Irradiation of novel MTR fuel plates in BR2

International Nuclear Information System (INIS)

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

2000-01-01

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

Loading method of core constituting elements

International Nuclear Information System (INIS)

Kasai, Shigeo

1976-01-01

Purpose: To provide a remote-controlled replacing method for core constituting elements in a liquid-metal cooling fast breeder, wherein particularly, the core constituting elements are prevented from being loaded on the core position other than as designated. Constitution: The method comprises a first step which determines a position of a suitable neutron shielding body in order to measure a reference level of complete insertion of the core constituting elements, a second step which inserts a gripper for a fuel exchanger, a third step which decides stroke dimensions of the complete insertion, and a fourth step which discriminates the core constituting elements to begin handling of fuel rods. The method further comprises a fifth step which determines a loading position of fuel rod, and a sixth step which inserts and loads fuel rods into the core. The method still further comprises a seventh step which compares and judges the dimension of loading stroke and the dimension of complete inserting stroke so that when coincided, loading is completed, and when not coincided, loading is not completed and then the cycle of the fourth step is repeated. (Kawakami, Y.)

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

Science.gov (United States)

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

2017-11-01

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

Optimized core loading sequence for Ukraine WWER-1000 reactors

International Nuclear Information System (INIS)

Dye, M.; Shah, H.

2015-01-01

Fuel Assemblies (WFAs) experienced mechanical damage of the grids during loading at both South Ukraine 2 (SU2) and South Ukraine 3 (SU3). The grids were damaged due to high lateral loads exceeding their strength limit. The high lateral loads were caused by a combination of distortion and stiffness of the mixed core fuel assemblies and significant fuel assembly-to-fuel assembly interaction combined with the core loading sequence being used. To prevent damage of the WFA grids during core loading, Westinghouse has developed a loading sequence technique and loading aides (smooth sided dummies and top nozzle loading guides) designed to minimize fuel assembly-to-fuel assembly interaction while maximizing the potential for successful loading (i.e., no fuel assembly damage and minimized loading time). The loading sequence technique accounts for cycle-specific core loading patterns and is based on previous Westinghouse WWER core loading experience and fundamental principles. The loading aids are developed to “open-up” the target core location or to provide guidance into a target core location. The Westinghouse optimized core loading sequence and smooth sided dummies were utilized during the successful loading of SU3 Cycle 25 mixed core in March 2015, with no instances of fuel assembly damage and yet still provided considerable time savings relative to the 2012 and 2013 SU3 reload campaigns. (authors)

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

International Nuclear Information System (INIS)

El-Morshedy, Salah El-Din

2010-01-01

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

Sipping test on a failed MTR fuel element

International Nuclear Information System (INIS)

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

2002-01-01

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

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

Directory of Open Access Journals (Sweden)

Paul J. T. Johnson

2015-06-01

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

Application of reactivity method to MTR fuel burn-up measurement

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

Taylor, N.P.

1984-01-01

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

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

Indian Academy of Sciences (India)

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

Early detection of coolant boiling in research reactors with MTR-type fuel

International Nuclear Information System (INIS)

Kozma, R.; Turkcan, E.; Verhoef, J.P.

1992-10-01

A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-07-01

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

Final disposition of MTR fuel

International Nuclear Information System (INIS)

Jonnson, Erik B.

1996-01-01

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

Structural assessment of TAPS core shroud under accident loads

International Nuclear Information System (INIS)

Bhasin, Vivek; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

1996-09-01

Over the last few years, the Core Shroud of Boiling Water Reactors (BWRs) operating in foreign countries, have developed cracks at weld locations. As a first step for assessment of structural safety of Tarapur Atomic Power Station (TAPS) core shroud, its detailed stress analysis was done for postulated accident loads. This report is concerned with structural assessment of core shroud, of BWR at TAPS, subjected to loads resulting from main steam line break (MSLB), recirculation line break (RLB) and safe shut down earthquake. The stress analysis was done for core shroud in healthy condition and without any crack since, visual examination conducted till now, do not indicate presence of any flaw. Dynamic structural analysis for MSLB and RLB events was done using dynamic load factor (DLF) method. The complete core shroud and its associated components were modelled and analysed using 3D plate/shell elements. Since, the components of core shroud are submerged in water, hence, hydrodynamic added mass was also considered for evaluation of natural frequencies. It was concluded that from structural point of view, adequate safety margin is available under all the accident loads. Nonlinear analysis was done to evaluate buckling/collapse load. The collapse/buckling load have sufficient margin against the allowable limits. The displacements are low hence, the insertion of control rod may not be affected. (author)

Static analysis of material testing reactor cores:critical core calculations

International Nuclear Information System (INIS)

Nawaz, A. A.; Khan, R. F. H.; Ahmad, N.

1999-01-01

A methodology has been described to study the effect of number of fuel plates per fuel element on critical cores of Material Testing Reactors (MTR). When the number of fuel plates are varied in a fuel element by keeping the fuel loading per fuel element constant, the fuel density in the fuel plates varies. Due to this variation, the water channel width needs to be recalculated. For a given number of fuel plates, water channel width was determined by optimizing k i nfinity using a transport theory lattice code WIMS-D/4. The dimensions of fuel element and control fuel element were determined using this optimized water channel width. For the calculated dimensions, the critical cores were determined for the given number of fuel plates per fuel element by using three dimensional diffusion theory code CITATION. The optimization of water channel width gives rise to a channel width of 2.1 mm when the number of fuel plates is 23 with 290 g ''2''3''5U fuel loading which is the same as in the case of Pakistan Reactor-1 (PARR-1). Although the decrease in number of fuel element results in an increase in optimal water channel width but the thickness of standard fuel element (SFE) and control fuel element (CFE) decreases and it gives rise to compact critical and equilibrium cores. The criticality studies of PARR-1 are in good agreement with the predictions

Reprocessing of MTR fuel at Dounreay

International Nuclear Information System (INIS)

Hough, N.

1997-01-01

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

Analysis of a Neutronic Computational Model for the Core of Material Testing Reactor MTR by Using SQUID Code

International Nuclear Information System (INIS)

Al-Taweel, M.H.

2015-01-01

It is a conventional practice in the design of nuclear reactor to introduce calculation of hot points to determine spatial variation for energy generated and then determine power distribution.The study had been carried out for core of a reactor type (MTR) by the neutronic code SQUID. In this study, we replace the reflector of the reactor by H 2 O instead of D 2 O as originally the reactor designed.From the study we conclude that the reactor can operates safely, to make sure of that we calculate the multiplication factor where their values ranged from (1.0854) when all control rods are up to (1.001)when three control rods are up.Also the values of hot points were calculated and compared with French documents results with D 2 O as a reflector where the difference is (0.19%), and with light water as reflector instead of heavy water was calculated.For different cases according to control rod position , the values of hot point ranged between (0.46) to (1.64) in case all control rods are up also the values of the average power distributed on different fuel cells were calculated in case of light water as reflector firstly with three control rods are down and the maximum value (2.13*10 -2 Μw).Secondly in case offour control rods are down, the maximum value (1.925*10 -2 Μw) we notice almost coincidence between the neutron flux distribution through the core of reactor and in different positions of control rods

MA-core loaded untuned RF compression cavity for HIRFL-CSR

International Nuclear Information System (INIS)

Mei Lirong; Xu Zhe; Yuan Youjin; Jin Peng; Bian Zhibin; Zhao Hongwei; Xia Jiawen

2012-01-01

To meet the requirements of high energy density physics and plasma physics research at HIRFL-CSR the goal of achieving a higher accelerating gap voltage was proposed. Therefore, a magnetic alloy (MA)-core loaded radio frequency (RF) cavity that can provide a higher accelerating gap voltage compared to standard ferrite loaded cavities has been studied at IMP. In order to select the proper magnetic alloy material to load the RF compression cavity, measurements of four different kinds of sample MA-cores have been carried out. By testing the small cores, the core composition was selected to obtain the desired performance. According to the theoretical calculation and simulation, which show reasonable consistency for the MA-core loaded cavity, the desired performance can be achieved. Finally about 1000 kW power will be needed to meet the requirements of 50 kV accelerating gap voltage by calculation.

Anticancer drug mithramycin interacts with core histones: An additional mode of action of the DNA groove binder

Directory of Open Access Journals (Sweden)

Amrita Banerjee

2014-01-01

Full Text Available Mithramycin (MTR is a clinically approved DNA-binding antitumor antibiotic currently in Phase 2 clinical trials at National Institutes of Health for treatment of osteosarcoma. In view of the resurgence in the studies of this generic antibiotic as a human medicine, we have examined the binding properties of MTR with the integral component of chromatin – histone proteins – as a part of our broad objective to classify DNA-binding molecules in terms of their ability to bind chromosomal DNA alone (single binding mode or both histones and chromosomal DNA (dual binding mode. The present report shows that besides DNA, MTR also binds to core histones present in chromatin and thus possesses the property of dual binding in the chromatin context. In contrast to the MTR–DNA interaction, association of MTR with histones does not require obligatory presence of bivalent metal ion like Mg2+. As a consequence of its ability to interact with core histones, MTR inhibits histone H3 acetylation at lysine 18, an important signature of active chromatin, in vitro and ex vivo. Reanalysis of microarray data of Ewing sarcoma cell lines shows that upon MTR treatment there is a significant down regulation of genes, possibly implicating a repression of H3K18Ac-enriched genes apart from DNA-binding transcription factors. Association of MTR with core histones and its ability to alter post-translational modification of histone H3 clearly indicates an additional mode of action of this anticancer drug that could be implicated in novel therapeutic strategies.

LMFBR core flowering response to an impulse load

International Nuclear Information System (INIS)

Brochard, D.; Petret, J.C.; Queval, J.C.; Gibert, R.J.

1993-01-01

Some incidental situations like MFCI (Meeting Fuel Coolant Incident) may induce a core flowering and lead to consider impulse loans applied to LMFBR core. These highly dynamic loads are very different considering their spatial repartition and their frequency content from the seismic loads which have been deeply studied. Recently, tests have been performed on the LMFBR core mock-up RAPSODIE in order to validate the calculation methods for centered impulse load. These tests consist in injecting water quickly in the mock-up through a specific device replacing the core central assembly. The influence of the injection pressure and the influence of the injection axial position have been investigate. During the tests, the top displacements of some assemblies have been measured. The aim of this paper is first to present the experimental device and the test results. Then a non linear numerical model is described; this model includes the impact between subassemblies and is based on an homogenization method allowing to take into account with accuracy the fluid structure interaction.The comparisons between calculation results an test results will finally be presented

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

MTR fuel inspection at CERCA

International Nuclear Information System (INIS)

Fanjas, Y.

1992-01-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Core/corona modeling of diode-imploded annular loads

Science.gov (United States)

Terry, R. E.; Guillory, J. U.

1980-11-01

The effects of a tenuous exterior plasma corona with anomalous resistivity on the compression and heating of a hollow, collisional aluminum z-pinch plasma are predicted by a one-dimensional code. As the interior ("core") plasma is imploded by its axial current, the energy exchange between core and corona determines the current partition. Under the conditions of rapid core heating and compression, the increase in coronal current provides a trade-off between radial acceleration and compression, which reduces the implosion forces and softens the pitch. Combined with a heuristic account of energy and momentum transport in the strongly coupled core plasma and an approximate radiative loss calculation including Al line, recombination and Bremsstrahlung emission, the current model can provide a reasonably accurate description of imploding annular plasma loads that remain azimuthally symmetric. The implications for optimization of generator load coupling are examined.

Preliminary developments of MTR plates with uranium nitride

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

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

Preliminary developments of MTR plates with uranium nitride

International Nuclear Information System (INIS)

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

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Coragem, Helio Boemer de Oliveira

1980-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Coragem, Helio Boemer de Oliveira

1980-07-01

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

Validation of the Nuclear Design Method for MOX Fuel Loaded LWR Cores

International Nuclear Information System (INIS)

Saji, E.; Inoue, Y.; Mori, M.; Ushio, T.

2001-01-01

The actual batch loading of mixed-oxide (MOX) fuel in light water reactors (LWRs) is now ready to start in Japan. One of the efforts that have been devoted to realizing this batch loading has been validation of the nuclear design methods calculating the MOX-fuel-loaded LWR core characteristics. This paper summarizes the validation work for the applicability of the CASMO-4/SIMULATE-3 in-core fuel management code system to MOX-fuel-loaded LWR cores. This code system is widely used by a number of electric power companies for the core management of their commercial LWRs. The validation work was performed for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. Each validation consists of two parts: analyses of critical experiments and core tracking calculations of operating plants. For the critical experiments, we have chosen a series of experiments known as the VENUS International Program (VIP), which was performed at the SCK/CEN MOL laboratory in Belgium. VIP consists of both BWR and PWR fuel assembly configurations. As for the core tracking calculations, the operating data of MOX-fuel-loaded BWR and PWR cores in Europe have been utilized

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

International Nuclear Information System (INIS)

Mueller, H.; Gruber, G.

1991-01-01

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

Reactivity worth of the thermal column of a MTR type swimming pool research reactor using low enriched uranium fuel

International Nuclear Information System (INIS)

Ali Khan, L.; Ahmad, N.

2002-01-01

The reactivity worth of the thermal column of a typical MTR type swimming pool research reactor using low enriched uranium fuel has been determined by modeling the core using standard computer codes. It was also measured experimentally by operating the reactor in the stall and open ends. The calculated value of the reactivity worth of the thermal column is about 14% greater than the experimentally determined value

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

DEFF Research Database (Denmark)

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

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

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

Science.gov (United States)

Day, Simon E.

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

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

International Nuclear Information System (INIS)

Day, S.E.

2006-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Day, S.E

2006-07-01

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

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

NARCIS (Netherlands)

Trip, H; Evers, ME; Driessen, AJM

2004-01-01

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

Mtr Extracellular Electron Transfer Pathways in Fe(III)-reducing or Fe(II)-oxidizing Bacteria: A Genomic Perspective

Energy Technology Data Exchange (ETDEWEB)

Shi, Liang; Rosso, Kevin M.; Zachara, John M.; Fredrickson, Jim K.

2012-12-01

Originally discovered in the dissimilatory metal-reducing bacterium Shewanella oneidensis MR-1 (MR-1), the Mtr (i.e., metal-reducing) pathway exists in all characterized strains of metal-reducing Shewanella. The protein components identified to date for the Mtr pathway of MR-1 include four multi-heme c-type cytochromes (c-Cyts), CymA, MtrA, MtrC and OmcA, and a porin-like, outer membrane protein MtrB. They are strategically positioned along the width of the MR-1 cell envelope to mediate electron transfer from the quinone/quinol pool in the inner-membrane to the Fe(III)-containing minerals external to the bacterial cells. A survey of microbial genomes revealed homologues of the Mtr pathway in other dissimilatory Fe(III)-reducing bacteria, including Aeromonas hydrophila, Ferrimonas balearica and Rhodoferax ferrireducens, and in the Fe(II)-oxidizing bacteria Dechloromonas aromatica RCB, Gallionella capsiferriformans ES-2 and Sideroxydans lithotrophicus ES-1. The widespread distribution of Mtr pathways in Fe(III)-reducing or Fe(II)-oxidizing bacteria emphasizes the importance of this type of extracellular electron transfer pathway in microbial redox transformation of Fe. Their distribution in these two different functional groups of bacteria also emphasizes the bi-directional nature of electron transfer reactions carried out by the Mtr pathways. The characteristics of the Mtr pathways may be shared by other pathways used by microorganisms for exchanging electrons with their extracellular environments.

Analysis of the Ford Nuclear Reactor LEU core

Energy Technology Data Exchange (ETDEWEB)

Rathkopf, J A; Drumm, C R; Martin, W R; Lee, J C [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

1983-09-01

This paper has summarized the current status of the effort to analyze the FNR HEU/LEU cores and to compare the calculated results with measurements. In general, calculated predictions of experimental results are quite good, especially for global parameters such as reactivity, as seen in the single HEU/LEU element substitution experiment and the LEU full core critical loading. Shim rod worths are predicted well for two of the rods but too high for a third rod possibly due to inaccurate thermal flux distribution calculation. The calculated thermal flux maps show excellent agreement with experiment throughout the FNR core. In the heavy water tank, however, experimental values for the thermal flux obtained by different methods are inconsistent among themselves as well as with the calculated finding. Work is under.way to use our computational tools to correct the discrepancies between the various measurement techniques and to improve the computational results for flux distribution and the rod worth experiment. Although uncertainties exist in our analysis, as evidenced by the discrepancies mentioned above, we consider our present calculational package to be a useful, reasonably accurate, and efficient system for performing analyses of MTR LEU/HEU core configurations.

Establishing a LEU MTR fuel manufacturing facility in South Africa

International Nuclear Information System (INIS)

Jamie, R.W.; Kocher, A.

2010-01-01

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

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Neutronic analysis of the conversion of HEU to LEU fuel for a 5-MW MTR core

International Nuclear Information System (INIS)

Pazirandeh, A.; Bartsch, G.

1987-01-01

In recent years, due to cessation of highly enriched uranium (HEU) fuel supply, practical steps have been taken to substitute HEU fuel in almost all research reactors by medium-enriched uranium or low-enriched uranium (LEU) fuels. In this study, a neutronic calculation of a 5-MW research reactor core fueled with HEU (93% 235 U) is presented. In order to assess the performance of the core with the LEU ( 235 U loadings were examined. The core consists of 22 standard fuel elements (SFEs) and 6 control fuel elements (CFEs). Each fuel elements has 18 curved plates of which two end plates are dummies. Initial 235 U content is 195 g 235 U/SFE and 9.7 g 235 U/CFE or /PFE. In all calculations the permitted changes to the fuel elements are (a) 18 active plates per SFE, (b) fuel plates assumed to be flat, and (c) 8 or 9 active plates per CFE

Planning a new research reactor for AECL: The MAPLE-MTR concept

International Nuclear Information System (INIS)

Lee, A.G.; Lidstone, R.F.; Donnelly, J.V.

1992-01-01

AECL Research is assessing its needs and options for future irradiation research facilities. A planning team has been assembled to identify the irradiation requirements for AECL's research programs and compile options for satisfying the irradiation requirements. The planning team is formulating a set of criteria to evaluate the options and will recommend a plan for developing an appropriate research facility. Developing the MAPLE Materials Test Reactor (MAPLE-MTR) concept to satisfy AECL's irradiation requirements is one option under consideration by the planning team. AECL is undertaking this planning phase because the NRU reactor is 35 years old and many components are nearing the end of their design life. This reactor has been a versatile facility for proof testing CANDU components and fuel designs because the CANDU irradiation environment was simulated quite well. However, the CANDU design has matured and the irradiation requirements have changed. Future research programs will emphasize testing CANDU components near or beyond their design limits. To provide these irradiation conditions, the NRU reactor needs to be upgraded. Upgrading and refurbishing the NRU reactor is being considered, but the potentially large costs and regulatory uncertainties make this option very challenging. AECL is also developing the MAPLE-MTR concept as a potential replacement for the NRU reactor. The MAPLE-MTR concept starts from the recent MAPLE-X10 design and licensing experience and adapts this technology to satisfy the primary irradiation requirements of AECL's research programs. This approach should enable AECL to minimize the need for major advances in nuclear technology (e.g., fuel design, heat transfer). The preliminary considerations for developing the MAPLE-MTR concept are presented in this report. A summary of AECL's research programs is presented along with their irradiation requirements. This is followed by a description of safety criteria that need to be taken into

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

International Nuclear Information System (INIS)

Huizenga, D. J.

2016-01-01

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

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

Science.gov (United States)

Beliaev, A S; Saffarini, D A

1998-12-01

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

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

International Nuclear Information System (INIS)

Coragem, Helio Boemer de Oliveira

1980-01-01

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

Fast core prediction simulator for load follow control

International Nuclear Information System (INIS)

Yim, Man Sung; Lee, Sang Hoon; Lee, Un Chul

1990-01-01

An operator-assisting system for the reactor core control under power changing operating condition was developed. The system is consisted of core simulator routine and Xenon and Iodine initial condition generation routine. The initial condition generation routine, without exactly knowing the core status, is capable of providing accurate number densities and axial offset conditions of Xenon and Iodine after several hours of predictor- corrector calculations using the plant instrumentation signals of power level and power axial offset. The core simulator routine, even with the two node core model, gives equivalently accurate results as the one-dimensional model for the core behaviour simulation under power changing condition and can provide proper control strategies for load follow operation. The core simulator can also be used by the operator to develop remedial actions to restore the distorted power distribution by using its prediction capability

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

Directory of Open Access Journals (Sweden)

Nicolle F. Som

2017-06-01

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

Transportation of spent MTR fuels

Energy Technology Data Exchange (ETDEWEB)

Raisonnier, D.

1997-08-01

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

Transportation of spent MTR fuels

International Nuclear Information System (INIS)

Raisonnier, D.

1997-01-01

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

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

Directory of Open Access Journals (Sweden)

Khedr Ahmed

2005-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-09-01

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

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.

Optimization programs for reactor core fuel loading exhibiting reduced neutron leakage

International Nuclear Information System (INIS)

Darilek, P.

1991-01-01

The program MAXIM was developed for the optimization of the fuel loading of WWER-440 reactors. It enables the reactor core reactivity to be maximized by modifying the arrangement of the fuel assemblies. The procedure is divided into three steps. The first step includes the passage from the three-dimensional model of the reactor core to the two-dimensional model. In the second step, the solution to the problem is sought assuming that the multiplying properties, or the reactivity in the zones of the core, vary continuously. In the third step, parameters of actual fuel assemblies are inserted in the ''continuous'' solution obtained. Combined with the program PROPAL for a detailed refinement of the loading, the program MAXIM forms a basis for the development of programs for the optimization of fuel loading with burnable poisons. (Z.M.). 16 refs

Methodology for thermal-hydraulics analysis of pool type MTR fuel research reactors

International Nuclear Information System (INIS)

Umbehaun, Pedro Ernesto

2000-01-01

This work presents a methodology developed for thermal-hydraulic analysis of pool type MTR fuel research reactors. For this methodology a computational program, FLOW, and a model, MTRCR-IEAR1 were developed. FLOW calculates the cooling flow distribution in the fuel elements, control elements, irradiators, and through the channels formed among the fuel elements and among the irradiators and reflectors. This computer program was validated against experimental data for the IEA-R1 research reactor core at IPEN-CNEN/SP. MTRCR-IEAR1 is a model based on the commercial program Engineering Equation Solver (EES). Besides the thermal-hydraulic analyses of the core in steady state accomplished by traditional computational programs like COBRA-3C/RERTR and PARET, this model allows to analyze parallel channels with different cooling flow and/or geometry. Uncertainty factors of the variables from neutronic and thermalhydraulic calculation and also from the fabrication of the fuel element are introduced in the model. For steady state analyses MTRCR-IEAR1 showed good agreement with the results of COBRA-3C/RERTR and PARET. The developed methodology was used for the calculation of the cooling flow distribution and the thermal-hydraulic analysis of a typical configuration of the IEA-R1 research reactor core. (author)

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

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

Directory of Open Access Journals (Sweden)

Chih-Chia Su

2015-04-01

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

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

International Nuclear Information System (INIS)

Iracane, Daniel; Cordier, Pierre-Yves

2009-01-01

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

Shock loading of reactor vessel following hypothetical core disruptive accident

International Nuclear Information System (INIS)

Srinivas, G.; Doshi, J.B.

1990-01-01

Hypothetical Core Disruptive Accident (HCDA) has been historically considered as the maximum credible accident in Fast Breeder Reactor systems. Environmental consequences of such an accident depends to a great extent on the ability of the reactor vessel to maintain integrity during the shock loading following an HCDA. In the present paper, a computational model of the reactor core and the surrounding coolant with a free surface is numerical technique. The equations for conservation of mass, momentum and energy along with an equation of state are considered in two dimensional cylindrical geometry. The reactor core at the end of HCDA is taken as a bubble of hot, vaporized fuel at high temperature and pressure, formed at the center of the reactor vessel and expanding against the surrounding liquid sodium coolant. The free surface of sodium at the top of the vessel and the movement of the core bubble-liquid coolant interface are tracked by Marker and Cell (MAC) procedure. The results are obtained for the transient pressure at the vessel wall and also for the loading on the roof plug by the impact of the slug of liquid sodium. The computer code developed is validated against a benchmark experiment chosen to be ISPRA experiment reported in literature. The computer code is next applied to predict the loading on the Indian Prototype Fast Breeder Reactor (PFBR) being developed at Kalpakkam

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

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

International Nuclear Information System (INIS)

Oliveira, Carlos Alberto de; Mattar Neto, Miguel

2011-01-01

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

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

Directory of Open Access Journals (Sweden)

S V Kulkarni

2015-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

EPR: High load variation performances with the 'Tmode' core control

International Nuclear Information System (INIS)

Grossetete, A.

2008-01-01

The load variation performances on a PWR are directly linked to the core control design. This design is mainly characterized by the definition of the control rod banks and the way to both perform the banks movements and to modify the core boron concentration by injection of boric acid or water. The following paper presents the principles of the T mode, the new fully automatic core control mode for the EPR which provides high performance in terms of maneuverability and optimizes the effluents. First, the paper describes the division of the control rods into two control banks (Pbank for temperature and Hbank for power distribution). Then typical movements of these banks during power changes are shown. Then, the principle of the 3 control loops (Tave, AO, Pmax), used to obtain these desired control rod movements, is given. Finally, a load following transient simulation is presented. (authors)

EPR: high load variation performances with the 'TMODE' core control

International Nuclear Information System (INIS)

Pairot, Frederic

2008-01-01

The load variation performances on a PWR are directly linked to the core control design. This design is mainly characterized by the definition of the control rod banks and the way to both perform the banks movements and to modify the core boron concentration by injection of boric acid or water. The following paper presents the principles of the T mode, the new fully automatic core control mode for the EPR which provides high performance in terms of maneuverability and optimizes the effluents. First, the paper describes the division of the control rods into two control banks (Pbank for temperature and Hbank for power distribution). Then typical movements of these banks during power changes are shown. Then, the principle of the 3 control loops (Tave, AO, Pmax), used to obtain these desired control rod movements, is given. Finally, a load following transient simulation is presented. (author)

Development of a VVER-1000 core loading pattern optimization program based on perturbation theory

International Nuclear Information System (INIS)

Hosseini, Mohammad; Vosoughi, Naser

2012-01-01

Highlights: ► We use perturbation theory to find an optimum fuel loading pattern in a VVER-1000. ► We provide a software for in-core fuel management optimization. ► We consider two objectives for our method (perturbation theory). ► We show that perturbation theory method is very fast and accurate for optimization. - Abstract: In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. Two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain the fuel integrity. Because of the numerous possible patterns of fuel assemblies in the reactor core, finding the best configuration is so important and challenging. Different techniques for optimization of fuel loading pattern in the reactor core have been introduced by now. In this study, a software is programmed in C language to find an order of the fuel loading pattern of a VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process launches by considering an initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. The results on a typical VVER-1000 reactor reveal that the method could reach to a pattern with an allowed radial power peaking factor and increases the cycle length 1.1 days, as well.

Core-Sheath Paraffin-Wax-Loaded Nanofibers by Electrospinning for Heat Storage.

Science.gov (United States)

Lu, Yuan; Xiao, Xiudi; Zhan, Yongjun; Huan, Changmeng; Qi, Shuai; Cheng, Haoliang; Xu, Gang

2018-04-18

Paraffin wax (PW) is widely used for smart thermoregulation materials due to its good thermal performance. However, the leakage and low thermal conductivity of PW hinder its application in the heat storage field. Accordingly, developing effective methods to address these issues is of great importance. In this study, we explored a facile approach to obtain PW-loaded core-sheath structured flexible nanofibers films via coaxial electrospinning technique. The PW as the core layer was successfully encapsulated by the sheath-layer poly(methyl methacrylate). The diameter of the fiber core increased from 395 to 848 nm as the core solution speed rate increased from 0.1 to 0.5 mL/h. In addition, it can be seen that higher core solution speed rate could lead to higher PW encapsulation efficiency according to the transmission electron microscopy results. The core-sheath nanofiber films, moreover, possessed the highest latent heat of 58.25 J/g and solidifying enthalpy of -56.49 J/g. In addition, we found that after 200 thermal cycles, there was little change in latent heat, which demonstrated that it is beneficial for the PW-loaded core-sheath structure to overcome the leakage issue and enhance thermal stability properties for the thermoregulation film.

A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

Energy Technology Data Exchange (ETDEWEB)

Makmal, T. [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel); Nuclear Physics and Engineering Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Aviv, O. [Radiation Safety Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Gilad, E., E-mail: gilade@bgu.ac.il [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel)

2016-10-21

A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections. - Highlights: • Simple, inexpensive, safe and flexible experimental setup that can be quickly deployed. • Experimental results are thoroughly corroborated against ORIGEN2 burnup code. • Experimental uncertainty of 9% and 5% deviation between measurements and simulations. • Very high burnup MTR fuel element is examined, with 60% depletion of {sup 235}U. • Impact of highly irregular irradiation regime on burnup evaluation is studied.

Fracture Behaviours in Compression-loaded Triangular Corrugated Core Sandwich Panels

Directory of Open Access Journals (Sweden)

Zaid N.Z.M.

2016-01-01

Full Text Available The failure modes occurring in sandwich panels based on the corrugations of aluminium alloy, carbon fibre-reinforced plastic (CFRP and glass fibre-reinforced plastic (GFRP are analysed in this work. The fracture behaviour of these sandwich panels under compressive stresses is determined through a series of uniform lateral compression performed on samples with different cell wall thicknesses. Compression test on the corrugated-core sandwich panels were conducted using an Instron series 4505 testing machine. The post-failure examinations of the corrugated-core in different cell wall thickness were conducted using optical microscope. Load-displacement graphs of aluminium alloy, GFRP and CFRP specimens were plotted to show progressive damage development with five unit cells. Four modes of failure were described in the results: buckling, hinges, delamination and debonding. Each of these failure modes may dominate under different cell wall thickness or loading condition, and they may act in combination. The results indicate that thicker composites corrugated-core panels tend can recover more stress and retain more stiffness. This analysis provides a valuable insight into the mechanical behaviour of corrugated-core sandwich panels for use in lightweight engineering applications.

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-15

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

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

Directory of Open Access Journals (Sweden)

Corinne E. Rouquette-Loughlin

2017-04-01

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

New options to fuel plate for MTR reactor

International Nuclear Information System (INIS)

Macedo, C.R.

1988-01-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Core-based criterion for extreme supermodular functions

Czech Academy of Sciences Publication Activity Database

Studený, Milan; Kroupa, Tomáš

2016-01-01

Roč. 206, č. 1 (2016), s. 122-151 ISSN 0166-218X R&D Projects: GA ČR GA13-20012S EU Projects: European Commission 622645 - OASIG Institutional support: RVO:67985556 Keywords : supermodular function * submodular function * core * conditional independence * generalized permutohedron * indecomposable polytope Subject RIV: BA - General Mathematics Impact factor: 0.956, year: 2016 http://library.utia.cas.cz/separaty/2016/MTR/studeny-0459059.pdf

Core of Coalition Games on MV-algebras

Czech Academy of Sciences Publication Activity Database

Kroupa, Tomáš

2011-01-01

Roč. 21, č. 3 (2011), s. 479-492 ISSN 0955-792X R&D Projects: GA MŠk 1M0572; GA ČR GA102/08/0567 Institutional research plan: CEZ:AV0Z10750506 Keywords : coalition game * core * MV-algebra Subject RIV: BA - General Mathematics Impact factor: 0.611, year: 2011 http://library.utia.cas.cz/separaty/2011/MTR/kroupa-0359839.pdf

Core losses of an inverter-fed permanent magnet synchronous motor with an amorphous stator core under no-load

Directory of Open Access Journals (Sweden)

Nicolas Denis

2016-05-01

Full Text Available In this paper, an interior permanent magnet synchronous motor (IPMSM with a stator core made of amorphous magnetic material (AMM is presented. The IPMSM is driven by a voltage source three-phase inverter with classical pulse width modulation (PWM control. The core losses under no-load condition are measured by experiment and compared to an equivalent IPMSM with a stator core made of NO steel. Under these conditions, the core losses are influenced by the stator, rotor and magnet shapes but also by the PWM carrier signal that implies a high frequency harmonic in the magnetic flux density. It is demonstrated that the AMM can reduce the core losses by about 56 %.

Machine learning of the reactor core loading pattern critical parameters

International Nuclear Information System (INIS)

Trontl, K.; Pevec, D.; Smuc, T.

2007-01-01

The usual approach to loading pattern optimization involves high degree of engineering judgment, a set of heuristic rules, an optimization algorithm and a computer code used for evaluating proposed loading patterns. The speed of the optimization process is highly dependent on the computer code used for the evaluation. In this paper we investigate the applicability of a machine learning model which could be used for fast loading pattern evaluation. We employed a recently introduced machine learning technique, Support Vector Regression (SVR), which has a strong theoretical background in statistical learning theory. Superior empirical performance of the method has been reported on difficult regression problems in different fields of science and technology. SVR is a data driven, kernel based, nonlinear modelling paradigm, in which model parameters are automatically determined by solving a quadratic optimization problem. The main objective of the work reported in this paper was to evaluate the possibility of applying SVR method for reactor core loading pattern modelling. The starting set of experimental data for training and testing of the machine learning algorithm was obtained using a two-dimensional diffusion theory reactor physics computer code. We illustrate the performance of the solution and discuss its applicability, i.e., complexity, speed and accuracy, with a projection to a more realistic scenario involving machine learning from the results of more accurate and time consuming three-dimensional core modelling code. (author)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Characteristic statistic algorithm (CSA) for in-core loading pattern optimization

International Nuclear Information System (INIS)

Liu Zhihong; Hu Yongming; Shi Gong

2007-01-01

To solve the problem of PWR in-core loading pattern optimization, a more suitable global optimization algorithm, i.e., Characteristic statistic algorithm (CSA), is used. The searching process of this algorithm and how to apply it to this problem are presented. Loading pattern optimization code SCYCLE is developed. Two different problems on real PWR models are calculated and the results are compared with other algorithms. It is shown that SCYCLE has high efficiency and good global performance on this problem. (authors)

NaF-loaded core-shell PAN-PMMA nanofibers as reinforcements for Bis-GMA/TEGDMA restorative resins.

Science.gov (United States)

Cheng, Liyuan; Zhou, Xuegang; Zhong, Hong; Deng, Xuliang; Cai, Qing; Yang, Xiaoping

2014-01-01

A kind of core-shell nanofibers containing sodium fluoride (NaF) was produced and used as reinforcing materials for dimethacrylate-based dental restorative resins in this study. The core-shell nanofibers were prepared by coaxial-electrospinning with polyacrylonitrile (PAN) and poly(methyl methacrylate) (PMMA) solutions as core and shell fluids, respectively. The produced PAN-PMMA nanofibers varied in fiber diameter and the thickness of PMMA shell depending on electrospinning parameters. NaF-loaded nanofibers were obtained by incorporating NaF nanocrystals into the core fluid at two loadings (0.8 or 1.0wt.%). Embedment of NaF nanocrystals into the PAN core did not damage the core-shell structure. The addition of PAN-PMMA nanofibers into Bis-GMA/TEGDMA clearly showed the reinforcement due to the good interfacial adhesion between fibers and resin. The flexural strength (Fs) and flexural modulus (Ey) of the composites decreased slightly as the thickness of PMMA shell increasing. Sustained fluoride releases with minor initial burst release were achieved from NaF-loaded core-shell nanofibers and the corresponding composites, which was quite different from the case of embedding NaF nanocrystals into the dental resin directly. The study demonstrated that NaF-loaded PAN-PMMA core-shell nanofibers were not only able to improve the mechanical properties of restorative resin, but also able to provide sustained fluoride release to help in preventing secondary caries. © 2013.

Statistic techniques of process control for MTR type

International Nuclear Information System (INIS)

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

2002-01-01

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

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

Directory of Open Access Journals (Sweden)

Jani Reddy Bolla

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

Harmonic Differential Quadrature Analysis of Soft-Core Sandwich Panels under Locally Distributed Loads

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Xinwei Wang

2016-11-01

Full Text Available Sandwich structures are widely used in practice and thus various engineering theories adopting simplifying assumptions are available. However, most engineering theories of beams, plates and shells cannot recover all stresses accurately through their constitutive equations. Therefore, the soft-core is directly modeled by two-dimensional (2D elasticity theory without any pre-assumption on the displacement field. The top and bottom faces act like the elastic supports on the top and bottom edges of the core. The differential equations of the 2D core are then solved by the harmonic differential quadrature method (HDQM. To circumvent the difficulties in dealing with the locally distributed load by point discrete methods such as the HDQM, a general and rigorous way is proposed to treat the locally distributed load. Detailed formulations are provided. The static behavior of sandwich panels under different locally distributed loads is investigated. For verification, results are compared with data obtained by ABAQUS with very fine meshes. A high degree of accuracy on both displacement and stress has been observed.

On Input Vector Representation for the SVR model of Reactor Core Loading Pattern Critical Parameters

International Nuclear Information System (INIS)

Trontl, K.; Pevec, D.; Smuc, T.

