of aging of Saudi production, projects include modernizing current injection water treatment facilities, installing a new seawater injection plant on the Persian Gulf, installing dewatering facilities in a number of locations and installing a pilot gas lift project. In addition, equipment orders indicate the new discoveries south of Riyadh may also need the assistance of water injection from inception of production

  8. Shrub expansion in SW Greenland

    DEFF Research Database (Denmark)

    Jørgensen, Rasmus Halfdan

    Arctic regions have experienced higher temperatures in recent decades, and the warming trend is projected to continue in the coming years. Arctic ecosystems are considered to be particularly vulnerable to climate change. Expansion of shrubs has been observed widely in tundra areas across the Arctic......, and has a range of ecosystem effects where it occurs. Shrub expansion has to a large extend been attributed to increasing temperatures over the past century, while grazing and human disturbance have received less attention. Alnus viridis ssp. crispa is a common arctic species that contributes...... to increasing shrub cover. Despite this, there is only limited experimental evidence that growth of the species responds to warming. Plant populations in fragmented and isolated locations could face problems adapting to a warming climate due to limited genetic variation and restricted migration from southern...

  9. Neutronic calculations of PARR-1 cores using LEU silicide fuel

    International Nuclear Information System (INIS)

    Arshad, M.; Bakhtyar, S.; Hayat, T.; Salahuddin, A.

    1991-08-01

    Detailed neutronic calculations have been carried out for different PARR-1 cores utilizing low enriched uranium (LEU) silicide fuel and operating at an upgraded power of 9 MW. The calculations include the search for critical loadings in open and stall ends of the pool, neutronic analysis of the first full equilibrium core and calculations cores. The burnup study of inventory have also been carried out. Further, the reactivity coefficients of the first full power operation core are evaluated for use in the accident analysis. 14 figs. (author)

  10. Assessment of CANDU-6 reactivity devices for DUPIC fuel

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Choi, Hang Bok

    1998-11-01

    Reactivity device characteristics for a CANDU 6 reactor loaded with DUPIC fuel have been assessed. The lattice parameters were generated by WIMS-AECL code and the core calculations were performed by RFSP code with a 3-dimensional full core model. The reactivity devices studied are the zone controller, adjusters, mechanical control absorber and shutoff rods. For the zone controller system, damping capability for spatial oscillation was investigated. For the adjusters, the restart capability was investigated. For the adjusters, the restart capability was investigated. The shin operation and power stepback calculation were also performed to confirm the compatibility of the current adjuster system. The mechanical control absorber was assessed for the function of compensating temperature reactivity feedback following a power reduction. And shutoff rods were also assessed to investigate the following a power reduction. And shutoff rods were also assessed to investigate the static reactivity worth. This study has shown that the current reactivity device system of CANDU-6 core with the DUPIC fuel. (author). 9 refs., 17 tabs., 7 figs

  11. Probabilistic method for evaluating reactivity margin of nuclear reactors

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko

    1984-01-01

    A probabilistic method is proposed that will permit in the design stage to estimate quantitatively the likelihood with which any or all design criteria applicable to a nuclear reactor are actually satisfied after its construction. The method is trially applied to the core reactivity balance problem of the experimental Very High Temperature Reactor, and calculations are performed on the probability with which a design study core will, upon construction, satisfy design criteria concerning (a) one rod stuck and (b) startup margin. The method should prove useful in making engineering judgments before approving reactor core design. (author)

  12. RELIABILITY OF LENTICULAR EXPANSION COMPENSATORS

    Directory of Open Access Journals (Sweden)

    Gabriel BURLACU,

    2011-11-01

    Full Text Available Axial lenticular compensators are made to take over the longitudinal heat expansion, shock , vibration and noise, made elastic connections for piping systems. In order to have a long life for installations it is necessary that all elements, including lenticular compensators, have a good reliability. This desire can be did by technology of manufactoring and assembly of compensators, the material for lenses and by maintenance.of compensator

  13. Effective neutron temperature measurements in well moderated reactor by the reactivity coefficient method

    International Nuclear Information System (INIS)

    Raisic, N.; Klinc, T.

    1968-11-01

    The ratio of the reactivity changes of a nuclear reactor produced by successive introduction of two different neutron absorbers in the reactor core, has been measured and information on effective neutron temperature at a particular point obtained. Boron was used as a l/v absorber and cadmium as an absorber sensiti ve to neutron temperature. Effective neutron temperature distribution has been deduced by moving absorbers across the reactor core and observing the corresponding reactivity changes. (author)

  14. Reactive diffusion and stresses in nanowires or nanorods

    International Nuclear Information System (INIS)

    Roussel, Manuel; Erdélyi, Zoltán; Schmitz, Guido

    2017-01-01

    Heterostructured nanowires are of prime interest in nowadays technology such as field-effect transistors, field emitters, batteries and solar cells. We consider their aging behavior and developed a model focusing on reactive diffusion in core-shell nanowires. A complete set of analytical equations is presented that takes into account thermodynamic driving forces, vacancy distribution, elastic stress and its plastic relaxation. This complete description of the reactive diffusion can be used in finite element simulations to investigate diffusion processes in various geometries. In order to show clearly the interplay between the cylindrical geometry, the reactive diffusion and the stresses developing in the nanowire, we investigate the formation of an intermetallic reaction product in various core-shell geometries. Emphasis is placed on showing how it is possible to control the kinetics of the reaction by applying an axial stress to the nanowires.

  15. The 3-dimensional core model DYN3D

    Energy Technology Data Exchange (ETDEWEB)

    Grundmann, U.; Mittag, S.; Rohde, U.

    1999-01-01

    Analyzing the safety margins in transients and accidents of nuclear reactors 3-dimensional models of the core were used to avoid conservative assumptions needed for point kinetics or 1-dimensional models. Therefore, the 3D code DYN3D has been developed for the analysis of reactivity initiated accidents (RIA) in thermal nuclear reactors. The power distributions are calculated with the help of nodal expansion methods (NEM) for hexagonal and Cartesian geometry. The fuel rod model and the thermohydraulic part provide fuel temperatures, coolant temperatures and densities as well as boron concentrations for the calculation of feedback effects on the basis of cross section libraries generated by cell codes. Safety relevant parameters like maximum fuel and cladding temperatures, critical heat flux and degree of cladding oxidation are estimated. DYN3D can analyze RIA initiated by moved control rods and/or perturbations of the coolant flow. Stationary and transient boundary conditions for the coolant flow, the core inlet temperatures and boron concentrations at the core inlet have to be given. For analyzing more complex transients the code DYN3D is coupled with the plant model ATHLET of the GRS. The extensive validation work accomplished for DYN3D is presented in several examples. Some applications of the code are described. (orig.) [Deutsch] Die Verwendung 3-dimensionaler Kernmodelle zur Untersuchung der Sicherheitsreserven bei Uebergangsprozessen und Stoerfaellen in Kernreaktoren vermeidet konservative Annahmen, die bei der Benutzung des Punktmodells oder 1-dimensionaler Modelle erforderlich sind. Aus diesen Gruenden wurde das 3-dimensionale Rechenprogramm DYN3D fuer die Untersuchung von Reaktivitaetsstoerfaellen in thermischen Reaktoren entwickelt. Die Leistungsverteilung wird mit nodalen Methoden fuer hexagonale oder kartesische Geometrie berechnet. Das Brennstabmodell und der thermohydraulische Teil von DYN3D liefert die Brennstofftemperaturen, Kuehlmitteltemperaturen

  16. MIMI: Multimodality, Multiresource, Information Integration Environment for Biomedical Core Facilities

    OpenAIRE

    Szymanski, Jacek; Wilson, David L.; Zhang, Guo-Qiang

    2007-01-01

    The rapid expansion of biomedical research has brought substantial scientific and administrative data management challenges to modern core facilities. Scientifically, a core facility must be able to manage experimental workflow and the corresponding set of large and complex scientific data. It must also disseminate experimental data to relevant researchers in a secure and expedient manner that facilitates collaboration and provides support for data interpretation and analysis. Administrativel...

  17. Spring 5 & reactive streams

    CERN Multimedia

    CERN. Geneva; Clozel, Brian

    2017-01-01

    Spring is a framework widely used by the world-wide Java community, and it is also extensively used at CERN. The accelerator control system is constituted of 10 million lines of Java code, spread across more than 1000 projects (jars) developed by 160 software engineers. Around half of this (all server-side Java code) is based on the Spring framework. Warning: the speakers will assume that people attending the seminar are familiar with Java and Spring’s basic concepts. Spring 5.0 and Spring Boot 2.0 updates (45 min) This talk will cover the big ticket items in the 5.0 release of Spring (including Kotlin support, @Nullable and JDK9) and provide an update on Spring Boot 2.0, which is scheduled for the end of the year. Reactive Spring (1h) Spring Framework 5.0 has been released - and it now supports reactive applications in the Spring ecosystem. During this presentation, we'll talk about the reactive foundations of Spring Framework with the Reactor project and the reactive streams specification. We'll al...

  18. Reactivity of nitriles

    International Nuclear Information System (INIS)

    Kukushkin, Yu.N.

    1987-01-01

    Reactivity of coordination nitriles in transition metal (Ru, Mo, W, Zr, Hf) complexes, namely: transformation of nitriles of the first coordination sphere into N-acyl-substituted amides, amidines, nitrile interaction; with water, alkalines, alcoholes, hydrogen, azide and cyanide ions is considered. Introduction of acetonitrile molecule to uranium (4)-carbon double bond is discussed

  19. Clojure reactive programming

    CERN Document Server

    Borges, Leonardo

    2015-01-01

    If you are a Clojure developer who is interested in using Reactive Programming to build asynchronous and concurrent applications, this book is for you. Knowledge of Clojure and Leiningen is required. Basic understanding of ClojureScript will be helpful for the web chapters, although it is not strictly necessary.

  20. A Universal Reactive Machine

    DEFF Research Database (Denmark)

    Andersen, Henrik Reif; Mørk, Simon; Sørensen, Morten U.

    1997-01-01

    Turing showed the existence of a model universal for the set of Turing machines in the sense that given an encoding of any Turing machine asinput the universal Turing machine simulates it. We introduce the concept of universality for reactive systems and construct a CCS processuniversal...

  1. Chemical Reactivity Test (CRT)

    Energy Technology Data Exchange (ETDEWEB)

    Zaka, F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-12-13

    The Chemical Reactivity Test (CRT) is used to determine the thermal stability of High Explosives (HEs) and chemical compatibility between (HEs) and alien materials. The CRT is one of the small-scale safety tests performed on HE at the High Explosives Applications Facility (HEAF).

  2. Reactive power compensating system

    Energy Technology Data Exchange (ETDEWEB)

    Williams, Timothy J. (Redondo Beach, CA); El-Sharkawi, Mohamed A. (Renton, WA); Venkata, Subrahmanyam S. (Seattle, WA)

    1987-01-01

    The reactive power of an induction machine is compensated by providing fixed capacitors on each phase line for the minimum compensation required, sensing the current on one line at the time its voltage crosses zero to determine the actual compensation required for each phase, and selecting switched capacitors on each line to provide the balance of the compensation required.

  3. Reactive Power Compensating System.

    Energy Technology Data Exchange (ETDEWEB)

    Williams, Timothy J.; El-Sharkawi, Mohamed A.; Venkata, Subrahmanyam S.