2008-01-01

Determination and optimization of reactor core loading pattern is an important factor in nuclear power plant operation. The goal is to minimize the amount of enriched uranium (fresh fuel) and burnable absorbers placed in the core, while maintaining nuclear power plant operational and safety characteristics. The usual approach to loading pattern optimization involves high degree of engineering judgment, a set of heuristic rules, an optimization algorithm and a computer code used for evaluating proposed loading patterns. The speed of the optimization process is highly dependent on the computer code used for the evaluation. Recently, we proposed a new method for fast loading pattern evaluation based on general robust regression model relying on the state of the art research in the field of machine learning. We employed Support Vector Regression (SVR) technique. SVR is a supervised learning method in which model parameters are automatically determined by solving a quadratic optimization problem. The preliminary tests revealed a good potential of the SVR method application for fast and accurate reactor core loading pattern evaluation. However, some aspects of model development are still unresolved. The main objective of the work reported in this paper was to conduct additional tests and analyses required for full clarification of the SVR applicability for loading pattern evaluation. We focused our attention on the parameters defining input vector, primarily its structure and complexity, and parameters defining kernel functions. All the tests were conducted on the NPP Krsko reactor core, using MCRAC code for the calculation of reactor core loading pattern critical parameters. The tested input vector structures did not influence the accuracy of the models suggesting that the initially tested input vector, consisted of the number of IFBAs and the k-inf at the beginning of the cycle, is adequate. The influence of kernel function specific parameters (σ for RBF kernel

Genetic algorithm for the optimization of the loading pattern for reactor core fuel management

International Nuclear Information System (INIS)

Zhou Sheng; Hu Yongming; zheng Wenxiang

2000-01-01

The paper discusses the application of a genetic algorithm to the optimization of the loading pattern for in-core fuel management with the NP characteristics. The algorithm develops a matrix model for the fuel assembly loading pattern. The burnable poisons matrix was assigned randomly considering the distributed nature of the poisons. A method based on the traveling salesman problem was used to solve the problem. A integrated code for in-core fuel management was formed by combining this code with a reactor physics code

Development of an RF accelerating structure loaded with multi-ring magnetic cores

International Nuclear Information System (INIS)

Morita, Yuichi; Kageyama, Tatsuya; Kato, Ichiro; Yamashita, Satoru

2012-01-01

In order to upgrade the J-PARC rings (RCS and MR) for more beam powers, the existing accelerating structures for both rings need to be improved for better performance especially in the long-term reliability. As a solution for this purpose, we have proposed a new accelerating structure loaded with multi-ring core modules. Each core module consists of three ring FINEMET cores with different radial sizes concentrically arranged and sandwiched between two glass epoxy plates with flow channels grooved on the surfaces. The Fe-based FINEMET cores are to be cooled with the turbulent flow of Fluorinert (chemically inert perfluorinated liquid). Therefore, the cores need neither impregnation nor coating with epoxy resin for anti corrosion. A half-gap cavity loaded with three core modules, which is a minimum configuration for the performance test, is under fabrication. Additionally, a high efficient solid state RF amplifier is under development. Thirty two amplifier modules, each of which is a push-pull class-D amplifier driven by power MOSFET hybrids, are combined to deliver RF power up to 60 kW (peak power with a duty factor of 50%) at frequencies 1.7 ± 0.2MHz. The amplitude of the RF output can be modulated by changing the voltage across the drain and source of the power MOSFET in proportion to the wave envelope. This paper reports the recent status of our R and D activities. (author)

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

International Nuclear Information System (INIS)

Abbate, P.; Madariaga, M.R.

1990-01-01

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

Machine Learning of the Reactor Core Loading Pattern Critical Parameters

Directory of Open Access Journals (Sweden)

Krešimir Trontl

2008-01-01

Full Text Available The usual approach to loading pattern optimization involves high degree of engineering judgment, a set of heuristic rules, an optimization algorithm, and a computer code used for evaluating proposed loading patterns. The speed of the optimization process is highly dependent on the computer code used for the evaluation. In this paper, we investigate the applicability of a machine learning model which could be used for fast loading pattern evaluation. We employ a recently introduced machine learning technique, support vector regression (SVR, which is a data driven, kernel based, nonlinear modeling paradigm, in which model parameters are automatically determined by solving a quadratic optimization problem. The main objective of the work reported in this paper was to evaluate the possibility of applying SVR method for reactor core loading pattern modeling. We illustrate the performance of the solution and discuss its applicability, that is, complexity, speed, and accuracy.

Machine Learning of the Reactor Core Loading Pattern Critical Parameters

International Nuclear Information System (INIS)

Trontl, K.; Pevec, D.; Smuc, T.

2008-01-01

The usual approach to loading pattern optimization involves high degree of engineering judgment, a set of heuristic rules, an optimization algorithm, and a computer code used for evaluating proposed loading patterns. The speed of the optimization process is highly dependent on the computer code used for the evaluation. In this paper, we investigate the applicability of a machine learning model which could be used for fast loading pattern evaluation. We employ a recently introduced machine learning technique, support vector regression (SVR), which is a data driven, kernel based, nonlinear modeling paradigm, in which model parameters are automatically determined by solving a quadratic optimization problem. The main objective of the work reported in this paper was to evaluate the possibility of applying SVR method for reactor core loading pattern modeling. We illustrate the performance of the solution and discuss its applicability, that is, complexity, speed, and accuracy

Analysis of ex-core detector response measured during nuclear ship Mutsu land-loaded core critical experiment

International Nuclear Information System (INIS)

Itagaki, M.; Abe, J.I.; Kuribayashi, K.

1987-01-01

There are some cases where the ex-core neutron detector response is dependent not only on the fission source distribution in a core but also on neutron absorption in the borated water reflector. For example, an unexpectedly large response variation was measured during the nuclear ship Mutsu land-loaded core critical experiment. This large response variation is caused largely by the boron concentration change associated with the change in control rod positioning during the experiment. The conventional Crump-Lee response calculation method has been modified to take into account this boron effect. The correction factor in regard to this effect has been estimated using the one-dimensional transport code ANISN. The detector response variations obtained by means of this new calculation procedure agree well with the measured values recorded during the experiment

Experience with mixed cores in the IRR-1

International Nuclear Information System (INIS)

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

1985-01-01

Over twenty mixed core configurations composed of 'old' (18 curved plate) and 'new' 23 flat plate) MTR type fuel elements were irradiated in the IRR-1 swimming pool reactor. The number of 'new' fuel elements in the core varied from one to twenty. To establish the safety of these configurations, thermohydraulic calculations were carried out to derive the maximum allowed hot channel power, determined by the onset of flow instabilities. A core is considered safe if its hot channel power, obtained from a two-dimensional neutronic calculation of power distribution in the core, does not exceed the maximum allowed value. The conservative nature of the assumptions used in the above safety evaluation procedure was verified by measurements of pressure drops vs. coolant flow rates as well as of temperature and neutron flux distributions. (author)

The Analysis of the Effect of Coolant Channel Width on Fuel Loading of the RSG-GAS Core

International Nuclear Information System (INIS)

Surbakti; Tukiran

2004-01-01

The RGS-GAS using uranium silicide fuel, plate type and 250 g U of loading is planned to increase the fuel loading to 300 g U even to 400 g U. The silicide fuel has advantages when increase the fuel loading in the same volume. Because of that case, it is necessary to analyze the effect of coolant channel width on fuel loading of the RSG-GAS core. Analyzing the effect the work which done is to generate cell and core calculation using WIMSD/4 and Batan-2DIFF codes. The WIMSD/4 code is used to generate cross section of core material and Batan-2DIFF is used to calculate the effective multiplication factor. The model that used in this calculation there are three kind of fuel loading namely, 250 g U, 300 g U and 400 g U. The coolant channel width is simulated from 1.75 mm to 2.55 mm. From that fuel loadings, it is analyzed which coolant channel width gave the best effective multiplication factor. From result of analysis showed that the best effective multiplication factor is on the coolant channel width of 2.55 mm for third of fuel loadings. (author)

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

International Nuclear Information System (INIS)

Schneider, B.; Hofmann, G.

1994-09-01

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

Operational and research activities of Tsing Hua open pool reactor

International Nuclear Information System (INIS)

Wang, T.-K.; Tseng, D.-L.; Chou, H.-P.; Onyang Minsun

1988-01-01

Tsing Hua Open Pool Reaction (THOR) is the first nuclear reactor to become operational in Taiwan. It reached its first critical on April 13, 1961. Until now, THOR has been operated successfully for 27 years. The major missions of THOR include radioisotope production, neutron activation analysis, nuclear science and engineering researches, education, and personnel training. The THOR was originally loaded with HEU MTR-type fuels. A gradual fuel replacing program using LEU TRIGA fuel to replace MTR started in 1977. By 1987, THOR was loaded with all TRIGA fuels. This paper gives a brief history of THOR, its current status, the core conversion work, some selected research topics, and its improvement plan. (author)

Dithranol-loaded lipid-core nanocapsules improve the photostability and reduce the in vitro irritation potential of this drug

Energy Technology Data Exchange (ETDEWEB)

Savian, Ana L. [Programa de Pós-Graduação em Ciências Farmacêuticas, Universidade Federal de Santa Maria, Av. Roraima, 1000, Santa Maria, RS 97105-900 (Brazil); Rodrigues, Daiane [Curso de Farmácia, Centro de Ciências da Saúde, Universidade Federal de Santa Maria, Av. Roraima, 1000, Santa Maria, RS 97105-900 (Brazil); Weber, Julia; Ribeiro, Roseane F. [Programa de Pós-Graduação em Ciências Farmacêuticas, Universidade Federal de Santa Maria, Av. Roraima, 1000, Santa Maria, RS 97105-900 (Brazil); Motta, Mariana H. [Curso de Farmácia, Centro de Ciências da Saúde, Universidade Federal de Santa Maria, Av. Roraima, 1000, Santa Maria, RS 97105-900 (Brazil); Schaffazick, Scheila R.; Adams, Andréa I.H. [Programa de Pós-Graduação em Ciências Farmacêuticas, Universidade Federal de Santa Maria, Av. Roraima, 1000, Santa Maria, RS 97105-900 (Brazil); Andrade, Diego F. de; Beck, Ruy C.R. [Programa de Pós-Graduação em Ciências Farmacêuticas, Universidade Federal do Rio Grande do Sul, Av. Ipiranga, 2752, Porto Alegre, RS 90610-000 (Brazil); and others

2015-01-01

Dithranol is a very effective drug for the topical treatment of psoriasis. However, it has some adverse effects such as irritation and stain in the skin that make its application and patient adherence to treatment difficult. The aims of this work were to prepare and characterize dithranol-loaded nanocapsules as well as to evaluate the photostability and the irritation potential of these nanocarriers. Lipid-core nanocapsules containing dithranol (0.5 mg/mL) were prepared by interfacial deposition of preformed polymer. EDTA (0.05%) or ascorbic acid (0.02%) was used as antioxidants. After preparation, dithranol-loaded lipid-core nanocapsules showed satisfactory characteristics: drug content close to the theoretical concentration, encapsulation efficiency of about 100%, nanometric mean size (230–250 nm), polydispersity index below 0.25, negative zeta potential, and pH values from 4.3 to 5.6. In the photodegradation study against UVA light, we observed a higher stability of the dithranol-loaded lipid-core nanocapsules comparing to the solution containing the free drug (half-life times around 4 and 1 h for the dithranol-loaded lipid-core nanocapsules and free drug solution containing EDTA, respectively; half-life times around 17 and 7 h for the dithranol-loaded lipid-core nanocapsules and free drug solution containing ascorbic acid, respectively). Irritation test by HET-CAM method was conducted to evaluate the safety of the formulations. From the results it was found that the nanoencapsulation of the drug decreased its toxicity compared to the effects observed for the free drug. - Highlights: • Strategy to prepare lipid-core nanocapsules containing dithranol • Evaluation of the nanoencapsulation effect on the photostability and irritation • Evaluation of the in vitro release of dithranol-loaded lipid-core nanocapsules.

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.

Westinghouse loading pattern search methodology for complex core designs

International Nuclear Information System (INIS)

Chao, Y.A.; Alsop, B.H.; Johansen, B.J.; Morita, T.

1991-01-01

Pressurized water reactor core designs have become more complex and must meet a plethora of design constraints. Trends have been toward longer cycles with increased discharge burnup, increased burnable absorber (BA) number, mixed BA types, reduced radial leakage, axially blanketed fuel, and multiple-batch feed fuel regions. Obtaining economical reload core loading patterns (LPs) that meet design criteria is a difficult task to do manually. Automated LP search tools are needed. An LP search tool cannot possibly perform an exhaustive search because of the sheer size of the combinatorial problem. On the other hand, evolving complexity of the design features and constraints often invalidates expert rules based on past design experiences. Westinghouse has developed a sophisticated loading pattern search methodology. This methodology is embodied in the LPOP code, which Westinghouse nuclear designers use extensively. The LPOP code generates a variety of LPs meeting design constraints and performs a two-cycle economic evaluation of the generated LPs. The designer selects the most appropriate patterns for fine tuning and evaluation by the design codes. This paper describes the major features of the LPOP methodology that are relevant to fulfilling the aforementioned requirements. Data and examples are also provided to demonstrate the performance of LPOP in meeting the complex design needs

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

Conceptual design of next generation MTR

Energy Technology Data Exchange (ETDEWEB)

Nagata, Hiroshi; Yamaura, Takayuki; Naka, Michihiro; Kawamata, Kazuo; Izumo, Hironobu; Hori, Naohiko; Nagao, Yoshiharu; Kusunoki, Tsuyoshi; Kaminaga, Masanori; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan); Mine, M [Hitachi-GE Nuclear Energy, Ltd., Hitachi, Ibaraki (Japan); Yamazaki, S [Kawasaki Heavy Industries, Ltd., Kobe, Hyogo (Japan); Ishikawa, S [NGK Insulators, Ltd., Nagoya, Aichi (Japan); Miura, K [Sukegawa Electric Co., Ltd., Takahagi, Ibaraki (Japan); Nakashima, S [Fuji Electric Co., Ltd., Tokyo (Japan); Yamaguchi, K [Chiyoda Technol Corp., Tokyo (Japan)

2012-03-15

Conceptual design of the high-performance and low-cost next generation materials testing reactor (MTR) which will be expected to construct in the nuclear power plant introduction countries, started from 2010 in JAEA and nuclear-related companies in Japan. The aims of this conceptual design are to achieve highly safe reactor, economical design, high availability factor and advanced irradiation utilization. One of the basic reactor concept was determined as swimming pool type, thermal power of 10MW and water cooled and moderated reactor with plate type fuel element same as the JMTR. It is expected that the research reactors are used for human resource development, progress of the science and technology, expansion of industry use, lifetime extension of LWRs and so on. (author)

Corrosion of MTR type fuel plates containing U3O8-Al cermet cores

International Nuclear Information System (INIS)

Durazzo, M.

1985-01-01

The fuel plate samples containing U 3 O 8 -Al cermet cores with concentrations from 10 to 90% of U 3 O 8 weight were fabricated. Samples with 58% of U 3 O 8 eight were fabricated using compacts with densities from 75 to 95% of theoretical density. The influences of U 3 O 8 concentration and porosity of compacted core on porosity and uniformity of core thickness are discussed. The U 3 O 8 -Al cores were submitted to corrosion tests and exposed to deionized water at temperatures of 30, 50, 70 and 90 0 C by cladding deffect produced artificially. The results shown that core corrosion is accompanied by hydrogen release. The total volum of released hydrogen and the time interval to observe the initiation of hydrogen releasing (incubation time) are depending on core pososity and absolute temperature. A mechanism for U 3 O 8 -Al core corrosion process is proposed and discussed. The cladding of fuel plate samples was submitted to corrosion tests under similar conditons of the IAE-R1 reactor operating at 2, 5 and 10 MW. (Author) [pt

Preliminary analysis of a large 1600 MWe PWR core loaded with 30% MOX fuel

International Nuclear Information System (INIS)

Polidoro, Franco; Corsetti, Edoardo; Vimercati, Giuliano

2011-01-01

The paper presents a full-core 3-D analysis of the performances of a large 1600 MWe PWR core, loaded with 30% MOX fuel, in accordance with the European Utility Requirements (EUR). These requirements state that the European next generation power plants have to be designed capable to use MOX (UO 2 - PuO 2 ) fuel assemblies up to 50% of the core, together with UO 2 fuel assemblies. The use of MOX assemblies has a significant impact on key physic parameters and on safety. A lot of studies have been carried out in the past to explore the feasibility of plutonium recycling strategies by loading LWR reactors with MOX fuel. Many of these works were based on lattice codes, in order to perform detailed analyses of the neutronic characteristics of MOX assemblies. With the aim to take into account their interaction with surrounding UO 2 fuel elements, and the global effects on the core at operational conditions, an integrated approach making use of a 3-D core simulation is required. In this light, the present study adopts the state-of-art numerical models CASMO-5 and SIMULATE-3 to analyze the behavior of the core fueled with 30% MOX and to compare it with that of a large PWR reference core, fueled with UO 2 . (author)

Fluid structure interaction studies on acoustic load response of light water nuclear reactor core internals under blowdown condition

International Nuclear Information System (INIS)

Moses Lemuel Raj, G.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

1998-12-01

Acoustic load evaluation within two phase medium and the related fluid-structure interaction analysis in case of Loss of Coolant Accidents (LOCA) for light water reactor systems is an important inter-disciplinary area. The present work highlights the development of a three-dimensional finite element code FLUSHEL to analyse LOCA induced depressurization problems for Pressurised Water Reactor (PWR) core barrel and Boiling Water Reactor (BWR) core shroud. With good comparison obtained between prediction made by the present code and the experimental results of HDR-PWR test problem, coupled fluid-structure interaction analysis of core shroud of Tarapur Atomic Power Station (TAPS) is presented for recirculation line break. It is shown that the acoustic load induced stresses in the core shroud are small and downcomer acoustic cavity modes are decoupled with the shell multi-lobe modes. Thus the structural integrity of TAPS core shroud for recirculation line break induced acoustic load is demonstrated. (author)

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

International Nuclear Information System (INIS)

Weiss, M.

2001-01-01

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

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

International Nuclear Information System (INIS)

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

1988-01-01

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

Preliminary concept of a zero power nuclear reactor core

International Nuclear Information System (INIS)

Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

2011-01-01

The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

Preliminary concept of a zero power nuclear reactor core

Energy Technology Data Exchange (ETDEWEB)

Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

Operating experience, measurements, and analysis of the LEU whole core demonstration at the FNR

International Nuclear Information System (INIS)

Weha, D.K.; Drumm, C.R.; King, J.S.; Martin, W.R.; Lee, J.C.

1984-01-01

The 2-MW Ford Nuclear Reactor at the University of Michigan is serving as the demonstration reactor for the MTR-type low enrichment (LEU) fuel for the Reduced Enrichment for Research and Test Reactor program. Operational experience gained through six months of LEU core operation and seven months of mixed HEU-LEU core operation is presented. Subcadmium flux measurements performed with rhodium self-powered neutron detectors and iron wire activations are compared with calculations. Measured reactivity parameters are compared for HEU and LEU cores. Finally, the benchmark calculations for several HEU, LEU, and mixed HEU-LEU FNR cores and the International Atomic Energy Agency (IAEA) benchmark problem are presented. (author)

MTR fuel testing in BR2

International Nuclear Information System (INIS)

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

2000-01-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-11-20

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-06-30

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

Non-linear Dynamic Analysis of Steel Hollow I-core Sandwich Panel under Air Blast Loading

Directory of Open Access Journals (Sweden)

Asghar Vatani Oskouei

2015-12-01

Full Text Available In this paper, the non-linear dynamic response of novel steel sandwich panel with hollow I-core subjected to blast loading was studied. Special emphasis is placed on the evaluation of midpoint displacements and energy dissipation of the models. Several parameters such as boundary conditions, strain rate, mesh dependency and asymmetrical loading are considered in this study. The material and geometric non-linearities are also considered in the numerical simulation. The results obtained are compared with available experimental data to verify the developed FE model. Modeling techniques are described in detail. According to the results, sandwich panels with hollow I-core allowed more plastic deformation and energy dissipation and less midpoint displacement than conventional I-core sandwich panels and also equivalent solid plate with the same weight and material.

Parameter studies to determine sensitivity of slug impact loads to properties of core surrounding structures

International Nuclear Information System (INIS)

Gvildys, J.

1985-01-01

A sensitivity study of the HCDA slug impact response of fast reactor primary containment to properties of core surrounding structures was performed. Parameters such as the strength of the radial shield material, mass, void, and compressibility properties of the gas plenum material, mass of core material, and mass and compressibility properties of the coolant were used as variables to determine the magnitude of the slug impact loads. The response of the reactor primary containment and the partition of energy were also given. A study was also performed using water as coolant to study the difference in slug impact loads

SEDRIO/INCORE, an automatic optimal loading pattern search system for PWR NPP reload core using an expert system

International Nuclear Information System (INIS)

Xian Chunyu; Zhang Zongyao

2003-01-01

The expert knowledge library for Daya Bay and Qinshan phase II NPP has been established based on expert knowledge, and the reload core loading pattern heuristic search is performed. The in-core fuel management code system INCORE that has been used in engineering design is employed for neutron calculation, and loading pattern is evaluated by using of cycle length and core radial power peaking factor. The developed system SEDRIO/INCORE has been applied in cycle 4 for unit 2 of Daya Bay NPP and cycle 4 for Phase II in Qinshan NPP. The application demonstrated that the loading patterns obtained by SEDRIO/INCORE system are much better than reference ones from the view of the radial power peak and the cycle length

LIGHT-WEIGHT LOAD-BEARING STRUCTURES REINFORCED BY CORE ELEMENTS MADE OF SEGMENTS AND A METHOD OF CASTING SUCH STRUCTURES

DEFF Research Database (Denmark)

2009-01-01

The invention relates to a light-weight load-bearing structure, reinforced by core elements (2) of a strong material constituting one or more compression or tension zones in the structure to be cast, which core (2) is surrounded by or adjacent to a material of less strength compared to the core (2......), where the core (2) is constructed from segments (1) of core elements (2) assembled by means of one or more prestressing elements (4). The invention further relates to a method of casting of light-weight load-bearing structures, reinforced by core elements (2) of a strong material constituting one...... or more compression or tension zones in the structure to be cast, which core (2) is surrounded by or adjacent to a material of less strength compared to the core (2), where the core (2) is constructed from segments (1) of core elements (2) assembled and hold together by means of one or more prestressing...

Safeguarding subcriticality during loading and shuffling operations in the higher density of the RSG-GAS's silicide core

International Nuclear Information System (INIS)

Sembiring, T.M.; Kuntoro, I.

2003-01-01

The core conversion program of the RSG-GAS reactor is to convert the all-oxide to all-silicide core. The silicide equilibrium core with fuel meat density of 3.55 gU cm -3 is an optimal core for RSG-GAS reactor and it can significantly increase the operation cycle length from 25 to 32 full power days. Nevertheless, the subcriticality of the shutdown core and the shutdown margin are lower than of the oxide core. Therefore, the deviation of subcriticality condition in the higher silicide core caused by the fuel loading and shuffling error should be reanalysed. The objective of this work is to analyse the sufficiency of the subcriticality condition of the shutdown core to face the worst condition caused by an error during loading and shuffling operations. The calculations were carried out using the 2-dimensional multigroup neutron diffusion code of Batan-FUEL. In the fuel handling error, the calculated results showed that the subcriticality condition of the shutdown higher density silicide equilibrium core of RSG-GAS can be maintained. Therefore, all fuel management steps are fixed in the present reactor operation manual can be applied in the higher silicide equilibrium core of RSG-GAS reactor. (author)

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

International Nuclear Information System (INIS)

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

1985-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-06-27

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Development of a core follow calculational system for research reactors

International Nuclear Information System (INIS)

Muller, E.Z.; Ball, G.; Joubert, W.R.; Schutte, H.C.; Stoker, C.C.; Reitsma, F.

1994-01-01

Over the last few years a comprehensive Pressurized Water Reactor and Materials Testing Reactor core analysis code system based on modern reactor physics methods has been under development by the Atomic Energy Corporation of South Africa. This system, known as OSCAR-3, will incorporate a customized graphical user interface and data management system to ensure user-friendliness and good quality control. The system has now reached the stage of development where it can be used for practical MTR core analyses. This paper describes the current capabilities of the components of the OSCAR-3 package, their integration within the package, and outlines future developments. 10 refs., 1 tab., 1 fig

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

International Nuclear Information System (INIS)

Brodda, B.G.; Fachinger, J.

1995-01-01

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

Fuel requirements for experimental devices in MTR reactors. A perturbation model for reactor core analysis

International Nuclear Information System (INIS)

Beeckmans de West-Meerbeeck, A.

1991-01-01

Irradiation in neutron absorbing devices, requiring high fast neutron fluxes in the core or high thermal fluxes in the reflector and flux traps, lead to higher density fuel and larger core dimensions. A perturbation model of the reactor core helps to estimate the fuel requirements. (orig.)

Performance of Estimation of distribution algorithm for initial core loading optimization of AHWR-LEU

International Nuclear Information System (INIS)

Thakur, Amit; Singh, Baltej; Gupta, Anurag; Duggal, Vibhuti; Bhatt, Kislay; Krishnani, P.D.

2016-01-01

Highlights: • EDA has been applied to optimize initial core of AHWR-LEU. • Suitable value of weighing factor ‘α’ and population size in EDA was estimated. • The effect of varying initial distribution function on optimized solution was studied. • For comparison, Genetic algorithm was also applied. - Abstract: Population based evolutionary algorithms now form an integral part of fuel management in nuclear reactors and are frequently being used for fuel loading pattern optimization (LPO) problems. In this paper we have applied Estimation of distribution algorithm (EDA) to optimize initial core loading pattern (LP) of AHWR-LEU. In EDA, new solutions are generated by sampling the probability distribution model estimated from the selected best candidate solutions. The weighing factor ‘α’ decides the fraction of current best solution for updating the probability distribution function after each generation. A wider use of EDA warrants a comprehensive study on parameters like population size, weighing factor ‘α’ and initial probability distribution function. In the present study, we have done an extensive analysis on these parameters (population size, weighing factor ‘α’ and initial probability distribution function) in EDA. It is observed that choosing a very small value of ‘α’ may limit the search of optimized solutions in the near vicinity of initial probability distribution function and better loading patterns which are away from initial distribution function may not be considered with due weightage. It is also observed that increasing the population size improves the optimized loading pattern, however the algorithm still fails if the initial distribution function is not close to the expected optimized solution. We have tried to find out the suitable values for ‘α’ and population size to be considered for AHWR-LEU initial core loading pattern optimization problem. For sake of comparison and completeness, we have also addressed the

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Characteristics of first loaded IG-110 graphite in HTTR core

International Nuclear Information System (INIS)

Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Sawa, Kazuhiro; Hanawa, Satoshi; Ishihara, Masahiro

2006-10-01

IG-110 graphite is a fine-grained isotropic and nuclear-grade graphite with excellent resistivity on both irradiation and corrosion and with high strength. The IG-110 graphite is used for the graphite components of High Temperature Engineering Test Reactor (HTTR) such as fuel and control rod guide blocks and support posts. In order to design and fabricate the graphite components in the HTTR, the Japan Atomic Energy Research Institute (the Japan Atomic Energy Agency at present) had established the graphite structural design code and design data on the basis of former research results. After the design code establishment, the IG-110 graphite components were fabricated and loaded in the HTTR core. This report summarized the characteristics of the first loaded IG-110 graphite as basic data for surveillance test, measuring material characteristics changed by neutron irradiation and oxidation. By comparing the design data, it was shown that the first loaded IG-110 graphite had excellent strength properties and enough safety margins to the stress limits in the design code. (author)

Partial thorium loading in the initial core of Kakrapar atomic power reactor

International Nuclear Information System (INIS)

Balakrishnan, M.R.

1993-01-01

The first unit of Kakrapar nuclear power station has gone critical with some thorium oxide fuel bundles loaded in its core. The thorium helps to flatten the power by reducing neutron flux in the centre of the reactor. However, the placing of the thorium had to be planned with care, because if the neutron flux at a point where a safety rod is located is depressed, the reactivity worth of the safety rod gets reduced. Using a dynamic programing approach, the Reactor Engineering Division of Bhabha Atomic Research Centre worked out a satisfactory configuration for loading the thorium bundles

Bulk coolant cavitation in LMFBR containment loading following a whole-core explosion

International Nuclear Information System (INIS)

Jones, A.V.

1977-01-01

An LMFBR core undergoing an explosion transmits energy to the containment in a series of pressure waves and the containment loading is determined by their cumulative effect. These pressure waves are modified by their interaction with the coolant through which they propagate. It is necessary to model both the induction of bulk cavitation by tension waves and the interaction of pressure waves with cavitated liquid in realistic containment loading calculations. This paper sets out the progress which has been achieved in such modelling and first indications for the effect of bulk coolant cavitation in LMFBR containment loading. Conclusions may be briefly summarised: 1) Bulk cavitation must be included in realistic containment loading calculations. 2) Phenomenological models of cavitated liquid without memory are inappropriate. The best approach is to model bubble dynamics directly, including at least momentum conservation and surface tension. 3) The containment loading resulting from a given explosion is sensitive to the state of preparation of the coolant. The number density of nucleation sites should therfore accompany the results of model tests. (Auth.)

Fuel loading method to exchangeable reactor core of BWR type reactor and its core

International Nuclear Information System (INIS)

Koguchi, Kazushige.

1995-01-01

In a fuel loading method for an exchangeable reactor core of a BWR type reactor, at least two kinds of fresh fuel assemblies having different reactivities between axial upper and lower portions are preliminarily prepared, and upon taking out fuel assemblies of advanced combustion and loading the fresh fuel assemblies dispersingly, they are disposed so as to attain a predetermined axial power distribution in the reactor. At least two kinds of fresh fuel assemblies have a content of burnable poisons different between the axial upper portion and lower portions. In addition, reactivity characteristics are made different at a region higher than the central boundary and a region lower than the central boundary which is set within a range of about 6/24 to 16/24 from the lower portion of the fuel effective length. There can be attained axial power distribution as desired such as easy optimization of the axial power distribution, high flexibility, and flexible flattening of the power distribution, and it requires no special change in view of the design and has a good economical property. (N.H.)

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

International Nuclear Information System (INIS)

Fleck, C.M.

1979-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-07-01

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

Mechanical and thermo-mechanical response of a lead-core bearing device subjected to different loading conditions

Directory of Open Access Journals (Sweden)

Zhelyazov Todor

2018-01-01

Full Text Available The contribution is focused on the numerical modelling, simulation and analysis of a lead-core bearing device for passive seismic isolation. An accurate finite element model of a lead-core bearing device is presented. The model is designed to analyse both mechanical and thermo-mechanical responses of the seismic isolator to different loading conditions. Specifically, the mechanical behaviour in a typical identification test is simulated. The response of the lead-core bearing device to circular sinusoidal paths is analysed. The obtained shear displacement – shear force relationship is compared to experimental data found in literature sources. The hypothesis that heating of the lead-core during cyclic loading affects the degrading phenomena in the bearing device is taken into account. Constitutive laws are defined for each material: lead, rubber and steel. Both predefined constitutive laws (in the used general–purpose finite element code and semi-analytical procedures aimed at a more accurate modelling of the constitutive relations are tested. The results obtained by finite element analysis are to be further used to calibrate a macroscopic model of the lead-core bearing device seen as a single-degree-of-freedom mechanical system.

One-pot synthesis and characterization of rhodamine derivative-loaded magnetic core-shell nanoparticles

Energy Technology Data Exchange (ETDEWEB)

Zhang Jin, E-mail: jzhang@eng.uwo.ca; Li Jiaxin [University of Western Ontario, Department of Chemical and Biochemical Engineering (Canada); Razavi, Fereidoon S. [Brock University, Department of Physics (Canada); Mumin, Abdul Md. [University of Western Ontario, Department of Chemical and Biochemical Engineering (Canada)

2011-05-15

A new method to produce elaborate nanostructure with magnetic and fluorescent properties in one entity is reported in this article. Magnetite (Fe{sub 3}O{sub 4}) coated with fluorescent silica (SiO{sub 2}) shell was produced through the one-pot reaction, in which one reactor was utilized to realize the synthesis of superparamagnetic core of Fe{sub 3}O{sub 4}, the formation of SiO{sub 2} coating through the condensation and polymerization of tetraethylorthosilicate (TEOS), and the encapsulation of tetramethyl rhodamine isothiocyanate-dextran (TRITC-dextran) within silica shell. Transmission electron microscopy (TEM), energy dispersive X-ray (EDX) analysis, and X-ray diffraction (XRD) were carried out to investigate the core-shell structure. The magnetic core of the core-shell nanoparticles is 60 {+-} 10 nm in diameter. The thickness of the fluorescent SiO{sub 2} shell is estimated at 15 {+-} 5 nm. In addition, the fluorescent signal of the SiO{sub 2} shell has been detected by the laser confocal scanning microscopy (LCSM) with emission wavelength ({lambda}{sub em}) at 566 nm. In addition, the magnetic properties of TRITC-dextran loaded silica-coating iron oxide nanoparticles (Fe{sub 3}O{sub 4}-SiO{sub 2} NPs) were studied. The hysteresis loop of the core-shell NPs measured at room temperature shows that the saturation magnetization (M{sub s}) is not reached even at the field of 70 kOe (7T). Meanwhile, the very low coercivity (H{sub c}) and remanent magnetization (M{sub r}) are 0.375 kOe and 6.6 emu/g, respectively, at room temperature. It indicates that the core-shell particles have the superparamagnetic properties. The measured blocking temperature (T{sub B}) of the TRITC-dextran loaded Fe{sub 3}O{sub 4}-SiO{sub 2} NPs is about 122.5 K. It is expected that the multifunctional core-shell nanoparticles can be used in bio-imaging.

Generalized perturbation theory error control within PWR core-loading pattern optimization

International Nuclear Information System (INIS)

Imbriani, J.S.; Turinsky, P.J.; Kropaczek, D.J.

1995-01-01

The fuel management optimization code FORMOSA-P has been developed to determine the family of near-optimum loading patterns for PWR reactors. The code couples the optimization technique of simulated annealing (SA) with a generalized perturbation theory (GPT) model for evaluating core physics characteristics. To ensure the accuracy of the GPT predictions, as well as to maximize the efficient of the SA search, a GPT error control method has been developed

Optimization of fuel core loading pattern design in a VVER nuclear power reactors using Particle Swarm Optimization (PSO)

International Nuclear Information System (INIS)

Babazadeh, Davood; Boroushaki, Mehrdad; Lucas, Caro

2009-01-01

The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor (K eff ) in order to extract the maximum energy, and keeping the local power peaking factor (P q ) lower than a predetermined value to maintain fuel integrity. In this research, a new strategy based on Particle Swarm Optimization (PSO) algorithm has been developed to optimize the fuel core loading pattern in a typical VVER. The PSO algorithm presents a simple social model by inspiration from bird collective behavior in finding food. A modified version of PSO algorithm for discrete variables has been developed and implemented successfully for the multi-objective optimization of fuel loading pattern design with constraints of keeping P q lower than a predetermined value and maximizing K eff . This strategy has been accomplished using WIMSD and CITATION calculation codes. Simulation results show that this algorithm can help in the acquisition of a new pattern without contravention of the constraints.

Conceptual study of axial offset fluctuations upon stepwise power changes in a thorium–plutonium core to improve load-following conditions

International Nuclear Information System (INIS)

Lau, Cheuk Wah; Dykin, Victor; Nylén, Henrik; Björk, Klara Insulander; Sandberg, Urban

2014-01-01

Highlights: • Thorium–plutonium mixed oxide to improve nuclear reactors load-following capability. • SIMULATE-3 was the main calculation tool. • The Ringhals-3 PWR unit in Sweden was used as a reference. • Lower xenon poisoning and shorter reactor dead time. - Abstract: The increased share of renewable energy, such as wind and solar power, will increase the demand for load-following power sources, and nuclear reactors could be one option. However, during rapid load-following events, traditional UOX cores could be restricted by the volatile oscillation of the power distribution. Therefore, a conceptual study on stability properties of Th-MOX PWR concerning axial offset power excursion during load-following events are investigated and discussed. The study is performed in SIMULATE-3 for a realistic PWR core (Ringhals-3) at the end of cycle, where the largest amplitude of the axial offset oscillations is expected. It is shown that the Th-MOX core possesses much better stability characteristics and shorter reactor dead time compared with a traditional UOX core, and the main reasons are the lower sensitivity to perturbations in the neutron spectrum, lower xenon poisoning and lower thermal neutron flux

Adaptation of adhesive post and cores to dentin after in vitro occlusal loading: evaluation of post material influence.

Science.gov (United States)

Dietschi, Dider; Ardu, Stefano; Rossier-Gerber, Anne; Krejci, Ivo

2006-12-01

Fatigue resistance of post and cores is critical to the long term behavior of restored nonvital teeth. The purpose of this in vitro trial was to evaluate the influence of the post material's physical properties on the adaptation of adhesive post and core restorations after cyclic mechanical loading. Composite post and cores were made on endodontically treated deciduous bovine teeth using 3 anisotropic posts (made of carbon, quartz, or quartz-and-carbon fibers) and 3 isotropic posts (zirconium, stainless steel, titanium). Specimens were submitted to 3 successive loading phases--250,000 cycles at 50 N, 250,000 at 75 N, and 500,000 at 100 N--at a rate of 1.5 Hz. Restoration adaptation was evaluated under SEM, before and during loading (margins) and after test completion (margins and internal interfaces). Six additional samples were fabricated for the characterization of interface micromorphology using confocal microscopy. Mechanical loading increased the proportion of marginal gaps in all groups; carbon fiber posts presented the lowest final gap proportion (7.11%) compared to other stiffer metal-ceramic or softer fiber posts (11.0% to 19.1%). For internal adaptation, proportions of debonding between dentin and core or cement varied from 21.69% (carbon post) to 47.37% (stainless steel post). Debonding at the post-cement interface occurred only with isotropic materials. Confocal microscopy observation revealed that gaps were generally associated with an incomplete hybrid layer and reduced resin tags. Regardless of their rigidity, metal and ceramic isotropic posts proved less effective than fiber posts at stabilizing the post and core structure in the absence of the ferrule effect, due to the development of more interfacial defects with either composite or dentin.

Synthesis, Characterization and Drug Loading of Multiresponsive p[NIPAm-co-PEGMA] (core/p[NIPAm-co-AAc] (Shell Nanogels with Monodisperse Size Distributions

Directory of Open Access Journals (Sweden)

Rajesh Raju

2018-03-01

Full Text Available We report the synthesis and properties of temperature- and pH-responsive p([NIPAm-co-PEGMA] (core/[NIPAm-co-AAc] (shell nanogels with narrow size distributions, tunable sizes and increased drug loading efficiencies. The core-shell nanogels were synthesized using an optimized two-stage seeded polymerization methodology. The core-shell nanogels show a narrow size distribution and controllable physico-chemical properties. The hydrodynamic sizes, charge distributions, temperature-induced volume phase transition behaviors, pH-responsive behaviors and drug loading capabilities of the core-shell nanogels were investigated using transmission electron microscopy, zeta potential measurements, dynamic light scattering and UV-Vis spectroscopy. The size of the core-shell nanogels was controlled by polymerizing NIPAm with crosslinker poly(ethylene glycol dimethacrylate (PEGDMA of different molecular weights (Mn-200, 400, 550 and 750 g/mol during the core synthesis. It was found that the swelling/deswelling kinetics of the nanogels was sharp and reversible; with its volume phase transition temperature in the range of 40–42 °C. Furthermore, the nanogels loaded with l-3,4-dihydroxyphenylalanine (L-DOPA, using a modified breathing-in mechanism, showed high loading and encapsulation efficiencies, providing potential possibilities of such nanogels for biomedical applications.