    1985-01-04

    The circuit was designed for the specific application of wind-driven induction generators. It has great potential for application in any situation where a varying reactive power load is present, such as with induction motors or generators, or for transmission network compensation.

  4. The iodine reactivity

    International Nuclear Information System (INIS)

    2003-01-01

    The iodine is an important element because it has long life isotopes (such as iodine 129) and a great mobility in natural media. Iodine presents a complex chemistry because of its volatility and its strong redox reactivity. The S.E.C.R. works to better understand the reactivity of this element in different natural, industrial or biological environments. It plays a part in thermochemical sites as a possible way of hydrogen formation. This seminar gives some aspects relative to the chemical reactivity of iodine, since its thermochemistry in the I/S cycles to produce hydrogen to its reactivity in the natural medium and its potential radiological impact. This document includes 4 presentations transparencies) dealing with: the 129 I cycle rejected in the low radioactive gaseous and liquid effluents of the La Hague reprocessing plant (C. Frechou); a bibliographic review of iodine retention in soils (F. Bazer-Bachi); the hydrogen production and the iodine/sulfur thermochemical cycle (role of iodine in the process); and the direct characterization by electro-spray ionization mass spectroscopy of iodine fixation by fulvic acids (P. Reiller, B. Amekraz, C. Moulin, V. Moulin)

  5. Upscaling of reactive flows

    NARCIS (Netherlands)

    Kumar, K.

    2012-01-01

    The thesis deals with the upscaling of reactive flows in complex geometry. The reactions which may include deposition or dissolution take place at a part of the boundary and depending on the size of the reaction domain, the changes in the pore structure that are due to the deposition process may or

  6. Effect of Strain on the Reactivity of Metal Surfaces

    DEFF Research Database (Denmark)

    Mavrikakis, Manos; Hammer, Bjørk; Nørskov, Jens Kehlet

    1998-01-01

    Self-consistent density functional calculations for the adsorption of O and CO, and the dissociation of CO on strained and unstrained Ru(0001) surfaces are used to show how strained metal surfaces have chemical properties that are significantly different from those of unstrained surfaces. Surface...... reactivity increases with lattice expansion, following a concurrent up-shift of the metal d states. Consequences for the catalytic activity of thin metal overlayers are discussed....

  7. Criticality experiment for No.2 core of DF-VI fast neutron criticality facility

    International Nuclear Information System (INIS)

    Yang Lijun; Liu Zhenhua; Yan Fengwen; Luo Zhiwen; Chu Chun; Liang Shuhong

    2007-01-01

    At the completion of the DF-VI fast neutron criticality facility, its core changed, and it was restarted and a series of experiments and measurements were made. According to the data from 29 criticality experiments, the criticality element number and mass were calculated, the control rod reactivity worth were measured by period method and rod compensate method, reactivity worth of safety rod and safety block were measured using reactivity instrument; the reactivity worth of outer elements and radial distribution of elements were measured too. Based on all the measurements mentioned above, safety operation parameters for core 2 in DF-VI fast neutron criticality facility were conformed. (authors)

  8. A concept of passive safety pressurized water reactor system with inherent matching nature of core heat generation and heat removal

    International Nuclear Information System (INIS)

    Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke

    1995-01-01

    The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation of scale-up capability in design were considered. The inherent matching nature of core heat generation and heat removal rate is introduced by the core with high reactivity coefficient for moderator density and low reactivity coefficient for fuel temperature (Doppler effect) and once-through steam generators (SGs). This nature makes the nuclear steam supply system physically-slave for the steam and energy conversion system by controlling feed water mass flow rate. The nature can be obtained by eliminating chemical shim and adopting in-vessel control rod drive mechanism (CRDM) units and a low power density core. In order to simplify the system, a large pressurizer, canned pumps, passive residual heat removal systems with air coolers as a final heat sink and passive coolant injection system are adopted and the functions of volume and boron concentration control and seal water supply are eliminated from the chemical and volume control system (CVCS). The emergency diesel generators and auxiliary component cooling system of 'safety class' for transferring heat to sea water as a final heat sink in emergency are also eliminated. All of systems are built in the containment except for the air coolers of the passive residual heat removal system. The analysis of the system revealed that the primary coolant expansion in 100% load reduction in 60 s can be mitigated in the pressurizer without actuating the pressure relief valves and the pressure in 50% load change in 30 s does not exceed the maximum allowable pressure in accidental conditions in regardless of pressure regulation. (author)

  9. A Power Series Expansion and Its Applications

    Science.gov (United States)

    Chen, Hongwei

    2006-01-01

    Using the power series solution of a differential equation and the computation of a parametric integral, two elementary proofs are given for the power series expansion of (arcsin x)[squared], as well as some applications of this expansion.

  10. Enhanced thermal expansion control rod drive lines for improving passive safety of fast reactors

    International Nuclear Information System (INIS)

    Edelmann, M.; Baumann, W.; Kuechle, M.; Kussmaul, G.; Vaeth, W.; Bertram, A.

    1992-01-01

    The paper presents a device for increasing the thermal expansion effect of control rod drive lines on negative reactivity feedback in fast reactors. The enhanced thermal expansion of this device can be utilized for both passive rod drop and forced insertion of absorbers in unprotected transients, e.g. ULOF. In this way the reactor is automatically brought into a permanently subcritical state and temperatures are kept well below the boiling point of the coolant. A prototype of such a device called ATHENa (German: Shut-down by THermal Expansion of Na) is presently under construction and will be tested. The paper presents the principle, design features and thermal properties of ATHENs as well as results of reactor dynamics calculations of ULOF's for EFR with enhanced thermal expansion control rod drive lines. (author)

  11. Spatial weighting of Doppler reactivity feedback

    International Nuclear Information System (INIS)

    Carew, J.F.; Diamond, D.J.; Todosow, M.

    1977-12-01

    The spatial weighting of the local Doppler feedback implicit in the determination of the core Doppler feedback reactivity has been investigated. Using a detailed planar PDQ7-II PWR model with local fuel-temperature feedback, the core Doppler spatial weight factor, S, has been determined for various control patterns and power levels. Assuming power-squared weighting of the local Doppler feedback, a simple analytic expression for S has been derived and, based on comparison with the PDQ7-II results, provides a convenient and accurate representation of the Doppler spatial weight factor. The sensitivity of these results to variations in the fuel rod heat transfer coefficients, fuel loading and the magnitude of the Doppler coefficient has also been evaluated. The dependence of the local Doppler coefficient on moderator temperature, boron concentration and control rod density has been determined and found to be weak. Selected comparisons with vendor analyses have been made and indicate general agreement

  12. Nuclear reactor core catcher

    International Nuclear Information System (INIS)

    1977-01-01

    A nuclear reactor core catcher is described for containing debris resulting from an accident causing core meltdown and which incorporates a method of cooling the debris by the circulation of a liquid coolant. (U.K.)

  13. Development of small, fast reactor core designs using lead-based coolant

    International Nuclear Information System (INIS)

    Cahalan, J. E.; Hill, R. N.; Khalil, H. S.; Wade, D. C.

    1999-01-01

    A variety of small (100 MWe) fast reactor core designs are developed, these include compact configurations, long-lived (15-year fuel lifetime) cores, and derated, natural circulation designs. Trade studies are described which identify key core design issues for lead-based coolant systems. Performance parameters and reactivity feedback coefficients are compared for lead-bismuth eutectic (LBE) and sodium-cooled cores of consistent design. The results of these studies indicate that the superior neutron reflection capability of lead alloys reduces the enrichment and burnup swing compared to conventional sodium-cooled systems; however, the discharge fluence is significantly increased. The size requirement for long-lived systems is constrained by reactivity loss considerations, not fuel burnup or fluence limits. The derated lead-alloy cooled natural circulation cores require a core volume roughly eight times greater than conventional compact systems. In general, reactivity coefficients important for passive safety performance are less favorable for the larger, derated configurations

  14. Seismic core shroud

    International Nuclear Information System (INIS)

    Puri, A.; Mullooly, J.F.

    1981-01-01

    A core shroud is provided, comprising: a coolant boundary, following the shape of the core boundary, for channeling the coolant through the fuel assemblies; a cylindrical band positioned inside the core barrel and surrounding the coolant boundary; and support members extending from the coolant boundary to the band, for transferring load from the coolant boundary to the band. The shroud may be assembled in parts using automated welding techniques, and it may be adjusted to fit the reactor core easily

  15. MTR fuel element burn-up measurements by the reactivity method

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  16. On the area expansion of magnetic flux tubes in solar active regions

    Energy Technology Data Exchange (ETDEWEB)

    Dudík, Jaroslav [DAMTP, CMS, University of Cambridge, Wilberforce Road, Cambridge CB3 0WA (United Kingdom); Dzifčáková, Elena [Astronomical Institute of the Academy of Sciences of the Czech Republic, Fričova 298, 251 65 Ondřejov (Czech Republic); Cirtain, Jonathan W., E-mail: J.Dudik@damtp.cam.ac.uk, E-mail: elena@asu.cas.cz [NASA Marshall Space Flight Center, VP 62, Huntsville, AL 35812 (United States)

    2014-11-20

    We calculated the three-dimensional (3D) distribution of the area expansion factors in a potential magnetic field, extrapolated from the high-resolution Hinode/SOT magnetogram of the quiescent active region NOAA 11482. Retaining only closed loops within the computational box, we show that the distribution of area expansion factors show significant structure. Loop-like structures characterized by locally lower values of the expansion factor are embedded in a smooth background. These loop-like flux tubes have squashed cross-sections and expand with height. The distribution of the expansion factors show an overall increase with height, allowing an active region core characterized by low values of the expansion factor to be distinguished. The area expansion factors obtained from extrapolation of the Solar Optical Telescope magnetogram are compared to those obtained from an approximation of the observed magnetogram by a series of 134 submerged charges. This approximation retains the general flux distribution in the observed magnetogram, but removes the small-scale structure in both the approximated magnetogram and the 3D distribution of the area expansion factors. We argue that the structuring of the expansion factor can be a significant ingredient in producing the observed structuring of the solar corona. However, due to the potential approximation used, these results may not be applicable to loops exhibiting twist or to active regions producing significant flares.

  17. Core Values | NREL

    Science.gov (United States)

    Core Values Core Values NREL's core values are rooted in a safe and supportive work environment guide our everyday actions and efforts: Safe and supportive work environment Respect for the rights physical and social environment Integrity Maintain the highest standard of ethics, honesty, and integrity

  18. Sidewall coring shell

    Energy Technology Data Exchange (ETDEWEB)

    Edelman, Ya A; Konstantinov, L P; Martyshin, A N

    1966-12-12

    A sidewall coring shell consists of a housing and a detachable core catcher. The core lifter is provided with projections, the ends of which are situated in another plane, along the longitudinal axis of the lifter. The chamber has corresponding projections.

  19. Sweden: Autonomous Reactivity Control (ARC) Systems

    International Nuclear Information System (INIS)

    Qvist, Staffan A.