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

International Nuclear Information System (INIS)

Joppen, F.

2000-01-01

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

Mixed PWR core loadings with inert matrix Pu-fuel assemblies

International Nuclear Information System (INIS)

Stanculescu, A.; Kasemeyer, U.; Paratte, J.-M.; Chawla, R.

1999-01-01

The most efficient way to enhance plutonium consumption in light water reactors is to eliminate the production of plutonium all together. This requirement leads to fuel concepts in which the uranium is replaced by an inert matrix. At PSI, studies have focused on employing ZrO 2 as inert matrix. Adding a burnable poison to such a fuel proves to be necessary. As a result of scoping studies, Er 2 O 3 was identified as the most suitable burnable poison material. The results of whole-core three-dimensional neutronics analyses indicated, for a present-day 1000 MW e pressurised water reactor, the feasibility of an asymptotic equilibrium four-batch cycle fuelled solely with the proposed PuO 2 -Er 2 O 3 -ZrO 2 inert matrix fuel (IMF). The present paper presents the results of more recent investigations related to 'real-life' situations, which call for transition configurations in which mixed IMF and UO 2 assembly loadings must be considered. To determine the influence of the introduction of IMF assemblies on the characteristics of a UO 2 -fuelled core, three-dimensional full-core calculations have been performed for a present-day 1000 MW e PWR containing up to 12 optimised IMF assemblies. (author)

A High-Order Theory for the Analysis of Circular Cylindrical Composite Sandwich Shells with Transversely Compliant Core Subjected to External Loads

DEFF Research Database (Denmark)

Rahmani, Omid; Khalili, S.M.R.; Thomsen, Ole Thybo

2012-01-01

A new model based on the high order sandwich panel theory is proposed to study the effect of external loads on the free vibration of circular cylindrical composite sandwich shells with transversely compliant core, including also the calculation of the buckling loads. In the present model......, which is based on a 3D elasticity solution for the core material, can be used as a benchmark in future studies of the free vibration and buckling of circular cylindrical composite sandwich shells with a transversely compliant core....

Hollow-Core FRP–Concrete–Steel Bridge Columns under Torsional Loading

Directory of Open Access Journals (Sweden)

Sujith Anumolu

2017-11-01

Full Text Available This paper presents the behavior of hollow-core fiber-reinforced polymer–concrete–steel (HC-FCS columns under cyclic torsional loading combined with constant axial load. The HC-FCS consists of an outer fiber-reinforced polymer (FRP tube and an inner steel tube, with a concrete shell sandwiched between the two tubes. The FRP tube was stopped at the surface of the footing, and provided confinement to the concrete shell from the outer direction. The steel tube was embedded into the footing to a length of 1.8 times the diameter of the steel tube. The longitudinal and transversal reinforcements of the column were provided by the steel tube only. A large-scale HC-FCS column with a diameter of 24 in. (610 mm and applied load height of 96 in. (2438 mm with an aspect ratio of four was investigated during this study. The study revealed that the torsional behavior of the HC-FCS column mainly depended on the stiffness of the steel tube and the interactions among the column components (concrete shell, steel tube, and FRP tube. A brief comparison of torsional behavior was made between the conventional reinforced concrete columns and the HC-FCS column. The comparison illustrated that both column types showed high initial stiffness under torsional loading. However, the HC-FCS column maintained the torsion strength until a high twist angle, while the conventional reinforced concrete column did not.

Core physics analysis in support of the FNR HEU-LEU demonstration experiment

Energy Technology Data Exchange (ETDEWEB)

Losey, David C; Brown, Forrest B; Martin, William R; Lee, John C [Department of Nuclear Engineering, University of Michigan (United States)

1983-08-01

A core neutronics analysis has been undertaken to assess the impact of low-enrichment fuel on the performance and utilization of the FNR As part of this analytic effort a computer code system has been assembled which will be of general use in analyzing research reactors with MTR-type fuel. The code system has been extensively tested and verified in calculations for the present high enrichment core. The analysis presented here compares the high-and-low enrichment fuels in batch and equilibrium core configurations which model the actual FNR operating conditions. The two fuels are compared for cycle length, fuel burnup, and flux and power distributions, as well as for the reactivity effects which are important in assessing the impact of LEU fuel on reactor shutdown margin. (author)

Core physics analysis in support of the FNR HEU-LEU demonstration experiment

International Nuclear Information System (INIS)

Losey, David C.; Brown, Forrest B.; Martin, William R.; Lee, John C.

1983-01-01

A core neutronics analysis has been undertaken to assess the impact of low-enrichment fuel on the performance and utilization of the FNR As part of this analytic effort a computer code system has been assembled which will be of general use in analyzing research reactors with MTR-type fuel. The code system has been extensively tested and verified in calculations for the present high enrichment core. The analysis presented here compares the high-and-low enrichment fuels in batch and equilibrium core configurations which model the actual FNR operating conditions. The two fuels are compared for cycle length, fuel burnup, and flux and power distributions, as well as for the reactivity effects which are important in assessing the impact of LEU fuel on reactor shutdown margin. (author)

Corrosion of cermet cores of fuel plates for nuclear research reactor

International Nuclear Information System (INIS)

Durazzo, M.; Ramanathan, L.V.

1984-01-01

Materials Testing Reactor (MTR) type fuel plates containing U 3 O 8 -Al cores and clad with Al are used in various research reactor. Preliminary investigations, where in the cladding of samples was drilled to simulate conditions of rupture due to pitting attack, revealed that considerable quantities of H 2 was evolved upon exposure of the core to water. The corrosion of cermets cores of different densities was characterized as a function of H 2 evolution that revealed 3 stages. A first stage consisting of an incubation period followed by initiation of H 2 evolution, a second stage with a constant rate of H 2 evolution and a third stage with a low rate of H 2 evolution. All 3 stages were found to vary as a function of cermet density and water temperature. (Author) [pt

The Intermediate Set and Limiting Superdi erential for Coalition Games: Between the Core and the Weber Set

Czech Academy of Sciences Publication Activity Database

Adam, Lukáš; Kroupa, T.

2017-01-01

Roč. 46, č. 4 (2017), s. 891-918 ISSN 0020-7276 R&D Projects: GA ČR GA15-00735S Institutional support: RVO:67985556 Keywords : coalition game * limiting superdi erential * intermediate set * core * Weber set Subject RIV: BA - General Mathematics OBOR OECD: Statistics and probability Impact factor: 0.713, year: 2016 http://library.utia.cas.cz/separaty/2016/MTR/adam-0467365.pdf

FRG-1: new millenium - new compact core

International Nuclear Information System (INIS)

Schreiner, P.; Knop, W.

2001-01-01

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

The Response of Clamped Shallow Sandwich Arches with Metallic Foam Cores to Projectile Impact Loading

Directory of Open Access Journals (Sweden)

Yanping Fan

Full Text Available Abstract The dynamic response and energy absorption capabilities of clamped shallow sandwich arches with aluminum foam core were numerically investigated by impacting the arches at mid-span with metallic foam projectiles. The typical deformation modes, deflection response, and core compression of sandwich arches obtained from the tests were used to validate the computation model. The resistance to impact loading was quantified by the permanent transverse deflection at mid-span of the arches as a function of projectile momentum. The sandwich arches have a higher shock resistance than the monolithic arches of equal mass, and shock resistance could be significantly enhanced by optimizing geometrical configurations. Meanwhile, decreasing the face-sheet thickness and curvature radius could enhance the energy absorption capability of the sandwich arches. Finite element calculations indicated that the ratio of loading time to structural response time ranged from 0.1 to 0.4. The projectile momentum, which was solely used to quantify the structural response of sandwich arches, was insufficient. These findings could provide guidance in conducting further theoretical studies and producing the optimal design of metallic sandwich structures subjected to impact loading.

Neutronic design of a 22 MW MTR type nuclear research reactor

International Nuclear Information System (INIS)

Khamis, I.; Khattab, K.; Soleman, I.; Ghazi, N.

2008-01-01

The neutronic design calculations of a 22 MW MTR type nuclear research reactor are conducted in this project. This reactor type is selected by the Arab Atomic Energy Commission in a cooperated project. The design calculations are conducted in two methods: The deterministic method, solving the neutron transport and diffusion equations using the WIMSD4 and the CITATION codes, and the probabilistic method using the MCNP code. Good agreements are noticed between the results of the multiplication factor and the neutron flux distribution which prove the accuracy of our models using the two methods. (authors)

Neutronic design of a 22 MW MTR type nuclear research reactor

International Nuclear Information System (INIS)

Khamis, I.; Khattab, K.; Soleman, I.; Ghazi, N.

2006-12-01

The neutronic design calculations of a 22 MW MTR type nuclear research reactor are conducted in this project. This reactor type is selected by the Arab Atomic Energy Commission in a cooperated project. The design calculations are conducted in two methods: The deterministic method, solving the neutron transport and diffusion equations using the WIMSD4 and the CITATION codes, and the probabilistic method using the MCNP code. Good agreements are noticed between the results of the multiplication factor and the neutron flux distribution which prove the accuracy of our models using the two methods. (author)

Effects of hydrophobic drug-polyesteric core interactions on drug loading and release properties of poly(ethylene glycol)-polyester-poly(ethylene glycol) triblock core-shell nanoparticles

International Nuclear Information System (INIS)

Khoee, Sepideh; Hassanzadeh, Salman; Goliaie, Bahram

2007-01-01

BAB amphiphilic triblock copolymers consisting of poly(ethylene glycol) (B) (PEG) as the hydrophilic segment and different polyesters (A) as the hydrophobic block were prepared by a polycondensation reaction as efficient model core-shell nanoparticles to assay the effect of interactions between the hydrophobic drug and the polyesteric core in terms of drug loading content and release profile. PEG-poly(hexylene adipate)-PEG (PEG-PHA-PEG) and PEG-poly(butylene adipate)-PEG (PEG-PBA-PEG) to PEG-poly(ethylene adipate)-PEG (PEG-PEA-PEG) core-shell type nanoparticles entrapping quercetin (an anticarcinogenic, allergy inhibitor and antibacterial agent), were prepared by a nanoprecipitation method and characterized by dynamic light scattering (DLS), transmission electron microscopy (TEM) and x-ray diffraction (XRD) techniques. It was found that the obtained nanoparticles showed a smooth surface and spherical shape with controllable sizes in the range of 64-74 nm, while drug loading varied from 7.24% to 19% depending on the copolymer composition and the preparation conditions. The in vitro release behaviour exhibited a sustained release and was affected by the polymer-drug interactions. UV studies revealed the presence of hydrogen bonding as the main existing interaction between quercetin and polyesters in the nanosphere cores

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Methotrexate diethyl ester-loaded lipid-core nanocapsules in aqueous solution increased antineoplastic effects in resistant breast cancer cell line.

Science.gov (United States)

Yurgel, Virginia C; Oliveira, Catiuscia P; Begnini, Karine R; Schultze, Eduarda; Thurow, Helena S; Leon, Priscila M M; Dellagostin, Odir A; Campos, Vinicius F; Beck, Ruy C R; Guterres, Silvia S; Collares, Tiago; Pohlmann, Adriana R; Seixas, Fabiana K

2014-01-01

Breast cancer is the most frequent cancer affecting women. Methotrexate (MTX) is an antimetabolic drug that remains important in the treatment of breast cancer. Its efficacy is compromised by resistance in cancer cells that occurs through a variety of mechanisms. This study evaluated apoptotic cell death and cell cycle arrest induced by an MTX derivative (MTX diethyl ester [MTX(OEt)2]) and MTX(OEt)2-loaded lipid-core nanocapsules in two MTX-resistant breast adenocarcinoma cell lines, MCF-7 and MDA-MB-231. The formulations prepared presented adequate granulometric profile. The treatment responses were evaluated through flow cytometry. Relying on the mechanism of resistance, we observed different responses between cell lines. For MCF-7 cells, MTX(OEt)2 solution and MTX(OEt)2-loaded lipid-core nanocapsules presented significantly higher apoptotic rates than untreated cells and cells incubated with unloaded lipid-core nanocapsules. For MDA-MB-231 cells, MTX(OEt)2-loaded lipid-core nanocapsules were significantly more efficient in inducing apoptosis than the solution of the free drug. S-phase cell cycle arrest was induced only by MTX(OEt)2 solution. The drug nanoencapsulation improved apoptosis induction for the cell line that presents MTX resistance by lack of transport receptors.

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Immobilisation of MTR waste in cement (product evaluation)

International Nuclear Information System (INIS)

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

1988-01-01

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

The THMIS-MTR observation of a active region filament

Science.gov (United States)

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

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

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

Directory of Open Access Journals (Sweden)

Dan eCoursolle

2012-02-01

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

An automated optimization of core fuel loading pattern for pressurized water reactors

International Nuclear Information System (INIS)

Chen Renji

1988-11-01

An optimum method was adopted to search for an optimum fuel loading pattern in pressurized water reactors. A radial power peak factor was chosen as the objective function of the optimum loading. The direct search method with shuffling rules is used to find optimum solution. The search for an optimum loading pattern with the smallest radial power peak by exchanging fuel assemblies was made. The search process is divided into two steps. In the first step fresh fuels or high reactivity fuels are arranged which are placed in core interior to have a reasonable fuel loading pattern. To further reduce the radial power peak factor, the second step will be necessary to rearrange the exposed lower reactivity fuel around the assemblies which has the radial power peak. In optimum process 1.5 group coarse mesh diffusion theory or two group nodal Green function diffusion theory is utilized to calculate the two dimensional power distribution after each shuffle. Also, above two methods are combinatively utilized to calculate the state quantity. It is not only true to save CPU time, but also can obtian exact results. Besides above function, the code MSOFEL is used to search critical boron concentration and to predict burn-up. The code has been written with FORTRAN-4. The optimum loading pattern was chosen for OCONEE and QINSHAN nuclear power plants as reference examples. The validity and feasibility of MSOFEL was demonstrated

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

Energy Technology Data Exchange (ETDEWEB)

Silva, Clayton Pereira da

2012-07-01

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

An intelligent nuclear reactor core controller for load following operations, using recurrent neural networks and fuzzy systems

International Nuclear Information System (INIS)

Boroushaki, M.; Ghofrani, M.B.; Lucas, C.; Yazdanpanah, M.J.

2003-01-01

In the last decade, the intelligent control community has paid great attention to the topic of intelligent control systems for nuclear plants (core, steam generator...). Papers mostly used approximate and simple mathematical SISO (single-input-single-output) model of nuclear plants for testing and/or tuning of the control systems. They also tried to generalize theses models to a real MIMO (multi-input-multi-output) plant, while nuclear plants are typically of complex nonlinear and multivariable nature with high interactions between their state variables and therefore, many of these proposed intelligent control systems are not appropriate for real cases. In this paper, we designed an on-line intelligent core controller for load following operations, based on a heuristic control algorithm, using a valid and updatable recurrent neural network (RNN). We have used an accurate 3-dimensional core calculation code to represent the real plant and to train the RNN. The results of simulation show that this intelligent controller can control the reactor core during load following operations, using optimum control rod groups manoeuvre and variable overlapping strategy. This methodology represents a simple and reliable procedure for controlling other complex nonlinear MIMO plants, and may improve the responses, comparing to other control systems

Development of silane grafted ZnO core shell nanoparticles loaded diglycidyl epoxy nanocomposites film for antimicrobial applications.

Science.gov (United States)

Suresh, S; Saravanan, P; Jayamoorthy, K; Ananda Kumar, S; Karthikeyan, S

2016-07-01

In this article a series of epoxy nanocomposites film were developed using amine functionalized (ZnO-APTES) core shell nanoparticles as the dispersed phase and a commercially available epoxy resin as the matrix phase. The functional group of the samples was characterized using FT-IR spectra. The most prominent peaks of epoxy resin were found in bare epoxy and in all the functionalized ZnO dispersed epoxy nanocomposites (ZnO-APTES-DGEBA). The XRD analysis of all the samples exhibits considerable shift in 2θ, intensity and d-spacing values but the best and optimum concentration is found to be 3% ZnO-APTES core shell nanoparticles loaded epoxy nanocomposites supported by FT-IR results. From TGA measurements, 100wt% residue is obtained in bare ZnO nanoparticles whereas in ZnO core shell nanoparticles grafted DGEBA residue percentages are 37, 41, 45, 46 and 52% for 0, 1, 3, 5 and 7% ZnO-APTES-DGEBA respectively, which is confirmed with ICP-OES analysis. From antimicrobial activity test, it was notable that antimicrobial activity of 7% ZnO-APTES core shell nanoparticles loaded epoxy nanocomposite film has best inhibition zone effect against all pathogens under study. Copyright © 2016 Elsevier B.V. All rights reserved.

MTR fuel plate qualification capabilities at SCK-CEN

International Nuclear Information System (INIS)

Koonen, E.; Jacquet, P.

2002-01-01

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

The sustained-release behavior and in vitro and in vivo transfection of pEGFP-loaded core-shell-structured chitosan-based composite particles

Science.gov (United States)

Wang, Yun; Lin, Fu-xing; Zhao, Yu; Wang, Mo-zhen; Ge, Xue-wu; Gong, Zheng-xing; Bao, Dan-dan; Gu, Yu-fang

2014-01-01

Novel submicron core-shell-structured chitosan-based composite particles encapsulated with enhanced green fluorescent protein plasmids (pEGFP) were prepared by complex coacervation method. The core was pEGFP-loaded thiolated N-alkylated chitosan (TACS) and the shell was pH- and temperature-responsive hydroxybutyl chitosan (HBC). pEGFP-loaded TACS-HBC composite particles were spherical, and had a mean diameter of approximately 120 nm, as measured by transmission electron microscopy and particle size analyzer. pEGFP showed sustained release in vitro for >15 days. Furthermore, in vitro transfection in human embryonic kidney 293T and human cervix epithelial cells, and in vivo transfection in mice skeletal muscle of loaded pEGFP, were investigated. Results showed that the expression of loaded pEGFP, both in vitro and in vivo, was slow but could be sustained over a long period. pEGFP expression in mice skeletal muscle was sustained for >60 days. This work indicates that these submicron core-shell-structured chitosan-based composite particles could potentially be used as a gene vector for in vivo controlled gene transfection. PMID:25364253

Fatigue behaviour of core-spun yarns containing filament by means of cyclic dynamic loading

Science.gov (United States)

Esin, S.; Osman, B.

2017-10-01

The behaviour of yarns under dynamic loading is important that leads to understand the growth characteristics which is exposed to repetitive loadings during usage of fabric made from these yarns. Fabric growth is undesirable property that originated from low resilience characteristics of fabric. In this study, the effects of the filament fineness and yarn linear density on fatigue behaviour of rigid-core spun yarns were determined. Cotton covered yarns containing different filament fineness of polyester (PET) draw textured yarns (DTY) (100d/36f, 100d/96f, 100d/144f, 100d/192f and 100d/333f) and yarn linear densities (37 tex, 30 tex, 25 tex and 21 tex) were manufactured by using a modified ring spinning system at the same spinning parameters. Repetitive loads were applied for 25 cycles at levels between 0.1 and 3 N. Dynamic modulus and dynamic strain of yarn samples were analyzed statistically. Results showed that filament fineness and yarn linear density have significance effect on dynamic modulus and dynamic strain after cyclic loading.

Automatic determination of BWR fuel loading patterns based on K.E. technique with core physics simulation

International Nuclear Information System (INIS)

Ikehara, T.; Tsuiki, M.; Takeshita, T.

1990-01-01

On the basis oof a computerized search method, a prototype for a fuel loading pattern expert system has been developed to support designers in core design for BWRs. The method was implemented by coupling rules and core physics simulators into an inference engine to establish an automated generate-and-test cycle. A search control mechanism, which prunes paths to be searched and selects appropriate rules through the interaction with the user, was also introduced to accomplish an effective search. The constraints in BWR core design are: (1) cycle length more than L, (2) core shutdown margin more than S, and (3) thermal margin more than T. Here L, S, and T are the specified minimum values. In this system, individual rules contain the manipulation to improve the core shutdown margin explicitly. Other items were taken into account only implicitly. Several applications to the test cases were carried out. It was found that the results were comparable with those obtained by human expert engineers. Broad applicability of the present method in the BWR core design domain was proved

Progress in the development of uranium silicide (U3Si2) fuel at BATAN

International Nuclear Information System (INIS)

Suripto, A.; Soentono, S.

1995-01-01

After successful fabrication of two full-size prototype fuel elements containing ∼3.0 gU/cm 3 in the form of U 3 Si 2 -Al dispersion now undergoing irradiation in the Reaktor Serba Guna G.A. Siwabessy (RSG-GAS) core since 1990, further development in U 3 Si 2 -A2 dispersion fuel element manufacturing has been pursued, whose progress in discussed in this paper, with a special attention on the use of much higher-loading aimed at obtaining a better understanding on the influence of higher-loading on fuel core and plate manufacturing and quality. At present, high-loading U 3 Si 2 -AI dispersion miniplates are being manufactured for preparing some mini-fuel elements to be test-irradiated in the new MTR in-pile loop of the RSG-GAS. (author)

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

International Nuclear Information System (INIS)

Abbate, P.

1990-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

Ultrasound-guided diagnostic breast biopsy methodology: retrospective comparison of the 8-gauge vacuum-assisted biopsy approach versus the spring-loaded 14-gauge core biopsy approach

Science.gov (United States)

2011-01-01

Background Ultrasound-guided diagnostic breast biopsy technology represents the current standard of care for the evaluation of indeterminate and suspicious lesions seen on diagnostic breast ultrasound. Yet, there remains much debate as to which particular method of ultrasound-guided diagnostic breast biopsy provides the most accurate and optimal diagnostic information. The aim of the current study was to compare and contrast the 8-gauge vacuum-assisted biopsy approach and the spring-loaded 14-gauge core biopsy approach. Methods A retrospective analysis was done of all ultrasound-guided diagnostic breast biopsy procedures performed by either the 8-gauge vacuum-assisted biopsy approach or the spring-loaded 14-gauge core biopsy approach by a single surgeon from July 2001 through June 2009. Results Among 1443 ultrasound-guided diagnostic breast biopsy procedures performed, 724 (50.2%) were by the 8-gauge vacuum-assisted biopsy technique and 719 (49.8%) were by the spring-loaded 14-gauge core biopsy technique. The total number of false negative cases (i.e., benign findings instead of invasive breast carcinoma) was significantly greater (P = 0.008) in the spring-loaded 14-gauge core biopsy group (8/681, 1.2%) as compared to in the 8-gauge vacuum-assisted biopsy group (0/652, 0%), with an overall false negative rate of 2.1% (8/386) for the spring-loaded 14-gauge core biopsy group as compared to 0% (0/148) for the 8-gauge vacuum-assisted biopsy group. Significantly more (P guided diagnostic breast biopsy procedure. Significantly more (P guided diagnostic breast biopsy procedure. Conclusions In appropriately selected cases, the 8-gauge vacuum-assisted biopsy approach appears to be advantageous to the spring-loaded 14-gauge core biopsy approach for providing the most accurate and optimal diagnostic information. PMID:21835024

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

International Nuclear Information System (INIS)

Smith, D.L.

1987-01-01

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

Four years Re-Use of low burned fuel assemblies from units 1 and 2 in core loadings of units 3 and 4 WWER-440 at Kozloduy NPP

International Nuclear Information System (INIS)

Stoyanova, I.; Antov, A.; Spasova, V.

2006-01-01

At the end of 2002 units 1 and 2 of Kozloduy NPP were shutdown before their design life time which left a large number of assemblies yet with a significant energy resources. A decision to load these assemblies into the cores of Units 3 and 4 during the next 4 cycles has been taken. In 2003, 43 assemblies from Unit 1 cycle 23 rd and 55 assemblies from Unit 2 cycle 24 th are loaded in the cores of units 3 and 4 respectively. In 2004, new 49 assemblies from Unit 1 cycle 23rd and new 55 assemblies from Unit 2 cycle 24th are loaded in the cores of units 3 and 4 respectively. In 2005, the next new 25 assemblies from Unit 1 cycle 23 rd and 66 assemblies from Unit 2 cycle 24th are loaded in the cores of units 3 and 4 respectively. In 2006, the next new 54 assemblies from Unit 1 cycle 23 rd and 52 assemblies from Unit 2 cycle 24 th + 2 assemblies from Unit 3 cycle 19th are loaded in the cores of Units 3 and 4 respectively. The SPPSHB computer code system is used for development and safety assessment of the fuel loading patterns of Units 3 and 4 at Kozloduy NPP with low burned assemblies from units 1 and 2 (Authors)

Three years re-use of low burned fuel assemblies from units 1 and 2 in core loading of units 3 and 4 WWER-440 at Kozloduy NPP

International Nuclear Information System (INIS)

Stoyanova, I.; Antov, A.; Spasova, V.

2006-01-01

At the end of 2002 Units 1 and 2 of Kozloduy NPP were shutdown before the end of their designed lifetime. This left a large number of assemblies yet with a significant energy resource. In 2003, 43 assemblies from Unit 1 and 55 assemblies from Unit 2 are loaded in the cores of Units 3 and 4 respectively. In 2004, new 49 assemblies from Unit 1 and new 55 assemblies from Unit 2 are loaded in the cores of Units 3 and 4 respectively. In 2005, the next new 25 assemblies from Unit 1 and 66 assemblies from Unit 2 are loaded in the cores of Units 3 and 4 respectively. The realized core loadings of Units 3 and 4 are illustrated. The large number of low burned assemblies after the shut down of Units 1 and 2 and their utilization during 4 consecutive cycles in the cores of Units 3 and 4 both do not permit to obtain in the last 3 cycles the typical mean burn up 35-36 MWd/kg U of 3.6% discharged assemblies

Time-dependent fracture probability of bilayer, lithium-disilicate-based glass-ceramic molar crowns as a function of core/veneer thickness ratio and load orientation

Science.gov (United States)

Anusavice, Kenneth J.; Jadaan, Osama M.; Esquivel–Upshaw, Josephine

2013-01-01

Recent reports on bilayer ceramic crown prostheses suggest that fractures of the veneering ceramic represent the most common reason for prosthesis failure. Objective The aims of this study were to test the hypotheses that: (1) an increase in core ceramic/veneer ceramic thickness ratio for a crown thickness of 1.6 mm reduces the time-dependent fracture probability (Pf) of bilayer crowns with a lithium-disilicate-based glass-ceramic core, and (2) oblique loading, within the central fossa, increases Pf for 1.6-mm-thick crowns compared with vertical loading. Materials and methods Time-dependent fracture probabilities were calculated for 1.6-mm-thick, veneered lithium-disilicate-based glass-ceramic molar crowns as a function of core/veneer thickness ratio and load orientation in the central fossa area. Time-dependent fracture probability analyses were computed by CARES/Life software and finite element analysis, using dynamic fatigue strength data for monolithic discs of a lithium-disilicate glass-ceramic core (Empress 2), and ceramic veneer (Empress 2 Veneer Ceramic). Results Predicted fracture probabilities (Pf) for centrally-loaded 1,6-mm-thick bilayer crowns over periods of 1, 5, and 10 years are 1.2%, 2.7%, and 3.5%, respectively, for a core/veneer thickness ratio of 1.0 (0.8 mm/0.8 mm), and 2.5%, 5.1%, and 7.0%, respectively, for a core/veneer thickness ratio of 0.33 (0.4 mm/1.2 mm). Conclusion CARES/Life results support the proposed crown design and load orientation hypotheses. Significance The application of dynamic fatigue data, finite element stress analysis, and CARES/Life analysis represent an optimal approach to optimize fixed dental prosthesis designs produced from dental ceramics and to predict time-dependent fracture probabilities of ceramic-based fixed dental prostheses that can minimize the risk for clinical failures. PMID:24060349

Time-dependent fracture probability of bilayer, lithium-disilicate-based, glass-ceramic, molar crowns as a function of core/veneer thickness ratio and load orientation.

Science.gov (United States)

Anusavice, Kenneth J; Jadaan, Osama M; Esquivel-Upshaw, Josephine F

2013-11-01

Recent reports on bilayer ceramic crown prostheses suggest that fractures of the veneering ceramic represent the most common reason for prosthesis failure. The aims of this study were to test the hypotheses that: (1) an increase in core ceramic/veneer ceramic thickness ratio for a crown thickness of 1.6mm reduces the time-dependent fracture probability (Pf) of bilayer crowns with a lithium-disilicate-based glass-ceramic core, and (2) oblique loading, within the central fossa, increases Pf for 1.6-mm-thick crowns compared with vertical loading. Time-dependent fracture probabilities were calculated for 1.6-mm-thick, veneered lithium-disilicate-based glass-ceramic molar crowns as a function of core/veneer thickness ratio and load orientation in the central fossa area. Time-dependent fracture probability analyses were computed by CARES/Life software and finite element analysis, using dynamic fatigue strength data for monolithic discs of a lithium-disilicate glass-ceramic core (Empress 2), and ceramic veneer (Empress 2 Veneer Ceramic). Predicted fracture probabilities (Pf) for centrally loaded 1.6-mm-thick bilayer crowns over periods of 1, 5, and 10 years are 1.2%, 2.7%, and 3.5%, respectively, for a core/veneer thickness ratio of 1.0 (0.8mm/0.8mm), and 2.5%, 5.1%, and 7.0%, respectively, for a core/veneer thickness ratio of 0.33 (0.4mm/1.2mm). CARES/Life results support the proposed crown design and load orientation hypotheses. The application of dynamic fatigue data, finite element stress analysis, and CARES/Life analysis represent an optimal approach to optimize fixed dental prosthesis designs produced from dental ceramics and to predict time-dependent fracture probabilities of ceramic-based fixed dental prostheses that can minimize the risk for clinical failures. Copyright © 2013 Academy of Dental Materials. All rights reserved.

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

International Nuclear Information System (INIS)

Silva, Clayton Pereira da

2012-01-01

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

On the optimization of thorium bundle distribution in the initial core loading for a PHWR

International Nuclear Information System (INIS)

Mishra, S.; Ray, S.; Kumar, A.N.; Modak, R.S.; Ganesan, S.

2009-01-01

If the initial core of Indian 220 MWe PHWR is loaded with all fresh Natural Uranium fuel, only about 70% Full Power can be drawn in the initial operation due to large power peaking. It is possible to load few tens of Th bundles at selected locations to get nearly full power without violating any safety limits. Finding the best possible locations is, however, a fairly complex and massive combinatorial optimization problem. Here, optimum solutions are obtained by a latest evolutionary algorithm called EDA implemented on the EKA built at Computational Research Laboratories (CRL) in Pune. The effect of varying the number of Th bundles on results is discussed. (author)

Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

International Nuclear Information System (INIS)

Park, Donghwan; Kim, Youngae; Kim, Sungmin

2016-01-01

CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future

Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

Energy Technology Data Exchange (ETDEWEB)

Park, Donghwan; Kim, Youngae; Kim, Sungmin [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2016-10-15

CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future.

Uncertainties assessment for safety margins evaluation in MTR reactors core thermal-hydraulic design

International Nuclear Information System (INIS)

Gimenez, M.; Schlamp, M.; Vertullo, A.

2002-01-01

This report contains a bibliographic review and a critical analysis of different methodologies used for uncertainty evaluation in research reactors core safety related parameters. Different parameters where uncertainties are considered are also presented and discussed, as well as their intrinsic nature regarding the way their uncertainty combination must be done. Finally a combined statistical method with direct propagation of uncertainties and a set of basic parameters as wall and DNB temperatures, CHF, PRD and their respective ratios where uncertainties should be considered is proposed. (author)

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

International Nuclear Information System (INIS)

Yuan, Liq-Ji

1987-01-01

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

Preparations to load, transport, receive, and store the damaged TMI-2 [Three Mile Island] reactor core

International Nuclear Information System (INIS)

Reno, H.W.; Schmitt, R.C.; Quinn, G.J.; Ayers, A.L. Jr.; Lilburn, B.J. Jr.; Uhl, D.L.

1986-03-01

The March 1979 incident at the Three Mile Island Nuclear Power Station (TMI) which damaged the core of the Unit 2 reactor resulted in numerous scientific and technical challenges. Some of those challenges involve removing, packaging, and transporting the core debris to the Idaho National Engineering Laboratory (INEL) for storage, examination, and preparation for final disposal. This paper highlights preparations for transporting the core debris from TMI to INEL and receiving and storing that material at INEL. Issues discussed include interfacing of equipment and facilities at TMI, loading operations, transportation activities using a newly designed cask, receiving and storing operations at INEL, and criticality control during storage. Key to the transportation effort was designing, testing, fabricating, and licensing two rail casks which individually provide double containment of the damaged fuel. 27 figs

MTR radiological database for SRS spent nuclear fuel facilities

International Nuclear Information System (INIS)

Blanchard, A.

2000-01-01

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

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

International Nuclear Information System (INIS)

Smith, D.L.

1985-12-01

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

LIGHT-WEIGHT LOAD-BEARING STRUCTURE

DEFF Research Database (Denmark)

2009-01-01

The invention relates to a light-weight load-bearing structure (1) with optimized compression zone (2), where along one or more compression zones (2) in the structure (1) to be cast a core (3) of strong concrete is provided, which core (3) is surrounded by concrete of less strength (4) compared...... to the core (3) of strong concrete. The invention also relates to a method of casting of light-weight load-bearing structures (1) with optimized compression zone (2) where one or more channels, grooves, ducts, pipes and/or hoses (5) formed in the load-bearing structure (1) serves as moulds for moulding one...... or more cores (3) of strong concrete in the light-weight load-bearing structure (1)....

Fast reactor core concepts to improve transmutation efficiency

International Nuclear Information System (INIS)

Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

2015-01-01

Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate

Feasibility of fully ceramic microencapsulated (FCM) replacement fuel assembly for OPR-1000 core fully loaded with FCM fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Lee, W.J.; Lee, K.H.; Kwon, H.; Chun, J.H.; Kim, Y.M. [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of); Venneri, F. [Ultra Safe Nuclear Corp., Los Alamos, NM (United States)

2014-07-01

The feasibility of replacing conventional UO{sub 2} fuel assemblies (FAs) of light water reactors with accident-tolerant fully ceramic microencapsulated (FCM) FAs has been explored referencing OPR-1000, 1000MW{sub e} PWR. An optimum FCM FA design, 16x16 FCM FA with Silicon Carbide-coated Zircaloy cladding, was selected based on core-level scoping analysis for five FCM FA design candidates screened from FA-level study. For the selected FCM FA design, detailed core following analysis from initial to equilibrium cores, initially fully loaded with the FCM FAs, was carried out to quantify core physics parameters. Using these parameters, the core thermal-hydraulics and coated fuel particle performance of the FCM core was assessed, and the safety margin and accident-tolerance of the FCM core was evaluated for limiting design- and beyond design-basis-accidents. From the study, it has been demonstrated that the FCM fuel is a viable option in replacing the OPR-1000 core with enhanced safety and accident tolerance while maintaining the core neutronics, thermal-hydraulics and mechanical compatibility. (author)

Efficient approach for simulating response of multi-body structure in reactor core subjected to seismic loading

International Nuclear Information System (INIS)

Zhang Hongkun; Cen Song; Wang Haitao; Cheng Huanyu

2012-01-01

An efficient 3D approach is proposed for simulating the complicated responses of the multi-body structure in reactor core under seismic loading. By utilizing the rigid-body and connector functions of the software Abaqus, the multi-body structure of the reactor core is simplified as a mass-point system interlinked by spring-dashpot connectors. And reasonable schemes are used for determining various connector coefficients. Furthermore, a scripting program is also complied for the 3D parametric modeling. Numerical examples show that, the proposed method can not only produce the results which satisfy the engineering requirements, but also improve the computational efficiency more than 100 times. (authors)

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

International Nuclear Information System (INIS)

Lashkari, A.; Khalafi, H.; Kazeminejad, H.

2013-01-01

Highlights: ► Kinetic parameters of Tehran research reactor mixed-core have been calculated. ► Burn-up effect on TRR kinetics parameters has been studied. ► Replacement of LEU-CFE with HEU-CFE in the TRR core has been investigated. ► Results of each mixed core were compared to the reference core. ► Calculation of kinetic parameters are necessary for reactivity and power excursion transient analysis. - Abstract: In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR P C package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change

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

International Nuclear Information System (INIS)

Koll, J.H.

1987-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Silva, Clayton Pereira da

2012-07-01

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

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)

Calculation analysis of the neutronic experimental data coming from the NUR reactor start-up

International Nuclear Information System (INIS)

Madariaga, M.; Villarino, E.; Relloso, J.; Rubio, R

1991-01-01

NUR is a new MTR reactor located in Argelia which became critical in march 1989. It is loaded with a 19 plates LEU Fe. This paper contains: a) Reactivity measurements in the first cores technical information about the Fe and some other data necessary for performing cell and reactor calculations b) calculation comparisons with the measured values (2-D and 3-D calculations) with an statistical analysis of the data set from the control rod calibration. (orig.)

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

International Nuclear Information System (INIS)

Hey, Martin A.; Gomez Marlasca, Fernando

2003-01-01

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

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

Directory of Open Access Journals (Sweden)

Gatsa O.