    2015-01-01

    The next generation of nuclear energy systems must be licensed, constructed, and operated in a manner that will provide a competitively priced supply of energy, keeping in consideration an optimum use of natural resources, while addressing nuclear safety, waste, and proliferation resistance, and the public perception concerns of the countries in which those systems are deployed. These issues are tightly interconnected, and the implementation of passive and inherent safety features is a high priority in all modern reactor designs since it helps to tackle many of the issues at once. To this end, the Autonomous Reactivity Control (ARC) system was developed to ensure excellent inherent safety performance of Generation-IV reactors while having a minimal impact on core performance and economic viability. This paper covers the principles for ARC system design and analysis, the problem of ensuring ARC system response stability and gives examples of the impact of installing ARC systems in well-known fast reactor core systems. It is shown that even with a relatively modest ARC installation, having a near-negligible impact on core performance during standard operation, cores such as the European Sodium Fast Reactor (ESFR) can be made to survive any postulated unprotected transient without coolant boiling or fuel melting

  20. Immune reactivities against gums.

    Science.gov (United States)

    Vojdani, Aristo; Vojdani, Charlene

    2015-01-01

    Different kinds of gums from various sources enjoy an extremely broad range of commercial and industrial use, from food and pharmaceuticals to printing and adhesives. Although generally recognized as safe by the US Food and Drug Administration (FDA), gums have a history of association with sensitive or allergic reactions. In addition, studies have shown that gums have a structural, molecular similarity to a number of common foods. A possibility exists for cross-reactivity. Due to the widespread use of gums in almost every aspect of modern life, the overall goal of the current investigation was to determine the degree of immune reactivity to various gum antigens in the sera of individuals representing the general population. The study was a randomized, controlled trial. 288 sera purchased from a commercial source. The sera was screened for immunoglobulin G (IgG) and immunoglobulin E (IgE) antibodies against extracts of mastic gum, carrageenan, xantham gum, guar gum, gum tragacanth, locust bean gum, and β-glucan, using indirect enzyme-linked immunosorbent assay (ELISA) testing. For each gum antigen, inhibition testing was performed on the 4 sera that showed the highest IgG and IgE immune reactivity against the different gums used in the study. Inhibition testing on these same sera for sesame albumin, lentil, corn, rice, pineapple, peanut, pea protein, shrimp, or kidney bean was used to determine the cross-reactivity of these foods with the gum. Of the 288 samples, 4.2%-27% of the specimens showed a significant elevation in IgG antibodies against various gums. Only 4 of 288, or 1.4%, showed a simultaneous elevation of the IgG antibody against all 7 gum extracts. For the IgE antibody, 15.6%-29.1% of the specimens showed an elevation against the various gums. A significant percentage of the specimens, 12.8%, simultaneously produced IgE antibodies against all 7 tested extracts. Overall, the percentage of elevation in IgE antibodies against different gum extracts, with

  1. Strain expansion-reduction approach

    Science.gov (United States)

    Baqersad, Javad; Bharadwaj, Kedar

    2018-02-01

    Validating numerical models are one of the main aspects of engineering design. However, correlating million degrees of freedom of numerical models to the few degrees of freedom of test models is challenging. Reduction/expansion approaches have been traditionally used to match these degrees of freedom. However, the conventional reduction/expansion approaches are only limited to displacement, velocity or acceleration data. While in many cases only strain data are accessible (e.g. when a structure is monitored using strain-gages), the conventional approaches are not capable of expanding strain data. To bridge this gap, the current paper outlines a reduction/expansion technique to reduce/expand strain data. In the proposed approach, strain mode shapes of a structure are extracted using the finite element method or the digital image correlation technique. The strain mode shapes are used to generate a transformation matrix that can expand the limited set of measurement data. The proposed approach can be used to correlate experimental and analytical strain data. Furthermore, the proposed technique can be used to expand real-time operating data for structural health monitoring (SHM). In order to verify the accuracy of the approach, the proposed technique was used to expand the limited set of real-time operating data in a numerical model of a cantilever beam subjected to various types of excitations. The proposed technique was also applied to expand real-time operating data measured using a few strain gages mounted to an aluminum beam. It was shown that the proposed approach can effectively expand the strain data at limited locations to accurately predict the strain at locations where no sensors were placed.

  2. Expansion of protein domain repeats.

    Directory of Open Access Journals (Sweden)

    Asa K Björklund

    2006-08-01

    Full Text Available Many proteins, especially in eukaryotes, contain tandem repeats of several domains from the same family. These repeats have a variety of binding properties and are involved in protein-protein interactions as well as binding to other ligands such as DNA and RNA. The rapid expansion of protein domain repeats is assumed to have evolved through internal tandem duplications. However, the exact mechanisms behind these tandem duplications are not well-understood. Here, we have studied the evolution, function, protein structure, gene structure, and phylogenetic distribution of domain repeats. For this purpose we have assigned Pfam-A domain families to 24 proteomes with more sensitive domain assignments in the repeat regions. These assignments confirmed previous findings that eukaryotes, and in particular vertebrates, contain a much higher fraction of proteins with repeats compared with prokaryotes. The internal sequence similarity in each protein revealed that the domain repeats are often expanded through duplications of several domains at a time, while the duplication of one domain is less common. Many of the repeats appear to have been duplicated in the middle of the repeat region. This is in strong contrast to the evolution of other proteins that mainly works through additions of single domains at either terminus. Further, we found that some domain families show distinct duplication patterns, e.g., nebulin domains have mainly been expanded with a unit of seven domains at a time, while duplications of other domain families involve varying numbers of domains. Finally, no common mechanism for the expansion of all repeats could be detected. We found that the duplication patterns show no dependence on the size of the domains. Further, repeat expansion in some families can possibly be explained by shuffling of exons. However, exon shuffling could not have created all repeats.

  3. A digital instrument for reactivity measurements in a nuclear reactor

    International Nuclear Information System (INIS)

    Chwaszczewski, S.

    1979-01-01

    An instrument for digital determination of the reactivity in nuclear reactors is described. It is based on the CAMAC standard apparatus, suitable for the use of pulse or current type neutron detectors and operates with prompt response and an output signal proportional to the core neutron flux. The measured data of neutron flux and reactivity can be registered by a digital display unit, an indicator, or, by request of the operator, a paper type punch. The algorithms used for reactivity calculation are considered and the results of numerical studies on those algorithms are discussed. The instrument has been used for determining the reactivity of the control elements in the fast-thermal assembly ANNA and in the research reactor MARIA. Some results of these measurements are given. (author)

  4. Digital instrument for reactivity measurements in a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chwaszczewski, S [Institute of Nuclear Research, Warsaw (Poland)

    1979-07-01

    An instrument for digital determination of the reactivity in nuclear reactors is described. It is based on the CAMAC standard apparatus, suitable for the use of pulse or current type neutron detectors and operates with prompt response and an output signal proportional to the core neutron flux. The measured data of neutron flux and reactivity can be registered by a digital display unit, an indicator, or, by request of the operator, a paper type punch. The algorithms used for reactivity calculation are considered and the results of numerical studies on those algorithms are discussed. The instrument has been used for determining the reactivity of the control elements in the fast-thermal assembly ANNA and in the research reactor MARIA. Some results of these measurements are given.

  5. Assessment of reactivity devices for CANDU-6 with DUPIC fuel

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Choi, Hang Bok

    1998-01-01

    Reactivity device characteristics for a CANDU-6 reactor loaded with DUPIC fuel have been assessed. A transport code WIMS-AECL and a three-dimensional diffusion code RFSP were used for the lattice parameter generation and the core calculation, respectively. Three major reactivity devices have been assessed for their inherent functions. For the zone controller system, damping capability for spatial oscillation was investigated. The restart capability of the adjuster system was investigated. The shim operation and power stepback calculation were also performed to confirm the compatibility of the current adjuster rod system. The mechanical control absorber was assessed for the capability to compensate the temperature reactivity feedback following a power reduction. This study has shown that the current reactivity device systems retain their functions when used in a DUPIC fuel CANDU reactor

  6. Isothermal temperature reactivity coefficient measurement in TRIGA reactor

    International Nuclear Information System (INIS)

    Zagar, T.; Ravnik, M.; Trkov, A.

    2002-01-01

    Direct measurement of an isothermal temperature reactivity coefficient at room temperatures in TRIGA Mark II research reactor at Jozef Stefan Institute in Ljubljana is presented. Temperature reactivity coefficient was measured in the temperature range between 15 o C and 25 o C. All reactivity measurements were performed at almost zero reactor power to reduce or completely eliminate nuclear heating. Slow and steady temperature decrease was controlled using the reactor tank cooling system. In this way the temperatures of fuel, of moderator and of coolant were kept in equilibrium throughout the measurements. It was found out that TRIGA reactor core loaded with standard fuel elements with stainless steel cladding has small positive isothermal temperature reactivity coefficient in this temperature range.(author)

  7. SCRAM reactivity calculations with the KIKO3D code

    International Nuclear Information System (INIS)

    Hordosy, G.; Kerszturi, A.; Maraczy, Cs.; Temesvari, E.

    1999-01-01

    Discrepancies between calculated static reactivities and measured reactivities evaluated with reactivity meters led to investigating SCRAM with the KIKO3D dynamic code, The time and space dependent neutron flux in the reactor core during the rod drop measurement was calculated by the KIKO3D nodal diffusion code. For calculating the ionisation chamber signals the Green function technique was applied. The Green functions of ionisation chambers were evaluated via solving the neutron transport equation in the reflector regions with the MCNP Monte Carlo code. The detector signals during asymmetric SCRAM measurements were calculated and compared with measured data using the inverse point kinetics transformation. The sufficient agreement validates the KIKO3D code to determine the reactivities after SCRAM. (Authors)

  8. Nuclear fuel reprocessing expansion strategies

    International Nuclear Information System (INIS)

    Gallagher, J.M.

    1975-01-01

    A description is given of an effort to apply the techniques of operations research and energy system modeling to the problem of determination of cost-effective strategies for capacity expansion of the domestic nuclear fuel reprocessing industry for the 1975 to 2000 time period. The research also determines cost disadvantages associated with alternative strategies that may be attractive for political, social, or ecological reasons. The sensitivity of results to changes in cost assumptions was investigated at some length. Reactor fuel types covered by the analysis include the Light Water Reactor (LWR), High-Temperature Gas-Cooled Reactor (HTGR), and the Fast Breeder Reactor (FBR)

  9. Thermal expansion of LATGS crystals

    International Nuclear Information System (INIS)

    Kassem, M.E.; Kandil, S.H.; Hamed, A.E.; Stankowska, J.

    1989-04-01

    The thermal expansion of triglycine sulphate crystals doped with L-α alanine (LATGS) has been studied around the phase transition temperature (30-60 deg. C) using thermomechanical analysis TMA. With increasing the content of admixture, the transition temperature (T c ) was shifted towards higher values, while the relative changes in the dimension of the crystals (ΔL/L 0 ) of the studied directions varied both in the para- and ferroelectric phases. The transition width in the case of doped crystals was found to be broad, and this broadening increases with increasing the content of L-α alanine. (author). 12 refs, 3 figs

  10. Contribution of thermal expansion and

    Directory of Open Access Journals (Sweden)

    O.I.Pursky

    2007-01-01

    Full Text Available A theoretical model is developed to describe the experimental results obtained for the isobaric thermal conductivity of rare gas solids (RGS. The isobaric thermal conductivity of RGS has been analysed within Debye approximation with regard to the effect of thermal expansion. The suggested model takes into consideration the fact that thermal conductivity is determined by U-processes while above the phonon mobility edge it is determined by "diffusive" modes migrating randomly from site to site. The mobility edge ω0 is determined from the condition that the phonon mean-free path restricted by the U-processes cannot be smaller than half of the phonon wavelength.

  11. Characterization of Coated Sand Cores from Two Different Binder Systems for Grey Iron Castings

    DEFF Research Database (Denmark)

    Nwaogu, Ugochukwu Chibuzoh; Tiedje, Niels Skat; Poulsen, Thomas

    or veining and metal penetration defects. The use of refractory coatings on cores is fundamental to obtaining acceptable casting surface quality and is used on resin bonded cores in production foundries. In this study new sol gel-coated sand cores made from coldbox and furan binder systems were investigated......Expansion defects on the surface of the castings include sand burn-in, metal penetration and/or veining, finning or scab. Veining or finning and metal penetration are of interest. These defects are associated with silica sand and result from the penetration of liquid metal into cracks formed during...... differential expansion of the core during heating. The rapid expansion of silica sand up to 600 oC and especially at 573 oC, where the α – β phase transformation occurs, is the cause of stresses in the core system. These stresses cause crack formation and metal melt flows into these cracks causing finning...

  12. Simple kinetic theory model of reactive collisions. IV. Laboratory fixed orientational cross sections

    International Nuclear Information System (INIS)

    Evans, G.T.

    1987-01-01

    The differential orientational cross section, obtainable from molecular beam experiments on aligned molecules, is calculated using the line-of-normals model for reactive collisions involving hard convex bodies. By means of kinetic theory methods, the dependence of the cross section on the angle of attack γ 0 is expressed in a Legendre function expansion. Each of the Legendre expansion coefficients is given by an integral over the molecule-fixed cross section and functions of the orientation dependent threshold energy

  13. An alternative solver for the nodal expansion method equations - 106

    International Nuclear Information System (INIS)

    Carvalho da Silva, F.; Carlos Marques Alvim, A.; Senra Martinez, A.

    2010-01-01

    An automated procedure for nuclear reactor core design is accomplished by using a quick and accurate 3D nodal code, aiming at solving the diffusion equation, which describes the spatial neutron distribution in the reactor. This paper deals with an alternative solver for nodal expansion method (NEM), with only two inner iterations (mesh sweeps) per outer iteration, thus having the potential to reduce the time required to calculate the power distribution in nuclear reactors, but with accuracy similar to the ones found in conventional NEM. The proposed solver was implemented into a computational system which, besides solving the diffusion equation, also solves the burnup equations governing the gradual changes in material compositions of the core due to fuel depletion. Results confirm the effectiveness of the method for practical purposes. (authors)

  14. A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor

    International Nuclear Information System (INIS)

    Yang, W.S.; Kim, T.K.; Grandy, C.

    2007-01-01

    This paper describes the core design and performance characteristics of a metal fuel core concept for a 1000 MWt Advanced Burner Reactor. A ternary metal fuel form of U-TRU-Zr was assumed with weapons grade plutonium feed for the startup core and TRU recovered from LWR spent fuel for the recycled equilibrium core. A compact burner core was developed by trade-off between the burnup reactivity loss and TRU conversion ratio, with a fixed cycle length of one-year. In the startup core, the average TRU enrichment is 15.5%, the TRU conversion ratio is 0.81, and the burnup reactivity loss over a cycle is 3.6% Δk. The heavy metal and TRU inventories are 13.1 and 2.0 metric tons, respectively. The average discharge burnup is 93 MWd/kg, and the TRU consumption rate is 55.5 kg/year. For the recycled equilibrium core, the average TRU enrichment is 22.1 %, the TRU conversion ratio is 0.73, and the burnup reactivity loss is 2.2% Δk. The TRU inventory and consumption rate are 2.9 metric tons and 81.6 kg/year, respectively. The evaluated reactivity coefficients provide sufficient negative feedbacks. The control systems provide shutdown margins that are more than adequate. The integral reactivity parameters for quasi-static reactivity balance analysis indicate favorable passive safety features, although detailed safety analyses are required to verify passive safety behavior. (authors)

  15. A feasibility study on wavelet transform for reactivity coefficient estimation

    International Nuclear Information System (INIS)

    Shimazu, Yoichiro

    2000-01-01

    Recently, a new method using Fourier transform has been introduced in place of the conventional method in order to reduce the time required for the measurement of moderator temperature coefficient in domestic PWRs. The basic concept of these methods is to eliminate noise in the reactivity signal. From this point of view, wavelet analysis is also known as an effective method. In this paper, we tried to apply this method to estimate reactivity coefficients of a nuclear reactor. The basic idea of the reactivity coefficient estimation is to analyze the ratios themselves of the corresponding expansion coefficients of the wavelet transform of the signals of reactivity and the relevant parameter. The concept requires no inverse wavelet transform. Based on numerical simulations, it is found that the method can reasonably estimate reactivity coefficient, for example moderator temperature coefficient, with less length of time sequence data than those required for Fourier transform method. We will continue this study to examine the validity of the estimation procedure for the actual reactor data and further to estimate the other reactivity coefficients. (author)

  16. What makes ecological systems reactive?

    Science.gov (United States)

    Snyder, Robin E

    2010-06-01

    Although perturbations from a stable equilibrium must ultimately vanish, they can grow initially, and the maximum initial growth rate is called reactivity. Reactivity thus identifies systems that may undergo transient population surges or drops in response to perturbations; however, we lack biological and mathematical intuition about what makes a system reactive. This paper presents upper and lower bounds on reactivity for an arbitrary linearized model, explores their strictness, and discusses their biological implications. I find that less stable systems (i.e. systems with long transients) have a smaller possible range of reactivities for which no perturbations grow. Systems with more species have a higher capacity to be reactive, assuming species interactions do not weaken too rapidly as the number of species increases. Finally, I find that in discrete time, reactivity is determined largely by mean interaction strength and neither discrete nor continuous time reactivity are sensitive to food web topology. 2010 Elsevier Inc. All rights reserved.

  17. Rotary core drills

    Energy Technology Data Exchange (ETDEWEB)

    1967-11-30

    The design of a rotary core drill is described. Primary consideration is given to the following component parts of the drill: the inner and outer tube, the core bit, an adapter, and the core lifter. The adapter has the form of a downward-converging sleeve and is mounted to the lower end of the inner tube. The lifter, extending from the adapter, is split along each side so that it can be held open to permit movement of a core. It is possible to grip a core by allowing the lifter to assume a closed position.

  18. Population increase in Kirtland's warbler and summer range expansion to Wisconsin and Michigan's Upper Peninsula, USA

    Science.gov (United States)

    John R. Probst; Deahn Donner; Carol I. Bocetti; Steve Sjogren

    2003-01-01

    The threatened Kirtland`s warbler Dendroica kirtlandii breeds in stands of young jack pine Pinus banksiana growing on well-drained soils in Michigan, USA. We summarize information documenting the range expansion of Kirtland`s warbler due to increased habitat management in the core breeding range in the Lower Peninsula of Michigan...

  19. Bearing for the reactivation

    International Nuclear Information System (INIS)

    Santamaria Alexandra

    2003-01-01

    Ecopetrol undertook an aggressive plan to reactivate the activities of seismic that allows fulfilling the goals proposed for this year (2003). Although the production registered a descent of 9%, the financial results throw utilities for $1.1 trillion pesos to the closing of September and contributions in bonuses for $1.2 trillions. The author also refers to the general balance, to the finances, raw production, taxes and transfers

  20. An evolutionary analysis of genome expansion and pathogenicity in Escherichia coli.

    Science.gov (United States)

    Bohlin, Jon; Brynildsrud, Ola B; Sekse, Camilla; Snipen, Lars

    2014-10-09

    There are several studies describing loss of genes through reductive evolution in microbes, but how selective forces are associated with genome expansion due to horizontal gene transfer (HGT) has not received similar attention. The aim of this study was therefore to examine how selective pressures influence genome expansion in 53 fully sequenced and assembled Escherichia coli strains. We also explored potential connections between genome expansion and the attainment of virulence factors. This was performed using estimations of several genomic parameters such as AT content, genomic drift (measured using relative entropy), genome size and estimated HGT size, which were subsequently compared to analogous parameters computed from the core genome consisting of 1729 genes common to the 53 E. coli strains. Moreover, we analyzed how selective pressures (quantified using relative entropy and dN/dS), acting on the E. coli core genome, influenced lineage and phylogroup formation. Hierarchical clustering of dS and dN estimations from the E. coli core genome resulted in phylogenetic trees with topologies in agreement with known E. coli taxonomy and phylogroups. High values of dS, compared to dN, indicate that the E. coli core genome has been subjected to substantial purifying selection over time; significantly more than the non-core part of the genome (pcoli genome size correlated with estimated HGT size (pcoli are largely attained through HGT. No associations were found between selective pressures operating on the E. coli core genome, as estimated using relative entropy, and genome size (p~0.98). On a larger time frame, genome expansion in E. coli, which is significantly associated with the acquisition of virulence factors, appears to be independent of selective forces operating on the core genome.

  1. High thermal expansion, sealing glass

    Science.gov (United States)

    Brow, R.K.; Kovacic, L.

    1993-11-16

    A glass composition is described for hermetically sealing to high thermal expansion materials such as aluminum alloys, stainless steels, copper, and copper/beryllium alloys, which includes between about 10 and about 25 mole percent Na[sub 2]O, between about 10 and about 25 mole percent K[sub 2]O, between about 5 and about 15 mole percent Al[sub 2]O[sub 3], between about 35 and about 50 mole percent P[sub 2]O[sub 5] and between about 5 and about 15 mole percent of one of PbO, BaO, and mixtures thereof. The composition, which may also include between 0 and about 5 mole percent Fe[sub 2]O[sub 3] and between 0 and about 10 mole percent B[sub 2]O[sub 3], has a thermal expansion coefficient in a range of between about 160 and 210[times]10[sup [minus]7]/C and a dissolution rate in a range of between about 2[times]10[sup [minus]7] and 2[times]10[sup [minus]9]g/cm[sup 2]-min. This composition is suitable to hermetically seal to metallic electrical components which will be subjected to humid environments over an extended period of time.

  2. Nonperturbative path integral expansion II

    International Nuclear Information System (INIS)

    Kaiser, H.J.

    1976-05-01

    The Feynman path integral representation of the 2-point function for a self-interacting Bose field is investigated using an expansion ('Path Integral Expansion', PIE) of the exponential of the kinetic term of the Lagrangian. This leads to a series - illustrated by a graph scheme - involving successively a coupling of more and more points of the lattice space commonly employed in the evaluation of path integrals. The values of the individual PIE graphs depend of course on the lattice constant. Two methods - Pade approximation and Borel-type extrapolation - are proposed to extract information about the continuum limit from a finite-order PIE. A more flexible PIE is possible by expanding besides the kinetic term a suitably chosen part of the interaction term too. In particular, if the co-expanded part is a mass term the calculation becomes only slightly more complicated than in the original formulation and the appearance of the graph scheme is unchanged. A significant reduction of the number of graphs and an improvement of the convergence of the PIE can be achieved by performing certain sums over an infinity of graph elements. (author)

  3. Power flattening and reactivity suppression strategies for the Canadian supercritical water reactor concept

    International Nuclear Information System (INIS)

    McDonald, M.; Colton, A.; Pencer, J.