2018-01-01

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

Experimental study of the mechanical behaviour of pin reinforced foam core sandwich materials under shear load

International Nuclear Information System (INIS)

Dimassi, M A; Brauner, C; Herrmann, A S

2016-01-01

Sandwich structures with a lightweight closed cell hard foam core have the potential to be used in primary structures of commercial aircrafts. Compared to honeycomb core sandwich, the closed cell foam core sandwich overcomes the issue of moisture take up and makes the manufacturing of low priced and highly integrated structures possible. However, lightweight foam core sandwich materials are prone to failure by localised external loads like low velocity impacts. Invisible cracks could grow in the foam core and threaten the integrity of the structure. In order to enhance the out-of-plane properties of foam core sandwich structures and to improve the damage tolerance (DT) dry fibre bundles are inserted in the foam core. The pins are infused with resin and co-cured with the dry fabric face sheets in an out-of-autoclave process. This study presents the results obtained from shear tests following DIN 53294-standard, on flat sandwich panels. All panels were manufactured with pin-reinforcement manufactured with the Tied Foam Core Technology (TFC) developed by Airbus. The effects of pin material (CFRP and GFRP) and pin volume fraction on the shear properties of the sandwich structure and the crack propagation were investigated and compared to a not pinned reference. It has been concluded that the pin volume fraction has a remarkable effect on the shear properties and damage tolerance of the observed structure. Increasing the pin volume fraction makes the effect of crack redirection more obvious and conserves the integrity of the structure after crack occurrence. (paper)

Development of MTR fuel plate with U-Al dispersion core constituents

International Nuclear Information System (INIS)

Bressiani, Jose Carlos

1979-01-01

This work is a contribution to the development of fuel plates for Research Nuclear Reaction Materials Test Reactors. The plates have the core constituted by dispersions of metallic uranium in aluminum. The main topics of this work are: 1) The preparation of uranium powder with particle sizes in the 53-105μm diameter range; 2) The mixture and cold-pressing of uranium and aluminum powders for different uranium concentrations; 3) The behavior of the dispersions in the roll milling conditions; 4) Blister, radiographic, metallographic and irradiation tests for quality control of the plates. The irradiation test was performed in the IEA-R1 swimming-pool reactor using a prototype with a dispersion of aluminum and natural uranium (45 w/o ), reaching an integrated neutron flux of 8.663 X 10 18 n/cm 2 , no visual changes being noticed after the completion of the experiment. The behavior of the uranium-aluminum reaction for dispersions with 45% w/o uranium also studied. X-ray diffraction experiments showed the formation of UAl 2 UAl 3 and UAl 4 , while energy dispersive analysis of X-rays(EDAX) demonstrated that the diffusion of aluminum in uranium is the mechanism responsible for that reaction. The activation energy for the U-Al reaction was determined by dilatometric experiments yielding 20.2 kcal/mol.The aluminum-uranium reaction reaches an end when extended to 96 h at 600 deg C, namely, when all the uranium is found in the UAl 4 composition. (author)

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

Paclitaxel loaded folic acid targeted nanoparticles of mixed lipid-shell and polymer-core: in vitro and in vivo evaluation.

Science.gov (United States)

Zhao, Peiqi; Wang, Hanjie; Yu, Man; Liao, Zhenyu; Wang, Xianhuo; Zhang, Fei; Ji, Wei; Wu, Bing; Han, Jinghua; Zhang, Haichang; Wang, Huaqing; Chang, Jin; Niu, Ruifang

2012-06-01

A functional drug carrier comprised of folic acid modified lipid-shell and polymer-core nanoparticles (FLPNPs) including poly(D,L-lactide-co-glycolide) (PLGA) core, PEGylated octadecyl-quaternized lysine modified chitosan (PEG-OQLCS) as lipid-shell, folic acid as targeting ligand and cholesterol was prepared and evaluated for targeted delivery of paclitaxel (PTX). Confocal microscopy analysis confirmed the coating of the lipid-shell on the polymer-core. Physicochemical characterizations of FLPNPs, such as particle size, zeta potential, morphology, encapsulation efficiency, and in vitro PTX release, were also evaluated. The internalization efficiency and targeting ability of FLPNPs were demonstrated by flow cytometry and confocal microscopy. PTX loaded FLPNPs showed a significantly higher cytotoxicity than the commercial PTX formulation (Taxol®). The intravenous administration of PTX encapsulated FLPNPs led to tumor regression and improvement of animal survival in a murine model, compared with that observed with Taxol® and biodistribution study showed that PTX concentration in tumor for PTX encapsulated FLPNPs was higher than other PTX formulations. Our data indicate that PTX loaded FLPNPs are a promising nano-sized drug formulation for cancer therapy. Copyright © 2012 Elsevier B.V. All rights reserved.

Analysis of Random-Loading HTR-PROTEUS Cores with Continuous Energy Monte Carlo Code Based on A Statistical Geometry Model

International Nuclear Information System (INIS)

Murata, Isao; Miyamaru, Hiroyuki

2008-01-01

Spherical elements have remarkable features in various applications in the nuclear engineering field. In 1990's, by the project of HTR-PROTEUS at PSI various pebble bed reactor experiments were conducted including cores with a lot of spherical fuel elements loaded randomly. In this study, criticality experiments of the random-loading HTR-PROTEUS cores were analyzed by MCNP-BALL, which could deal with a random arrangement of spherical fuel elements exactly with a statistical geometry model. As a result of analysis, the calculated effective multiplication factors were in fairly good agreement with the measurements within about 0.5%Δk/k. In comparison with other numerical analysis, our effective multiplication factors were between the experimental values and the VSOP calculations. To investigate the discrepancy of the effective multiplication factors between the experiments and calculations, sensitivity analyses were performed. As the result, the sensitivity of impurity boron concentration was fairly large. The reason of the present slight overestimation was not made clear at present. However, the presently existing difference was thought to be related to the impurity boron concentration, not to the modelling of the reactor and the used nuclear data. From the present study, it was confirmed that MCNP-BALL would have an advantage to conventional transport codes by comparing with their numerical results and the experimental values. As for the criticality experiment of PROTEUS, we would conclude that the two cores of Core 4.2 and 4.3 could be regarded as an equivalent experiment of a reference critical core, which was packed in the packing fraction of RLP. (authors)

Analysis of Random-Loading HTR-PROTEUS Cores with Continuous Energy Monte Carlo Code Based on A Statistical Geometry Model

Energy Technology Data Exchange (ETDEWEB)

Murata, Isao; Miyamaru, Hiroyuki [Division of Electrical, Electronic and Information Engineering, Osaka University, Yamada-oka 2-1, Suita, Osaka, 565-0871 (Japan)

2008-07-01

Spherical elements have remarkable features in various applications in the nuclear engineering field. In 1990's, by the project of HTR-PROTEUS at PSI various pebble bed reactor experiments were conducted including cores with a lot of spherical fuel elements loaded randomly. In this study, criticality experiments of the random-loading HTR-PROTEUS cores were analyzed by MCNP-BALL, which could deal with a random arrangement of spherical fuel elements exactly with a statistical geometry model. As a result of analysis, the calculated effective multiplication factors were in fairly good agreement with the measurements within about 0.5%DELTAk/k. In comparison with other numerical analysis, our effective multiplication factors were between the experimental values and the VSOP calculations. To investigate the discrepancy of the effective multiplication factors between the experiments and calculations, sensitivity analyses were performed. As the result, the sensitivity of impurity boron concentration was fairly large. The reason of the present slight overestimation was not made clear at present. However, the presently existing difference was thought to be related to the impurity boron concentration, not to the modelling of the reactor and the used nuclear data. From the present study, it was confirmed that MCNP-BALL would have an advantage to conventional transport codes by comparing with their numerical results and the experimental values. As for the criticality experiment of PROTEUS, we would conclude that the two cores of Core 4.2 and 4.3 could be regarded as an equivalent experiment of a reference critical core, which was packed in the packing fraction of RLP. (authors)

A fast-running core prediction model based on neural networks for load-following operations in a soluble boron-free reactor

Energy Technology Data Exchange (ETDEWEB)

Jang, Jin-wook [Korea Atomic Energy Research Institute, P.O. Box 105, Yusong, Daejon 305-600 (Korea, Republic of)], E-mail: Jinwook@kaeri.re.kr; Seong, Seung-Hwan [Korea Atomic Energy Research Institute, P.O. Box 105, Yusong, Daejon 305-600 (Korea, Republic of)], E-mail: shseong@kaeri.re.kr; Lee, Un-Chul [Department of Nuclear Engineering, Seoul National University, Shinlim-Dong, Gwanak-Gu, Seoul 151-742 (Korea, Republic of)

2007-09-15

A fast prediction model for load-following operations in a soluble boron-free reactor has been proposed, which can predict the core status when three or more control rod groups are moved at a time. This prediction model consists of two multilayer feedforward neural network models to retrieve the axial offset and the reactivity, and compensation models to compensate for the reactivity and axial offset arising from the xenon transient. The neural network training data were generated by taking various overlaps among the control rod groups into consideration for training the neural network models, and the accuracy of the constructed neural network models was verified. Validation results of predicting load following operations for a soluble boron-free reactor show that this model has a good capability to predict the positions of the control rods for sustaining the criticality of a core during load-following operations to ensure that the tolerable axial offset band is not exceeded and it can provide enough corresponding time for the operators to take the necessary actions to prevent a deviation from the tolerable operating band.

A fast-running core prediction model based on neural networks for load-following operations in a soluble boron-free reactor

International Nuclear Information System (INIS)

Jang, Jin-wook; Seong, Seung-Hwan; Lee, Un-Chul

2007-01-01

A fast prediction model for load-following operations in a soluble boron-free reactor has been proposed, which can predict the core status when three or more control rod groups are moved at a time. This prediction model consists of two multilayer feedforward neural network models to retrieve the axial offset and the reactivity, and compensation models to compensate for the reactivity and axial offset arising from the xenon transient. The neural network training data were generated by taking various overlaps among the control rod groups into consideration for training the neural network models, and the accuracy of the constructed neural network models was verified. Validation results of predicting load following operations for a soluble boron-free reactor show that this model has a good capability to predict the positions of the control rods for sustaining the criticality of a core during load-following operations to ensure that the tolerable axial offset band is not exceeded and it can provide enough corresponding time for the operators to take the necessary actions to prevent a deviation from the tolerable operating band

Code systems for effective and precise calculation of the basic neutron characteristics, core loading optimization, analysis and estimation of the operation regimes of WWER type reactors

International Nuclear Information System (INIS)

Apostolov, T.; Ivanov, K.; Prodanova, R.; Manolova, M.; Petrova, T.; Alekova, G.

1993-01-01

Two directions for investigations are suggested: 1) Analysis and evaluation of the real loading patterns and operational regimes for Kozloduy NPP WWER-440 and WWER-1000 in the frame of the recent safety criteria and nuclear power plant operating limits. 2) Development of modern code system for WWER type reactor core analysis with advanced features: new design and materials for fuel and control rods, increasing the fuel enrichment, using the integral and discrete burnable absorbers etc. The fuel technology design evolution maximizes the fuel utilization efficiency, improves operation performance and enhances safety margins. By the joint efforts of specialists from INRNE, Sofia (BG) and KAB, Berlin (GE), the codes NESSEL-IV-EC, PYTHIA and DERAB have been developed and verified. In the frame of the PHARE programme the joint project ASPERCA has been proposed intended for reactor physics calculations with PHYBER-WWER code for safety enhancement and operation reliability improvement. In-core fuel management benchmarks for 4 cycles of unit 2 (WWER-440) and 2 cycles of unit 5 (WWER-1000) have been performed. The coordination of burnable absorber design implementation, low leakage loadings usage, reloading enrichment increase and steel content reduction in the core have made the reactor core analysis more demanding and the definition of loading patterns - more difficult. This complexity requires routine use of three-dimensional fast accurate core model with extended and updated cross section libraries. To meet the needs of WWER advanced loading patterns and in-core fuel management improvements the HEXANES code systems is being developed and qualified. Some test calculations have been carried out by the HEXANES code system investigating the influence of Gd in the fuel on the main reactor physics parameters. For reevaluation of the core safety-related design limits forming the basis of licensing procedure, the code DYN3D/M2 is used. 16 refs., 3 figs. (author)

ROSA full-core and DNBR capabilities

International Nuclear Information System (INIS)

Gibcus, H.P.M.; Verhagen, F.C.M.; Wakker, P.H.

2013-01-01

The latest developments of the ROSA (Reloading Optimization by Simulated Annealing) code system with an emphasis on the first full-core version and the minimum DNBR (Departure from Nucleate Boiling Ratio) as a new optimization parameter are presented. Designing the core loading pattern of nuclear power plants is becoming a more and more complex task. This task becomes even more complicated if asymmetries in the core loading pattern arise, for instance due to damaged fuel assemblies. For over almost 2 decades ROSA, NRG's (Nuclear Research and consultancy Group) loading pattern optimization code system for PWRs, has proven to be a valuable tool to reactor operators in accomplishing this task. To improve the use of ROSA for designing asymmetric loading patterns, NRG has developed a full-core version of ROSA besides the original quarter-core version which requires rotational symmetry in the computational domain. The extension of ROSA with DNBR as an optimization parameter is part of ROSA's continuous development. (orig.)

ROSA full-core and DNBR capabilities

International Nuclear Information System (INIS)

Gibcus, H.P.M.; Verhagen, F.C.M.; Wakker, P.H.

2012-01-01

This paper presents the latest developments of the ROSA (Reloading Optimization by Simulated Annealing) code system with an emphasis on the first full-core version and the minimum DNBR (Departure from Nucleate Boiling Ratio) as a new optimization parameter. Designing the core loading pattern of nuclear power plants is becoming a more and more complex task. This task becomes even more complicated if asymmetries in the core loading pattern arise, for instance due to damaged fuel assemblies. For over almost two decades ROSA, NRG's (Nuclear Research and consultancy Group) loading pattern optimization code system for PWRs, has proven to be a valuable tool to reactor operators in accomplishing this task. To improve the use of ROSA for designing asymmetric loading patterns, NRG has developed a full-core version of ROSA besides the original quarter-core version which requires rotational symmetry in the computational domain. The extension of ROSA with DNBR as an optimization parameter is part of ROSA's continuous development. (orig.)

Improving the calculated core stability by the core nuclear design optimization

International Nuclear Information System (INIS)

Partanen, P.

1995-01-01

Three different equilibrium core loadings for TVO II reactor have been generated in order to improve the core stability properties at uprated power level. The reactor thermal power is assumed to be uprated from 2160 MW th to 2500 MW th , which moves the operating point after a rapid pump rundown where the core stability has been calculated from 1340 MW th and 3200 kg/s to 1675 MW th and 4000 kg/s. The core has been refuelled with ABB Atom Svea-100 -fuel, which has 3,64% w/o U-235 average enrichment in the highly enriched zone. PHOENIX lattice code has been used to provide the homogenized nuclear constants. POLCA4 static core simulator has been used for core loadings and cycle simulations and RAMONA-3B program for simulating the dynamic response to the disturbance for which the stability behaviour has been evaluated. The core decay ratio has been successfully reduced from 0,83 to 0,55 mainly by reducing the power peaking factors. (orig.) (7 figs., 1 tab.)

Fabrication, fabrication control and in-core follow up of 4 LEU leader fuel elements based on U3Si2 in RECH-1

International Nuclear Information System (INIS)

Chavez, J.C.; Barrera, M.; Olivares, L.; Lisboa, J.

1999-01-01

The RECH-1 MTR reactor has been converted from HEU to MEU (45% enrichment) and the decision to a LEU (20% enrichment) conversion was taken some years ago. This LEU conversion decision involved a local fuel development and fabrication based on U 3 Si 2 -Al dispersion fuel, and a fabrication qualification stage that resulted in four fuel elements fully complying with established fabrication standards for this type of fuel. This report-presents relevant points of these four leaders fuel elements fabrication, in particular a fuel plate core homogeneity control development. A summary of the intended in core follow-up studies for the leaders fuel elements is also presented here. (author)

Antibiotic-loaded, silver core-embedded mesoporous silica nanovehicles as a synergistic antibacterial agent for the treatment of drug-resistant infections.

Science.gov (United States)

Wang, Yao; Ding, Xiali; Chen, Yuan; Guo, Mingquan; Zhang, Yan; Guo, Xiaokui; Gu, Hongchen

2016-09-01

Drug-resistant bacterial infections have become one of the most serious risks in public health as they make the conventional antibiotics less efficient. There is an urgent need for developing new generations of antibacterial agents in this field. In this work, a nanoplatform of LEVO-loaded and silver core-embedded mesoporous silica nanovehicles (Ag@MSNs@LEVO) is demonstrated as a synergistic antibacterial agent for the treatment of drug-resistant infections both in vitro and in vivo. The combination of the inner Ag core and the loaded antibiotic drug in mesopores endows the single-particle nanoplatform with a synergistic effect on killing the drug-resistant bacteria. The nanoplatform of Ag@MSNs@LEVO exhibits superior antibacterial activity to LEVO-loaded MSNs (MSNs@LEVO) and silver core-embedded MSNs (Ag@MSNs) in vitro. In the in vivo acute peritonitis model, the infected drug-resistant Escherichia coli GN102 in peritoneal cavity of the mice is reduced by nearly three orders of magnitude and the aberrant pathological feature of spleen and peritoneum disappears after treatment with Ag@MSNs@LEVO. Importantly, this nanopaltform renders no obvious toxic side effect to the mice during the tested time. There is no doubt that this study strongly indicates a promising potential of Ag@MSNs@LEVO as a synergistic and safety therapy tool for the clinical drug-resistant infections. Copyright © 2016 Elsevier Ltd. All rights reserved.

The loadings and strength of nuclear power plant structures in core damage accidents

International Nuclear Information System (INIS)

Varpasuo, P.

1994-01-01

The reactor cavity of VVER-91 NPP is a thick-walled, cylindrical reinforced concrete structure. In case of molten core-water reaction during the severe reactor accident the load carrying capacity of the cavity structure is of interest against the short impulse type loading caused by the steam explosion phenomenon. The assumed size of the impulse was 20 kPa-s and the duration was 10 ms. This investigation was divided in several phases. First, the elastic response of the cavity was determined using the ABAQUS code. Next, the static response of the cavity was evaluated using elasto-plastic properties of reinforcement and concrete and also taking into account the cracking of the concrete. This analysis was done with the aid of ABAQUS/STANDARD and ANSYS codes and the obtained results agreed reasonably with each other. In order to obtain a qualitative picture of the behaviour of the structure under the impulse load a simplified single degree of freedorn model was developed. The hoop reinforcement of the cavity was taken as an elasto-plastic spring and the wall concrete acted as a mass. Using this model the suitable amount of hoop reinforcement was determined. In next phase, the dynamic analysis of the structure was attempted using elasto-plastic material properties and concrete cracking. (13 refs., 57 figs.)

Web-based Core Design System Development

International Nuclear Information System (INIS)

Moon, So Young; Kim, Hyung Jin; Yang, Sung Tae; Hong, Sun Kwan

2011-01-01

The selection of a loading pattern is one of core design processes in the operation of a nuclear power plant. A potential new loading pattern is identified by selecting fuels that to not exceed the major limiting factors of the design and that satisfy the core design conditions for employing fuel data from the existing loading pattern of the current operating cycle. The selection of a loading pattern is also related to the cycle plan of an operating nuclear power plant and must meet safety and economic requirements. In selecting an appropriate loading pattern, all aspects, such as input creation, code runs and result processes are processed as text forms manually by a designer, all of which may be subject to human error, such as syntax or running errors. Time-consuming results analysis and decision-making processes are the most significant inefficiencies to avoid. A web-based nuclear plant core design system was developed here to remedy the shortcomings of an existing core design system. The proposed system adopts the general methodology of OPR1000 (Korea Standard Nuclear Power Plants) and Westinghouse-type plants. Additionally, it offers a GUI (Graphic User Interface)-based core design environment with a user-friendly interface for operators. It reduces human errors related to design model creation, computation, final reload core model selection, final output confirmation, and result data validation and verification. Most significantly, it reduces the core design time by more than 75% compared to its predecessor

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

International Nuclear Information System (INIS)

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

1990-08-01

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

The effects of mechanical loading on tendons--an in vivo and in vitro model study.

Directory of Open Access Journals (Sweden)

Jianying Zhang

Full Text Available Mechanical loading constantly acts on tendons, and a better understanding of its effects on the tendons is essential to gain more insights into tendon patho-physiology. This study aims to investigate tendon mechanobiological responses through the use of mouse treadmill running as an in vivo model and mechanical stretching of tendon cells as an in vitro model. In the in vivo study, mice underwent moderate treadmill running (MTR and intensive treadmill running (ITR regimens. Treadmill running elevated the expression of mechanical growth factors (MGF and enhanced the proliferative potential of tendon stem cells (TSCs in both patellar and Achilles tendons. In both tendons, MTR upregulated tenocyte-related genes: collagen type I (Coll. I ∼10 fold and tenomodulin (∼3-4 fold, but did not affect non-tenocyte-related genes: LPL (adipocyte, Sox9 (chondrocyte, Runx2 and Osterix (both osteocyte. However, ITR upregulated both tenocyte (Coll. I ∼7-11 fold; tenomodulin ∼4-5 fold and non-tenocyte-related genes (∼3-8 fold. In the in vitro study, TSCs and tenocytes were stretched to 4% and 8% using a custom made mechanical loading system. Low mechanical stretching (4% of TSCs from both patellar and Achilles tendons increased the expression of only the tenocyte-related genes (Coll. I ∼5-6 fold; tenomodulin ∼6-13 fold, but high mechanical stretching (8% increased the expression of both tenocyte (Coll. I ∼28-50 fold; tenomodulin ∼14-48 fold and non-tenocyte-related genes (2-5-fold. However, in tenocytes, non-tenocyte related gene expression was not altered by the application of either low or high mechanical stretching. These findings indicate that appropriate mechanical loading could be beneficial to tendons because of their potential to induce anabolic changes in tendon cells. However, while excessive mechanical loading caused anabolic changes in tendons, it also induced differentiation of TSCs into non-tenocytes, which may lead to the development

Research on loading pattern optimization for VVER reactor

International Nuclear Information System (INIS)

Tran Viet Phu; Nguyen Thi Mai Huong; Nguyen Huu Tiep; Ta Duy Long; Tran Vinh Thanh; Tran Hoai Nam

2017-01-01

A study on fuel loading pattern optimization of a VVER reactor was performed. In this study, a core physics simulator was developed based on a multi-group diffusion theory for the use in the problem of fuel loading optimization of VVER reactors. The core simulator could handle the triangular meshes of the core and the computational speed is fast. Verification of the core simulator was confirmed against a benchmark problem of a VVER-1000 reactor. Several optimization methods such as DS, SA, TS and a combination of them were investigated and implemented in coupling with the core simulator. Calculations was performed for optimizing the fuel loading pattern of the core using these methods based on a benchmark core model in comparison with the reference core. Comparison among these methods have shown that a combination of SA+TS is the most effective for the problem of fuel loading pattern optimization. Advanced methods are being researched continuously. (author)

Conversion program in Sweden

Energy Technology Data Exchange (ETDEWEB)

Jonsson, E.B. [Studsvik Nuclear AB, Nykoeping (Sweden)

1997-08-01

The conversion of the Swedish 50 MW R2 reactor from HEU to LEU fuel has been successfully accomplished over a 16 cycles long process. The conversion started in January 1991 with the introduction of 6 LEU assemblies in the 8*8 core. The first all LEU core was loaded in March 1993 and physics measurements were performed for the final licensing reports. A total of 142 LEU fuel assemblies have been irradiated up until September 1994 without any fuel incident. The operating licence for the R2 reactor was renewed in mid 1994 taking into account new fuel type. The Swedish Nuclear Inspectorate (SKI) pointed out one crucial problem with the LEU operation, that the back end of the LEU fuel cycle has not yet been solved. For the HEU fuel Sweden had the reprocessing alternative. The country is now relying heavily on the success of the USDOEs Off Site Fuels Policy to take back the spent fuel from the research reactors. They have in the meantime increased their intermediate storage facilities. There is, however, a limit both in time and space for storage of MTR-type of assemblies in water. The penalty of the lower thermal neutron flux in LEU cores has been reduced by improvements of the new irradiation rigs and by fine tuning the core calculations. The Studsvik code package, CASMO-SIMULATE, widely used for ICFM in LWRs has been modified to suit the compact MTR type of core.

Core-shell nanocarriers with high paclitaxel loading for passive and active targeting

Science.gov (United States)

Jin, Zhu; Lv, Yaqi; Cao, Hui; Yao, Jing; Zhou, Jianping; He, Wei; Yin, Lifang

2016-06-01

Rapid blood clearance and premature burst release are inherent drawbacks of conventional nanoparticles, resulting in poor tumor selectivity. iRGD peptide is widely recognized as an efficient cell membrane penetration peptide homing to αVβ3 integrins. Herein, core-shell nanocapsules (NCs) and iRGD-modified NCs (iRGD-NCs) with high drug payload for paclitaxel (PTX) were prepared to enhance the antitumor activities of chemotherapy agents with poor water solubility. Improved in vitro and in vivo tumor targeting and penetration were observed with NCs and iRGD-NCs; the latter exhibited better antitumor activity because iRGD enhanced the accumulation and penetration of NCs in tumors. The NCs were cytocompatible, histocompatible, and non-toxic to other healthy tissues. The endocytosis of NCs was mediated by lipid rafts in an energy-dependent manner, leading to better cytotoxicity of PTX against cancer cells. In contrast with commercial product, PTX-loaded NCs (PTX-NCs) increased area under concentration-time curve (AUC) by about 4-fold, prolonged mean resident time (MRT) by more than 8-fold and reduced the elimination rate constant by greater than 68-fold. In conclusion, the present nanocarriers with high drug-loading capacity represent an efficient tumor-targeting drug delivery system with promising potential for cancer therapy.

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

Science.gov (United States)

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

2018-01-01

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

Core mechanics and configuration behavior of advanced LMFBR core restraint concepts

International Nuclear Information System (INIS)

Fox, J.N.; Wei, B.C.

1978-02-01

Core restraint systems in LMFBRs maintain control of core mechanics and configuration behavior. Core restraint design is complex due to the close spacing between adjacent components, flux and temperature gradients, and irradiation-induced material property effects. Since the core assemblies interact with each other and transmit loads directly to the core restraint structural members, the core assemblies themselves are an integral part of the core restraint system. This paper presents an assessment of several advanced core restraint system and core assembly concepts relative to the expected performance of currently accepted designs. A recommended order for the development of the advanced concepts is also presented

Investigation of cutaneous penetration properties of stearic acid loaded to dendritic core-multi-shell (CMS) nanocarriers.

Science.gov (United States)

Lohan, S B; Icken, N; Teutloff, C; Saeidpour, S; Bittl, R; Lademann, J; Fleige, E; Haag, R; Haag, S F; Meinke, M C

2016-03-30

Dendritic core-multi shell (CMS) particles are polymer based systems consisting of a dendritic polar polyglycerol polymer core surrounded by a two-layer shell of nonpolar C18 alkyl chains and hydrophilic polyethylene glycol. Belonging to nanotransport systems (NTS) they allow the transport and storage of molecules with different chemical characters. Their amphipihilic character CMS-NTS permits good solubility in aqueous and organic solutions. We showed by multifrequency electron paramagnetic resonance (EPR) spectroscopy that spin-labeled 5-doxyl stearic acid (5DSA) can be loaded into the CMS-NTS. Furthermore, the release of 5DSA from the carrier into the stratum corneum of porcine skin was monitored ex vivo by EPR spectroscopy. Additionally, the penetration of the CMS-NTS into the skin was analyzed by fluorescence microscopy using indocarbocyanine (ICC) covalently bound to the nanocarrier. Thereby, no transport into the viable skin was observed, whereas the CMS-NTS had penetrated into the hair follicles down to a depth of 340 μm ± 82 μm. Thus, it could be shown that the combined application of fluorescence microscopy and multi-frequency EPR spectroscopy can be an efficient tool for investigating the loading of spin labeled drugs to nanocarrier systems, drug release and penetration into the skin as well as the localization of the NTS in the skin. Copyright © 2016 Elsevier B.V. All rights reserved.

The sustained-release behavior and in vitro and in vivo transfection of pEGFP-loaded core-shell-structured chitosan-based composite particles

Directory of Open Access Journals (Sweden)

Wang Y

2014-10-01

Full Text Available Yun Wang,1 Fu-xing Lin,2 Yu Zhao,1 Mo-zhen Wang,2 Xue-wu Ge,2 Zheng-xing Gong,1 Dan-dan Bao,1 Yu-fang Gu1 1Department of Plastic Surgery, First Affiliated Hospital of Anhui Medical University, 2CAS Key Laboratory of Soft Matter Chemistry, Department of Polymer Science and Engineering, University of Science and Technology of China, Hefei, Anhui, People’s Republic of China Abstract: Novel submicron core-shell-structured chitosan-based composite particles ­encapsulated with enhanced green fluorescent protein plasmids (pEGFP were prepared by complex coacervation method. The core was pEGFP-loaded thiolated N-alkylated chitosan (TACS and the shell was pH- and temperature-responsive hydroxybutyl chitosan (HBC. pEGFP-loaded TACS-HBC composite particles were spherical, and had a mean diameter of approximately 120 nm, as measured by transmission electron microscopy and particle size analyzer. pEGFP showed sustained release in vitro for >15 days. Furthermore, in vitro transfection in human embryonic kidney 293T and human cervix epithelial cells, and in vivo transfection in mice skeletal muscle of loaded pEGFP, were investigated. Results showed that the expression of loaded pEGFP, both in vitro and in vivo, was slow but could be sustained over a long period. pEGFP expression in mice skeletal muscle was sustained for >60 days. This work indicates that these submicron core-shell-structured chitosan-based composite particles could potentially be used as a gene vector for in vivo controlled gene transfection. Keywords: gene therapy, gene transfection, hydroxybutyl chitosan, thiolated N-alkylated chitosan, pEGFP, complex coacervation

AP1000 core design with 50% MOX loading

International Nuclear Information System (INIS)

Fetterman, Robert J.

2009-01-01

The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core design and a mixed MOX/UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance

AP1000 core design with 50% MOX loading

International Nuclear Information System (INIS)

Fetterman, Robert J.

2008-01-01

The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core and a mixed MOX / UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

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

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.

MR spectroscopy (MRS) and magnetisation transfer imaging (MTI), lesion load and clinical scores in early relapsing remitting multiple sclerosis: a combined cross-sectional and longitudinal study

International Nuclear Information System (INIS)

Bellmann-Strobl, J.; Paul, F.; Aktas, O.; Zipp, F.; Stiepani, H.; Bohner, G.; Klingebiel, R.; Wuerfel, J.; Warmuth, C.; Wandinger, K.P.

2009-01-01

The purpose of this study was to correlate magnetic resonance imaging (MRI)-based lesion load assessment with clinical disability in early relapsing remitting multiple sclerosis (RRMS). Seventeen untreated patients (ten women, seven men; mean age 33.0±7.9 years) with the initial diagnosis of RRMS were included for cross-sectional as well as longitudinal (24 months) clinical and MRI-based assessment in comparison with age-matched healthy controls. Conventional MR sequences, MR spectroscopy (MRS) and magnetisation transfer imaging (MTI) were performed at 1.5 T. Lesion number and volume, MRS and MTI measurements for lesions and normal appearing white matter (NAWM) were correlated to clinical scores [Expanded Disability Status Scale (EDSS), Multiple Sclerosis Functional Composite (MSFC)] for monitoring disease course after treatment initiation (interferon β-1a). MTI and MRS detected changes [magnetisation transfer ratio (MTR), N-acetylaspartate (NAA)/creatine ratio] in NAWM over time. EDSS and lesional MTR increases correlated throughout the disease course. Average MTR of NAWM raised during the study (p<0.05) and correlated to the MSFC score (r=0.476, p<0.001). At study termination, NAA/creatine ratio of NAWM correlated to the MSFC score (p<0.05). MTI and MRS were useful for initial disease assessment in NAWM. MTI and MRS correlated with clinical scores, indicating potential for monitoring the disease course and gaining new insights into treatment-related effects. (orig.)

MR spectroscopy (MRS) and magnetisation transfer imaging (MTI), lesion load and clinical scores in early relapsing remitting multiple sclerosis: a combined cross-sectional and longitudinal study

Energy Technology Data Exchange (ETDEWEB)

Bellmann-Strobl, J.; Paul, F.; Aktas, O.; Zipp, F. [Charite - University Medicine Berlin and Max Delbrueck Center for Molecular Medicine, Cecilie Vogt Clinic for Neurology, Berlin (Germany); Stiepani, H.; Bohner, G.; Klingebiel, R. [Charite - University Medicine Berlin, Department of Neuroradiology, Berlin (Germany); Wuerfel, J. [Charite - University Medicine Berlin and Max Delbrueck Center for Molecular Medicine, Cecilie Vogt Clinic for Neurology, Berlin (Germany); University Schleswig-Holstein, Institute of Neuroradiology, Campus Luebeck, Kiel (Germany); Warmuth, C. [Charite - University Medicine Berlin, Department of Radiology, Berlin (Germany); Wandinger, K.P. [Charite - University Medicine Berlin, Department of Neurology, Berlin (Germany)

2009-08-15

The purpose of this study was to correlate magnetic resonance imaging (MRI)-based lesion load assessment with clinical disability in early relapsing remitting multiple sclerosis (RRMS). Seventeen untreated patients (ten women, seven men; mean age 33.0{+-}7.9 years) with the initial diagnosis of RRMS were included for cross-sectional as well as longitudinal (24 months) clinical and MRI-based assessment in comparison with age-matched healthy controls. Conventional MR sequences, MR spectroscopy (MRS) and magnetisation transfer imaging (MTI) were performed at 1.5 T. Lesion number and volume, MRS and MTI measurements for lesions and normal appearing white matter (NAWM) were correlated to clinical scores [Expanded Disability Status Scale (EDSS), Multiple Sclerosis Functional Composite (MSFC)] for monitoring disease course after treatment initiation (interferon {beta}-1a). MTI and MRS detected changes [magnetisation transfer ratio (MTR), N-acetylaspartate (NAA)/creatine ratio] in NAWM over time. EDSS and lesional MTR increases correlated throughout the disease course. Average MTR of NAWM raised during the study (p<0.05) and correlated to the MSFC score (r=0.476, p<0.001). At study termination, NAA/creatine ratio of NAWM correlated to the MSFC score (p<0.05). MTI and MRS were useful for initial disease assessment in NAWM. MTI and MRS correlated with clinical scores, indicating potential for monitoring the disease course and gaining new insights into treatment-related effects. (orig.)

AP1000 core design with 50% MOX loading

Energy Technology Data Exchange (ETDEWEB)

Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)

2008-07-01

The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core and a mixed MOX / UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

AP1000 core design with 50% MOX loading

Energy Technology Data Exchange (ETDEWEB)

Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)], E-mail: fetterrj@westinghouse.com

2009-04-15

The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core design and a mixed MOX/UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance.

Full MOX core design in ABWR

International Nuclear Information System (INIS)

Ihara, Toshiteru; Mochida, Takaaki; Izutsu, Sadayuki; Fujimaki, Shingo

2003-01-01

Electric Power Development Co., Ltd. (EPDC) has been investigating an ABWR plant for construction at Oma-machi in Aomori Prefecture. The reactor, termed FULL MOX-ABWR will have its reactor core eventually loaded entirely with mixed-oxide (MOX) fuel. Extended use of MOX fuel in the plant is expected to play important roles in the country's nuclear fuel recycling policy. MOX fuel bundles will initially be loaded only to less than one-third of the reactor, but will be increased to cover its entire core eventually. The number of MOX fuel bundles in the core thus varies anywhere from 0 to 264 for the initial cycle and, 0 to 872 for equilibrium cycles. The safety design of the FULL MOX-ABWR briefly stated next considers any probable MOX loading combinations out of such MOX bundle usage scheme, starting from full UO 2 to full MOX cores. (author)

HTR core physics analysis at NRG

International Nuclear Information System (INIS)

Kuijper, J.C.; Haas, J.B.M. de; Oppe, J.

2002-01-01

Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANTHERMIX the 3-D steady-state and transient core physics code PANTHER has been interfaced with the HTR thermal hydraulics code THERMIX to enable core follow and transient analyses on both pebble bed and block type HTR systems. Recently the capabilities of PANTHERMIX have been extended with the possibility to simulate the flow of pebbles through the core cavity and the (re)loading of pebbles on top of the core.The PANTHERMIX code system is being applied for the benchmark exercises for the Chinese HTR-10 and Japanese HTTR first criticality, calculating the critical loading, control rod worth and the isothermal temperature coefficients at zero power conditions. Also core physics calculations have been performed on an early version the South African PBMR design. The reactor physics properties of the reactor at equilibrium core loading have been studied as well as a selected run-in scenario, starting form fresh fuel. The recently developed reload option of PANTHERMIX was used extensively in these analyses. The examples shown demonstrate the capabilities of PANTHERMIX for performing steady-state and transient HTR core physics analyses. However, additional validation, especially for transient analyses, remains desirable. (author)

Performance of CASTOR {sup registered} HAW cask cold trials for loading, Transport and storage of HAW canisters

Energy Technology Data Exchange (ETDEWEB)

Wilmsmeier, Marco; Horn, Thomas; Graf, Wilhelm [GNS Gesellschaft fuer Nuklear-Service mbH (Germany)

2009-07-01

With over 30 years of experience in the design, manufacturing, assembly and loading of CASTOR {sup registered} casks, GNS is one of the worldwide leading suppliers of casks for the transport and storage of spent fuel assemblies as well as for canisters with vitrified high active wastes (meanwhile over 1.000 casks loaded and stored and more than 1.500 ordered). GNS's products are used at around 30 sites worldwide for a wide range of inventories from pressurised and boiling water reactor fuels (PWR, VVER and BWR, RBMK), thorium high-temperature reactor fuels (THTR) and research reactor fuels (MTR) to vitrified high active wastes (HAW) from reprocessing plants. GNS is responsible for all nuclear wastes resulting from German Nuclear Power Plants and assists and/or performs in the loading and dispatch of CASTOR {sup registered} casks as well as their transport to and storage at central interim storage facilities and local interim storage areas. (orig.)

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

International Nuclear Information System (INIS)

Smith, D.L.

1987-10-01

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

FFTF initial fuel loading, preanalyses, and comparison with preliminary results

International Nuclear Information System (INIS)

Rothrock, R.B.; Daughtry, J.W.; Zimmerman, B.D.; Petrowicz, N.E.; Bennett, R.A.; Ombrellaro, P.A.

1980-02-01

Disadvantages of conventional loading from the center out were circumvented by loading one trisector at a time, and connecting the control rod drivelines in each sector after it was loaded so that the rods could be operated during the loading of subsequent trisectors. This sequence was interrupted once during the loading of the final sector, to achieve initial criticality at an approximately minimum critical loading and to measure absolute subcriticality by the rod drop technique. An in-core detector was preferable to the standard FTR ex-core detectors for monitoring the initial fuel loading. Consequently, special fission chambers were installed in an instrument thimble near the core center to monitor the initial fuel loading

Application of nonlinear nodal diffusion method for a small research reactor

International Nuclear Information System (INIS)

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

2014-01-01

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

The PWR loading pattern optimization in X-IMAGE

International Nuclear Information System (INIS)

Stevens, J.G.; Smith, K.S.; Rempe, K.R.; Downar, T.J.