    2015-01-01

    The Canadian supercritical water-cooled reactor (SCWR) is a conceptual heavy water moderated, supercritical light water cooled pressure tube reactor. In contrast to current heavy water power reactors, the Canadian SCWR will be a batch fuelled reactor. Associated with batch fuelling is a large beginning-of-cycle excess reactivity. Furthermore, radial power peaking arising as a consequence of batch refuelling must be mitigated in some way. In this paper, burnable neutron absorber (BNA) added to fuel and absorbing rods inserted into the core are considered for reactivity management and power flattening. A combination of approaches appears adequate to reduce the core radial power peaking, while also providing reactivity suppression. (author)

  4. Development of a computer-aided digital reactivity computer system for PWRs

    International Nuclear Information System (INIS)

    Chung, S.-K.; Sung, K.-Y.; Kim, D.; Cho, D.-Y.

    1993-01-01

    Reactor physics tests at initial startup and after reloading are performed to verify nuclear design and to ensure safety operation. Two kinds of reactivity computers, analog and digital, have been widely used in the pressurized water reactor (PWR) core physics test. The test data of both reactivity computers are displayed only on the strip chart recorder, and these data are managed by hand so that the accuracy of the test results depends on operator expertise and experiences. This paper describes the development of the computer-aided digital reactivity computer system (DRCS), which is enhanced by system management software and an improved system for the application of the PWR core physics test

  5. GAPER-1D, 1-D Multigroup 1. Order Perturbation Transport Theory for Reactivity Coefficient

    International Nuclear Information System (INIS)

    Koch, P.K.

    1976-01-01

    1 - Description of problem or function: Reactivity coefficients are computed using first-order transport perturbation theory for one- dimensional multi-region reactor assemblies. The number of spatial mesh-points and energy groups is arbitrary. An elementary synthesis scheme is employed for treatment of two- and three-dimensional problems. The contributions to the change in inverse multiplication factor, delta(1/k), from perturbations in the individual capture, net fission, total scattering, (n,2n), inelastic scattering, and leakage cross sections are computed. A multi-dimensional prompt neutron lifetime calculation is also available. 2 - Method of solution: Broad group cross sections for the core and perturbing or sample materials are required as input. Scalar neutron fluxes and currents, as computed by SN transport calculations, are then utilized to solve the first-order transport perturbation theory equations. A synthesis scheme is used, along with independent SN calculations in two or three dimensions, to treat a multi- dimensional assembly. Spherical harmonics expansions of the angular fluxes and scattering source terms are used with leakage and anisotropic scattering treated in a P1 approximation. The angular integrations in the perturbation theory equations are performed analytically. Various reactivity coefficients and material worths are then easily computed at specified positions in the assembly. 3 - Restrictions on the complexity of the problem: The formulation of the synthesis scheme used for two- and three-dimensional problems assumes that the fluxes and currents were computed by the DTF4 code (NESC Abstract 209). Therefore, fluxes and currents from two- or three-dimensional transport or diffusion theory codes cannot be used

  6. Research on plasma core reactors

    International Nuclear Information System (INIS)

    Jarvis, G.A.; Barton, D.M.; Helmick, H.H.; Bernard, W.; White, R.H.

    1977-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with 1-m-diam by 1-m-long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF 6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000-cm 3 aluminum canister in the central region was fueled with UF 6 gas and fission density distributions determined. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation

  7. Hepatitis B virus reactivation during immunosuppressive therapy: Appropriate risk stratification.

    Science.gov (United States)

    Seto, Wai-Kay

    2015-04-28

    Our understanding of hepatitis B virus (HBV) reactivation during immunosuppresive therapy has increased remarkably during recent years. HBV reactivation in hepatitis B surface antigen (HBsAg)-positive individuals has been well-described in certain immunosuppressive regimens, including therapies containing corticosteroids, anthracyclines, rituximab, antibody to tumor necrosis factor (anti-TNF) and hematopoietic stem cell transplantation (HSCT). HBV reactivation could also occur in HBsAg-negative, antibody to hepatitis B core antigen (anti-HBc) positive individuals during therapies containing rituximab, anti-TNF or HSCT.For HBsAg-positive patients, prophylactic antiviral therapy is proven to the effective in preventing HBV reactivation. Recent evidence also demonstrated entecavir to be more effective than lamivudine in this aspect. For HBsAg-negative, anti-HBc positive individuals, the risk of reactivations differs with the type of immunosuppression. For rituximab, a prospective study demonstrated the 2-year cumulative risk of reactivation to be 41.5%, but prospective data is still lacking for other immunosupressive regimes. The optimal management in preventing HBV reactivation would involve appropriate risk stratification for different immunosuppressive regimes in both HBsAg-positive and HBsAg-negative, anti-HBc positive individuals.

  8. Spatially Resolved Quantification of the Surface Reactivity of Solid Catalysts.

    Science.gov (United States)

    Huang, Bing; Xiao, Li; Lu, Juntao; Zhuang, Lin

    2016-05-17

    A new property is reported that accurately quantifies and spatially describes the chemical reactivity of solid surfaces. The core idea is to create a reactivity weight function peaking at the Fermi level, thereby determining a weighted summation of the density of states of a solid surface. When such a weight function is defined as the derivative of the Fermi-Dirac distribution function at a certain non-zero temperature, the resulting property is the finite-temperature chemical softness, termed Fermi softness (SF ), which turns out to be an accurate descriptor of the surface reactivity. The spatial image of SF maps the reactive domain of a heterogeneous surface and even portrays morphological details of the reactive sites. SF analyses reveal that the reactive zones on a Pt3 Y(111) surface are the platinum sites rather than the seemingly active yttrium sites, and the reactivity of the S-dimer edge of MoS2 is spatially anisotropic. Our finding is of fundamental and technological significance to heterogeneous catalysis and industrial processes demanding rational design of solid catalysts. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  9. Ultra-low thermal expansion realized in giant negative thermal expansion materials through self-compensation

    OpenAIRE

    Fei-Ran Shen; Hao Kuang; Feng-Xia Hu; Hui Wu; Qing-Zhen Huang; Fei-Xiang Liang; Kai-Ming Qiao; Jia Li; Jing Wang; Yao Liu; Lei Zhang; Min He; Ying Zhang; Wen-Liang Zuo; Ji-Rong Sun

    2017-01-01

    Materials with zero thermal expansion (ZTE) or precisely tailored thermal expansion are in urgent demand of modern industries. However, the overwhelming majority of materials show positive thermal expansion. To develop ZTE or negative thermal expansion (NTE) materials as compensators has become an important challenge. Here, we present the evidence for the realization of ultra-low thermal expansion in Mn–Co–Ge–In particles. The bulk with the Ni2In-type hexagonal structure undergoes giant NTE o...

  10. HYDRATE CORE DRILLING TESTS

    Energy Technology Data Exchange (ETDEWEB)

    John H. Cohen; Thomas E. Williams; Ali G. Kadaster; Bill V. Liddell

    2002-11-01

    The ''Methane Hydrate Production from Alaskan Permafrost'' project is a three-year endeavor being conducted by Maurer Technology Inc. (MTI), Noble, and Anadarko Petroleum, in partnership with the U.S. DOE National Energy Technology Laboratory (NETL). The project's goal is to build on previous and ongoing R&D in the area of onshore hydrate deposition. The project team plans to design and implement a program to safely and economically drill, core and produce gas from arctic hydrates. The current work scope includes drilling and coring one well on Anadarko leases in FY 2003 during the winter drilling season. A specially built on-site core analysis laboratory will be used to determine some of the physical characteristics of the hydrates and surrounding rock. Prior to going to the field, the project team designed and conducted a controlled series of coring tests for simulating coring of hydrate formations. A variety of equipment and procedures were tested and modified to develop a practical solution for this special application. This Topical Report summarizes these coring tests. A special facility was designed and installed at MTI's Drilling Research Center (DRC) in Houston and used to conduct coring tests. Equipment and procedures were tested by cutting cores from frozen mixtures of sand and water supported by casing and designed to simulate hydrate formations. Tests were conducted with chilled drilling fluids. Tests showed that frozen core can be washed out and reduced in size by the action of the drilling fluid. Washing of the core by the drilling fluid caused a reduction in core diameter, making core recovery very difficult (if not impossible). One successful solution was to drill the last 6 inches of core dry (without fluid circulation). These tests demonstrated that it will be difficult to capture core when drilling in permafrost or hydrates without implementing certain safeguards. Among the coring tests was a simulated hydrate

  11. Safety characteristics of the US advanced liquid metal reactor core

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Gyorey, G.L.; Lipps, A.J.; Wu, T.

    1991-01-01

    The U.S. Advanced Liquid Metal Reactor (ALMR) design employs innovative, passive features to provide an unprecedented level of public safety and the ability to demonstrate this safety to the public. The key features employed in the core design to produce the desired passive safety characteristics are: a small core with a tight restraint system, the use of metallic U-Pu-Zr fuel, control rod withdrawal limiters, and gas expansion modules. In addition, the reactor vessel and closure are designed to have the capability to withstand, with large margins, the maximum possible core disruptive accident without breach and radiological release. (author)

  12. Comparative sodium void effects for different advanced liquid metal reactor fuel and core designs

    International Nuclear Information System (INIS)

    Dobbin, K.D.; Kessler, S.F.; Nelson, J.V.; Gedeon, S.R.; Omberg, R.P.

    1991-01-01

    An analysis of metal-, oxide-, and nitride-fueled advanced liquid metal reactor cores was performed to investigate the calculated differences in sodium void reactivity, and to determine the relationship between sodium void reactivity and burnup reactivity swing using the three fuel types. The results of this analysis indicate that nitride fuel has the least positive sodium void reactivity for any given burnup reactivity swing. Thus, it appears that a good design compromise between transient overpower and loss of flow response is obtained using nitride fuel. Additional studies were made to understand these and other nitride advantages. (author)

  13. Automatic optimization of core loading patterns to maximize cycle energy production within operational constraints

    International Nuclear Information System (INIS)

    Hobson, G.H.; Turinsky, P.J.

    1986-01-01

    Computational capability has been developed to automatically determine the core loading pattern which minimizes fuel cycle costs for a pressurized water reactor. Equating fuel cycle cost minimization with core reactivity maximization, the objective is to determine the loading pattern which maximizes core reactivity at end-of-cycle while satisfying the power peaking constraint throughout the cycle and region average discharge burnup limit. The method utilizes a two-dimensional, coarse mesh, finite difference scheme to evaluate core reactivity and fluxes for an initial reference loading pattern as a function of cycle burnup. First order perturbation theory is applied to determine the effects of assembly shuffling on reactivity, power distribution, and end-of-cycle burnup

  14. Comparisons of significant parameters for a standard 20% enriched and FLIP 70% enriched TRIGA core

    International Nuclear Information System (INIS)

    Ringle, John C.; Anderson, Terrance V.; Johnson, Arthur G.