1993-01-01

The design of reactor core loading patterns is difficult due to the staggering number of patterns. The integer nature and nonlinear neutronic response of core design preclude simple prescriptions for generation of the feasible patterns, much less optimization among feasible candidates. Fortunately, recent developments in optimization, graphical user interfaces (GUIs), and the speed and low cost of engineering workstations combine to make loading pattern automation possible. The optimization module SIMAN has been added to X-IMAGE to automatically generate high-quality core loadings

Critical experiments on enriched uranium graphite moderated cores

International Nuclear Information System (INIS)

Kaneko, Yoshihiko; Akino, Fujiyoshi; Kitadate, Kenji; Kurokawa, Ryosuke

1978-07-01

A variety of 20 % enriched uranium loaded and graphite-moderated cores consisting of the different lattice cells in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments systematically. In the present report, the experimental results for homogeneously or heterogeneously fuel loaded cores and for simulation core of the experimental reactor for a multi-purpose high temperature reactor are filed so as to be utilized for evaluating the accuracy of core design calculation for the experimental reactor. The filed experimental data are composed of critical masses of uranium, kinetic parameters, reactivity worths of the experimental control rods and power distributions in the cores with those rods. Theoretical analyses are made for the experimental data by adopting a simple ''homogenized cylindrical core model'' using the nuclear data of ENDF/B-III, which treats the neutron behaviour after smearing the lattice cell structure. It is made clear from a comparison between the measurement and the calculation that the group constants and fundamental methods of calculations, based on this theoretical model, are valid for the homogeneously fuel loaded cores, but not for both of the heterogeneously fuel loaded cores and the core for simulation of the experimental reactor. Then, it is pointed out that consideration to semi-homogeneous property of the lattice cells for reactor neutrons is essential for high temperature graphite-moderated reactors using dispersion fuel elements of graphite and uranium. (author)

Reactor core performance estimating device

International Nuclear Information System (INIS)

Tanabe, Akira; Yamamoto, Toru; Shinpuku, Kimihiro; Chuzen, Takuji; Nishide, Fusayo.

1995-01-01

The present invention can autonomously simplify a neural net model thereby enabling to conveniently estimate various amounts which represents reactor core performances by a simple calculation in a short period of time. Namely, a reactor core performance estimation device comprises a nerve circuit net which divides the reactor core into a large number of spacial regions, and receives various physical amounts for each region as input signals for input nerve cells and outputs estimation values of each amount representing the reactor core performances as output signals of output nerve cells. In this case, the nerve circuit net (1) has a structure of extended multi-layered model having direct coupling from an upper stream layer to each of downstream layers, (2) has a forgetting constant q in a corrected equation for a joined load value ω using an inverse error propagation method, (3) learns various amounts representing reactor core performances determined using the physical models as teacher signals, (4) determines the joined load value ω decreased as '0' when it is to less than a predetermined value upon learning described above, and (5) eliminates elements of the nerve circuit net having all of the joined load value decreased to 0. As a result, the neural net model comprises an autonomously simplifying means. (I.S.)

Analysis and Testing of Load Characteristics for Rotary-Percussive Drilling of Lunar Rock Simulant with a Lunar Regolith Coring Bit

Directory of Open Access Journals (Sweden)

Peng Li

2017-01-01

Full Text Available Based on an optimized lunar regolith coring bit (LRCB configuration, the load characteristics of rotary-percussive drilling of lunar rock simulant in a laboratory environment are analyzed to determine the effects of the drilling parameters (the rotational velocity, the penetration rate, and the percussion frequency on the drilling load. The process of rotary drilling into lunar rock using an LRCB is modeled as an interaction between an elemental blade and the rock. The rock’s fracture mechanism during different stages of the percussive mechanism is analyzed to create a load forecasting model for the cutting and percussive fracturing of rock using an elemental blade. Finally, a model of the load on the LRCB is obtained from the analytic equation for the bit’s cutting blade distribution; experimental verification of the rotary-impact load characteristics for lunar rock simulant with different parameters is performed. The results show that the penetrations per revolution (PPR are the primary parameter influencing the drilling load. When the PPR are fixed, increasing the percussion frequency reduces the drilling load on the rock. Additionally, the variation pattern of the drilling load of the bit is in agreement with that predicted by the theoretical model. This provides a research basis for subsequent optimization of the drilling procedure and online recognition of the drilling process.

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.

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Web buckling behavior under in-plane compression and shear loads for web reinforced composite sandwich core

Science.gov (United States)

Toubia, Elias Anis

Sandwich construction is one of the most functional forms of composite structures developed by the composite industry. Due to the increasing demand of web-reinforced core for composite sandwich construction, a research study is needed to investigate the web plate instability under shear, compression, and combined loading. If the web, which is an integral part of the three dimensional web core sandwich structure, happens to be slender with respect to one or two of its spatial dimensions, then buckling phenomena become an issue in that it must be quantified as part of a comprehensive strength model for a fiber reinforced core. In order to understand the thresholds of thickness, web weight, foam type, and whether buckling will occur before material yielding, a thorough investigation needs to be conducted, and buckling design equations need to be developed. Often in conducting a parametric study, a special purpose analysis is preferred over a general purpose analysis code, such as a finite element code, due to the cost and effort usually involved in generating a large number of results. A suitable methodology based on an energy method is presented to solve the stability of symmetrical and specially orthotropic laminated plates on an elastic foundation. Design buckling equations were developed for the web modeled as a laminated plate resting on elastic foundations. The proposed equations allow for parametric studies without limitation regarding foam stiffness, geometric dimensions, or mechanical properties. General behavioral trends of orthotropic and symmetrical anisotropic plates show pronounced contribution of the elastic foundation and fiber orientations on the buckling resistance of the plate. The effects of flexural anisotropy on the buckling behavior of long rectangular plates when subjected to pure shear loading are well represented in the model. The reliability of the buckling equations as a design tool is confirmed by comparison with experimental results

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Effect of core ceramic grinding on fracture behaviour of bilayered zirconia veneering ceramic systems under two loading schemes.

Science.gov (United States)

Jian, Yu-Tao; Tang, Tian-Yu; Swain, Michael V; Wang, Xiao-Dong; Zhao, Ke

2016-12-01

The aim of this in vitro study was to evaluate the effect of core ceramic grinding on the fracture behaviour of bilayered zirconia under two loading schemes. Interfacial surfaces of sandblasted zirconia disks (A) were ground with 80 (B), 120 (C) and 220 (D) grit diamond discs, respectively. Surface roughness and topographic analysis were performed using a confocal scanning laser microscope (CSLM) and a scanning electron microscopy (SEM). Relative monoclinic content was evaluated using X-ray diffraction analysis (XRD) then reevaluated after simulated veneer firing. Biaxial fracture strength (σ) and Weibull modulus (m) were calculated either with core in compression (subgroup Ac-Dc) or in tension (subgroup At-Dt). Facture surfaces were examined by SEM and energy dispersive X-ray spectroscopy (EDS). Maximum tensile stress at fracture was estimated by finite element analysis. Statistical data analysis was performed using Kruskal-Wallis and one-way ANOVA at a significance level of 0.05. As grit size of the diamond disc increased, zirconia surface roughness decreased (pgrinding. No difference in initial (p=0.519 for subgroups Ac-Dc) and final fracture strength (p=0.699 for subgroups Ac-Dc; p=0.328 for subgroups At-Dt) was found among the four groups for both loading schemes. While coarse grinding slightly increased final fracture strength reliability (m) for subgroups Ac-Dc. Two different modes of fracture were observed according to which material was on the bottom surface. Components of the liner porcelain remained on the zirconia surface after fracture for all groups. Technician grinding changed surface topography of zirconia ceramic material, but was not detrimental to the bilayered system strength after veneer application. Coarse grinding slightly improved the fracture strength reliability of the bilayered system tested with core in compression. It is recommended that veneering porcelain be applied directly after routine lab grinding of zirconia ceramic, and its

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.)

Influence of fuel assembly loading pattern and fuel burnups upon leakage neutron flux spectra from light water reactor core (Joint research)

International Nuclear Information System (INIS)

Kojima, Kensuke; Okumura, Keisuke; Kosako, Kazuaki; Torii, Kazutaka

2016-01-01

At the decommissioning of light water reactors (LWRs), it is important to evaluate an amount of radioactivity in the ex-core structures such as a reactor containment vessel, radiation shieldings, and so on. It is thought that the leakage neutron spectra in these radioactivation regions, which strongly affect the induced radioactivity, would be changed by different reactor core configurations such as fuel assembly loading pattern and fuel burnups. This study was intended to evaluate these effects. For this purpose, firstly, partial neutron currents on the core surfaces were calculated for some core configurations. Then, the leakage neutron flux spectra in major radioactivation regions were calculated based on the provided currents. Finally, influence of the core configurations upon the neutron flux spectra was evaluated. As a result, it has been found that the influence is small on the spectrum shapes of neutron fluxes. However, it is necessary to pay attention to the facts that intensities of the leakage neutron fluxes are changed by the configurations and that intensities and spectrum shapes of the leakage neutron fluxes are changed depending on the angular direction around the core. (author)

A review of PFR core distortion experience

International Nuclear Information System (INIS)

Brook, A.J.

1984-01-01

Neutron induced voidage (NIV) swelling and irradiation creep, acting together or individually, produce deformation in core components exposed to a fast neutron flux and can lead to mechanical interaction between them. Today the nature of these processes is reasonably well understood, and reactor designers have two options in attempting to accomodate them: either by employing a flexible free standing design in which contact loadings are low but in which distortion may be high, or more commonly, by some type of restrained core in which inter-component loadings are high, but where distortion is relatively small. The aims of this paper are: a. to describe briefly the various operational limits of core and core component distortion and how they arise, for which a brief description of reactor construction is necessary; b. to outline how the problems of inter-component contact loadings are overcome for the interactive core; c. to describe some other potential problems which arise either from absolute swelling, or from differential swelling between components; of particular relevance here is the problem of contact loadings between absorber rods and their guide tubes; d. to comment on the degree of agreement with, and the feedback provided by, PIE findings; e. to show how the results of the work influence reactor operators and the reload program

Methotrexate diethyl ester-loaded lipid-core nanocapsules in aqueous solution increased antineoplastic effects in resistant breast cancer cell line

Directory of Open Access Journals (Sweden)

Yurgel VC

2014-03-01

Full Text Available Virginia C Yurgel,1,* Catiuscia P Oliveira,2,* Karine R Begnini,1 Eduarda Schultze,1 Helena S Thurow,1 Priscila MM Leon,1 Odir A Dellagostin,1 Vinicius F Campos,1 Ruy CR Beck,2 Silvia S Guterres,2 Tiago Collares,1 Adriana R Pohlmann,2–4 Fabiana K Seixas11Programa de Pós-Graduação em Biotecnologia (PPGB, Grupo de Pesquisa em Oncologia Celular e Molecular, Laboratório de Genômica Funcional, Biotecnologia/Centro de Desenvolvimento Tecnológico, Universidade Federal de Pelotas, Pelotas, Rio Grande do Sul, Brazil; 2Programa de Pós-Graduação em Ciências Farmacêuticas, Faculdade de Farmácia, Universidade Federal do Rio Grande do Sul, Porto Alegre, Rio Grande do Sul, Brazil; 3Departamento de Química Orgânica, Instituto de Química, Universidade Federal do Rio Grande do Sul, Porto Alegre, Rio Grande do Sul, Brazil; 4Centro de Nanociência e Nanotecnologia, CNANO-UFRGS, Universidade Federal do Rio Grande do Sul, Porto Alegre, Rio Grande do Sul, Brazil*These authors contributed equally to this workAbstract: Breast cancer is the most frequent cancer affecting women. Methotrexate (MTX is an antimetabolic drug that remains important in the treatment of breast cancer. Its efficacy is compromised by resistance in cancer cells that occurs through a variety of mechanisms. This study evaluated apoptotic cell death and cell cycle arrest induced by an MTX derivative (MTX diethyl ester [MTX(OEt2] and MTX(OEt2-loaded lipid-core nanocapsules in two MTX-resistant breast adenocarcinoma cell lines, MCF-7 and MDA-MB-231. The formulations prepared presented adequate granulometric profile. The treatment responses were evaluated through flow cytometry. Relying on the mechanism of resistance, we observed different responses between cell lines. For MCF-7 cells, MTX(OEt2 solution and MTX(OEt2-loaded lipid-core nanocapsules presented significantly higher apoptotic rates than untreated cells and cells incubated with unloaded lipid-core nanocapsules. For MDA-MB-231

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

International Nuclear Information System (INIS)

Thomasson, J.

1998-01-01

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

Two optimal control methods for PWR core control

International Nuclear Information System (INIS)

Karppinen, J.; Blomsnes, B.; Versluis, R.M.

1976-01-01

The Multistage Mathematical Programming (MMP) and State Variable Feedback (SVF) methods for PWR core control are presented in this paper. The MMP method is primarily intended for optimization of the core behaviour with respect to xenon induced power distribution effects in load cycle operation. The SVF method is most suited for xenon oscillation damping in situations where the core load is unpredictable or expected to stay constant. Results from simulation studies in which the two methods have been applied for control of simple PWR core models are presented. (orig./RW) [de

Hyaluronic acid/poly(lactic-co-glycolic acid) core/shell fiber meshes loaded with epigallocatechin-3-O-gallate as skin tissue engineering scaffolds.

Science.gov (United States)

Lee, Eun Ji; Lee, Jong Ho; Jin, Linhua; Jin, Oh Seong; Shin, Yong Cheol; Sang, Jin Oh; Lee, Jaebeom; Hyon, Suong-Hyu; Han, Dong-Wook

2014-11-01

In this study, hyaluronic acid (HA)/poly(lactic-co-glycolic acid, PLGA) core/shell fiber meshes loaded with epigallocatechin-3-O-gallate (EGCG) (HA/PLGA-E) for application to tissue engineering scaffolds for skin regeneration were prepared via coaxial electrospinning. Physicochemical properties of HA/PLGA-E core/shell fiber meshes were characterized by SEM, Raman spectroscopy, contact angle, EGCG release profiling and in vitro degradation. Biomechanical properties of HA/PLGA-E meshes were also investigated by a tensile strength test. SEM images showed that HA/PLGA-E fiber meshes had a three-dimensional interconnected pore structure with an average fiber diameter of about 1270 nm. Raman spectra revealed that EGCG was uniformly dispersed in the PLGA shell of meshes. HA/PLGA-E meshes showed sustained EGCG release patterns by controlled diffusion and PLGA degradation over 4 weeks. EGCG loading did not adversely affect the tensile strength and elastic modulus of HA/PLGA meshes, while increased their hydrophilicity and surface energy. Attachment of human dermal fibroblasts on HA/PLGA-E meshes was appreciably increased and their proliferation was steadily retained during the culture period. These results suggest that HA/PLGA-E core/shell fiber meshes can be potentially used as scaffolds supporting skin regeneration.

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-07-01

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

Investigation of failure mechanisms for HTGR core supports

International Nuclear Information System (INIS)

Bennett, J.G.; Ju, F.D.; Anderson, C.A.

1976-12-01

The report is concerned with potential instabilities of High-Temperature Gas-Cooled Reactor Cores supported by graphite columns. Two failure mechanisms are investigated in detail: that of torsional buckling of the entire core-column assemblage and that of column failure alone. A torsional model of the core-column assemblage is described and static buckling loads are calculated. Dynamic instability of the model to seismic loadings is also investigated. Individual column failure is examined using nonlinear graphite behavior and safety factors for static loading situations are given and compared to values given by conventional design formulas. A model of a cracked graphite column is given and buckling loads are computed for columns using a combined column and fracture mechanics analysis. A finite element analysis of a cracked graphite column is presented

A diurnal resonance in the ocean tide and in the earth's load response due to the resonant free 'core nutation'

Science.gov (United States)

Wahr, J. M.; Sasao, T.

1981-01-01

The effects of the oceans, which are subject to a resonance due to a free rotational eigenmode of an elliptical, rotating earth with a fluid outer core having an eigenfrequency of (1 + 1/460) cycle/day, on the body tide and nutational response of the earth to the diurnal luni-tidal force are computed. The response of an elastic, rotating, elliptical, oceanless earth with a fluid outer core to a given load distribution on its surface is first considered, and the tidal sea level height for equilibrium and nonequilibrium oceans is examined. Computations of the effects of equilibrium and nonequilibrium oceans on the nutational and deformational responses of the earth are then presented which show small but significant perturbations to the retrograde 18.6-year and prograde six-month nutations, and more important effects on the earth body tide, which is also resonant at the free core notation eigenfrequency.

Fuel assembly and reactor core

International Nuclear Information System (INIS)

Yuchi, Yoko; Aoyama, Motoo; Haikawa, Katsumasa; Yamanaka, Akihiro; Koyama, Jun-ichi.

1996-01-01

In a fuel assembly of a BWR type reactor, a region substantially containing burnable poison is divided into an upper region and a lower region having different average concentrations of burnable poison along a transverse cross section perpendicular to the axial direction. The ratio of burnable poison contents of both regions is determined to not more than 80%, and the average concentration of the burnable poison in the lower region is determined to not less than 9% by weight. An infinite multiplication factor at an initial stage of the burning of the fuel assembly is controlled effectively by the burnable poisons. Namely, the ratio of the axial power can be controlled by the distribution of the enrichment degree of uranium fuels and the distribution of the burnable poison concentration in the axial direction. Since the average enrichment degree of the reactor core has to be increased in order to provide an initially loaded reactor core at high burnup degree. Distortion of the power distribution in the axial direction of the reactor core to which fuel assemblies at high enrichment degree are loaded is flattened to improve thermal margin, to extend continuous operation period and increase a burnup degree upon take-out thereby improving fuel economy without worsening the reactor core characteristics of the initially loaded reactor core. (N.H.)

Stochastic optimization of loading pattern for PWR

International Nuclear Information System (INIS)

Smuc, T.; Pevec, D.

1994-01-01

The application of stochastic optimization methods in solving in-core fuel management problems is restrained by the need for a large number of proposed solutions loading patterns, if a high quality final solution is wanted. Proposed loading patterns have to be evaluated by core neutronics simulator, which can impose unrealistic computer time requirements. A new loading pattern optimization code Monte Carlo Loading Pattern Search has been developed by coupling the simulated annealing optimization algorithm with a fast one-and-a-half dimensional core depletion simulator. The structure of the optimization method provides more efficient performance and allows the user to empty precious experience in the search process, thus reducing the search space size. Hereinafter, we discuss the characteristics of the method and illustrate them on the results obtained by solving the PWR reload problem. (authors). 7 refs., 1 tab., 1 fig

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

International Nuclear Information System (INIS)

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

1993-01-01

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

SMART core preliminary nuclear design-II

Energy Technology Data Exchange (ETDEWEB)

Lee, Jeong Chan; Ji, Seong Kyun; Chang, Moon Hee

1997-06-01

Three loading patterns for 330 MWth SMART core are constructed for 25, 33 and 29 CRDMs, and one loading pattern for larger 69-FA core with 45 CRDMs is also constructed for comparison purpose. In this study, the core consists of 57 reduced height Korean Optimized Fuel Assemblies (KOFAs) developed by KAERI. The enrichment of fuel is 4.95 w/o. As a main burnable poison, 35% B-10 enriched B{sub 4}C-Al{sub 2}O{sub 3} shim is used. To control stuck rod worth, some gadolinia bearing fuel rods are used. The U-235 enrichment of the gadolinia bearing fuel rods is 1.8 w/o as used in KOFA. All patterns return cycle length of about 3 years. Three loading patterns except 25-CRDM pattern satisfy cold shutdown condition of keff {<=} 0.99 without soluble boron. These three patterns also satisfy the refueling condition of keff {<=} 0.95. In addition to the construction of loading pattern, an editing module of MASTER PPI files for rod power history generation is developed and rod power histories are generated for 29-CRDM loading pattern. Preliminary Fq design limit is suggested as 3.71 based on KOFA design experience. (author). 9 tabs., 45 figs., 16 refs.

Transformer sound level caused by core magnetostriction and winding stress displacement variation

Directory of Open Access Journals (Sweden)

Chang-Hung Hsu

2017-05-01

Full Text Available Magnetostriction caused by the exciting variation of the magnetic core and the current conducted by the winding wired to the core has a significant result impact on a power transformer. This paper presents the sound of a factory transformer before on-site delivery for no-load tests. This paper also discusses the winding characteristics from the transformer full-load tests. The simulation and the measurement for several transformers with capacities ranging from 15 to 60 MVA and high voltage 132kV to low voltage 33 kV are performed. This study compares the sound levels for transformers by no-load test (core/magnetostriction and full-load test (winding/displacement ε. The difference between the simulated and the measured sound levels is about 3dB. The results show that the sound level depends on several parameters, including winding displacement, capacity, mass of the core and windings. Comparative results of magnetic induction of cores and the electromagnetic force of windings for no-load and full-load conditions are examined.

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

Energy Technology Data Exchange (ETDEWEB)

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

1966-05-15

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

Fuel management for off-load annual refuelling of the D-HHT 600 MW(e) reference core

Energy Technology Data Exchange (ETDEWEB)

Hansen, U

1973-03-16

The reference design for the Dragon-HHT reactor has been optimised for on-load continuous refuelling. The possiblity to operate the reactor on a discontinuous annual reloading schedule might prove of interest and/or necessity. In this paper the influence of an annual 4-batch fuel management scheme on the core physics and fuel cycle economics is investigated. The results of the present investigation give a good indication of the relative merits of the two fuel management schemes. Although a broader parameter survey and a more detailed scrutinising of special cases would be desirable, we feel that the main conclusions are correct and that the principle differences have been elicited.

Face/core interface fracture characterization of mixed mode bending sandwich specimens

DEFF Research Database (Denmark)

Quispitupa, Amilcar; Berggreen, Christian; Carlsson, L.A.

2011-01-01

and PVC H45, H100 and H250 foam core materials were evaluated. A methodology to perform precracking on fracture specimens in order to achieve a sharp and representative crack front is outlined. The mixed mode loading was controlled in the mixed mode bending (MMB) test rig by changing the loading......Debonding of the core from the face sheets is a critical failure mode in sandwich structures. This paper presents an experimental study on face/core debond fracture of foam core sandwich specimens under a wide range of mixed mode loading conditions. Sandwich beams with E‐glass fibre face sheets...... application point (lever arm distance). Finite element analysis was performed to determine the mode‐mixity at the crack tip. The results showed that the face/core interface fracture toughness increased with increased mode II loading. Post failure analysis of the fractured specimens revealed that the crack...

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

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.

Development of lycopene-loaded lipid-core nanocapsules: physicochemical characterization and stability study

International Nuclear Information System (INIS)

Santos, Priscilla Pereira dos; Paese, Karina; Guterres, Silvia Stanisçuaski; Pohlmann, Adriana Raffin; Costa, Tania Hass; Jablonski, André; Flôres, Simone Hickmann; Rios, Alessandro de Oliveira

2015-01-01

The objective of this study was to develop lycopene-loaded lipid-core nanocapsules (Lyc-LNCs) by the interfacial deposition of preformed poly(ε-caprolactone) (PCL). Lyc extract (93.9 %) was obtained from tomatoes, and the organic phase was prepared with polymer (PCL), caprylic/capric triglycerides, sorbitan monostearate, and Lyc in a mixture of acetone and ethanol under magnetic stirring at 40 °C. The organic phase was injected into an aqueous phase containing polysorbate 80, and the suspension was concentrated under reduced pressure. The formulation with a Lyc concentration of 85 µg/mL was characterized in terms of size distribution, zeta potential, encapsulation efficiency, pH, viscosity, and color. The Lyc-LNC formulation presented stable values for the z-average (193 ± 4.7 nm) and zeta potential (−11.5 ± 0.40 mV). Despite the lower pH, Lyc content, and color change of the suspension, the nanocapsules showed satisfactory stability, presenting around 50 % Lyc content after 14 days of storage at room temperature (25 °C)

Development of lycopene-loaded lipid-core nanocapsules: physicochemical characterization and stability study

Science.gov (United States)

dos Santos, Priscilla Pereira; Paese, Karina; Guterres, Silvia Stanisçuaski; Pohlmann, Adriana Raffin; Costa, Tania Hass; Jablonski, André; Flôres, Simone Hickmann; Rios, Alessandro de Oliveira

2015-02-01

The objective of this study was to develop lycopene-loaded lipid-core nanocapsules (Lyc-LNCs) by the interfacial deposition of preformed poly(ɛ-caprolactone) (PCL). Lyc extract (93.9 %) was obtained from tomatoes, and the organic phase was prepared with polymer (PCL), caprylic/capric triglycerides, sorbitan monostearate, and Lyc in a mixture of acetone and ethanol under magnetic stirring at 40 °C. The organic phase was injected into an aqueous phase containing polysorbate 80, and the suspension was concentrated under reduced pressure. The formulation with a Lyc concentration of 85 µg/mL was characterized in terms of size distribution, zeta potential, encapsulation efficiency, pH, viscosity, and color. The Lyc-LNC formulation presented stable values for the z-average (193 ± 4.7 nm) and zeta potential (-11.5 ± 0.40 mV). Despite the lower pH, Lyc content, and color change of the suspension, the nanocapsules showed satisfactory stability, presenting around 50 % Lyc content after 14 days of storage at room temperature (25 °C).

Development of lycopene-loaded lipid-core nanocapsules: physicochemical characterization and stability study

Energy Technology Data Exchange (ETDEWEB)

Santos, Priscilla Pereira dos [Universidade Federal do Rio Grande do Sul (UFRGS), Instituto de Ciência e Tecnologia de Alimentos (Brazil); Paese, Karina; Guterres, Silvia Stanisçuaski [Universidade Federal do Rio Grande do Sul, Programa de Pós-Graduação em Ciências Farmacêuticas, Faculdade de Farmácia (Brazil); Pohlmann, Adriana Raffin [Universidade Federal do Rio Grande do Sul (UFRGS), Departamento de Química Orgânica, Instituto de Química (Brazil); Costa, Tania Hass [Universidade Federal do Rio Grande do Sul (UFRGS), Departamento de Química, Instituto de Química (Brazil); Jablonski, André [Universidade Federal do Rio Grande do Sul (UFRGS), Departamento de Engenharia de Minas (Brazil); Flôres, Simone Hickmann; Rios, Alessandro de Oliveira, E-mail: alessandro.rios@ufrgs.br [Universidade Federal do Rio Grande do Sul (UFRGS), Instituto de Ciência e Tecnologia de Alimentos (Brazil)

2015-02-15

The objective of this study was to develop lycopene-loaded lipid-core nanocapsules (Lyc-LNCs) by the interfacial deposition of preformed poly(ε-caprolactone) (PCL). Lyc extract (93.9 %) was obtained from tomatoes, and the organic phase was prepared with polymer (PCL), caprylic/capric triglycerides, sorbitan monostearate, and Lyc in a mixture of acetone and ethanol under magnetic stirring at 40 °C. The organic phase was injected into an aqueous phase containing polysorbate 80, and the suspension was concentrated under reduced pressure. The formulation with a Lyc concentration of 85 µg/mL was characterized in terms of size distribution, zeta potential, encapsulation efficiency, pH, viscosity, and color. The Lyc-LNC formulation presented stable values for the z-average (193 ± 4.7 nm) and zeta potential (−11.5 ± 0.40 mV). Despite the lower pH, Lyc content, and color change of the suspension, the nanocapsules showed satisfactory stability, presenting around 50 % Lyc content after 14 days of storage at room temperature (25 °C)

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)

In core instrumentation for online nuclear heating measurements of material testing reactor

International Nuclear Information System (INIS)

Reynard, C.; Andre, J.; Brun, J.; Carette, M.; Janulyte, A.; Merroun, O.; Zerega, Y.; Lyoussi, A.; Bignan, G.; Chauvin, J-P.; Fourmentel, D.; Glayse, W.; Gonnier, C.; Guimbal, P.; Iracane, D.; Villard, J.-F.

2010-01-01

The present work focuses on nuclear heating. This work belongs to a new advanced research program called IN-CORE which means 'Instrumentation for Nuclear radiations and Calorimetry Online in REactor' between the LCP (University of Provence-CNRS) and the CEA (French Atomic Energy Commission) - Jules Horowitz Reactor (JHR) program. This program started in September 2009 and is dedicated to the conception and the design of an innovative mobile experimental device coupling several sensors and ray detectors for on line measurements of relevant physical parameters (photonic heating, neutronic flux ...) and for an accurate parametric mapping of experimental channels in the JHR Core. The work presented below is the first step of this program and concerns a brief state of the art related to measurement methods of nuclear heating phenomena in research reactor in general and MTR in particular. A special care is given to gamma heating measurements. A first part deals with numerical codes and models. The second one presents instrumentation divided into various kinds of sensor such as calorimeter measurements and gamma ionization chamber measurements. Their basic principles, characteristics such as metrological parameters, operating mode, disadvantages/advantages, ... are discussed. (author)

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

International Nuclear Information System (INIS)

Hassel, H.W.

1991-01-01

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

Conceptual study of advanced PWR core design

International Nuclear Information System (INIS)

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

1997-09-01

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

Conceptual study of advanced PWR core design

Energy Technology Data Exchange (ETDEWEB)

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

1997-09-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-15

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

Improvement of daily load-following operation for boiling water reactors

International Nuclear Information System (INIS)

Kiguchi, Takashi; Kurihara, Kunitoshi; Sakurai, Mikio; Joge, Toshio; Asami, Kazuo.

1980-01-01

Recently, with the increase of the proportion of nuclear power generation to the total amount of power generation of electric power systems, the needs of daily load-following operation of nuclear power stations have heightened, accordingly the study on the method of daily load-following operation has been carried out for BWRs. In this study, by the combined use of the flow rate control of core coolnat being operated easily and the operation of control rods, the BWR system with the daily load-following performance of 100% power output in daytime and 50% power output at night was the target of development. For the purpose, the change of core characteristics during load-following was grasped analytically, and the range of load change was investigated. At the same time, as the first stage of developing operation control and monitoring system, the reactor output-adjusting device which makes generator output automatically follow the target load change pattern by the flow rate control of core coolnat, and the equipment for monitoring core performance on line were developed. The analysis of the method of daily load-following operation in present-day BWRs, the study on the improvement of load-following operation performance, the reactor output-adjusting device are described. (Kako, I.)

Successful completion of a time sensitive MTR and TRIGA Indonesian shipment

International Nuclear Information System (INIS)

Anne, Catherine; Patterson, John; Messick, Chuck

2005-01-01

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

Application of Computational Intelligence Methods to In-Core Fuel Management

International Nuclear Information System (INIS)

Erdogan, A.

2001-01-01

In this study, a computer program package has been developed which supports the in-core fuel management activities for pressurized water reactors, generates and recommends an optimum loading pattern to ensure safe and efficient reactor operation. A search for an optimum fuel loading pattern must be conducted in the space of several core parameters such as power distribution, which is an excessively time consuming computational process. Global core calculation codes take a relatively long time to do the task. The time interval necessary for the iterative process was reduced by using an artificial neural network estimator for the calculations. In this way, it was possible to analyze more loading patterns in the same time interval and the probability of finding a desired optimum was increased. As a case study, the core of the Almaraz Nuclear Plant of Spain, a pressurized water reactor, was modeled for the core calculation code system. The 2-group cross sections for the fuel assembly types were calculated and stored for later usage with the diffusion code. 2000 loading patterns were generated by placing fuel assemblies to random positions in the core, and for each pattern the power distribution and effective multiplication factor (k e ff) were calculated with the diffusion code. At the next stage, 500 of the loading patterns were introduced to the neural network as input data for the training process. The remaining 1500 patterns were used to validate the neural network implementation. It was shown that the neural network estimates the power distribution and the K effective within acceptable error limits. To complete the system, a loading pattern generator was developed. This module consists of a set of rules and an algorithm that places the fuel assemblies to core positions. The neural network estimated the power distribution and k e ff for the loading patterns that were generated by this module. The patterns that have a maximum power fraction lower than, and a minimum

Internal loading of an inhomogeneous compressible Earth with phase boundaries

Science.gov (United States)

Defraigne, P.; Dehant, V.; Wahr, J. M.

1996-01-01

The geoid and the boundary topography caused by mass loads inside the earth were estimated. It is shown that the estimates are affected by compressibility, by a radially varying density distribution, and by the presence of phase boundaries with density discontinuities. The geoid predicted in the chemical boundary case is 30 to 40 percent smaller than that predicted in the phase case. The effects of compressibility and radially varying density are likely to be small. The inner core-outer core topography for loading inside the mantle and for loading inside the inner core were computed.

Fracture Characterization of Sandwich Face/Core Interfaces

DEFF Research Database (Denmark)

Manca, Marcello

of load transfer between the faces and the core layer is lost, the debonds are considered as primary damage initiators. Under fatigue loading the debonds may evolve into cracks that cause a reduction in structural performance and consequent failure. At present most structural design is based on “life-time...... of sandwich structures is defects that are introduced in the manufacturing process. It is inevitable that areas of the face sheets will not fully adhere to the core resulting in defects known as “debonds”. Debonds can also be induced in-service due to e.g. localised impact loading or overloading. As the means...... such result it is important to devise new experimental and analytical techniques to establish the multi-mode fracture characteristics of sandwich plate structures and accordingly develop methods to inhibit defect propagation. This thesis deals with characterization of fracture between face and core...

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)

ACVP-02: Plasma SIV/SHIV RNA Viral Load Measurements through the AIDS and Cancer Virus Program Quantitative Molecular Diagnostics Core | Frederick National Laboratory for Cancer Research

Science.gov (United States)

The SIV plasma viral load assay performed by the Quantitative Molecular Diagnostics Core (QMDC) utilizes reagents specifically designed to detect and accurately quantify the full range of SIV/SHIV viral variants and clones in common usage in the rese

Containment loading during severe core damage accidents

International Nuclear Information System (INIS)

Fermandjian, J.; Evrard, J.M.; Cenerino, C.; Berthion, Y.; Carvallo, G.

1984-11-01

The objective of the article is to study the influence of the state of the reactor cavity (dry or flooded) and of the corium coolability on the thermal-hydraulics in the containment in the case of an accident sequence involving core melting and subsequent containment basemat erosion, in a 900 MWe PWR unit. Calculations are performed by using the JERICHO thermal hydraulics code

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Sensitivity analysis of reflector types and impurities in 10 MW MTR type nuclear research reactor

International Nuclear Information System (INIS)

Khattab, K.; Khamis, I.

2007-01-01

The 2-D and 3-D neutronics models for 10 MW nuclear research reactor of MTR type have been developed and presented in this paper. Our results agree very well with the results of seven countries mentioned in the IAEA-TECDOC-233. To study the effect of reflector types on the reactor effective multiplication factor, five types of reflectors such as pure beryllium, beryllium, heavy water, carbon and water are selected for this study. The pure beryllium is found to be the most efficient reflector in this group. The effect of the most important impurities, which exist on the beryllium reflector such as iron, silicon and aluminium on the reactor multiplication factor, have been analyzed as well. It is found that the iron impurity affects the reactor multiplication factor the most compared to silicon and aluminium impurities. (author)

FUEL BURN-UP CALCULATION FOR WORKING CORE OF THE RSG-GAS RESEARCH REACTOR AT BATAN SERPONG

Directory of Open Access Journals (Sweden)

Tukiran Surbakti

2017-12-01

Full Text Available The neutronic parameters are required in the safety analysis of the RSG-GAS research reactor. The RSG-GAS research reactor, MTR (Material Testing Reactor type is used for research and also in radioisotope production. RSG-GAS has been operating for 30 years without experiencing significant obstacles. It is managed under strict requirements, especially fuel management and fuel burn-up calculations. The reactor is operated under the supervision of the Regulatory Body (BAPETEN and the IAEA (International Atomic Energy Agency. In this paper, the experience of managing RSG-GAS core fuels will be discussed, there are hundred possibilities of fuel placements on the reactor core and the strategy used to operate the reactor will be crucial. However, based on strict calculation and supervision, there is no incorrect placement of the fuels in the core. The calculations were performed on working core by using the WIMSD-5B computer code with ENDFVII.0 data file to generate the macroscopic cross-section of fuel and BATAN-FUEL code were used to obtain the neutronic parameter value such as fuel burn-up fractions. The calculation of the neutronic core parameters of the RSG-GAS research reactor was carried out for U3Si2-Al fuel, 250 grams of mass, with an equilibrium core strategy. The calculations show that on the last three operating cores (T90, T91, T92, all fuels meet the safety criteria and the fuel burn-up does not exceed the maximum discharge burn-up of 59%. Maximum fuel burn-up always exists in the fuel which is close to the position of control rod.

Material dimensionality effects on the nanoindentation behavior of Al/a-Si core-shell nanostructures

Energy Technology Data Exchange (ETDEWEB)

Fleming, Robert A. [Department of Mechanical Engineering, University of Arkansas, Fayetteville, AR 72701 (United States); Center for Advanced Surface Engineering, University of Arkansas, Fayetteville, AR 72701 (United States); Goss, Josue A. [Center for Advanced Surface Engineering, University of Arkansas, Fayetteville, AR 72701 (United States); Zou, Min, E-mail: mzou@uark.edu [Department of Mechanical Engineering, University of Arkansas, Fayetteville, AR 72701 (United States); Center for Advanced Surface Engineering, University of Arkansas, Fayetteville, AR 72701 (United States)

2017-08-01

Highlights: • Nanoindentation behavior of Al/a-Si core-shell nanostructures were studied. • 3D core confinement enables significant deformation recovery beyond elastic limit. • As the confinement is reduced, the deformation recovery is reduced or suppressed. • Atomistic simulations suggest core confinement affects dislocation dynamics. • 3D confinement has the highest percentage of dislocation removal after unloading. - Abstract: The nanoindentation behavior of hemispherical Al/a-Si core-shell nanostructures (CSNs), horizontally-aligned Al/a-Si core-shell nanorods (CSRs) with various lengths, and an Al/a-Si layered thin film has been studied to understand the effects of geometrical confinement of the Al core on the CSN deformation behavior. When loaded beyond the elastic limit, the CSNs have an unconventional load-displacement behavior with no residual displacement after unloading, resulting in no net shape change after indentation. This behavior is enabled by dislocation activities within the confined Al core, as indicated by discontinuous indentation signatures (load-drops and load-jumps) observed in the load-displacement data. When the geometrical confinement of the core is slightly reduced, as in the case of CSRs with the shortest rod length, the discontinuous indentation signatures and deformation resistance are heavily reduced. Further decreases in core confinement result in conventional nanoindentation behavior, regardless of geometry. Supporting molecular dynamics simulations show that dislocations nucleated in the core of a CSN are more effectively removed during unloading compared to CSRs, which supports the hypothesis that the unique deformation resistance of Al/a-Si CSNs are enabled by 3-dimensional confinement of the Al core.

Material dimensionality effects on the nanoindentation behavior of Al/a-Si core-shell nanostructures

International Nuclear Information System (INIS)

Fleming, Robert A.; Goss, Josue A.; Zou, Min

2017-01-01

Highlights: • Nanoindentation behavior of Al/a-Si core-shell nanostructures were studied. • 3D core confinement enables significant deformation recovery beyond elastic limit. • As the confinement is reduced, the deformation recovery is reduced or suppressed. • Atomistic simulations suggest core confinement affects dislocation dynamics. • 3D confinement has the highest percentage of dislocation removal after unloading. - Abstract: The nanoindentation behavior of hemispherical Al/a-Si core-shell nanostructures (CSNs), horizontally-aligned Al/a-Si core-shell nanorods (CSRs) with various lengths, and an Al/a-Si layered thin film has been studied to understand the effects of geometrical confinement of the Al core on the CSN deformation behavior. When loaded beyond the elastic limit, the CSNs have an unconventional load-displacement behavior with no residual displacement after unloading, resulting in no net shape change after indentation. This behavior is enabled by dislocation activities within the confined Al core, as indicated by discontinuous indentation signatures (load-drops and load-jumps) observed in the load-displacement data. When the geometrical confinement of the core is slightly reduced, as in the case of CSRs with the shortest rod length, the discontinuous indentation signatures and deformation resistance are heavily reduced. Further decreases in core confinement result in conventional nanoindentation behavior, regardless of geometry. Supporting molecular dynamics simulations show that dislocations nucleated in the core of a CSN are more effectively removed during unloading compared to CSRs, which supports the hypothesis that the unique deformation resistance of Al/a-Si CSNs are enabled by 3-dimensional confinement of the Al core.

Fast Flux Test Facility core restraint system performance

International Nuclear Information System (INIS)

Hecht, S.L.; Trenchard, R.G.

1990-02-01

Characterizing Fast Flux Test Facility (FFTF) core restraint system performance has been ongoing since the first operating cycle. Characterization consists of prerun analysis for each core load, in-reactor and postirradiation measurements of subassembly withdrawal loads and deformations, and using measurement data to fine tune predictive models. Monitoring FFTF operations and performing trend analysis has made it possible to gain insight into core restraint system performance and head off refueling difficulties while maximizing component lifetimes. Additionally, valuable information for improved designs and operating methods has been obtained. Focus is on past operating experience, emphasizing performance improvements and avoidance of potential problems. 4 refs., 12 figs., 2 tabs

Development of a core management tool for MYRRHA

International Nuclear Information System (INIS)

Jalůvka, David; Van den Eynde, Gert; Vandewalle, Stefan

2013-01-01

Highlights: • An in-core fuel management tool is being developed for the flexible irradiation machine MYRRHA. • Specific issues of the MYRRHA in-core fuel management are briefly discussed. • The tool addresses the loading pattern optimization problem. • Illustrative in-core fuel management optimization problems are solved using the tool. - Abstract: MYRRHA is an advanced multi-purpose irradiation facility under development at SCK• CEN in Mol, Belgium. In order to ensure an economical and safe operation of the reactor, an in-core fuel management tool is being developed within the project to address the loading pattern optimization problem. In the paper, the current version of the tool – its architecture and design, unique features, and the field of its application, are presented. In the second part of the paper, the tool’s capabilities are demonstrated on simple MYRRHA in-core fuel management optimization problems

Economic optimization of PWR cores with ROSA

International Nuclear Information System (INIS)

Verhagen, F.C.M.; Wakker, P.H.

2005-01-01

The core-loading pattern is decisive for fuel cycle economics, fuel safety parameters and economic planning for future cycles. ROSA, NRG's loading pattern optimization code system for PWRs, has proven for over a decade to be a valuable tool to reactor operators for improving their fuel management economics. ROSA uses simulated annealing as loading pattern optimization technique, in combination with an extremely fast 3-D neutronics code for loading pattern calculations. The code is continuously extended with new optimization parameters and rules. This paper outlines recent developments of the ROSA code system and discusses results of PWR specific applications of ROSA. Core designs with a large variety of challenging constraints have been realized with ROSA. As a typical example, for the 193 assembly, Vantage 5H/RFA-2 fueled TVA's Watts Bar unit 1, a cycle 4 core with 76 feed assemblies was designed. This was followed by a high-energy cycle 5 with only 77 feed assemblies and approximately 535 days of natural cycle length. Subsequently, an economical core using 72 bundles was designed for cycle 6. This resulted in considerable savings in the cost of feed assemblies for reloads. The typical accuracy of ROSA compared to results of license codes in within ±0.02 for normalized assembly powers, ±0.03 for maximum enthalpy rise hot channel factor (F ΔH ), and ±3 days for natural cycle length. (author)

Transient analysis of house load operation for LNPP

International Nuclear Information System (INIS)

Shi Junying; Zheng Bin

2000-01-01

The author analysis the transient of house load operation for Ling'ao Nuclear Power Plant by using the methods of dynamic simulation and closed loops of primary and secondary system. The transient of house load operation from 100% FP is the most severe that can occur on the unit in normal operation because it causes immediately shedding of 95% of turbine load and requires the unit to operate steadily at reduced power. The results show that the transient can be successful both at beginning of core life and manual house load operation. However, more attentions must be paid to automatic house load operation caused by grid fault at toward end of core life because the success of the transient could be threatened by the actuation of the protection of high flux and high flux rate

Containment loadings due to hydrogen burning in LWR core meltdown accidents

International Nuclear Information System (INIS)

Cybulskis, P.

1981-01-01

The potential pressure loadings due to hydrogen burning under conditions representative of meltdown accident conditions are examined for a variety of PWR and BWR containment designs. For the PWR, the large dry, ice condenser, as well as subatmospheric containments are considered. For the BWR, MARK I, II, and III pressure suppression containments are evaluated. The key factors considered are: free volume, design pressure, extend to hydrogen generation, and the flammability of the atmosphere under a range of accident conditions. The potential for and the possible implications of hydrogen detonation are also considered. The results of these analyses show that the accumulation and rapid burning of the quantities of hydrogen that would be generated during core meltdown accidents will lead to pressures above design levels in all of the containments considered. As would be expected, containments characterized by small volumes and/or low design pressures are the most vulnerable to damage due to hydrogen burning. Large volume, high pressure designs may also be threatened but offer significantly more potential for accomodating hydrogen burns. The attainment of detonable hydrogen mixtures is made easier by smaller containment volumes. Detonable mixtures are also possible in the larger volume containments, but imply the accumulation of hydrogen for long periods of time without prior ignition. Hydrogen detonations, if they occur, would probably challenge the integrity of any of the containments considered. (orig.)

Loading pattern optimization by multi-objective simulated annealing with screening technique

International Nuclear Information System (INIS)

Tong, K. P.; Hyun, C. L.; Hyung, K. J.; Chang, H. K.

2006-01-01

This paper presents a new multi-objective function which is made up of the main objective term as well as penalty terms related to the constraints. All the terms are represented in the same functional form and the coefficient of each term is normalized so that each term has equal weighting in the subsequent simulated annealing optimization calculations. The screening technique introduced in the previous work is also adopted in order to save computer time in 3-D neutronics evaluation of trial loading patterns. For numerical test of the new multi-objective function in the loading pattern optimization, the optimum loading patterns for the initial and the cycle 7 reload PWR core of Yonggwang Unit 4 are calculated by the simulated annealing algorithm with screening technique. A total of 10 optimum loading patterns are obtained for the initial core through 10 independent simulated annealing optimization runs. For the cycle 7 reload core one optimum loading pattern has been obtained from a single simulated annealing optimization run. More SA optimization runs will be conducted to optimum loading patterns for the cycle 7 reload core and results will be presented in the further work. (authors)

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

International Nuclear Information System (INIS)

Bouchard, J.

2000-01-01

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

Optimization of refueling loading pattern of uranium zirconium hydride research reactor

International Nuclear Information System (INIS)

Chen Wei; Xie Zhongsheng; Chen Da

1999-01-01

The orthogonal design method is used in the optimization of in-core fuel management. A code package of in-core fuel management in hexagonal geometry HEX-ORTH is developed. The loading pattern after the end of 3 cycle of Xi'an Pulsed Reactor is optimized using the HEX-ORTH. The optimistic loading pattern of the core are obtained as the objective function is Max(k eff BOC )

Dental responsibility loadings and the relative value of dental services.

Science.gov (United States)

Teusner, D N; Ju, X; Brennan, D S

2017-09-01

To estimate responsibility loadings for a comprehensive list of dental services, providing a standardized unit of clinical work effort. Dentists (n = 2500) randomly sampled from the Australian Dental Association membership (2011) were randomly assigned to one of 25 panels. Panels were surveyed by questionnaires eliciting responsibility loadings for eight common dental services (core items) and approximately 12 other items unique to that questionnaire. In total, loadings were elicited for 299 items listed in the Australian Dental Schedule 9th Edition. Data were weighted to reflect the age and sex distribution of the workforce. To assess reliability, regression models assessed differences in core item loadings by panel assignment. Estimated loadings were described by reporting the median and mean. Response rate was 37%. Panel composition did not vary by practitioner characteristics. Core item loadings did not vary by panel assignment. Oral surgery and endodontic service areas had the highest proportion (91%) of services with median loadings ≥1.5, followed by prosthodontics (78%), periodontics (76%), orthodontics (63%), restorative (62%) and diagnostic services (31%). Preventive services had median loadings ≤1.25. Dental responsibility loadings estimated by this study can be applied in the development of relative value scales. © 2017 Australian Dental Association.

Fuel assembly for FBR type reactor and reactor core thereof

International Nuclear Information System (INIS)

Kobayashi, Kaoru.

1998-01-01

The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

An automatic procedure for optimizing fuel loading in consideration of the effect of burnup nonuniformity in assembly

International Nuclear Information System (INIS)

Wang Guoli.

1988-01-01

The effect of burnup nonuniformity across the assembly on optimizing fuel loading in core is investigated. Some new rules which can be used for optimizing fuel loading in the core are proposed. New automatic procedure for optimizing fuel loading in the core is described

Optical techniques for in-core measurements

International Nuclear Information System (INIS)

Brichard, B.

2007-01-01

The in-situ measurement of dimensional changes is a key issue for advanced irradiation programs in Material Test Reactors. It is for example crucial to monitor the changes of the dimensions of nuclear fuel assemblies as well as those of mechanically stressed structural material samples during in-pile irradiations. Different techniques already exist to carry out such measurements but they all come with a number of drawbacks. SCK-CEN and CEA have therefore decided to share the development of a measurement system that was never applied before in the core of a nuclear reactor. It relies on optical dimensional measurements and brings along unprecedented non-intrusiveness combined with high resolution. A clear advantage in using compact optical sensors results in a more efficient occupation of the irradiation volume available for target testings as well as a significant reduction of the gamma-heating associated with the in-pile instrumentation. The objectives of these shared studies are to design, develop, test and qualify an in-pile dimensional measurement system based on optical techniques, with the goal to implement this system in future MTR irradiation experiments. In 2006, we focussed our activities on sensor analysis, selection of the sensor prototypes, procurement and first irradiation experiment

Toward full MOX core design

International Nuclear Information System (INIS)

Rouviere, G.; Guillet, J.L.; Bruna, G.B.; Pelet, J.

1999-01-01

This paper presents a selection of the main preliminary results of a study program sponsored by COGEMA and currently carried out by FRAMATOME. The objective of this study is to investigate the feasibility of full MOX core loading in a French 1300 MWe PWR, a recent and widespread standard nuclear power plant. The investigation includes core nuclear design, thermal hydraulic and systems aspects. (authors)

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)

ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.

Energy Technology Data Exchange (ETDEWEB)

Lell, R. M.; Morman, J. A.; Schaefer, R.W.; McKnight, R.D.; Nuclear Engineering Division

2009-02-23

ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide, U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split table machine) were replaced by Pu-U-Mo plates containing 27% {sup 240}Pu in the plutonium

A core-shell albumin copolymer nanotransporter for high capacity loading and two-step release of doxorubicin with enhanced anti-leukemia activity.

Science.gov (United States)

Wu, Yuzhou; Ihme, Susann; Feuring-Buske, Michaela; Kuan, Seah Ling; Eisele, Klaus; Lamla, Markus; Wang, Yanran; Buske, Christian; Weil, Tanja

2013-06-01

The native transportation protein serum albumin represents an attractive nano-sized transporter for drug delivery applications due to its beneficial safety profile. Existing albumin-based drug delivery systems are often limited by their low drug loading capacity as well as noticeable drug leakage into the blood circulation. Therefore, a unique albumin-derived core-shell doxorubicin (DOX) delivery system based on the protein denaturing-backfolding strategy was developed. 28 DOX molecules were covalently conjugated to the albumin polypeptide backbone via an acid sensitive hydrazone linker. Polycationic and pegylated human serum albumin formed two non-toxic and enzymatically degradable protection shells around the encapsulated DOX molecules. This core-shell delivery system possesses notable advantages, including a high drug loading capacity critical for low administration doses, a two-step drug release mechanism based on pH and the presence of proteases, an attractive biocompatibility and narrow size distribution inherited from the albumin backbone, as well as fast cellular uptake and masking of epitopes due to a high degree of pegylation. The IC50 of these nanoscopic onion-type micelles was found in the low nanomolar range for Hela cells as well as leukemia cell lines. In vivo data indicate its attractive potential as anti-leukemia treatment suggesting its promising profile as nanomedicine drug delivery system. Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Mechanical properties of chemically bonded sand core materials dipped in sol-gel coating impregnated with filter

DEFF Research Database (Denmark)

Nwaogu, Ugochukwu Chibuzoh; Tiedje, Niels Skat

2012-01-01

A novel sol-gel coating impregnated with filter dust was applied on chemically bonded sand core materials by dipping. After curing, the strengths of the core materials were measured under uniaxial loading using a new strength testing machine (STM). The STM presents the loading history as a force-...... of the chemically bonded sand core materials, a combination of flexural and compression tests is suggested for improving the casting quality. © 2012 W. S. Maney & Son Ltd.......A novel sol-gel coating impregnated with filter dust was applied on chemically bonded sand core materials by dipping. After curing, the strengths of the core materials were measured under uniaxial loading using a new strength testing machine (STM). The STM presents the loading history as a force...... the strengths were increased under compression. The mode of fracture of the chemically bonded sand core materials was observed to be intergranular through the binder. The stiffness of the chemically bonded sand core materials was determined. For better understanding of the mechanical properties...

In-core nuclear fuel management optimization of VVER1000 using perturbation theory

International Nuclear Information System (INIS)

Hosseini, Mohammad; Vosoughi, Naser

2011-01-01

In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain fuel integrity. Because of the numerous possible patterns of the fuel assemblies in the reactor core, finding the best configuration is so important and complex. Different methods for optimization of fuel loading pattern in the core have been introduced so far. In this study, a software is programmed in C ⧣ language to find an order of the fuel loading pattern of the VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process lunches by considering the initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. It shall be noticed that the designed algorithm is performed by just shuffling the fuel assemblies. The obtained results by employing the mentioned method on a typical reactor reveal that this method has a high precision in achieving a pattern with an allowable radial power peaking factor. (author)

Comparison of adherence to generic multi-tablet regimens vs. brand multi-tablet and brand single-tablet regimens likely to incorporate generic antiretroviral drugs by breaking or not fixed-dose combinations in HIV-infected patients.

Science.gov (United States)

Rwagitinywa, Joseph; Lapeyre-Mestre, Maryse; Bourrel, Robert; Montastruc, Jean-Louis; Sommet, Agnès

2018-03-05

Adherence to antiretroviral (ARV) is crucial to achieve viral load suppression in HIV-infected patients. This study aimed to compare adherence to generic multi-tablet regimens (MTR) vs. brand MTR likely to incorporate ARV drugs without breaking fixed-dose combinations (FDC) and brand single-tablet regimens (STR) likely to incorporate generics by breaking the FDC. Patients aged of 18 years or over exposed to one of the generic or the brand of lamivudine (3TC), zidovudine/lamivudine (AZT/TC), nevirapine (NVP), or efavirenz (EFV), or the brand STR of efavirenz/emtricitabine/tenofovir (EFV/FTC/TDF). Adherence was measured by medication possession ratio (MPR) using both defined daily dose (DDD) and daily number of tablet recommended for adults (DNT). Adherence to generic MTR vs. brand MTR and brand STR was compared using Kruskal-Wallis. The overall median adherence was 0.97 (IQR 0.13) by DNT method and 0.97 (0.14) by DDD method. Adherence in patients exposed to generic MTR (n = 165) vs. brand MTR (n = 481) and brand STR (n = 470) was comparable by DNT and DDD methods. In conclusion, adherence to generic MTR was high and comparable with adherence to brand MTR and to STR. Utilization of DDD instead DNT to measure the MPR led to small but nonsignificant difference that has no clinical impact. © 2018 Société Française de Pharmacologie et de Thérapeutique.

Sensitivity analysis of reflector types and impurities in a 10 MW MTR type nuclear research reactor

International Nuclear Information System (INIS)

Khattab, K.; Khamis, I.

2008-01-01

The 2-D and 3-D neutronics models for 10 MW nuclear research reactor of MTR type have been developed and presented in this paper. Our results agree very well with the results of seven countries mentioned in the IAEA-TECDOC-233. To study the effect of reflector types on the reactor effective multiplication factor, five types of reflectors such as pure beryllium, beryllium, heavy water, carbon and water are selected for this study. The pure beryllium is found to be the most efficient reflector in this group. The effect of the most important impurities, which exist on the beryllium reflector such as iron, silicon and aluminium on the reactor multiplication factor, have been analyzed as well. It is found that the iron impurity affects the reactor multiplication factor the most compared to silicon and aluminium impurities. (author)

In-core fuel management: New challenges

International Nuclear Information System (INIS)

Kolmayer, A.; Vallee, A.; Mondot, J.

1992-01-01

Experience accumulated by pressurized water reactor (PWR) utilities allows them to improve their strategies in the use of eventual margins to core design limits. They are used for nuclear steam supply system (NSSS) power upgrading, to improve operating margins, or to adapt fuel management to specific objectives. As a result, in-core fuel management strategies have become very diverse: UO 2 or mixed-oxide loading, out-in or in-out fuel loading patterns, extended or annual cycle lengths with margins on design limits such as moderator temperature coefficients, boron concentrations, or peaking factors. Perspectives also appear concerning use of existing plutonium stocks or actinide incineration. Burnable poisons are most often needed to satisfactorily achieve these goals. Among them, gadolinia are now largely used, owing to their excellent performance. More than 24 Framatome first cores and reloads, representing more than 3000 gadolinia-bearing rods, have been irradiated since 1983

Recent enhancements of the INSIGHT integrated in-core fuel management tool

International Nuclear Information System (INIS)

Akio, Yamamoto

2001-01-01

Recent enhancements of the INSIGHT system are described in this paper. The INSIGHT system is an integrated in-core fuel management tool for pressurized water reactors (PWRs) runs on UNIX workstations. The INSIGHT system provides various capabilities which contribute to reduce fuel cycle cost and workload of in-core fuel management tasks, i.e. core follow calculations, interactive loading pattern design, automated multicycle analysis and interface between detailed core calculation codes. To minimize engineers' workload, most of input data for analysis modules are automatically generated by the INSIGHT system through specification of calculation conditions in the graphic user interface. Recent enhancements of the INSIGHT system are mainly focused to improve efficiency of loading pattern optimization and flexibility of multicycle analyses. To increase optimization efficiency, a parallel calculation capability, various optimization theories, extension of heuristic rules, screening by neural networks and so on were incorporated in the loading pattern optimization module. The multicycle analyses module was rewritten to increase flexibility such as cycle dependent specification of loading pattern search methods and so on. The INSIGHT system is currently used by Japanese utilities not only for regular in-core fuel management tasks but also for strategic fuel management studies to reduce fuel cycle cost

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)

Exploratory Analysis of a GGSN’s PDP Context Signaling Load

Directory of Open Access Journals (Sweden)

Florian Metzger

2014-01-01

Full Text Available This paper takes an exploratory look on control plane signaling in a mobile cellular core network. In contrast to most contributions in this field, our focus does not lie on the wireless or user-oriented parts of the network, but on signaling in the core network. In an investigation of core network data we take a look at statistics related to GTP tunnels and their signaling. Based on the results thereof we propose a definition of load at the GGSN and create an initial load queuing model. We find signs of user devices putting burden on the core network through their behavior.

Application of MSHIM core control strategy for westinghouse AP1000 nuclear power plant

International Nuclear Information System (INIS)

Onoue, Masaaki; Kawanishi, Tomohiro; Carlson, William R.; Morita, Toshio

2003-01-01

Westinghouse has developed a new core control strategy, in which two independently moving Rod Cluster Control Assembly (RCCA) groups are utilized for core control; one group for reactivity/temperature control, the other for axial power distribution (Axial Offset) control. This control procedure eliminates the need for Chemical Shim adjustments during power maneuvers, such as load follow, and is designated MSHIM (Mechanical Shim). This core control strategy is utilized in the AP1000. In the AP1000, it is possible to perform MSHIM load follow maneuvers for up to 95% of cycle life without changing the soluble boron concentration in the moderator. This core control strategy has been evaluated, via computer simulations, to provide appropriate margins to core and fuel design limits during normal operation maneuvers (including load follow operations) and also during anticipated Condition II accident transients. The primary benefits of MSHIM as a control strategy are as follows; Power change operation can be totally automated due to the elimination of boron concentration adjustments. Full load follow capability is achievable for up to more than 95% of cycle life. Load follow operations performed solely by mechanical devices results in a significant reduction in the boron system requirements and a significant reduction in daily effluent to be processed. (author)

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)

Minor actinides incineration by loading moderated targets in fast reactor

International Nuclear Information System (INIS)

Wu Hongchun; Sato, Daisuke; Takeda, Toshikazu

2000-01-01

The effect of hydrogen concentration and loaded mass of minor actinides (MAs) in the target on the core performance and MAs transmutation rate was analyzed in this paper. An optimum core was proposed which has 96 MAs target assemblies of which MAs fuel pins per assembly is 38 with the composition ratio U/MA/Zr/H of 1/4/10/50. This optimized core offers good core performance and can transmute MAs very effectively, the transmutation rate was about 67% (939 kg) and the incinerate (transmute by fission) rate was about 35% (489 kg) through 3 years of reactor operation. It is about 2-3 times larger than current transmutation method that MAs are loaded homogeneously in the PWR and fast reactor core. (author)

Design of a reactor core in the Oma Full MOX-ABWR

International Nuclear Information System (INIS)

Hama, Teruo

1999-01-01

The Electric Power Development Co., Ltd. has progressed a construction plan on an improved boiling-water reactor aiming at loading of MOX fuel in all reactor cores (full MOX-ABWR) at Oma-cho, Aomori prefecture, which is a last stage on application of approval on establishment at present. Here were described on outlines of reactor core in the full MOX-ABWR and its safety evaluation. For the full MOX-ABWR loading MOX fuel assembly into all reactor core, thermal and mechanical design analysis of fuel bars and core design analysis were conducted. As a result, it was confirmed that judgement standards in mixed core of MOX fuel and uranium fuel were also applicable as well as that in uranium fuel. (G.K.)

Dynamic response of cylindrical ACS support structures to core energy release

International Nuclear Information System (INIS)

Kennedy, J.M.; Belytschko, T.B.

1985-01-01

The code SAFE/RAS is applied to the analysis of a new design concept for the above-core structures when subjected to the loads of a core disruptive accident. The analysis involves the determination of the postbuckling response of a thin cylinder loaded both axially and vertically. The effects of variation of cylinder thickness and fluid-structure interaction are investigated

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

Directory of Open Access Journals (Sweden)

Mateusz Kurzawski

2015-03-01

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

MCNP calculation of the critical H_3BO_3 concentrations for the first fuel loading into the reactor core of NPP MO-3-4 units

International Nuclear Information System (INIS)

Vrban, B.; Lueley, J.; Farkas, G.; Hascik, J.; Hinca, R.; Petriska, M.; Slugen, V.

2012-01-01

The purpose of the analysis was the determination of critical H_3BO_3 concentrations for the first fuel loading into the reactor core of MO34 units using 2"n"d generation fuel during the first start-up of new unit using calculation code MCNP 1.60. H_3BO_3 concentrations were computed for the given temperature of the primary circuit and position of the 6"t"h safety control rod group. Because of the very first start-up of these units, detailed analyses of active-core parameters are required by National Regulatory Authority and needed for safe operation of nuclear facility. (authors)

The PWR cores management

International Nuclear Information System (INIS)

Barral, J.C.; Rippert, D.; Johner, J.

2000-01-01

During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

Study on ant colony optimization for fuel loading pattern problem

International Nuclear Information System (INIS)

Kishi, Hironori; Kitada, Takanori

2013-01-01

Modified ant colony optimization (ACO) was applied to the in-core fuel loading pattern (LP) optimization problem to minimize the power peaking factor (PPF) in the modeled 1/4 symmetry PWR core. Loading order was found to be important in ACO. Three different loading orders with and without the adjacent effect between fuel assemblies (FAs) were compared, and it was found that the loading order from the central core is preferable because many selections of FAs to be inserted are available in the core center region. LPs were determined from pheromone trail and heuristic information, which is a priori knowledge based on the feature of the problem. Three types of heuristic information were compared to obtain the desirable performance of searching LPs with low PPF. Moreover, mutation operation, such as the genetic algorithm (GA), was introduced into the ACO algorithm to avoid searching similar LPs because heuristic information used in ACO tends to localize the searching space in the LP problem. The performance of ACO with some improvement was compared with those of simulated annealing and GA. In conclusion, good performance can be achieved by setting proper heuristic information and mutation operation parameter in ACO. (author)

Nonlinear seismic analysis of a graphite reactor core

International Nuclear Information System (INIS)

Laframboise, W.L.; Desmond, T.P.

1988-01-01

Design and construction of the Department of Energy's N-Reactor located in Richland, Washington was begun in the late 1950s and completed in the early 1960s. Since then, the reactor core's structural integrity has been under review and is considered by some to be a possible safety concern. The reactor core is moderated by graphite. The safety concern stems from the degradation of the graphite due to the effects of long-term irradiation. To assess the safety of the reactor core when subjected to seismic loads, a dynamic time-history structural analysis was performed. The graphite core consists of 89 layers of numerous graphite blocks which are assembled in a 'lincoln-log' lattice. This assembly permits venting of steam in the event of a pressure tube rupture. However, such a design gives rise to a highly nonlinear structure when subjected to earthquake loads. The structural model accounted for the nonlinear interlayer sliding and for the closure and opening of gaps between the graphite blocks. The model was subjected to simulated earthquake loading, and the time-varying response of selected elements critical to safety were monitored. The analytically predicted responses (displacements and strains) were compared to allowable responses to assess margins of safety. (orig.)

Effects of Two Types of Melatonin-Loaded Nanocapsules with Distinct Supramolecular Structures: Polymeric (NC) and Lipid-Core Nanocapsules (LNC) on Bovine Embryo Culture Model.

Science.gov (United States)

Komninou, Eliza Rossi; Remião, Mariana Härter; Lucas, Caroline Gomes; Domingues, William Borges; Basso, Andrea Cristina; Jornada, Denise Soledade; Deschamps, João Carlos; Beck, Ruy Carlos Ruver; Pohlmann, Adriana Raffin; Bordignon, Vilceu; Seixas, Fabiana Kömmling; Campos, Vinicius Farias; Guterres, Silvia Stanisçuaski; Collares, Tiago

2016-01-01

Melatonin has been used as a supplement in culture medium to improve the efficiency of in vitro produced mammalian embryos. Through its ability to scavenge toxic oxygen derivatives and regulate cellular mRNA levels for antioxidant enzymes, this molecule has been shown to play a protective role against damage by free radicals, to which in vitro cultured embryos are exposed during early development. In vivo and in vitro studies have been performed showing that the use of nanocapsules as active substances carriers increases stability, bioavailability and biodistribution of drugs, such as melatonin, to the cells and tissues, improving their antioxidant properties. These properties can be modulated through the manipulation of formula composition, especially in relation to the supramolecular structures of the nanocapsule core and the surface area that greatly influences drug release mechanisms in biological environments. This study aimed to evaluate the effects of two types of melatonin-loaded nanocapsules with distinct supramolecular structures, polymeric (NC) and lipid-core (LNC) nanocapsules, on in vitro cultured bovine embryos. Embryonic development, apoptosis, reactive oxygen species (ROS) production, and mRNA levels of genes involved in cell apoptosis, ROS and cell pluripotency were evaluated after supplementation of culture medium with non-encapsulated melatonin (Mel), melatonin-loaded polymeric nanocapsules (Mel-NC) and melatonin-loaded lipid-core nanocapsules (Mel-LNC) at 10-6, 10-9, and 10-12 M drug concentrations. The highest hatching rate was observed in embryos treated with 10-9 M Mel-LNC. When compared to Mel and Mel-NC treatments at the same concentration (10-9 M), Mel-LNC increased embryo cell number, decreased cell apoptosis and ROS levels, down-regulated mRNA levels of BAX, CASP3, and SHC1 genes, and up-regulated mRNA levels of CAT and SOD2 genes. These findings indicate that nanoencapsulation with LNC increases the protective effects of melatonin

Effects of Two Types of Melatonin-Loaded Nanocapsules with Distinct Supramolecular Structures: Polymeric (NC and Lipid-Core Nanocapsules (LNC on Bovine Embryo Culture Model.

Directory of Open Access Journals (Sweden)

Eliza Rossi Komninou

Full Text Available Melatonin has been used as a supplement in culture medium to improve the efficiency of in vitro produced mammalian embryos. Through its ability to scavenge toxic oxygen derivatives and regulate cellular mRNA levels for antioxidant enzymes, this molecule has been shown to play a protective role against damage by free radicals, to which in vitro cultured embryos are exposed during early development. In vivo and in vitro studies have been performed showing that the use of nanocapsules as active substances carriers increases stability, bioavailability and biodistribution of drugs, such as melatonin, to the cells and tissues, improving their antioxidant properties. These properties can be modulated through the manipulation of formula composition, especially in relation to the supramolecular structures of the nanocapsule core and the surface area that greatly influences drug release mechanisms in biological environments. This study aimed to evaluate the effects of two types of melatonin-loaded nanocapsules with distinct supramolecular structures, polymeric (NC and lipid-core (LNC nanocapsules, on in vitro cultured bovine embryos. Embryonic development, apoptosis, reactive oxygen species (ROS production, and mRNA levels of genes involved in cell apoptosis, ROS and cell pluripotency were evaluated after supplementation of culture medium with non-encapsulated melatonin (Mel, melatonin-loaded polymeric nanocapsules (Mel-NC and melatonin-loaded lipid-core nanocapsules (Mel-LNC at 10-6, 10-9, and 10-12 M drug concentrations. The highest hatching rate was observed in embryos treated with 10-9 M Mel-LNC. When compared to Mel and Mel-NC treatments at the same concentration (10-9 M, Mel-LNC increased embryo cell number, decreased cell apoptosis and ROS levels, down-regulated mRNA levels of BAX, CASP3, and SHC1 genes, and up-regulated mRNA levels of CAT and SOD2 genes. These findings indicate that nanoencapsulation with LNC increases the protective effects of

BNL program in support of LWR degraded-core accident analysis

International Nuclear Information System (INIS)

Ginsberg, T.; Greene, G.A.

1982-01-01

Two major sources of loading on dry watr reactor containments are steam generatin from core debris water thermal interactions and molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON (Muir, 1981) and MARCH (Wooton, 1980) computer codes. Progress in the two programs is described in this paper. 8 figures

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)

Vibrational characterization of hexagonal duct core assemblies under various support conditions

International Nuclear Information System (INIS)

Bartholf, L.W.; Julyk, L.J.; Ryan, J.A.

1989-03-01

Analysis of the dynamic response of advanced Liquid Metal Reactor (LMR) core internals to seismic excitation requires a significant number of simplifying assumptions and idealizations to economically meet the constraints of present-day computer limitations. Fluid coupling and nonlinearities associated with inter-assembly lateral support stiffness and clearances of a large cluster of core internal assemblies are some of the factors that complicate the analytical procedure (Moran, 1976). Well defined test data were needed to quantify these and other uncertainties associated with the use of analytical or numerical computer codes used in the seismic design and analysis of reactor cores. The purpose of the present experimental program was to supplement existing data, such as reported in (Sasaki and Muto, 1983), by developing vibrational characteristics of core assemblies over a range of parameters relative to LMR conceptual designs. The parameters selected for this program were variations in number and location of restraints, restraint-pad to duct-load-pad clearances, and input forcing frequency and g-level. Feature tests were conducted to characterize load pad stiffness and coefficient of restitution, and to calibrate load pads to measure inter-assembly across-flat impact loads. Simulated full-size LMR hexagonal duct core assemblies were used in vibration tests. A single assembly and a row of five assemblies were tested in air to establish modal characteristics and forced response behavior. 2 refs., 7 figs., 1 tab

Transjugular Core Liver Biopsy with a 19-Gauge Spring-Loaded Cutting Needle

International Nuclear Information System (INIS)

Choh, Jeffery; Dolmatch, Bart; Safadi, Rami; Long, Phil; Geisinger, Michael; Lammert, Gary; Dempsey, James

1998-01-01

One hundred and five sequential transjugular core liver biopsies (TJLBx) were performed in 101 patients with coagulopathy and/or ascites using the 19-gauge Quick-Core Biopsy (QCB) needle. Two-hundred and seventy-three cores were obtained in 295 passes (92.5%). One-hundred and two of the 105 procedures (97.1%) led to a histopathologic diagnosis. One of the three nondiagnostic biopsies was done because of severe autolysis of the liver. There was one subcapsular hematoma, one hepatic arteriovenous fistula, and one liver capsular puncture. Two minor neck hematomas occurred. One death was reported (unrelated to the procedure). QCB needle TJLBx is an effective and relatively safe way to obtain core liver samples

Process and equipment for monitoring flux distribution in a nuclear reactor outside the core

International Nuclear Information System (INIS)

Graham, K.F.; Gopal, R.

1977-01-01

This concerns the monitoring system for axial flux distribution during the whole load operating range lying outside the core of, for example, a PWR. Flux distribution cards can be produced continuously. The core is divided into at least three sections, which are formed by dividing it at right angles to the longitudinal axis, and the flux is measured outside the core using adjacent detectors. Their output signals are calibrated by amplifiers so that the load distribution in the associated sections is reproduced. A summation of the calibrated output signals and the formation of a mean load signal takes place in summing stages. For monitoring, this is compared with a value which corresponds to the maximum permissible load setting. Apart from this the position of the control rods in the core can be taken into account by multiplication of the mean load signals by suitable peak factors. The distribution of monitoring positions or the position of the detectors can be progressive or symmetrical along the axis. (DG) 891 HP [de

Effect of varying geometrical parameters of trapezoidal corrugated-core sandwich structure

Directory of Open Access Journals (Sweden)

Zaid N.Z.M.

2017-01-01

Full Text Available Sandwich structure is an attractive alternative that increasingly used in the transportation and aerospace industry. Corrugated-core with trapezoidal shape allows enhancing the damage resistance to the sandwich structure, but on the other hand, it changes the structural response of the sandwich structure. The aim of this paper is to study the effect of varying geometrical parameters of trapezoidal corrugated-core sandwich structure under compression loading. The corrugated-core specimen was fabricated using press technique, following the shape of trapezoidal shape. Two different materials were used in the study, glass fibre reinforced plastic (GFRP and carbon fibre reinforced plastic (CFRP. The result shows that the mechanical properties of the core in compression loading are sensitive to the variation of a number of unit cells and the core thickness.

Assessing and Reducing Miscellaneous Electric Loads (MELs) in Lodging

Energy Technology Data Exchange (ETDEWEB)

Rauch, Emily M.

2011-09-01

Miscellaneous electric loads (MELs) are the loads outside of a building's core functions of heating, ventilating, air conditioning, lighting, and water heating. This report reviews methods to reduce MELs in lodging.

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

International Nuclear Information System (INIS)

Fridman, E.; Kliem, S.

2011-01-01

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

Preliminary considerations on the startup phase for the ASTRID core

International Nuclear Information System (INIS)

Mignot, G.

2015-01-01

This paper presents preliminary considerations on the startup phase for the ASTRID core, as well as an overview of the different steps before reaching the optimised equilibrium core. The start-up phase is assumed to cover the period between loading the dummy core into the reactor (for commissioning tests) and achieving the optimised equilibrium core. Four main stages are considered: a first stage of start-up tests before fuel core loading, a second stage related to zero power and power ramp-up tests, a third stage corresponding to the transition from the first core to the equilibrium contractual core, and the last stage to reach the optimised performance for the equilibrium core. In the two last stages, a sub-assembly surveillance plan based on post-irradiation examinations is taken into account. As this work is in its preliminary stages, the first scenarios shown for the start-up phase must not be considered as the ASTRID reference scenarios. The scenarios strongly depend on the assumptions considered in the analysis, whereas those discussed in this paper aim at outlining the content and the duration of the starting phases for the ASTRID core, which will be useful in subsequently assessing the core sub-assembly fabrication needs. Assumptions for the start-up phase will be updated in accordance with progress on the ASTRID core design development and core qualification programme. (author)

Application of the robust design concept for fuel loading pattern

International Nuclear Information System (INIS)

Endo, Tomohiro; Ohori, Kazuma; Yamamoto, Akio

2011-01-01

Application of the robust design concept for fuel loading pattern design is proposed as a new approach to improve the prediction accuracy of core characteristics. The robust design is a design concept that establishes a resistant (robust) system for perturbations or noises, by properly setting design variables. In order to apply the concept of robust design to fuel loading pattern design, we focus on a theoretical approach based on the higher order perturbation method. This approach indicates that the eigenvalue separation is one of the effective indices to measure the robustness of a designed fuel loading pattern. In order to verify the effectiveness of the eigenvalue separation as an index of robustness, numerical analysis is carried out for typical 3-loop PWR cores, and we evaluated the correlation between the eigenvalue separation and the variation of relative assembly power due to the perturbation of the cross section. The numerical results show that the variation of relative power decreases as the eigenvalue separation increases; thus, it is confirmed that the eigenvalue separation is an effective index of robustness. Based on the eigenvalue separation of a fuel loading pattern, we discuss design guidelines of a fuel loading pattern to improve the robustness. For example, if each fuel assembly has independent uncertainty on its cross section, the robustness of the core can be enhanced by increasing the relative power at the center of the core. The proposed guidelines will be useful to design a loading pattern that has robustness for uncertainties due to cross section, calculation method, and so on. (author)

Design studies for the Mark-III core of experimental multi-purpose VHTR

International Nuclear Information System (INIS)

Yasuno, Takehiko; Miyamoto, Yoshiaki; Mitake, Susumu; Shindo, Ryuiti; Arai, Taketoshi

1979-08-01

The Mark-III core in the first conceptual design made in 1975 is a fundamental core for VHTR. Subsequently, further design studies were made fuel loading scheme and control rod withdrawal sequence for the core to increase its safety margin (shutdown margin, etc.) and operational margin (minimum Reynolds number, maximum fuel temperature, etc.). It was shown that the Mark-III should exhibit the performance expected of VHTR, unless changes are made in the preconditions for its nuclear, thermal-hydraulic design. Also, the needs as below were indicated: (1) reasonable core design criteria and guidelines, (2) fuel-loading-scheme requirements in fuel management, fuel misloading and reactor operation, (3) confirmation on precision of the core design method and its further refinement. (author)

Fuel assemblies mechanical behaviour improvements based on design changes and loading patterns computational analyses

International Nuclear Information System (INIS)

Marin, J.; Aullo, M.; Gutierrez, E.