    1978-01-01

    A comparison is made between the 20% and 70% enriched cores. The initial start-up data for both cores show the FLIP needs ∼3.8 times the 235 U mass as the 20% core just to go critical. Operational configurations for both cores indicate a need for ∼33% additional fuel above initial critical for adequate maneuvering excess. The fuel element worths are higher in the central core locations for the 20% elements while the peripheral element worths are about the same (with some thermal flux peaking in the FLIP perheral elements). Pulsing comparisons of the two cores show significant differences in reactivity insertions and power peaks. (author)

  15. Aspects of cell calculations in deterministic reactor core analysis

    International Nuclear Information System (INIS)

    Varvayanni, M.; Savva, P.; Catsaros, N.

    2011-01-01

    Τhe capability of achieving optimum utilization of the deterministic neutronic codes is very important, since, although elaborate tools, they are still widely used for nuclear reactor core analyses, due to specific advantages that they present compared to Monte Carlo codes. The user of a deterministic neutronic code system has to make some significant physical assumptions if correct results are to be obtained. A decisive first step at which such assumptions are required is the one-dimensional cell calculations, which provide the neutronic properties of the homogenized core cells and collapse the cross sections into user-defined energy groups. One of the most crucial determinations required at the above stage and significantly influencing the subsequent three-dimensional calculations of reactivity, concerns the transverse leakages, associated to each one-dimensional, user-defined core cell. For the appropriate definition of the transverse leakages several parameters concerning the core configuration must be taken into account. Moreover, the suitability of the assumptions made for the transverse cell leakages, depends on earlier user decisions, such as those made for the core partition into homogeneous cells. In the present work, the sensitivity of the calculated core reactivity to the determined leakages of the individual cells constituting the core, is studied. Moreover, appropriate assumptions concerning the transverse leakages in the one-dimensional cell calculations are searched out. The study is performed examining also the influence of the core size and the reflector existence, while the effect of the decisions made for the core partition into homogenous cells is investigated. In addition, the effect of broadened moderator channels formed within the core (e.g. by removing fuel plates to create space for control rod hosting) is also examined. Since the study required a large number of conceptual core configurations, experimental data could not be available for

  16. ESFR core optimization and uncertainty studies

    International Nuclear Information System (INIS)

    Rineiski, A.; Vezzoni, B.; Zhang, D.; Marchetti, M.; Gabrielli, F.; Maschek, W.; Chen, X.-N.; Buiron, L.; Krepel, J.; Sun, K.; Mikityuk, K.; Polidoro, F.; Rochman, D.; Koning, A.J.; DaCruz, D.F.; Tsige-Tamirat, H.; Sunderland, R.

    2015-01-01

    In the European Sodium Fast Reactor (ESFR) project supported by EURATOM in 2008-2012, a concept for a large 3600 MWth sodium-cooled fast reactor design was investigated. In particular, reference core designs with oxide and carbide fuel were optimized to improve their safety parameters. Uncertainties in these parameters were evaluated for the oxide option. Core modifications were performed first to reduce the sodium void reactivity effect. Introduction of a large sodium plenum with an absorber layer above the core and a lower axial fertile blanket improve the total sodium void effect appreciably, bringing it close to zero for a core with fresh fuel, in line with results obtained worldwide, while not influencing substantially other core physics parameters. Therefore an optimized configuration, CONF2, with a sodium plenum and a lower blanket was established first and used as a basis for further studies in view of deterioration of safety parameters during reactor operation. Further options to study were an inner fertile blanket, introduction of moderator pins, a smaller core height, special designs for pins, such as 'empty' pins, and subassemblies. These special designs were proposed to facilitate melted fuel relocation in order to avoid core re-criticality under severe accident conditions. In the paper further CONF2 modifications are compared in terms of safety and fuel balance. They may bring further improvements in safety, but their accurate assessment requires additional studies, including transient analyses. Uncertainty studies were performed by employing a so-called Total Monte-Carlo method, for which a large number of nuclear data files is produced for single isotopes and then used in Monte-Carlo calculations. The uncertainties for the criticality, sodium void and Doppler effects, effective delayed neutron fraction due to uncertainties in basic nuclear data were assessed for an ESFR core. They prove applicability of the available nuclear data for ESFR

  17. Programming Reactive Extensions and LINQ

    CERN Document Server

    Liberty, Jesse

    2011-01-01

    Pro Reactive Extensions and LINQ is a deep dive into the next important technology for .NET developers: Reactive Extensions. This in-depth tutorial goes beyond what is available anywhere else to teach how to write WPF, Silverlight, and Windows Phone applications using the Reactive Extensions (Rx) to handle events and asynchronous method calls. Reactive programming allows you to turn those aspects of your code that are currently imperative into something much more event-driven and flexible. For this reason, it's sometimes referred to as LINQ for Events. Reactive programming hinges on the concep

  18. Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

    2009-01-01

    In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)

  19. Overview of PEC core design and requirements for PEC core restraint systems

    International Nuclear Information System (INIS)

    Cecchini, F.

    1984-01-01

    The Italian PEC reactor is an experimental loop type fast reactor of 120 MW thermal. Its main purpose is the in-pile development of fast reactor fuel. The mechanical principles in PEC core design and current modifications to ensure a safe seismic perturbation and shutdown are discussed in this paper. These anti-seismic modifications are aimed to limit the extent of reactivity perturbation during the seismic event and to guarantee control rod entry at any time during the seismic event

  20. Evaluation of reactivity and Xe behavior during daily load following operation

    International Nuclear Information System (INIS)

    Sakamoto, Yasunori; Araki, Tsuneyasu; Yamamoto, Fumiaki

    1992-01-01

    A boiling water reactor (BWR) has an excellent load following capability provided by a core flow control, which is used for changing a reactor power level and for compensating the subsequent Xe concentration change. The core characteristics during load following operations are investigated in detail, using our reactor core simulator. Comparisons of changes of the Doppler reactivity, the void reactivity and the Xe reactivity during transients are performed. Also the features of Xe transient during load following operations are shown. It has been shown that the core flow change required to compensate the Xe reactivity change produces much greater change of the void reactivity than that required for power level changes, and that the resulting local power change in the lower part of the core is greater than that in the upper part, because the Xe concentration change in the lower part is hardly compensated by the core flow control. Also the effects of power level changes, cycle patterns, and initial concentration of Xe and I on the Xe transient behavior have been investigated. (author)

  1. Expansion of passive safety function

    International Nuclear Information System (INIS)

    Inai, Nobuhiko; Nei, Hiromichi; Kumada, Toshiaki.

    1995-01-01

    Expansion of the use of passive safety functions is proposed. Two notions are presented. One is that, in the design of passive safety nuclear reactors where aversion of active components is stressed, some active components are purposely introduced, by which a system is built in such a way that it behaves in an apparently passive manner. The second notion is that, instead of using a passive safety function alone, a passive safety function is combined with some active components, relating the passivity in the safety function with enhanced controllability in normal operation. The nondormant system which the authors propose is one example of the first notion. This is a system in which a standby safety system is a portion of the normal operation system. An interpretation of the nondormant system via synergetics is made. As an example of the second notion, a PIUS density lock aided with active components is proposed and is discussed

  2. Use of compensation assemblies in the first core of SNR-300

    International Nuclear Information System (INIS)

    Billaux, M.; De Wouters, R.; Pilate, S.; Vandenberg, C.

    1975-01-01

    For the SNR-300 reactor, the use of thin fuel pins was limited to the first core. A direct consequence of changing from the cycle reloading scheme to a complete irradiation without refueling operation is an increase of the initial excess reactivity and plutonium investment. The new system of special assemblies conceived to compensate for the too high reactivity of the first core is described: fixed absorbers, made of B 4 C pins, and sodium diluents, consisting simply of hollow wrapper tubes [fr

  3. Isolation and Spectroscopic Characterization of Reactive Species in Atmospheric and Interstellar Reactions

    Science.gov (United States)

    Relph, Rachael A.

    2011-12-01

    A critical element to the study of chemical reactions is the characterization of reaction intermediates. Methods have been developed to isolate these transient species in the gas phase and when combined with infrared spectroscopy have proven to be excellent tools for determining the structure and reactivity of key intermediates. The studies presented here exploit these technologies to better understand the chemistry of species involved in atmospheric and interstellar reactions. An excellent example of their utility is in the study of the formation of proton hydrates and HONO in the upper atmosphere by sequential addition of water molecules onto the nitrosonium ion. This reaction only proceeds to products after addition of the fourth water molecule, and isolation and characterization of the intermediate trihydrate, NO+(H 2O)3, shows that this species is formed in three isomeric forms, each with a different water network that controls the degree of bond formation between the nitrosonium ion and an activated water molecule. Many isomeric structures are also seen in the clustering reactions of acetylene which may be a mechanism for the formation of benzene cation in interstellar space. The spectroscopy of the trimer, (C2H2)3 + indicates that this species exists in two major isomer classes; covalent forms, one of which may be benzene, and an ion-molecule complex, comprised of a loosely bound acetylene on a dimer core. Interestingly, this dimer core is different from the cyclobutadiene-like structure observed in dimerized acetylene, and proves to be a robust species on the potential energy surface as it survives further clustering events. Two structural isomers of CO3 -and NO3 - are also investigated, and found to have drastically different infrared spectra which are analyzed in the context of their electronic structure. Isomers in these systems are prepared under different expansion conditions which accounts for their unique spectral signatures.

  4. 3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

    International Nuclear Information System (INIS)

    Choudhry, Nakul; Riyas, A.; Devan, K.; Mohanakrishnan, P.

    2013-01-01

    Highlights: ► Enhanced burnup potential of existing prototype fast breeder reactor core is studied. ► By increasing the Pu enrichment, fuel burnup can be increased in existing PFBR core. ► Enhanced burnup increase economy and reduce load of fuel fabrication and reprocessing. ► Beginning of life reactivity is suppressed by increasing the number of diluents. ► Absorber rod worth requirements can be achieved by increasing 10 B enrichment. -- Abstract: Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500 MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100 GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134 GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10 B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs

  5. On the Equisummability of Hermite and Fourier Expansions

    Indian Academy of Sciences (India)

    We prove an equisummability result for the Fourier expansions and Hermite expansions as well as special Hermite expansions. We also prove the uniform boundedness of the Bochner-Riesz means associated to the Hermite expansions for polyradial functions.

  6. Development of core design and analyses technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

    1999-03-01

    Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

  7. Development of core design and analyses technology for integral reactor

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

    1999-03-01

    Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

  8. Reactivity insertion accident analysis

    International Nuclear Information System (INIS)

    Moreira, J.M.L.; Nakata, H.; Yorihaz, H.

    1990-04-01

    The correct prediction of postulated accidents is the fundamental requirement for the reactor licensing procedures. Accident sequences and severity of their consequences depend upon the analysis which rely on analytical tools which must be validated against known experimental results. Present work presents a systematic approach to analyse and estimate the reactivity insertion accident sequences. The methodology is based on the CINETHICA code which solves the point-kinetics/thermohydraulic coupled equations with weighted temperature feedback. Comparison against SPERT experimental results shows good agreement for the step insertion accidents. (author) [pt

  9. The core paradox.

    Science.gov (United States)

    Kennedy, G. C.; Higgins, G. H.