2001-01-01

In the past few years, incomplete RCCA insertion events (IRI) have been taking place at some nuclear plants. Large guide thimble distortion caused by high compressive loads together with the irradiation induced material creep and growth, is considered as the primary cause of those events. This disturbing phenomenon is worsened when some fuel assemblies are deformed to the extent that they push the neighbouring fuel assemblies and the distortion is transmitted along the core. In order to better understand this mechanism, ENUSA has developed a methodology based on finite element core simulation to enable assessments on the propensity of a given core loading pattern to propagate the distortion along the core. At the same time, the core loading pattern could be decided interacting with nuclear design to obtain the optimum response under both, nuclear and mechanical point of views, with the objective of progressively attenuating the core distortion. (author)

The adequacy of methods used for the approval of high burnup core loading

International Nuclear Information System (INIS)

Sonnenburg, H.G.

2002-01-01

New fuel assembly designs and new core loading strategies are foreseen by most utilities, optimising the use of nuclear fuel in nuclear power plants. Increasing the burn-up to high values above 50 MWd/kg affects the fuel and cladding conditions, which could have safety relevant consequences. It is the task of the safety authorities to assess the impact of these changes with respect to compliance with safety regulations. Usually this assessment is based on code analyses which contain models developed at a time when the burn-up was significantly lower. Because the high burn-up is accompanied with the development of new phenomena like the rim effect on fuel pellets, the codes' models need to be revised for the representation of these new phenomena. The objective of this paper is to present a review of the knowledge base of the fuel phenomena under high-burn-up conditions as seen from safety aspects. The safety relevant fuel rod phenomena will be discussed. It will further provide an assessment of the limitations of the methodologies so far applied in the context of LOCA and RIA transients. The recently started research activities in Germany to improve the methodologies will be presented. (author)

Design of full MOX core in ABWR

International Nuclear Information System (INIS)

Kinoshita, Y.; Hirose, T.; Sasagawa, M.; Sakuma, T

1999-01-01

A Full MOX-ABWR, loaded with mixed-oxide (MOX) fuels of up to 100% of the core, is planned. Increased MOX fuel utilization will result in greater savings of uranium. Studies on the fuel rod thermal-mechanical design, the core design and the safety evaluation have been made, and the results are summarized in this paper. To sum it all up, the safety of the Full MOX-ABWR has been confirmed through design evaluations adequately considering the MOX fuel and core characteristics. (author)

Face/core mixed mode debond fracture toughness characterization using the modified TSD test method

DEFF Research Database (Denmark)

Berggreen, Christian; Quispitupa, Amilcar; Costache, Andrei

2014-01-01

The modified tilted sandwich debond (TSD) test method is used to examine face/core debond fracture toughness of sandwich specimens with glass/polyester face sheets and PVC H45 and H100 foam cores over a large range of mode-mixities. The modification was achieved by reinforcing the loaded face sheet....... The fracture process was inspected visually during and after testing. For specimens with H45 core the crack propagated in the core. For specimens with an H100 core, the crack propagated between the resin-rich layer and the face sheet. © The Author(s) 2013 Reprints and permissions: sagepub...... with a steel bar, and fracture testing of the test specimens was conducted over a range of tilt angles. The fracture toughness exhibited mode-mixity phase angle dependence, especially for mode II dominated loadings; although, the fracture toughness remained quite constant for mode I dominated crack loadings...

Advantages of iron core in a tokamak

International Nuclear Information System (INIS)

Bettis, E.S.; Ballou, J.K.; Becraft, W.R.; Peng, Y.K.M.; Watts, H.L.

1977-01-01

A quantitative comparison of the iron core vs air core concepts was carried out on a preliminary basis by using a representative tokamak reactor design with the following self-consistent reference parameters. In the area of plasma engineering, poloidal field and MHD equilibrium considerations with an unsaturated iron core is discussed. The question of proper poloidal field coils to maintain D-shaped plasmas of relatively high anti β (7%) with a saturated iron core is also discussed. Estimates of the required iron core size, volt seconds, magnetic flux and its influence on force loading on the superconducting toroidal field coils are shown. Conceptual designs of the mechanical structure of an iron core device are presented. Favorable impacts on the OH power supply cost and complexity are indicated

A Minimum Shuffle Core Design Strategy for ESBWR

International Nuclear Information System (INIS)

Karve, A.A.; Fawcett, R.M.

2008-01-01

The Economic Simplified Boiling Water Reactor (ESBWR) is GEH's next evolution of advanced BWR technology. There are 1132 fuel bundles in the core and the thermal power is 4500 MWt. Similar to conventional plants there is an outage after a specified period of operation, when the plant shuts down. During the outage a specified fraction of fuel bundles are discharged from the core, it is loaded with the same fraction of fresh fuel, and fuel is shuffled to obtain an optimum core design that meets the goals for a successful operation of the next cycle. The discharge, load, and the associated shuffles are time-consuming and expensive tasks that impact the overall outage schedule and costs. Therefore, there is an incentive to keep maneuvers to a minimum and to perform them more efficiently. The benefits for a large core, such as the ESBWR with 1132 fuel bundles, are escalated. This study focuses on a core reload design strategy to minimize the total number of shuffles during an outage. A traditional equilibrium cycle is used as a reference basis, which sets the reference number of shuffles. In the minimum shuffle core design however, a set of two equilibrium cycles (N and N+1, referred to as a 'bi- equilibrium' cycle) is envisioned where the fresh fuel of cycle N (that becomes the once-burnt fuel of cycle N+1) ideally does not move in the two cycles. The cost of fuel efficiency is determined for obtaining such a core loading by comparing it to the traditional equilibrium cycle. There are several additional degrees of freedom when designing a bi-equilibrium cycle that could be utilized, and the potential benefits of these flexibilities are assessed. In summary, the feasibility of a minimum shuffle fuel cycle and core design for an ESBWR is studied. The cost of fuel efficiency is assessed in comparison to the traditional design. (authors)

Research on Shock Responses of Three Types of Honeycomb Cores

Science.gov (United States)

Peng, Fei; Yang, Zhiguang; Jiang, Liangliang; Ren, Yanting

2018-03-01

The shock responses of three kinds of honeycomb cores have been investigated and analyzed based on explicit dynamics analysis. According to the real geometric configuration and the current main manufacturing methods of aluminum alloy honeycomb cores, the finite element models of honeycomb cores with three different cellular configurations (conventional hexagon honeycomb core, rectangle honeycomb core and auxetic honeycomb core with negative Poisson’s ratio) have been established through FEM parametric modeling method based on Python and Abaqus. In order to highlight the impact response characteristics of the above three honeycomb cores, a 5 mm thick panel with the same mass and material was taken as contrast. The analysis results showed that the peak values of longitudinal acceleration history curves of the three honeycomb cores were lower than those of the aluminum alloy panel in all three reference points under the loading of a longitudinal pulse pressure load with the peak value of 1 MPa and the pulse width of 1 μs. It could be concluded that due to the complex reflection and diffraction of stress wave induced by shock in honeycomb structures, the impact energy was redistributed which led to a decrease in the peak values of the longitudinal acceleration at the measuring points of honeycomb cores relative to the panel.

Nanoengineering of methylene blue loaded silica encapsulated magnetite nanospheres and nanocapsules for photodynamic therapy

Energy Technology Data Exchange (ETDEWEB)

Andhariya, Nidhi [Bhavnagar University, Department of Physics (India); Chudasama, Bhupendra, E-mail: bnchudasama@gmail.com [Thapar University, School of Physics and Materials Science (India); Mehta, R. V. [Bhavnagar University, Department of Physics (India); Upadhyay, R. V. [Charotar University of Science and Technology, P.D. Patel Institute of Applied Sciences (India)

2011-09-15

Core-shell nanostructures have emerged as an important class of functional materials with potential applications in diverse fields, especially in health sciences. In this article, nanoengineering of novel magnetic colloidal dispersion containing surface modifiable silica with a core of single domain magnetite nanoparticles loaded with photosensitizer (PS) drug 'Methylene blue' (MB) has been described. Magnetite core is produced by the well-established chemical coprecipitation technique and silica shell is formed over it by the modified hydrolysis and condensation of TEOS (tetraethyl orthosilicate). Conditions for reaction kinetics have been established to tailor the core-shell structures in the form of nanospheres and nanocapsules. MB is loaded into the nanostructures by demethylation reaction. The major conclusion drawn from this study is that the synthesis route yields stable, non-aggregated MB loaded superparamagnetic magnetite-silica nanostructures with tailored morphology, tunable loading, and excellent magnetic properties.

An expert system for PWR core operation management

Energy Technology Data Exchange (ETDEWEB)

Ida, Toshio; Masuda, Masahiro; Nishioka, Hiromasa

1988-01-01

Planning for restartup after planned or unplanned reactor shutdown and load-follow operations is an important task in the core operation management of pressurized water reactors (PWRs). These planning problems have been solved by planning experts using their expertise and the computational prediction of core behavior. Therefore, the quality of the plan and the time consumed in the planning depend heavily on the skillfulness of the planning experts. A knowledge engineering approach has been recently considered as a promising means to solve such complicated planning problems. Many knowledge-based systems have been developed so far, and some of them have already been applied because of their effectiveness. The expert system REPLEX has been developed to aid core management engineers in making a successful plan for the restartup or the load-follow operation of PWRs within a shorter time. It can maintain planning tasks at a high-quality level independent of the skillfulness of core management engineers and enhance the efficiency of management. REPLEX has an explanation function that helps user understanding of plans. It could be a useful took, therefore, for the training of core management engineers.

An expert system for PWR core operation management

International Nuclear Information System (INIS)

Ida, Toshio; Masuda, Masahiro; Nishioka, Hiromasa.

1988-01-01

Planning for restartup after planned or unplanned reactor shutdown and load-follow operations is an important task in the core operation management of pressurized water reactors (PWRs). These planning problems have been solved by planning experts using their expertise and the computational prediction of core behavior. Therefore, the quality of the plan and the time consumed in the planning depend heavily on the skillfulness of the planning experts. A knowledge engineering approach has been recently considered as a promising means to solve such complicated planning problems. Many knowledge-based systems have been developed so far, and some of them have already been applied because of their effectiveness. The expert system REPLEX has been developed to aid core management engineers in making a successful plan for the restartup or the load-follow operation of PWRs within a shorter time. It can maintain planning tasks at a high-quality level independent of the skillfulness of core management engineers and enhance the efficiency of management. REPLEX has an explanation function that helps user understanding of plans. It could be a useful took, therefore, for the training of core management engineers

FRG compact core - one year experience

International Nuclear Information System (INIS)

Knop, W.; Schreiner, P.

2001-01-01

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

Core-shell microparticles for protein sequestration and controlled release of a protein-laden core.

Science.gov (United States)

Rinker, Torri E; Philbrick, Brandon D; Temenoff, Johnna S

2017-07-01

Development of multifunctional biomaterials that sequester, isolate, and redeliver cell-secreted proteins at a specific timepoint may be required to achieve the level of temporal control needed to more fully regulate tissue regeneration and repair. In response, we fabricated core-shell heparin-poly(ethylene-glycol) (PEG) microparticles (MPs) with a degradable PEG-based shell that can temporally control delivery of protein-laden heparin MPs. Core-shell MPs were fabricated via a re-emulsification technique and the number of heparin MPs per PEG-based shell could be tuned by varying the mass of heparin MPs in the precursor PEG phase. When heparin MPs were loaded with bone morphogenetic protein-2 (BMP-2) and then encapsulated into core-shell MPs, degradable core-shell MPs initiated similar C2C12 cell alkaline phosphatase (ALP) activity as the soluble control, while non-degradable core-shell MPs initiated a significantly lower response (85+19% vs. 9.0+4.8% of the soluble control, respectively). Similarly, when degradable core-shell MPs were formed and then loaded with BMP-2, they induced a ∼7-fold higher C2C12 ALP activity than the soluble control. As C2C12 ALP activity was enhanced by BMP-2, these studies indicated that degradable core-shell MPs were able to deliver a bioactive, BMP-2-laden heparin MP core. Overall, these dynamic core-shell MPs have the potential to sequester, isolate, and then redeliver proteins attached to a heparin core to initiate a cell response, which could be of great benefit to tissue regeneration applications requiring tight temporal control over protein presentation. Tissue repair requires temporally controlled presentation of potent proteins. Recently, biomaterial-mediated binding (sequestration) of cell-secreted proteins has emerged as a strategy to harness the regenerative potential of naturally produced proteins, but this strategy currently only allows immediate amplification and re-delivery of these signals. The multifunctional, dynamic

Load bearing capacities and elastic-plastic behavior of reactor vessel internals

International Nuclear Information System (INIS)

Watanabe, Keita; Nagase, Ryuichi

2017-01-01

Radial Support Keys (RSKs) are installed at the bottom of Reactor Vessel Internal (RVI) of Pressurized Water Reactor (PWR) and fit into Core Support Lugs of Reactor Pressure Vessel (RPV). This structure provides reactor core horizontal support and transmits the loads between RVI and RPV. RSK is one of the critical parts of RVI from the view point of earthquake-proof safety. In order to assure the structural integrity of Nuclear Reactor in case of massive earthquake, load bearing capacities of RSK are confirmed by static loading tests with reduced-scale mockups. In addition, collapse loads of actual components calculated by Limit Analyses are conservative enough compared to the load bearing capacities confirmed by the test. Thus, the methodology to calculate collapse load by Limit Analysis is applicable to evaluation of structural integrity for RSK. (author)

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Assessment of MRI Parameters as Imaging Biomarkers for Radiation Necrosis in the Rat Brain

Energy Technology Data Exchange (ETDEWEB)

Wang Silun [Department of Radiology, Johns Hopkins University School of Medicine, Baltimore, MD (United States); Tryggestad, Erik [Department of Radiation Oncology, Johns Hopkins University School of Medicine, Baltimore, MD (United States); Zhou Tingting [Department of Radiology, Johns Hopkins University School of Medicine, Baltimore, MD (United States); Armour, Michael [Department of Radiation Oncology, Johns Hopkins University School of Medicine, Baltimore, MD (United States); Wen Zhibo [Department of Radiology, Johns Hopkins University School of Medicine, Baltimore, MD (United States); Department of Radiology, Zhujiang Hospital, Southern Medical University, Guangzhou, Guangdong (China); Fu Dexue [Department of Oncology, Johns Hopkins University School of Medicine, Baltimore, MD (United States); Ford, Eric [Department of Radiation Oncology, Johns Hopkins University School of Medicine, Baltimore, MD (United States); Zijl, Peter C.M. van [Department of Radiology, Johns Hopkins University School of Medicine, Baltimore, MD (United States); F. M. Kirby Research Center for Functional Brain Imaging, Kennedy Krieger Institute, Baltimore, Maryland (United States); Zhou Jinyuan, E-mail: jzhou@mri.jhu.edu [Department of Radiology, Johns Hopkins University School of Medicine, Baltimore, MD (United States); F. M. Kirby Research Center for Functional Brain Imaging, Kennedy Krieger Institute, Baltimore, Maryland (United States)

2012-07-01

Purpose: Radiation necrosis is a major complication of radiation therapy. We explore the features of radiation-induced brain necrosis in the rat, using multiple MRI approaches, including T{sub 1}, T{sub 2}, apparent diffusion constant (ADC), cerebral blood flow (CBF), magnetization transfer ratio (MTR), and amide proton transfer (APT) of endogenous mobile proteins and peptides. Methods and Materials: Adult rats (Fischer 344; n = 15) were irradiated with a single, well-collimated X-ray beam (40 Gy; 10 Multiplication-Sign 10 mm{sup 2}) in the left brain hemisphere. MRI was acquired on a 4.7-T animal scanner at {approx}25 weeks' postradiation. The MRI signals of necrotic cores and perinecrotic regions were assessed with a one-way analysis of variance. Histological evaluation was accomplished with hematoxylin and eosin staining. Results: ADC and CBF MRI could separate perinecrotic and contralateral normal brain tissue (p < 0.01 and < 0.05, respectively), whereas T{sub 1}, T{sub 2}, MTR, and APT could not. MRI signal intensities were significantly lower in the necrotic core than in normal brain for CBF (p < 0.001) and APT (p < 0.01) and insignificantly higher or lower for T{sub 1}, T{sub 2}, MTR, and ADC. Histological results demonstrated coagulative necrosis within the necrotic core and reactive astrogliosis and vascular damage within the perinecrotic region. Conclusion: ADC and CBF are promising imaging biomarkers for identifying perinecrotic regions, whereas CBF and APT are promising for identifying necrotic cores.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Loading procedures for shipment of irradiated fuel

Energy Technology Data Exchange (ETDEWEB)

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

1974-07-01

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

Loading procedures for shipment of irradiated fuel

International Nuclear Information System (INIS)

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

1974-01-01

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

Stress analysis of fuel assemblies under seismic load

International Nuclear Information System (INIS)

Kiselev, A.; Krutko, E.; Kiselev, I.; Tutnov, A.

2011-01-01

One of the important parts of fuel assemblies (FA) safety validation is their strength estimation under the dynamic loads, such as the vibration effects caused by the work of reactor units and the seismic exposure of an earthquake, leading to extreme inertia loads on all elements of the NPP. Taking into account structural features of FA and a very large mass, the exposure of seismic loads can lead to significant deformation of fuel assemblies. It is necessary to assess the magnitude of the force interaction between the FA in case of an earthquake to estimate the strength and performance of fuel assemblies. It is also necessary to compute FA bending forms and maximum values for further RPS control rods inserting time estimation, and for disassembly possibility justification of the core and individual FA after the earthquake. The problem of WWER-1000 core dynamic behavior modeling with TVS-2M fuel assemblies under the seismic loads exposure using the finite element method is described. Each fuel assembly is represented by equivalent rod finite element model. The reactor core is simulated by 163 fuel assemblies in accordance with the reactor core construction. Stiffness characteristics of fuel assemblies are determined on the results of a series of static and dynamic TVS-2M FA field tests. The special algorithm was developed to consider the fuel rod slippage effect during deformation. The special contact elements are introduced into the model of the core to take into account the interaction of fuel assemblies with their neighbors and with core barrel. Solution of the dynamic equilibrium equations system of finite element model is implemented by direct integration using the explicit scheme. Parallel algorithms for numerical integration on multiprocessor computers with graphics processing unit is developed to improve the efficiency of calculations. Values of nodes displacement in finite element model of reactor core as a function of seismic excitation time are obtained

Continuous firefly algorithm applied to PWR core pattern enhancement

International Nuclear Information System (INIS)

Poursalehi, N.; Zolfaghari, A.; Minuchehr, A.; Moghaddam, H.K.

2013-01-01

Highlights: ► Numerical results indicate the reliability of CFA for the nuclear reactor LPO. ► The major advantages of CFA are its light computational cost and fast convergence. ► Our experiments demonstrate the ability of CFA to obtain the near optimal loading pattern. -- Abstract: In this research, the new meta-heuristic optimization strategy, firefly algorithm, is developed for the nuclear reactor loading pattern optimization problem. Two main goals in reactor core fuel management optimization are maximizing the core multiplication factor (K eff ) in order to extract the maximum cycle energy and minimizing the power peaking factor due to safety constraints. In this work, we define a multi-objective fitness function according to above goals for the core fuel arrangement enhancement. In order to evaluate and demonstrate the ability of continuous firefly algorithm (CFA) to find the near optimal loading pattern, we developed CFA nodal expansion code (CFANEC) for the fuel management operation. This code consists of two main modules including CFA optimization program and a developed core analysis code implementing nodal expansion method to calculate with coarse meshes by dimensions of fuel assemblies. At first, CFA is applied for the Foxholes test case with continuous variables in order to validate CFA and then for KWU PWR using a decoding strategy for discrete variables. Results indicate the efficiency and relatively fast convergence of CFA in obtaining near optimal loading pattern with respect to considered fitness function. At last, our experience with the CFA confirms that the CFA is easy to implement and reliable

Continuous firefly algorithm applied to PWR core pattern enhancement

Energy Technology Data Exchange (ETDEWEB)

Poursalehi, N., E-mail: npsalehi@yahoo.com [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of); Zolfaghari, A.; Minuchehr, A.; Moghaddam, H.K. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of)

2013-05-15

Highlights: ► Numerical results indicate the reliability of CFA for the nuclear reactor LPO. ► The major advantages of CFA are its light computational cost and fast convergence. ► Our experiments demonstrate the ability of CFA to obtain the near optimal loading pattern. -- Abstract: In this research, the new meta-heuristic optimization strategy, firefly algorithm, is developed for the nuclear reactor loading pattern optimization problem. Two main goals in reactor core fuel management optimization are maximizing the core multiplication factor (K{sub eff}) in order to extract the maximum cycle energy and minimizing the power peaking factor due to safety constraints. In this work, we define a multi-objective fitness function according to above goals for the core fuel arrangement enhancement. In order to evaluate and demonstrate the ability of continuous firefly algorithm (CFA) to find the near optimal loading pattern, we developed CFA nodal expansion code (CFANEC) for the fuel management operation. This code consists of two main modules including CFA optimization program and a developed core analysis code implementing nodal expansion method to calculate with coarse meshes by dimensions of fuel assemblies. At first, CFA is applied for the Foxholes test case with continuous variables in order to validate CFA and then for KWU PWR using a decoding strategy for discrete variables. Results indicate the efficiency and relatively fast convergence of CFA in obtaining near optimal loading pattern with respect to considered fitness function. At last, our experience with the CFA confirms that the CFA is easy to implement and reliable.

A report on the first loading of ZENITH

Energy Technology Data Exchange (ETDEWEB)

Absalom, R M; Barclay, F R; Cameron, I R; Della Loggia, V E; Hage, W; Kinchin, G H; Sanders, J E; Tiren, L I; Wilson, D J [Industrial Power Reactors Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1960-12-15

The loading to criticality of the first core of the zero energy reactor ZENITH, and the results of the subsequent experimental programme are described. The composition of the reactor core corresponds to graphite: U235 and Th232: U235 atomic ratios of 3843 and 9.51 respectively. (author)

A report on the first loading of ZENITH

International Nuclear Information System (INIS)

Absalom, R.M.; Barclay, F.R.; Cameron, I.R.; Della Loggia, V.E.; Hage, W.; Kinchin, G.H.; Sanders, J.E.; Tiren, L.I.; Wilson, D.J.

1960-12-01

The loading to criticality of the first core of the zero energy reactor ZENITH, and the results of the subsequent experimental programme are described. The composition of the reactor core corresponds to graphite: U235 and Th232: U235 atomic ratios of 3843 and 9.51 respectively. (author)

Physical events that occur in the reactor core during load changes; Les effets physiques sur le coeur mis en jeu lors des variations de puissance

Energy Technology Data Exchange (ETDEWEB)

Paulin, Ph. [Electricite de France (EDF/DPN/UNIE/GECC), 93 - Saint-Denis (France); Golfier, H. [CEA Saclay (DEN-DANS/DM2S/SERMA/LPEC), 91 - Gif-sur-Yvette (France)

2007-05-15

The reactor core control aims at mastering 2 important parameters that are relevant for reactor availability and safety. First, the reactivity that sets the power output and secondly, the power map in order to handle hot spots. In PWR-type reactors, physical events such as moderator or fuel temperature changes, xenon concentration, that are important for both parameters, evolve during load changes but also during power plateaus and are dependent on burn-up. In this article temperature effect and xenon poisoning are analysed and their impact are assessed along an irradiation campaign through a core neutronic simulation and data from instrumentation. Xenon oscillations are particularly well illustrated. The counter-reactions of the means used for reactor controlling: soluble boron and control rods, are also analysed. (A.C.)

The effect of rehearsal rate and memory load on verbal working memory.

Science.gov (United States)

Fegen, David; Buchsbaum, Bradley R; D'Esposito, Mark

2015-01-15

While many neuroimaging studies have investigated verbal working memory (WM) by manipulating memory load, the subvocal rehearsal rate at these various memory loads has generally been left uncontrolled. Therefore, the goal of this study was to investigate how mnemonic load and the rate of subvocal rehearsal modulate patterns of activity in the core neural circuits underlying verbal working memory. Using fMRI in healthy subjects, we orthogonally manipulated subvocal rehearsal rate and memory load in a verbal WM task with long 45-s delay periods. We found that middle frontal gyrus (MFG) and superior parietal lobule (SPL) exhibited memory load effects primarily early in the delay period and did not exhibit rehearsal rate effects. In contrast, we found that inferior frontal gyrus (IFG), premotor cortex (PM) and Sylvian-parietal-temporal region (area Spt) exhibited approximately linear memory load and rehearsal rate effects, with rehearsal rate effects lasting through the entire delay period. These results indicate that IFG, PM and area Spt comprise the core articulatory rehearsal areas involved in verbal WM, while MFG and SPL are recruited in a general supervisory role once a memory load threshold in the core rehearsal network has been exceeded. Copyright © 2014 Elsevier Inc. All rights reserved.

Fire resistance of extruded hollow-core slabs

DEFF Research Database (Denmark)

Hertz, Kristian Dahl; Giuliani, Luisa; Sørensen, Lars Schiøtt

2016-01-01

Prefabricated extruded hollow-core slabs are preferred building components for floor structures in several countries. It is therefore important to be able to document the fire resistance of these slabs proving fulfilment of standard fire resistance requirements of 60- and 120 minutes found in most...... a standard fire test and from a thorough examination of the comprehensive test documentation available on fire exposed hollow-core slabs. Mechanisms for loss of load-bearing capacity are clarified, and evidence of the fire resistance is found. For the first time the mechanisms responsible for loss of load......-bearing capacity are identified and test results and calculation approach are for the first time Applied in accordance with each other for assessment of fire resistance of the structure....

Nuclear fuel pellet loading machine

International Nuclear Information System (INIS)

Kee, R.W.; Denero, J.V.

1975-01-01

An apparatus for loading nuclear fuel pellets on trays for transfer in a system is described. A conveyor supplies pellets from a source to a loading station. When the pellets reach a predetermined position at the loading station, a manual or automatically operated arm pushes the pellets into slots on a tray and this process is repeated until pellet sensing switches detect that the tray is full. Thereupon, the tray is lowered onto a belt or other type conveyor and transferred to other apparatus in the system, such as a furnace for sintering, and in some cases, reduction of UO 2 . 2 to UO 2 . The pellets are retained on the tray and subsequently loaded directly into fuel rods to be used in the reactor core. (auth)

Demonstration of Load-Follow Simulation with VERA-CS and Standalone BISON

Energy Technology Data Exchange (ETDEWEB)

Stimpson, Shane G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-01-24

In this report, load-follow simulations using VERA-CS with one-way coupling to standalone BISON has been demonstrated including both a single rod with a full cycle of load-follow operations and a quarter-core model with a single month of load-follow.

MTHFR C677T and MTR A2756G polymorphisms and the homocysteine lowering efficacy of different doses of folic acid in hypertensive Chinese adults

Directory of Open Access Journals (Sweden)

Qin Xianhui

2012-01-01

Full Text Available Abstract Background This study aimed to investigate if the homocysteine-lowering efficacy of two commonly used physiological doses (0.4 mg/d and 0.8 mg/d of folic acid (FA can be modified by individual methylenetetrahydrofolate reductase (MTHFR C677T and/or methionine synthase (MTR A2756G polymorphisms in hypertensive Chinese adults. Methods A total of 480 subjects with mild or moderate essential hypertension were randomly assigned to three treatment groups: 1 enalapril only (10 mg, control group; 2 enalapril-FA tablet [10:0.4 mg (10 mg enalapril combined with 0.4 mg of FA, low FA group]; and 3 enalapril-FA tablet (10:0.8 mg, high FA group, once daily for 8 weeks. Results After 4 or 8 weeks of treatment, homocysteine concentrations were reduced across all genotypes and FA dosage groups, except in subjects with MTR 2756AG /GG genotype in the low FA group at week 4. However, compared to subjects with MTHFR 677CC genotype, homocysteine concentrations remained higher in subjects with CT or TT genotype in the low FA group (P P P = 0.005, but not in the low FA group (CC 9.9% vs. TT 11.2%, P = 0.989. Conclusions This study demonstrated that MTHFR C677T polymorphism can not only affect homocysteine concentration at baseline and post-FA treatment, but also can modify therapeutic responses to various dosages of FA supplementation.

Field testing of stiffened deep cement mixing piles under lateral cyclic loading

Science.gov (United States)

Raongjant, Werasak; Jing, Meng

2013-06-01

Construction of seaside and underground wall bracing often uses stiffened deep cement mixed columns (SDCM). This research investigates methods used to improve the level of bearing capacity of these SDCM when subjected to cyclic lateral loading via various types of stiffer cores. Eight piles, two deep cement mixed piles and six stiffened deep cement mixing piles with three different types of cores, H shape cross section prestressed concrete, steel pipe, and H-beam steel, were embedded though soft clay into medium-hard clay on site in Thailand. Cyclic horizontal loading was gradually applied until pile failure and the hysteresis loops of lateral load vs. lateral deformation were recorded. The lateral carrying capacities of the SDCM piles with an H-beam steel core increased by 3-4 times that of the DCM piles. This field research clearly shows that using H-beam steel as a stiffer core for SDCM piles is the best method to improve its lateral carrying capacity, ductility and energy dissipation capacity.

Progress of full MOX core design in ABWR

International Nuclear Information System (INIS)

Izutsu, S.; Sasagawa, M.; Aoyama, M.; Maruyama, H.; Suzuki, T.

2000-01-01

Full MOX ABWR core design has been made, based on the MOX design concept of 8x8 bundle configuration with a large central water rod, 40 GWd/t maximum bundle exposure, and the compatibility with 9x9 high-burnup UO 2 bundles. Core performance on shutdown margin and thermal margin of the MOX-loaded core is similar to that of UO 2 cores for the range from full UO 2 core to full MOX core. Safety analyses based on its safety parameters and MOX property have shown its conformity to the design criteria in Japan. In order to confirm the applicability of the nuclear design method to full MOX cores, Tank-type Critical Assembly (TCA) experiment data have been analyzed on criticality, power distribution and β eff /l measurements. (author)

Analysis of addition of the safety rods at RSG-GAS core

International Nuclear Information System (INIS)

S, Tukiran; S, Tagor Malem; K, Iman

2002-01-01

The silicide fuel loading of the RSG-GAS core is planned to increase from 250 gU to 300 gU. Increasing of fuel loading will prolong the operation cycle length from 25 days to 32,5 days, but ability of reactivity compensation by control rods system decreased because the reactivity shut-down margin is available only 1,03 %, expectation is 2.2 %. One of solutions is added two safety control rods in B-3 and G-10 positions the aim of installing two safety rods (BKP) in RSG-GAS core is to increase core safety margin. So before using the safety control rods in the RSG-GAS core, it is necessary to know its performance, one of the tests showing its performance is to measure the reactivity of the safety control rods. Measurement of safety control rods were done to know each reactivity worth of safety control rods at middle cycle so that the safety rod be used in the RSG-GAS core. Measurement done by using calibration control rods with couple compensation method which always using in the RSG-GAS core to measure the existing control rods. The results of measurement showed that two safety rods (BKP01 and BKP02) have reactivity worth of 93.5 cent and 87.5 cent, respectively. the total reactivity worth of safety control rods is 1.38%. So the two safety rods can be used to increase safety margin of the RSG-GAS core if the fuel is exchanged to 300 gU of loading

Calculations of slurry pump jet impingement loads

International Nuclear Information System (INIS)

Wu, T.T.

1996-01-01

This paper presents a methodology to calculate the impingement load in the region of a submerged turbulent jet where a potential core exits and the jet is not fully developed. The profile of the jet flow velocities is represented by a piece-wise linear function which satisfies the conservation of momentum flux of the jet flow. The adequacy of the of the predicted jet expansion is further verified by considering the continuity of the jet flow from the region of potential core to the fully developed region. The jet impingement load can be calculated either as a direct impingement force or a drag force using the jet velocity field determined by the methodology presented

Robust feedback-linearization control for axial power distribution in pressurized water reactors during load-following operation

OpenAIRE

M. Zaidabadi nejad; G.R. Ansarifar

2018-01-01

Improved load-following capability is one of the most important technical tasks of a pressurized water reactor. Controlling the nuclear reactor core during load-following operation leads to some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking: the core is subjected to sharp and large variation of local power density during transients. Axial offset (AO) is the parameter usually used to represent the core power peaking. One of the impor...

Power-level regulation and simulation of nonlinear pressurized water reactor core with xenon oscillation using H-infinity loop shaping control

Directory of Open Access Journals (Sweden)

Li Gang

2016-01-01

Full Text Available This investigation is to solve the power-level control issue of a nonlinear pressurized water reactor core with xenon oscillations. A nonlinear pressurized water reactor core is modeled using the lumped parameter method, and a linear model of the core is then obtained through the small perturbation linearization way. The H∞loop shapingcontrolis utilized to design a robust controller of the linearized core model.The calculated H∞loop shaping controller is applied to the nonlinear core model. The nonlinear core model and the H∞ loop shaping controller build the nonlinear core power-level H∞loop shaping control system.Finally, the nonlinear core power-level H∞loop shaping control system is simulatedconsidering two typical load processes that are a step load maneuver and a ramp load maneuver, and simulation results show that the nonlinear control system is effective.

Use of TRIGA flip fuel for improved in-core irradiations

Energy Technology Data Exchange (ETDEWEB)

Whittemore, W L [General Atomic Co., San Diego, CA (United States)

1974-07-01

Use of standard TRIGA fuel (20% enriched uranium) in a reactor provides a suitable facility for in-core irradiations. However, large numbers of in-core samples irradiated for long periods (many months) can be handled more economically with a TRIGA loaded with FLIP fuel. As an example, ten or more in-core thermionic devices (each worth 50 to 80 cents with respect to a water-filled position) were irradiated in the Mark III TRIGA at General Atomic Company for 18 months with only a modest change in excess reactivity due to core burnup. A core loading of FLIP fuel has been added to the General Atomic Mark F reactor in order to provide numerous in-core irradiation sites for the production of radioisotopes. Since the worth of a 500-gram sample of a molybdenum compound (used for the production of {sup 99}Mo) is about 25 to 50 cents with respect to a water-filled position, use of a FLIP- TRIGA core will permit the irradiation of more than 5 kilograms of a molybdenum compound. A procedure is under development for the production of {sup 99}Mo with relatively high specific activity. Several techniques to concentrate {sup 99}Mo have been tested experimentally. The results will be reported. (author)

A non-algorithmic approach to the In-core-fuel management problem of a PWR core

International Nuclear Information System (INIS)

Kimhy, Y.

1992-03-01

The primary objective of a commercial nuclear power plant operation is to produce electricity a low cost while satisfying safety constraints imposed on the operating conditions. Design of a fuel reload cycle for the current generation nuclear power plant represents a multistage process with a series of design decisions taken at various time points. Of these stages, reload core design is an important stage, due to its impact on safety and economic plant performance parameters. Overall. performance of the plant during the power production cycle depends on chosen fresh fuel parameters, as well as specific fuel configuration of the reactor core. The motivation to computerize generation and optimization of fuel reload configurations follows from some reasons: first, reload is performed periodically and requires manipulation of a large amount of data. second, in recent years, more complicated fuel loading patterns were developed and implemented following changes in fuel design and/or operational requirements, such as, longer cycles, advanced burnable poison designs, low leakage loading patterns and reduction of irradiation-induced damage of the pressure vessel. An algorithmic approach to the problem was generally adopted. The nature of the reload design process is a 'heuristic' search performed manually by a fuel manager. The knowledge used by the fuel manager is mostly accumulated experience in reactor physics and core calculations. These features of the problem and the inherent disadvantage of the algorithmic method are the main reasons to explore a non-algorithmic approach for solving the reload configuration problem. Several features of the 'solutions space' ( a collection of acceptable final configurations ) are emphasized in this work: 1) the space contain numerous number of entities (> 25) that are distributed un homogeneously, 2) the lack of a monotonic objective function decrease the probability to find an isolated optimum configuration by depth first search or

Finite element based stress analysis of BWR internals exposed to accident loads

Energy Technology Data Exchange (ETDEWEB)

Altstadt, E; Weiss, F P; Werner, M; Willschuetz, H G

1998-10-01

During a hypothetical accident the reactor pressure vessel internals of boiling water reactors can be exposed to considerable loads resulting from temperature gradients and pressure waves. Three dimensional FE models were developed for the core shroud, the upper and the lower core supporting structure, the steam separator pipes and the feed water distributor. The models of core shroud, upper core structure and lower core structure were coupled by means of the substructure technique. All FE models can be used for thermal and for structural mechanical analyses. As an example the FE analysis for the case of a station black-out scenario (loss of power supply for the main circulating pumps) with subsequent emergency core cooling is demonstrated. The transient temperature distributions within the core shroud and within the steam dryer pipes as well were calculated based on the fluid temperatures and the heat transfer coefficients provided by thermo-hydraulic codes. At the maximum temperature gradients in the core shroud, the mechanical stress distribution was computed in a static analysis with the actual temperature field being the load. (orig.)

Deep gray matter demyelination detected by magnetization transfer ratio in the cuprizone model.