    1973-01-01

    Rebuttal of suggestions from various critics attempting to provide an escape from the seeming paradox originated by Higgins and Kennedy's (1971) proposed possibility that the liquid in the outer core was thermally stably stratified and that this stratification might prove a powerful inhibitor to circulation of the outer core fluid of the kind postulated for the generation of the earth's magnetic field. These suggestions are examined and shown to provide no reasonable escape from the core paradox.

  10. Establishing the long-term fuel management scheme using point reactivity model

    International Nuclear Information System (INIS)

    Park, Yong-Soo; Kim, Jae-Hak; Lee, Young-Ouk; Song, Jae-Woong; Zee, Sung-Kyun

    1994-01-01

    A new approach to establish the long-term fuel management scheme is presented in this paper. The point reactivity model is used to predict the core average reactivity. An attempt to calculate batchwise power fraction is introduced through the two-dimensional nodal power algorithm based on the modified one-group diffusion equation and the number of fuel assemblies on the core periphery. Suggested is an empirical formula to estimate the radial leakage reactivity with ripe core design experience reflected. This approach predicts the cycle lengths and the discharge burnups of individual fuel batches up to an equilibrium core when the proper input data such as batch enrichment, batch size, type and content of burnable poison and reloading strategies are given. Eight benchmark calculations demonstrate that the new approach used in this study is reasonably accurate and highly efficient for the purpose of scoping calculation when compared with design code predictions. (author)

  11. Analysis of void reactivity measurements in full MOX BWR physics experiments

    International Nuclear Information System (INIS)

    Ando, Yoshihira; Yamamoto, Toru; Umano, Takuya

    2008-01-01

    In the full MOX BWR physics experiments, FUBILA, four 9x9 test assemblies simulating BWR full MOX assemblies were located in the center of the core. Changing the in-channel moderator condition of the four assemblies from 0% void to 40% and 70% void mock-up, void reactivity was measured using Amplified Source Method (ASM) technique in the subcritical cores, in which three fission chambers were located. ASM correction factors necessary to express the consistency of the detector efficiency between measured core configurations were calculated using collision probability cell calculation and 3D-transport core calculation with the nuclear data library, JENDL-3.3. Measured reactivity worth with ASM correction factor was compared with the calculated results obtained through a diffusion, transport and continuous energy Monte Carlo calculation respectively. It was confirmed that the measured void reactivity worth was reproduced well by calculations. (author)

  12. Study of reactivity of fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Rammsy, J.E.M.

    1985-01-01

    The reactor physics calculations of a 19 module Fluidized Bed Nuclear Reactor using Leopard and Odog codes are performed. The behaviour of the reactor was studied by calculating the reactivity of the reactor as a function of the parameters governing the operational and accidental conditions of the reactor. The effects of temperature, pressure, and vapor generation in the core on the reactivity are calculated. Also the start up behaviour of the reactor is analyzed. For the purpose of the study of a prototype research reactor, the calculations on a one module reactor have been performed. (Author) [pt

  13. Nuclear reactor core flow baffling

    International Nuclear Information System (INIS)

    Berringer, R.T.

    1979-01-01

    A flow baffling arrangement is disclosed for the core of a nuclear reactor. A plurality of core formers are aligned with the grids of the core fuel assemblies such that the high pressure drop areas in the core are at the same elevations as the high pressure drop areas about the core periphery. The arrangement minimizes core bypass flow, maintains cooling of the structure surrounding the core, and allows the utilization of alternative beneficial components such as neutron reflectors positioned near the core

  14. Reactivity monitoring for safety purposes on the UK prototype fast reactor

    International Nuclear Information System (INIS)

    Lord, D.J.; Wilkes, D.J.

    1987-01-01

    The small size and high rating of the liquid metal cooled fast breeder reactor (LMFBR) make the provision of safety related instrumentation for individual subassemblies both difficult and expensive. Global monitoring of the core is thus very attractive. Reactivity monitoring is an important part of such global monitoring. Reactivity monitoring on a short timescale (a few seconds) is used on the UK Prototype Fast Reactor (PFR) as a trip parameter and long-term reactivity monitoring is being developed as a means of providing early warning of slowly developing faults. Results are presented from PFR to demonstrate the capabilities of reactivity monitoring in an operational fast reactor power station. (author)

  15. Sediment Core Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Provides instrumentation and expertise for physical and geoacoustic characterization of marine sediments.DESCRIPTION: The multisensor core logger measures...

  16. [Tissular expansion in giant congenital nevi treatment].

    Science.gov (United States)

    Nguyen Van Nuoi, V; Francois-Fiquet, C; Diner, P; Sergent, B; Zazurca, F; Franchi, G; Buis, J; Vazquez, M-P; Picard, A; Kadlub, N

    2014-08-01

    Surgical management of giant melanotic naevi remains a surgical challenge. Tissue expansion provides tissue of the same quality for the repair of defects. The aim of this study is to review tissular expansion for giant melanotic naevi. We conducted a retrospective study from 2000 to 2012. All children patients who underwent a tissular expansion for giant congenital naevi had been included. Epidemiological data, surgical procedure, complication rate and results had been analysed. Thirty-tree patients had been included; they underwent 61 procedures with 79 tissular-expansion prosthesis. Previous surgery, mostly simple excision had been performed before tissular expansion. Complete naevus excision had been performed in 63.3% of the cases. Complications occurred in 45% of the cases, however in 50% of them were minor. Iterative surgery increased the complication rate. Tissular expansion is a valuable option for giant congenital naevus. However, complication rate remained high, especially when iterative surgery is needed. Copyright © 2013 Elsevier Masson SAS. All rights reserved.

  17. Simulation of rod drop experiments in the initial cores of Loviisa and Mochovce

    International Nuclear Information System (INIS)

    Kaloinen, E.; Kyrki-Rajamaeki, R.; Wasastjerna, F.

    1999-01-01

    Interpretation of rod drop measurements during startup tests of the Loviisa reactors has earlier been studied with two-dimensional core calculations using a spatial prompt jump approximation. In these calculations the prediction for the reactivity meter reading was lower than the measured values by 25%. Another approach to solve the problem is simulation of the rod drop experiment with dynamic core calculations coupled with out of core calculations to estimate the response of ex-core ionization chambers for the reactivity meter. This report described the calculations performed with the three-dimensional dynamic code HEXTRAN for prediction of the reactivity meter readings in rod drop experiments in initial cores of the WWER-440 reactors. (Authors)

  18. Core design methods for advanced LMFBRs

    International Nuclear Information System (INIS)

    Chandler, J.C.; Marr, D.R.; McCurry, D.C.; Cantley, D.A.

    1977-05-01

    The multidiscipline approach to advanced LMFBR core design requires an iterative design procedure to obtain a closely-coupled design. HEDL's philosophy requires that the designs should be coupled to the extent that the design limiting fuel pin, the design limiting duct and the core reactivity lifetime should all be equal and should equal the fuel residence time. The design procedure consists of an iterative loop involving three stages of the design sequence. Stage 1 consists of general mechanical design and reactor physics scoping calculations to arrive at an initial core layout. Stage 2 consists of detailed reactor physics calculations for the core configuration arrived at in Stage 1. Based upon the detailed reactor physics results, a decision is made either to alter the design (Stage 1) or go to Stage 3. Stage 3 consists of core orificing and detailed component mechanical design calculations. At this point, an assessment is made regarding design adequacy. If the design is inadequate the entire procedure is repeated until the design is acceptable

  19. Simulation program for multiple expansion Stirling machines

    International Nuclear Information System (INIS)

    Walker, G.; Weiss, M.; Fauvel, R.; Reader, G.; Bingham, E.R.

    1992-01-01

    Multiple expansion Stirling machines have been a topic of interest at the University of Calgary for some years. Recently a second-order computer simulation program with integral graphics package for Stirling cryocoolers with up to four stages of expansion were developed and made available to the Stirling community. Adaptation of the program to multiple expansion Stirling power systems is anticipated. This paper briefly introduces the program and presents a specimen result

  20. Semiclassical expansions for confined N fermion systems

    International Nuclear Information System (INIS)

    Krivine, H.; Martorell, J.; Casas, M.

    1989-01-01

    A new derivation of the Wigner Kirkwood expansion for N-fermion systems is presented, showing explicitly the connection to the WKB approximation for a single level. This allows to study separately the two ansatz required to obtain the semiclassical expansions: the asymptotic expansions in powers of ℎ and the smoothing of quantal effects. We discuss the one dimensional and three dimensional, with spherical symmetry, cases. Applications for standard potentials used in nuclear physics are described in detail

  1. Thermal and hygroscopic expansion characteristics of bamboo

    OpenAIRE

    Huang, Puxi; Chang, Wen-shao; Ansell, Martin P.; Bowen, Chris R.; Chew, John Y. M.; Adamak, Vana i

    2017-01-01

    The expansion and contraction of bamboo caused by temperature and moisture variations must be evaluated\\ud if bamboo is to be utilised as a building material. However, detailed expansion data, especially data in the ascent and\\ud descent processes of temperature and moisture are unexplored. The aim of this study is to investigate the expansion\\ud characteristics of Phyllostachys edulis (Moso bamboo) in ascent and descent processes of temperature and moisture.\\ud The measurement of linear ther...

  2. $\\delta$-Expansion at Finite Temperature

    OpenAIRE

    Ramos, Rudnei O.

    1996-01-01

    We apply the $\\delta$-expansion perturbation scheme to the $\\lambda \\phi^{4}$ self-interacting scalar field theory in 3+1 D at finite temperature. In the $\\delta$-expansion the interaction term is written as $\\lambda (\\phi^{2})^{ 1 + \\delta}$ and $\\delta$ is considered as the perturbation parameter. We compute within this perturbative approach the renormalized mass at finite temperature at a finite order in $\\delta$. The results are compared with the usual loop-expansion at finite temperature.

  3. Chromatic Derivatives, Chromatic Expansions and Associated Spaces

    OpenAIRE

    Ignjatovic, Aleksandar

    2009-01-01

    This paper presents the basic properties of chromatic derivatives and chromatic expansions and provides an appropriate motivation for introducing these notions. Chromatic derivatives are special, numerically robust linear differential operators which correspond to certain families of orthogonal polynomials. Chromatic expansions are series of the corresponding special functions, which possess the best features of both the Taylor and the Shannon expansions. This makes chromatic derivatives and ...

  4. Thermal expansion in small metallic particles

    International Nuclear Information System (INIS)

    Ivanov, A.S.

    1985-01-01

    An anomalously low thermal expansion observable in small particles is attributed to extending effect of the shell. It is shown that the coefficient of thermal expansion of the oxide-film-coated aluminium particles calculated using elastic constants and coefficients of thermal expansion of massive materials agres well with those measured experimentally. The linear dilatation of the shell, its stress to rupture and the values of the structural tension are estimated vs the temperature

  5. Study of the multiplication factor in the core of Saclay

    International Nuclear Information System (INIS)

    Jacrot, B.; Netter, F.; Raievski, V.

    1953-01-01

    Several methods were studied for the measure of the multiplication factor strength in a core, by experiences in subcritical regime. These methods are applied to the determination of the effect on the reactivity of such different parameters of the battery that: heavy water level, position of the regulating plates. These results are used to establish an experimental relation between the time of the rise of the divergent core and the factor of effective multiplication. It is also given the application of these methods to the assessment of the power of the core. (author) [fr

  6. Optimal burnable poison utilization in PWR core reload design

    International Nuclear Information System (INIS)

    Downar, T.J.