Directory of Open Access Journals (Sweden)

Sveinung Fjær

Full Text Available In multiple sclerosis (MS, the correlation between lesion load on conventional magnetic resonance imaging (MRI and clinical disability is weak. This clinico-radiological paradox might partly be due to the low sensitivity of conventional MRI to detect gray matter demyelination. Magnetization transfer ratio (MTR has previously been shown to detect white matter demyelination in mice. In this study, we investigated whether MTR can detect gray matter demyelination in cuprizone exposed mice. A total of 54 female C57BL/6 mice were split into one control group ( and eight cuprizone exposed groups ([Formula: see text]. The mice were exposed to [Formula: see text] (w/w cuprizone for up to six weeks. MTR images were obtained at a 7 Tesla Bruker MR-scanner before cuprizone exposure, weekly for six weeks during cuprizone exposure, and once two weeks after termination of cuprizone exposure. Immunohistochemistry staining for myelin (anti-Proteolopid Protein and oligodendrocytes (anti-Neurite Outgrowth Inhibitor Protein A was obtained after each weekly scanning. Rates of MTR change and correlations between MTR values and histological findings were calculated in five brain regions. In the corpus callosum and the deep gray matter a significant rate of MTR value decrease was found, [Formula: see text] per week ([Formula: see text] and [Formula: see text] per week ([Formula: see text] respectively. The MTR values correlated to myelin loss as evaluated by immunohistochemistry (Corpus callosum: [Formula: see text]. Deep gray matter: [Formula: see text], but did not correlate to oligodendrocyte density. Significant results were not found in the cerebellum, the olfactory bulb or the cerebral cortex. This study shows that MTR can be used to detect demyelination in the deep gray matter, which is of particular interest for imaging of patients with MS, as deep gray matter demyelination is common in MS, and is not easily detected on conventional clinical MRI.

Theoretical prediction on corrugated sandwich panels under bending loads

Science.gov (United States)

Shu, Chengfu; Hou, Shujuan

2018-05-01

In this paper, an aluminum corrugated sandwich panel with triangular core under bending loads was investigated. Firstly, the equivalent material parameters of the triangular corrugated core layer, which could be considered as an orthotropic panel, were obtained by using Castigliano's theorem and equivalent homogeneous model. Secondly, contributions of the corrugated core layer and two face panels were both considered to compute the equivalent material parameters of the whole structure through the classical lamination theory, and these equivalent material parameters were compared with finite element analysis solutions. Then, based on the Mindlin orthotropic plate theory, this study obtain the closed-form solutions of the displacement for a corrugated sandwich panel under bending loads in specified boundary conditions, and parameters study and comparison by the finite element method were executed simultaneously.

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.

In-core fuel element temperature and flow measurment of HFETR

International Nuclear Information System (INIS)

Chen Daolong; Jiang Pei

1988-02-01

The HFETR in-core fuel element temperature-flow measurement facility and its measurement system are expounded. The applications of the instrumented fuel element to stationary and transient states measurements during the lift of power, the operation test of all lifetime at first load, and the deepening burn-up test at second load are described. The method of determination of the hot point temperature under the fin is discussed. The error analysis is made. The fuel element out-of-pile water deprivation test is described. The development of this measurement facility and succesful application have made important contribution to high power and deep burn-up safe operation at two load, in-core fuel element irradiation, and varied investigation of HFETR. After operation at two loads, the integrated power of this instrumented fuel element arrives at 90.88 MWd, its maximum point burn-up is about 64.9%, so that the economy of fuel use of HFETR is raised very much

Supporting system for the core restraint of nuclear reactors

International Nuclear Information System (INIS)

Kaser, A.

1973-01-01

The core restraint of water cooled nuclear reactors which is needed to direct the flow of the coolant through the core can be manufactured only in a moderate wall thickness. Thus, the majority of the loads have to be transmitted to the core barrel which is more rigid. The patent refers to a system of circumferential and vertical support members most of which are free to move relatively to each other, thus reducing thermal stresses during operation. (P.K.)

The reprocessing of irradiated MTR fuel and the nuclear material accountancy - Dounreay, UKAEA

Energy Technology Data Exchange (ETDEWEB)

Barrett, T.R.; Harrison, R. [UKAEA, Nuclear Materials Control Dep., Dounreay (United Kingdom)

1997-07-01

The reprocessing of irradiated HEU MTR fuel is a sensible part of a safeguards regime. It brings together fuel otherwise scattered around the world into a concerted accountancy and protection arrangement. From a nuclear material accountants view the overall accountancy performance has been excellent. While investigations have been required for a few individual MUFs or trends, very little effort has required to be expended by the Nuclear Materials Control Department. That is a definition of a 'good plant'; it operates, measures and records input and output streams, and then the accountancy falls into place. As identified in this paper, the accountancy of the nuclear material processed in the plant is well founded and sound. The accountancy results over several decades confirm the adequacy of the safeguards arrangements at Dounreay. The processing makes good commercial sense and meets the current philosophy of recycling valuable resource materials. The risk of operating the full fuel cycle are less than those of extended storage of irradiated fuel at disparate diverse locations. The reprocessing at Dounreay accords with all of these philosophies. The assessed risk is at a very low level, well within published UK HSE 'tolerability of risk' regulatory guidelines. The impact of the operations are similarly low within the guidelines, for the operators and for the general public. (author)

The reprocessing of irradiated MTR fuel and the nuclear material accountancy - Dounreay, UKAEA

International Nuclear Information System (INIS)

Barrett, T.R.; Harrison, R.

1997-01-01

The reprocessing of irradiated HEU MTR fuel is a sensible part of a safeguards regime. It brings together fuel otherwise scattered around the world into a concerted accountancy and protection arrangement. From a nuclear material accountants view the overall accountancy performance has been excellent. While investigations have been required for a few individual MUFs or trends, very little effort has required to be expended by the Nuclear Materials Control Department. That is a definition of a 'good plant'; it operates, measures and records input and output streams, and then the accountancy falls into place. As identified in this paper, the accountancy of the nuclear material processed in the plant is well founded and sound. The accountancy results over several decades confirm the adequacy of the safeguards arrangements at Dounreay. The processing makes good commercial sense and meets the current philosophy of recycling valuable resource materials. The risk of operating the full fuel cycle are less than those of extended storage of irradiated fuel at disparate diverse locations. The reprocessing at Dounreay accords with all of these philosophies. The assessed risk is at a very low level, well within published UK HSE 'tolerability of risk' regulatory guidelines. The impact of the operations are similarly low within the guidelines, for the operators and for the general public. (author)

Failure analysis of bolted joints in foam-core sandwich composites

DEFF Research Database (Denmark)

Zabihpoor, M.; Moslemian, Ramin; Afshin, M.

2008-01-01

This study represents an effort to predict the bearing strength, failure modes, and failure load of bolted joints in foam-core sandwich composites. The studied joints have been used in a light full composite airplane. By using solid laminates, a new design for the joint zone is developed. These s......This study represents an effort to predict the bearing strength, failure modes, and failure load of bolted joints in foam-core sandwich composites. The studied joints have been used in a light full composite airplane. By using solid laminates, a new design for the joint zone is developed...

Radial power distribution shaping within a PWR fuel assembly utilizing asymmetrically loaded gadolinia-bearing fuel pins

International Nuclear Information System (INIS)

Stone, I.Z.

1992-01-01

As in-core fuel management designs evolve to meet the demands of increasing energy output, more innovative methods are developed to maintain power peaking within acceptable thermal margin limits. In-core fuel management staff must utilize various loading pattern strategies such as cross-core movement of fuel assemblies, multibatch enrichment schemes, and burnable absorbers as the primary means of controlling the radial power distribution. The utilization of fresh asymmetrically loaded gadolinia-bearing assemblies as a fuel management tool provides an additional means of controlling the radial power distribution. At Siemens Nuclear Power Corporation (SNP), fresh fuel assemblies fabricated with asymmetrically loaded gadolinia-bearing fuel rods have been used successfully for several cycles of reactor operation. Asymmetric assemblies are neutronically modeled using the same tools and models that SNP uses to model symmetrically loaded gadolinia-bearing fuel assemblies. The CASMO-2E code is used to produce the homogenized macroscopic assembly cross sections for the nodal core simulator. Optimum fuel pin locations within the asymmetrical assembly are determined using the pin-by-pin PDQ7 assembly core model for each new assembly design. The optimum pin location is determined by the rod loading that minimizes the peak-to-average pin power

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

International Nuclear Information System (INIS)

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

1980-01-01

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

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)

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)

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

Nuclear data uncertainties propagation methods in Boltzmann/Bateman coupled problems: Application to reactivity in MTR

International Nuclear Information System (INIS)

Frosio, Thomas; Bonaccorsi, Thomas; Blaise, Patrick

2016-01-01

Highlights: • Hybrid methods are developed for uncertainty propagation. • These methods take into account the flux perturbation in the coupled problem. • We show that OAT and MC methods give coherent results, except for Pearson correlations. • Local sensitivity analysis is performed. - Abstract: A novel method has been developed to calculate sensitivity coefficients in coupled Boltzmann/Bateman problem for nuclear data (ND) uncertainties propagation on the reactivity. Different uncertainty propagation methodologies, such as One-At-a-Time (OAT) and hybrid Monte-Carlo/deterministic methods have been tested and are discussed on an actual example of ND uncertainty problem on a Material Testing Reactor (MTR) benchmark. Those methods, unlike total Monte Carlo (MC) sampling for uncertainty propagation and quantification (UQ), allow obtaining sensitivity coefficients, as well as Bravais–Pearson correlations values between Boltzmann and Bateman, during the depletion calculation for global neutronics parameters such as the effective multiplication coefficient. The methodologies are compared to a pure MC sampling method, usually considered as the “reference” method. It is shown that methodologies can seriously underestimate propagated variances, when Bravais–Pearson correlations on ND are not taken into account in the UQ process.

Selective separation of actinides and long-lived fission products from 1 AW MTR liquid waste: pilot plant tests part II

International Nuclear Information System (INIS)

Grossi, G.; Marrocchelli, A.; Pietrelli, L.; Calle, C.; Gili, M.; Luce, A.; Troiani, F.

1992-01-01

In Italy there are some 120 m 3 of liquid High-level radioactive Wastes coming from MTR, Candu and EPK River fuel elements reprocessing. These High-level radioactive wastes contain a large amount of chemicals and inert salts together with cesium, strontium and transuranium elements. Transuranium elements and strontium are separated from the inert salts by means of a selective precipitation while Cesium is adsorbed on synthetic zeolithes (AZE Process) or precipitated with sodium Tetraphenyl borate (NaTPB) (ATE process). The benchscale experiments have confirmed the feasibility of selective separation processes and have showed that decontamination efficiency for strontium, plutonium and cesium were, respectively, 100, 5000 and 1000. This second part of the CEC final report describes Searse pilot plant tests with cold experiments. 37 Refs.; 17 Figs.; 16 Tabs

Serial contrast-enhanced magnetic resonance and magnetization transfer in the study of patients with multiple sclerosis

International Nuclear Information System (INIS)

Rovira, A.; Alonso, J.; Cucurella, G.; Nos, C.; Tintore, M.; Pedraza, S.; Rio, J.; Montalban, X.

1997-01-01

To demonstrate the changes in the magnetization transfer ratio (MTR) of different demyelinating plaques, correlating them with the baseline values in T1-weighted contrast-enhanced magnetic resonance (MR) sequences in order to relate them more closely to the underlying disease. The study was based on 33 demyelinating plaques obtained from six patients clinically diagnosed as having remitting-recurring multiple sclerosis (MS). All the patients underwent two MR studies at a 3 to 5-month interval, including contrast-enhanced T1 and T2- weighted sequences and magnetization transfer images. The latter were used to calculate the MTR for each of the demyelinating plaques included in the study. The statistical analysis of the results obtained revealed statistically significant between initial MTR values and those of subsequent T1-weighted sequences. The MTR demonstrate significant differences between plaques according to contrast-enhanced T1-weigh tes sequences, probably indicating variable degrees of edema, demyelination and tissue destruction. These differences should be taken into account to enable the use of T1-weighted sequences to quantify the lesion load in MS patients. (Author) 35 refs

Serial contrast-enhanced magnetic resonance and magnetization transfer in the study of patients with multiple sclerosis; Resonancia magnetica con contraste y transferencia de magnetizacion en el estudio seriado de pacientes con esclerosis multiple

Energy Technology Data Exchange (ETDEWEB)

Rovira, A; Alonso, J; Cucurella, G; Nos, C; Tintore, M; Pedraza, S; Rio, J; Montalban, X [Hospital General i Universitari Vall d` Hebron. Barcelona (Spain)

1998-12-31

To demonstrate the changes in the magnetization transfer ratio (MTR) of different demyelinating plaques, correlating them with the baseline values in T1-weighted contrast-enhanced magnetic resonance (MR) sequences in order to relate them more closely to the underlying disease. The study was based on 33 demyelinating plaques obtained from six patients clinically diagnosed as having remitting-recurring multiple sclerosis (MS). All the patients underwent two MR studies at a 3 to 5-month interval, including contrast-enhanced T1 and T2- weighted sequences and magnetization transfer images. The latter were used to calculate the MTR for each of the demyelinating plaques included in the study. The statistical analysis of the results obtained revealed statistically significant between initial MTR values and those of subsequent T1-weighted sequences. The MTR demonstrate significant differences between plaques according to contrast-enhanced T1-weigh tes sequences, probably indicating variable degrees of edema, demyelination and tissue destruction. These differences should be taken into account to enable the use of T1-weighted sequences to quantify the lesion load in MS patients. (Author) 35 refs.

Designing an application for managing distribution transformer load

Directory of Open Access Journals (Sweden)

Olga Liliana Sánchez

2006-09-01

Full Text Available This paper presents a software application for distribution transformer load management and calculating the elec- tricity distribution network service quality index (i.e. EDS and EFS indexes. Transformer operation state, calculating core and coil losses and selecting distribution transformers based on technical-economic criteria using daily-load curves are the load management strategies presented here. The application consists of a programme allowing to data acquisition input, an Excel-based spread-sheet having the equations for calculating load management data and a user-interface presenting the load management application’s results. The application was validated by technical-economic evaluation of three distribution transformers when supplying demand.

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)

Heysham II/Torness AGR core integrity

International Nuclear Information System (INIS)

Birch, A.L.; Hampson, J.D.

1985-01-01

The design and construction process for the Heysham II/Torness AGR core structures is presented. The design intent utilizing all past experience in designing and building AGR core structures is described. The major aspects of the design criteria and the design conditions are outlined to demonstrate how the integrity of the Heysham II/Torness core is assured. Since no recognized codes of practice for graphite core design exist, the National Nuclear Corporation (NNC) have conceived design criteria utilizing reserve factors based on their design experience. Target reserve factors are defined for particular loading conditions including the ultimate 'safe-shutdown earthquake'. The substantial programme of computer analysis and RandD work to substantiate the design, including seismic qualification, is described. In keeping with their responsibility for the detailed core structure design and the fuel path geometry (guide tube system), NNC attach great importance to design/manufacture/construction liaison, which is demonstrated in the quality assurance section. (author)

Generation of multigroup cross-sections from micro-group ones in code system SUHAM-U used for VVER-1000 reactor core calculations with MOX loading

Energy Technology Data Exchange (ETDEWEB)

Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)

2005-07-01

At the present time, the new code system SUHAM-U for calculation of the neutron-physical processes in nuclear reactor core with triangular and square lattices based both on the modern micro-group (about 7000 groups) cross-sections library of code system UNK and on solving the multigroup (up to 89 groups) neutron transport equation by Surface Harmonics Method is elaborated. In this paper the procedure for generation of multigroup cross-sections from micro-group ones for calculation of VVER-1000 reactor core with MOX loading is described. The validation has consisted in computing VVER-1000 fuel assemblies with uranium and MOX fuel and has shown enough high accuracy under corresponding selection of the number and boundaries of the energy groups. This work has been fulfilled in the frame of ISTC project 'System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels'.

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

International Nuclear Information System (INIS)

Khedr, A.

2009-01-01

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

Preliminary experience and near future utilization programmes of the MPR-30 fueled by LEU [low enriched uranium

International Nuclear Information System (INIS)

Arbie, B.; Soentono, S.

1987-01-01

The MTR type reactor MPR-30 G.A. Siwabessy, located at PUSPIPTEK Serpong has recently reached its first criticality. This multipurpose reactor is supposed to be the first MTR type reactor in the world that is designed and constructed to be fueled by low enriched uranium. Preliminary experience covering the approach to the first criticality and the excess reactivity loading as well as some thermal hydraulics and power ascension tests are briefly presented and discussed. The near future utilization programmes during and after commissioning are also presented. (Author)

Multicycle Optimization of Advanced Gas-Cooled Reactor Loading Patterns Using Genetic Algorithms

International Nuclear Information System (INIS)

Ziver, A. Kemal; Carter, Jonathan N.; Pain, Christopher C.; Oliveira, Cassiano R.E. de; Goddard, Antony J. H.; Overton, Richard S.

2003-01-01

A genetic algorithm (GA)-based optimizer (GAOPT) has been developed for in-core fuel management of advanced gas-cooled reactors (AGRs) at HINKLEY B and HARTLEPOOL, which employ on-load and off-load refueling, respectively. The optimizer has been linked to the reactor analysis code PANTHER for the automated evaluation of loading patterns in a two-dimensional geometry, which is collapsed from the three-dimensional reactor model. GAOPT uses a directed stochastic (Monte Carlo) algorithm to generate initial population members, within predetermined constraints, for use in GAs, which apply the standard genetic operators: selection by tournament, crossover, and mutation. The GAOPT is able to generate and optimize loading patterns for successive reactor cycles (multicycle) within acceptable CPU times even on single-processor systems. The algorithm allows radial shuffling of fuel assemblies in a multicycle refueling optimization, which is constructed to aid long-term core management planning decisions. This paper presents the application of the GA-based optimization to two AGR stations, which apply different in-core management operational rules. Results obtained from the testing of GAOPT are discussed

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Fission yields and cross section uncertainty propagation in Boltzmann/Bateman coupled problems: Global and local parameters analysis with a focus on MTR

International Nuclear Information System (INIS)

Frosio, Thomas; Bonaccorsi, Thomas; Blaise, Patrick

2016-01-01

Highlights: • Nuclear data uncertainty propagation for neutronic quantities in coupled problems. • Uncertainties are detailed for local isotopic concentrations and local power maps. • Correlations are built between space areas of the core and for different burnups. - Abstract: In a previous paper, a method was investigated to calculate sensitivity coefficients in coupled Boltzmann/Bateman problem for nuclear data (ND) uncertainties propagation on the reactivity. Different methodologies were discussed and applied on an actual example of multigroup cross section uncertainty problem for a 2D Material Testing Reactor (MTR) benchmark. It was shown that differences between methods arose from correlations between input parameters, as far as the method enables to take them into account. Those methods, unlike Monte Carlo (MC) sampling for uncertainty propagation and quantification (UQ), allow obtaining sensitivity coefficients, as well as correlations values between nuclear data, during the depletion calculation for the parameters of interest. This work is here extended to local parameters such as power factors and isotopic concentrations. It also includes fission yield (FY) uncertainty propagation, on both reactivity and power factors. Furthermore, it introduces a new methodology enabling to decorrelate direct and transmutation terms for local quantities: a Monte-Carlo method using built samples from a multidimensional Gaussian law is used to extend the previous studies, and propagate fission yield uncertainties from the CEA’s COMAC covariance file. It is shown that, for power factors, the most impacting ND are the scattering reactions, principally coming from 27 Al and (bounded hydrogen in) H 2 O. The overall effect is a reduction of the propagated uncertainties throughout the cycle thanks to negatively correlated terms. For fission yield (FY), the results show that neither reactivity nor local power factors are strongly affected by uncertainties. However, they

WWER core pattern enhancement using adaptive improved harmony search

International Nuclear Information System (INIS)

Nazari, T.; Aghaie, M.; Zolfaghari, A.; Minuchehr, A.; Norouzi, A.

2013-01-01

Highlights: ► The classical and improved harmony search algorithms are introduced. ► The advantage of IHS is demonstrated in Shekel's Foxholes. ► The CHS and IHS are compared with other Heuristic algorithms. ► The adaptive improved harmony search is applied for two cases. ► Two cases of WWER core are optimized in BOC FA pattern. - Abstract: The efficient operation and fuel management of PWRs are of utmost importance. Core performance analysis constitutes an essential phase in core fuel management optimization. Finding an optimum core arrangement for loading of fuel assemblies, FAs, in a nuclear core is a complex problem. In this paper, application of classical harmony search (HS) and adaptive improved harmony search (IHS) in loading pattern (LP) design, for pressurized water reactors, is described. In this analysis, finding the best core pattern, which attains maximum multiplication factor, k eff , by considering maximum allowable power picking factors (PPF) is the main objective. Therefore a HS based, LP optimization code is prepared and CITATION code which is a neutronic calculation code, applied to obtain effective multiplication factor, neutron fluxes and power density in desired cores. Using adaptive improved harmony search and neutronic code, generated LP optimization code, could be applicable for PWRs core with many numbers of FAs. In this work, at first step, HS and IHS efficiencies are compared with some other heuristic algorithms in Shekel's Foxholes problem and capability of the adaptive improved harmony search is demonstrated. Results show, efficient application of IHS. At second step, two WWER cases are studied and then IHS proffered improved core patterns with regard to mentioned objective functions.

Analysis of syntactic foam – GFRP sandwich composites for flexural loads

Science.gov (United States)

Paul, Daniel; Velmurugan, R.; Jayaganthan, R.; Gupta, N. K.; Manzhirov, A. V.

2018-04-01

The use of glass microballoon (GMB) — epoxy syntactic foams as a sandwich core material is studied. The skins and foam core are fabricated and joined instantaneously unlike the procedures followed in the previous studies. Each successive layer of the sandwich is fabricated when the previous layer is in a semi-gelled state. These sandwich samples are characterized for their properties under flexural loading. The failure modes and mechanical properties are carefully investigated. The change in fabrication technique results in a significant increase in the load bearing pattern of the sandwich. In earlier studies, debonding was found to occur prematurely since the bonding between the skins and core is the weakest plane. Using the current technique, core cracking occurs first, followed by skin fiber breaking and debonding happens at the end. This ensures that the load carrying phase of the structure is extended considerably. The sandwich is also analytically studied using Reddy’s higher order shear deformation theory. A higher order theory is selected as the sandwich can no longer be considered as a thin beam and thus shear effects also need to be considered in addition to bending effects.

Loads on reactor pressure vessel internals induced by low-pressure waves

International Nuclear Information System (INIS)

Benkert, J.; Mika, C.; Stegemann, D.; Valero, M.

1978-02-01

Departing from the conservation theorems for mass and impulse the computer code DRUWE has been developed which allows to calculate loads on the core shell with simplifying assumptions for the first period just after the rupture has opened. It can be supposed that the whole rupture cross section is set free within 15 msec. The calculation progresses in a way that for a core shell the local, timely pressure- and load development, respectively, the total dynamic load as well as the moments acting on the fixing of the core shell, can be calculated. The required input data are merely geometric data on the concept of the pressure vessel and its components as well as the effective subcooling of the fluid. By means of some parameters the programm development can be controlled in a way that the results are available in form of listings or diagrams, respectively, as well as in form of card decks for following investigations, e.g. solidity calculations. (orig./RW) [de

Improvement of SSR core design for ABWR-II

International Nuclear Information System (INIS)

Moriwaki, Masanao; Aoyama, Motoo; Okada, Hiroyuki; Kitamura, Hideya; Sakurada, Koichi; Tanabe, Akira

2003-01-01

In order to enhance the spectral shift effect in the ABWR-II reactor, a novel core design to bring out better performance of spectral shift rods (SSRs) is studied. The SSR is a new type of water rod, in which the water level develops naturally during operation and changes according to the coolant flow rate through the channel. By using the SSR, the average moderator density, which is directly related to core reactivity, can be controlled over a wide range by the core flow rate. In the new SSR core design, two types of SSR bundles, in which settings for the SSR water levels are different, are utilized and loaded according to flow distribution in the core. This two-region SSR core design allows wide variation in the average SSR water level, thus improving fuel economy. Enhancement of SSR function in the two-region SSR core increases the uranium saving factor by about 25%, from the 6% of the conventional uniform SSR core to about 8%. (author)

Upper-bound fission product release assessment for large break LOCA in CANFLEX bundle reactor core

International Nuclear Information System (INIS)

Oh, Duk Ju; Lee, Kang Moon

1996-07-01

Quarter-core gap inventory assessment for CANDU-6 reactor core loaded with CANFLEX fuel bundles has been performed as one of the licensing safety analyses required for 24 natural uranium CANFLEX bundle irradiation in CANDU-6 reactor. The quarter-core gap inventory for the CANFLEX bundle core is 5 - 10 times lower than that for the standard bundle core, depending on the half-life of the isotope. The lower gap inventory of the CANFLEX bundle core is attributed to the lower linear power of the CANFLEX bundle compared with the standard bundle. However, the whole core total inventories for both the CANFLEX and standard bundle cores are nearly the same. The 6 - 8 times lower upper-bound fission product releases of the CANFLEX bundle core for large break LOCA than those of the standard bundle core imply that the loading of 24 natural uranium CANFLEX bundles would improve the predicted consequences of the postulated accident described in the Wolsung 2 safety report. 2 tabs., 6 figs., 3 refs. (Author)

First in-core measurement results obtained with the innovative mobile calorimeter CALMOS inside the OSIRIS material testing reactor

International Nuclear Information System (INIS)

Carcreff, Hubert; Salmon, Laurent; Courtaux, Cedric

2014-01-01

Nuclear heating rate inside an MTR has to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. An innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70 MWth OSIRIS reactor operated by CEA. Thanks to a new calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. Development of the calorimetric probe required manufacturing and irradiation of mock-ups in the ex-core area, where nuclear heating rate does not exceed 2 W.g -1 . The calorimeter working mode, the different measurement procedures, main modeling and ex-core experimental results have been already presented in previous papers. In this paper, we present in-core results obtained from 2011 to 2013 with the final device. For the first time, this new experimental measurement system was operated in several experimental locations, with nominal in-core thermal hydraulic conditions, nominal neutron flux and nuclear heating rate up to 6 W.g -1 (in graphite). After a brief presentation of the displacement system specificities, first nuclear heating distributions are presented and discussed. The Finite Element model of the calorimeter was upgraded in order to match calculated temperatures with measured ones. This 'validated' model allowed to estimate a Kc factor which tends to correct small nonlinearities when heating rate is calculated from the 'calibration method'. A comparison is made between nuclear heating rates determined from 'calibration' and 'zero methods'. In addition, an evaluation of the global uncertainty associated to the measurements is detailed. Finally, a comparison is made with available measurements obtained from previous calorimeters. (authors)

Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Maldonado, Ivan [Univ. of Tennessee, Knoxville, TN (United States)

2016-04-14

The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate (“plank”) fuel. Proposal to FY12 NEUP entitled “Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors” was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project’s success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

International Nuclear Information System (INIS)

Petrovic, Bojan; Maldonado, Ivan

2016-01-01

The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate ('plank') fuel. Proposal to FY12 NEUP entitled 'Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors' was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project's success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

Research on Operation and Control Strategy of 600MW PWR in Load Follow

Energy Technology Data Exchange (ETDEWEB)

Qu, Bing Yang; Cao, Xin Rong [Harbin Engineering University, Harbin (China); Li, Han Chen [China Nuclear Power Engineering Co., Beijing (China)

2014-08-15

600MW Pressurized Water Reactor (PWR) is designed to operate in Constant Axial Offset Control (CAOC) strategy with base load originally. By calculations over a typical load follow scenario '12-3-6-3 {sup (}100-50-100%FP) via the CASMO-4E and SIMULATE-3 package, values of core operating parameter have been examined. With the progress of the nuclear power industry, advanced reactors are considered to have a good performance in load follow, economy and flexibility. Under the premise of fuel loading and structural dimensions unchanged, two independent control rod groups M and AO are used in 600MW pressurized water reactor to provide fine control of both the core reactivity and axial power distribution, which is named ' Improved G strategy .' The influences of different control rod distributions, composition materials, and overlap steps had in power changes have been examined in a comparative study to choose the optimal one.Then we simulate a range of load follow scenarios of the redesigned 600MW core without adjusting soluble boron concentration in the begin, middle and end of first cycle. This paper additionally demonstrated the moderator temperature coefficient and shutdown margin values of the reactor in Improved G strategy to compare with the thermal safety design criteria. It's demonstrated that adequate adjustment of control rod groups enable the core to perform load follow through Improved G strategy in 80% of cycle and save a large volume of liquid effluent particularly toward the end of cycle.

Face/core debond fatigue crack growth characterization using the sandwich mixed mode bending specimen

DEFF Research Database (Denmark)

Manca, Marcello; Quispitupa, Amilcar; Berggreen, Christian

2012-01-01

Face/core fatigue crack growth in foam-cored sandwich composites is examined using the mixed mode bending (MMB) test method. The mixed mode loading at the debond crack tip is controlled by changing the load application point in the MMB test fixture. Sandwich specimens were manufactured using H45...... and H100 PVC foam cores and E-glass/polyester face sheets. All specimens were pre-cracked in order to define a sharp crack front. The static debond fracture toughness for each material configuration was measured at different mode-mixity phase angles. Fatigue tests were performed at 80% of the static...

PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

International Nuclear Information System (INIS)

Bi, G.; Liu, C.; Si, S.

2012-01-01

This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade 233 U-Thorium (U 3 ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade 233 U extracted from burnt PuThOX fuel was used to fabrication of U 3 ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U 3 ThOX mixed core, the well designed U 3 ThOX FAs with 1.94 w/o fissile uranium (mainly 233 U) were located on the periphery of core as a blanket region. U 3 ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U 3 ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U 3 ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U 3 ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U 3 ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared

Wave loads on offshore wind turbines: Accurate tools and structural response

DEFF Research Database (Denmark)

Bredmose, Henrik

2014-01-01

Can the design models for offshore wind turbine wave loads be improved? And how will that change the overall load picture? Core questions of the Wave Loads project which was finalised in 2013 with two PhD theses, response calculations for jackets and monopiles, a detailed set of experiments and a3D...... coupled CFD wave solver...

HTR-proteus pebble bed experimental program core 4: random packing with a 1:1 moderator-to-fuel pebble ratio

Energy Technology Data Exchange (ETDEWEB)

Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Snoj, Luka [Jozef Stefan Inst. (IJS), Ljubljana (Slovenia); Lengar, Igor [Jozef Stefan Inst. (IJS), Ljubljana (Slovenia); Koberl, Oliver [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

2014-03-01

In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORE 4: RANDOM PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

Energy Technology Data Exchange (ETDEWEB)

John D. Bess; Leland M. Montierth

2013-03-01

In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering

Fast three-dimensional core optimization based on modified one-group model

Energy Technology Data Exchange (ETDEWEB)

Freire, Fernando S. [ELETROBRAS Termonuclear S.A. - ELETRONUCLEAR, Rio de Janeiro, RJ (Brazil). Dept. GCN-T], e-mail: freire@eletronuclear.gov.br; Martinez, Aquilino S.; Silva, Fernando C. da [Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear], e-mail: aquilino@con.ufrj.br, e-mail: fernando@con.ufrj.br

2009-07-01

The optimization of any nuclear reactor core is an extremely complex process that consumes a large amount of computer time. Fortunately, the nuclear designer can rely on a variety of methodologies able to approximate the analysis of each available core loading pattern. Two-dimensional codes are usually used to analyze the loading scheme. However, when particular axial effects are present in the core, two-dimensional analysis cannot produce good results and three-dimensional analysis can be required at all time. Basically, in this paper are presented the major advantages that can be found when one use the modified one-group diffusion theory coupled with a buckling correction model in optimization process. The results of the proposed model are very accurate when compared to benchmark results obtained from detailed calculations using three-dimensional nodal codes (author)

Fast three-dimensional core optimization based on modified one-group model

International Nuclear Information System (INIS)

Freire, Fernando S.; Martinez, Aquilino S.; Silva, Fernando C. da

2009-01-01

The optimization of any nuclear reactor core is an extremely complex process that consumes a large amount of computer time. Fortunately, the nuclear designer can rely on a variety of methodologies able to approximate the analysis of each available core loading pattern. Two-dimensional codes are usually used to analyze the loading scheme. However, when particular axial effects are present in the core, two-dimensional analysis cannot produce good results and three-dimensional analysis can be required at all time. Basically, in this paper are presented the major advantages that can be found when one use the modified one-group diffusion theory coupled with a buckling correction model in optimization process. The results of the proposed model are very accurate when compared to benchmark results obtained from detailed calculations using three-dimensional nodal codes (author)

Development of UCMS for Analysis of Designed and Measured Core Power Distribution

International Nuclear Information System (INIS)

Moon, Sang Rae; Hong, Sun Kwan; Yang, Sung Tae

2009-01-01

In this study, reactor core loading patterns were determined by calculating and verifying the factors affecting peak power and important core safety variables were reconciled with their design criteria using a newly designed unified core management system. Core loading patterns are designed for quadrant cores under the assumption that the power distribution of the reactor core is the same among symmetric fuel assemblies within the core. Actual core power distributions measured during core operation may differ slightly from their designed data. Reactor engineers monitor these differences between the designed and measured data by performing a surveillance procedure every month according to the technical specification requirements. It is difficult to monitor overall power distribution behavior throughout the assemblies using the current procedure because it requires the reactor engineer to compare the designed data with only the maximum value of the power peaking factor and the relative power density. It is necessary to enhance this procedure to check the primary variables such as core power distribution, because long cycle operation, high burnup, power up-rate, and improved fuel can change the environment in the core. To achieve this goal, a web-based Unified Core Management System (UCMS) was developed. To build the UCMS, a database system was established using reactor design data such as that in the Nuclear Design Report (NDR) and automated core analysis codes for all light water reactor power plants. The UCMS is designed to help reactor engineers to monitor important core variables and core safety margins by comparing the measured core power distribution with designed data for each fuel assembly during the cycle operation in nuclear power plants

Localized Effects in the Nonlinear Behavior of Sandwich Panels with a Transversely Flexible Core

DEFF Research Database (Denmark)

Frostig, Y.; Thomsen, Ole Thybo

2005-01-01

This paper presents the results of an investigation of the role of localized effects within the geometrically nonlinear domain on structural sandwich panels with a "compliant" core. Special emphasis is focused on the nonlinear response near concentrated loads and stiffened core regions. The adopted...... nonlinear analysis approach incorporates the effects of the vertical flexibility of the core, and it is based on the approach of the High-order Sandwich Panel Theory (HSAPT). The results demonstrate that the effects of localized loads, when taken into the geometrically nonlinear domain, change the response...... of the panel from a strength problem controlled by stress constraints into a stability problem with unstable limit point behavior when force-controlled loads are applied. The stability problem emerge as the nonlinear response develops with the formation of a small number of buckling waves in the compressed...

Flexural fatigue failures and lives of Eco-Core sandwich beams

International Nuclear Information System (INIS)

Hossain, Mohammad Mynul; Shivakumar, Kunigal

2014-01-01

Highlights: • Eco-Core sandwich beam is flexural fatigue tested to study its fatigue response. • The core showed three failure types: damage onset, progression and final failure. • These failures were found to be represented by 1%, 5% and 7% change in compliance. • The fatigue stress-life (S–N) relationship follows a power low, σ max /σ ct = A o N α . • The fatigue failure was by multiple vertical cracks followed by 45° shear failure. - Abstract: Eco-Core is a class of syntactic foam made from small volume of high char yield binder and large volume of a class of flyash for fire resistance application. Very little or no flexural fatigue data of this class of core material is reported in the open literature. This paper presents a flexural fatigue response of Eco-Core in a glass/vinyl ester composite face sheet sandwich beam. A four-point loaded flexural test specimen was designed and tested in static and fatigue loadings to cause tension failure in the core. The fatigue test was conducted at maximum cyclic stress (σ max ) ranged from 0.7σ ct to 0.9σ ct , where σ ct is the static flexural strength of the core. The sinusoidal loading frequency of 2 Hz with the stress ratio of 0.1 was used. Flexural fatigue failure modes of Eco-Core sandwich beam were classified: damage onset (single tension crack), damage progression (multiple tension cracks) and ultimate failure (a combination of tension and shear). These failures were characterized by 1%, 5% and 7% changes in compliance that corresponds to N 1% , N 5% and N 7% lives. The fatigue stress-life (S–N) relationship was found to follow the well-known power law equation, σ max /σ ct = A o N α . The constants A o and α were established for all three types of failures. The endurance limit was established based on 1 million cycles limit and it was found to be 0.65σ ct , 0.70σ ct and 0.71σ ct , respectively for the three modes of failure. Flexural fatigue and static failure modes of Eco-Core sandwich

Analysis and Behaviour of Sandwich Panels with Profiled Metal Facings under Transverse Load

Directory of Open Access Journals (Sweden)

M. Budescu

2004-01-01

Full Text Available Sandwich panels with thin steel facings and polyurethane core combine the load-carrying capacity of metal facings and protection functions with core properties. The core separates the two facings and keeps them in a stable condition, transmits shear between external layers, provides most of the shear rigidity and occasionally makes of useful contribution to the bending stiffness of the sandwich construction as a whole [1]. An experimental program on sandwich panels has been organized to prove that the mechanical properties of core and interface satisfy the load-carrying requirements for structural sandwich panels. The analysis of sandwich panels with deep profiles facings for cladding elements, respectively the roof constructions, has been carried out according to the European design norms [1], [5].

Pebble bed test reactor in peu-a-peu load

International Nuclear Information System (INIS)

Kranz, L.

1988-03-01

The presented work deals with a new type of load model for high temperature reactors with spherical fuels: the peu-a-peu load system. Using this load system the reactor core is only filled partially in the beginning of the power operation. But it has to be a critical base core. With proceeding burn-off the reactor is filled up with further fuel elements the way that it stays always just critically. When the reactor is filled up completely with fuel elements, the reactor operation has to be interrupted and the reactor has to be discharged. Afterwards a new cycle can start like the one just described. A reference reactor with 100 MW thermal power is investigated in this work in detail and should make clear the way of function of the load system and the base idea of 'simplicity and safety'. The improvement proposal to use again a part of the fuel elements of a cycle for the next cycle minimizes the higher specific uranium need of a peu-a-peu reactor decisively. (orig.) [de

Core seismic methods verification report

International Nuclear Information System (INIS)

Olsen, B.E.; Shatoff, H.D.; Rakowski, J.E.; Rickard, N.D.; Thompson, R.W.; Tow, D.; Lee, T.H.

1979-12-01

This report presents the description and validation of the analytical methods for calculation of the seismic loads on an HTGR core and the core support structures. Analytical modeling, integration schemes, parameter assignment, parameter sensitivity, and correlation with test data are key topics which have been covered in detail. Much of the text concerns the description and the results of a series of scale model tests performed to obtain data for code correlation. A discussion of scaling laws, model properties, seismic excitation, instrumentation, and data reduction methods is also presented, including a section on the identification and calculation of statistical errors in the test data