    1986-01-01

    A method was developed for determining the optimal distribution and depletion of burnable poisons in a Pressurized Water Reactor core. The well-known Haling depletion technique is used to achieve the end-of-cycle core state where the fuel assembly arrangement is configured in the absence of all control poison. The soluble and burnable poison required to control the core reactivity and power distribution are solved for as unknown variables while step depleting the cycle in reverse with a target power distribution. The method was implemented in the NRC approved licensing code SIMULATE

  7. Detonative propagation and accelerative expansion of the Crab Nebula shock front.

    Science.gov (United States)

    Gao, Yang; Law, Chung K

    2011-10-21

    The accelerative expansion of the Crab Nebula's outer envelope is a mystery in dynamics, as a conventional expanding blast wave decelerates when bumping into the surrounding interstellar medium. Here we show that the strong relativistic pulsar wind bumping into its surrounding nebula induces energy-generating processes and initiates a detonation wave that propagates outward to form the current outer edge, namely, the shock front, of the nebula. The resulting detonation wave, with a reactive downstream, then provides the needed power to maintain propagation of the shock front. Furthermore, relaxation of the curvature-induced reduction of the propagation velocity from the initial state of formation to the asymptotic, planar state of Chapman-Jouguet propagation explains the observed accelerative expansion. Potential richness in incorporating reactive fronts in the description of various astronomical phenomena is expected. © 2011 American Physical Society

  8. Can Psychiatric Rehabilitation Be Core to CORE?

    Science.gov (United States)

    Olney, Marjorie F.; Gill, Kenneth J.

    2016-01-01

    Purpose: In this article, we seek to determine whether psychiatric rehabilitation principles and practices have been more fully incorporated into the Council on Rehabilitation Education (CORE) standards, the extent to which they are covered in four rehabilitation counseling "foundations" textbooks, and how they are reflected in the…

  9. Methods and techniques of nuclear in-core fuel management

    International Nuclear Information System (INIS)

    Jong, A.J. de.

    1992-04-01

    Review of methods of nuclear in-core fuel management (the minimal critical mass problem, minimal power peaking) and calculational techniques: reactorphysical calculations (point reactivity models, continuous refueling, empirical methods, depletion perturbation theory, nodal computer programs); optimization techniques (stochastic search, linear programming, heuristic parameter optimization). (orig./HP)

  10. Core optimization studies for a small heating reactor

    International Nuclear Information System (INIS)

    Galperin, A.

    1986-11-01

    Small heating reactor cores are characterized by a high contribution of the leakage to the neutron balance and by a large power density variation in the axial direction. A limited number of positions is available for the control rods, which are necessary to satisfy overall reactivity requirements subject to a safety related constraint on the maximum worth of each rod. Design approaches aimed to improve safety and fuel utilization performance of the core include separation of the cooling and moderating functions of the water with the core in order to reduce hot-to-cold reactivity shift and judicious application of the axial Gd zoning aimed to improve the discharge burnup distribution. Several design options are analyzed indicating a satisfactory solution of the axial burnup distribution problem. The feasibility of the control rod system including zircaloy, stainless steel, natural boron and possibly enriched boron rods is demonstrated. A preliminary analysis indicates directions for further improvements of the core performance by an additional reduction of the hot-to-cold reactivity shift and by a reduction of the depletion reactivity swing adopting a higher gadolinium concentration in the fuel or a two-batch fuel management scheme. (author)

  11. Reactive documentation system

    Science.gov (United States)

    Boehnlein, Thomas R.; Kramb, Victoria

    2018-04-01

    Proper formal documentation of computer acquired NDE experimental data generated during research is critical to the longevity and usefulness of the data. Without documentation describing how and why the data was acquired, NDE research teams lose capability such as their ability to generate new information from previously collected data or provide adequate information so that their work can be replicated by others seeking to validate their research. Despite the critical nature of this issue, NDE data is still being generated in research labs without appropriate documentation. By generating documentation in series with data, equal priority is given to both activities during the research process. One way to achieve this is to use a reactive documentation system (RDS). RDS prompts an operator to document the data as it is generated rather than relying on the operator to decide when and what to document. This paper discusses how such a system can be implemented in a dynamic environment made up of in-house and third party NDE data acquisition systems without creating additional burden on the operator. The reactive documentation approach presented here is agnostic enough that the principles can be applied to any operator controlled, computer based, data acquisition system.

  12. Disjoint sum expansion method in FTA

    International Nuclear Information System (INIS)

    Ruan Keqiang

    1987-01-01

    An expansion formula for transforming boolean algebraic expressions into disjoint form was proved. Based on this expansion formula, a method for transforming system failure function into disjoint form was devised. The fact that the expansion can be done for several elements simulatneously makes the method flexible and fast. Some examples from fault tree analysis (FTA) and network analysis were examined by the new method to show its algorithm and its merit. Besides, by means of the proved expansion formula some boolean algebraic relations can proved very easily

  13. Thermal expansion of L-ascorbic acid

    Science.gov (United States)

    Nicolaï, B.; Barrio, M.; Tamarit, J.-Ll.; Céolin, R.; Rietveld, I. B.

    2017-04-01

    The specific volume of vitamin C has been investigated by X-ray powder diffraction as a function of temperature from 110 K up to complete degradation around 440 K. Its thermal expansion is relatively small in comparison with other organic compounds with an expansivity α v of 1.2(3) × 10-4 K-1. The structure consists of strongly bound molecules in the ac plane through a dense network of hydrogen bonds. The thermal expansion is anisotropic. Along the b axis, the expansion has most leeway and is about 10 times larger than in the other directions.

  14. The exponential function expansion of the intra-nodal cross sections for the spectral history gradient correction

    International Nuclear Information System (INIS)

    Cho, J. Y.; Noh, J. M.; Cheong, H. K.; Choo, H. K.

    1998-01-01

    In order to simplify the previous spectral history effect correction based on the polynomial expansion nodal method, a new spectral history effect correction is proposed. The new spectral history correction eliminates four microscopic depletion points out of total 13 depletion points in the previous correction by approximating the group cross sections with exponential function. The neutron flux to homogenize the group cross sections for the correction of the spectral history effect is calculated by the analytic function expansion nodal method in stead of the conventional polynomial expansion nodal method. This spectral history correction model is verified against the three MOX benchmark cores: a checkerboard type, a small core with 25 fuel assemblies, and a large core with 177 fuel assemblies. The benchmark results prove that this new spectral history correction model is superior to the previous one even with the reduced number of the local microscopic depletion points

  15. Device for supporting a nuclear reactor core

    International Nuclear Information System (INIS)

    Costes, D.

    1976-01-01

    The core of a light-water reactor which is enclosed in a prestressed concrete pressure vessel and held within a diffuser basket is supported by a device consisting of a cylindrical shell which surrounds the basket and is rigidly fixed to a plurality of frusto-conical skirts having concurrent axes and located substantially at right angles to the axis of the reactor core. The small base of each skirt is rigidly fixed to the shell and the large base is anchored in openings formed in the reactor vessel for the penetration of coolant inlet and outlet pipes. The top portion of the shell is secured to the top portion of the diffuser basket, a flat surface being formed on the shell at the point of connection with each frusto-conical skirt so as to ensure rigid suspension while permitting thermal expansion

  16. Thermoelastic expansion in prompt-critical neutron pulse idealized in a fissile metallic sphere

    International Nuclear Information System (INIS)

    Barroso, D.E.G.; Ribeiro, S.G.

    1985-01-01

    Prompt critical pulses in solid and homogeneous spheres of enriched uranium (93%) and metallic plutonium are studied. The feedback mechanism of the negative inserted reactivity is given by the elastic expansion due to the increase of the temperature in the sphere. Thermomechanical behavior and the capability of the system to become subcritical without a very large increase of energy released in the pulse are analysed. The neutronic and thermoelasticity equations are solved in the time. (M.C.K.) [pt

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

    International Nuclear Information System (INIS)

    Relloso, J.M.

    1990-01-01

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

  18. Reactor Core Design and Analysis for a Micronuclear Power Source

    Directory of Open Access Journals (Sweden)

    Hao Sun

    2018-03-01

    Full Text Available Underwater vehicle is designed to ensure the security of country sea boundary, providing harsh requirements for its power system design. Conventional power sources, such as battery and Stirling engine, are featured with low power and short lifetime. Micronuclear reactor power source featured with higher power density and longer lifetime would strongly meet the demands of unmanned underwater vehicle power system. In this paper, a 2.4 MWt lithium heat pipe cooled reactor core is designed for micronuclear power source, which can be applied for underwater vehicles. The core features with small volume, high power density, long lifetime, and low noise level. Uranium nitride fuel with 70% enrichment and lithium heat pipes are adopted in the core. The reactivity is controlled by six control drums with B4C neutron absorber. Monte Carlo code MCNP is used for calculating the power distribution, characteristics of reactivity feedback, and core criticality safety. A code MCORE coupling MCNP and ORIGEN is used to analyze the burnup characteristics of the designed core. The results show that the core life is 14 years, and the core parameters satisfy the safety requirements. This work provides reference to the design and application of the micronuclear power source.

  19. Improvement of Cycle Dependent Core Model for NPP Simulator

    International Nuclear Information System (INIS)

    Song, J. S.; Koo, B. S.; Kim, H. Y. and others

    2003-11-01

    The purpose of this study is to establish automatic core model generation system and to develop 4 cycle real time core analysis methodology with 5% power distribution and 500 pcm reactivity difference criteria for nuclear power plant simulator. The standardized procedure to generate database from ROCS and ANC, which are used for domestic PWR core design, was established for the cycle specific simulator core model generation. An automatic data interface system to generate core model also established. The system includes ARCADIS which edits group constant and DHCGEN which generates interface coupling coefficient correction database. The interface coupling coefficient correction method developed in this study has 4 cycle real time capability and accuracies of which the maximum differences between core design results are within 103 pcm reactivity, 1% relative power distribution and 6% control rod worth. A nuclear power plant core simulation program R-MASTER was developed using the methodology and applied by the concept of distributed client system in simulator. The performance was verified by site acceptance test in Simulator no. 2 in Kori Training Center for 30 initial condition generation and 27 steady state, transient and postulated accident situations

  20. Improvement of Cycle Dependent Core Model for NPP Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Song, J. S.; Koo, B. S.; Kim, H. Y. and others

    2003-11-15

    The purpose of this study is to establish automatic core model generation system and to develop 4 cycle real time core analysis methodology with 5% power distribution and 500 pcm reactivity difference criteria for nuclear power plant simulator. The standardized procedure to generate database from ROCS and ANC, which are used for domestic PWR core design, was established for the cycle specific simulator core model generation. An automatic data interface system to generate core model also established. The system includes ARCADIS which edits group constant and DHCGEN which generates interface coupling coefficient correction database. The interface coupling coefficient correction method developed in this study has 4 cycle real time capability and accuracies of which the maximum differences between core design results are within 103 pcm reactivity, 1% relative power distribution and 6% control rod worth. A nuclear power plant core simulation program R-MASTER was developed using the methodology and applied by the concept of distributed client system in simulator. The performance was verified by site acceptance test in Simulator no. 2 in Kori Training Center for 30 initial condition generation and 27 steady state, transient and postulated accident situations.