WorldWideScience

Sample records for pik physical model reactor

  1. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  2. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  3. High-density ultracold neutron sources for the WWR-M and PIK reactors

    Science.gov (United States)

    Serebrov, A. P.; Fomin, A. K.; Kharitonov, A. G.; Lyamkin, V. A.; Prudnikov, D. V.; Ivanov, S. A.; Erykalov, A. N.; Onegin, M. S.; Gridnev, K. A.

    2016-01-01

    It is proposed to equip the PIK and WWR-M research reactors at the Petersburg Nuclear Physics Institute (PNPI) with high-density ultracold neutron (UCN) sources, where UCNs will be obtained based on the effect of their accumulation in superfluid helium (due to the specific features of this quantum fluid). The maximum UCN storage time in superfluid helium is obtained at temperatures on the order of 1 K. These sources are expected to yield UCN densities of 103-104 cm-3, i.e., approximately three orders of magnitude higher than the density from existing UCN sources throughout the world. The development of highest intensity UCN sources will make PNPI an international center of fundamental UCN research.

  4. High-density ultracold neutron sources for the WWR-M and PIK reactors

    Energy Technology Data Exchange (ETDEWEB)

    Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Fomin, A. K.; Kharitonov, A. G.; Lyamkin, V. A.; Prudnikov, D. V.; Ivanov, S. A.; Erykalov, A. N.; Onegin, M. S. [National Research Centre “Kurchatov Institute”, Petersburg Nuclear Physics Institute (Russian Federation); Gridnev, K. A. [St. Petersburg State University (Russian Federation)

    2016-01-15

    It is proposed to equip the PIK and WWR-M research reactors at the Petersburg Nuclear Physics Institute (PNPI) with high-density ultracold neutron (UCN) sources, where UCNs will be obtained based on the effect of their accumulation in superfluid helium (due to the specific features of this quantum fluid). The maximum UCN storage time in superfluid helium is obtained at temperatures on the order of 1 K. These sources are expected to yield UCN densities of 10{sup 3}–10{sup 4} cm{sup –3}, i.e., approximately three orders of magnitude higher than the density from existing UCN sources throughout the world. The development of highest intensity UCN sources will make PNPI an international center of fundamental UCN research.

  5. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  6. UCN sources at external beams of thermal neutrons. An example of PIK reactor

    CERN Document Server

    Lychagin, E V; Muzychka, A Yu; Nekhaev, G V; Nesvizhevsky, V V; Onegin, M S; Sharapov, E I; Strelkov, A V

    2015-01-01

    We consider ultracold neutron (UCN) sources based on a new method of UCN production in superfluid helium (4He). The PIK reactor is chosen as a perspective example of the application of this idea, which consists of installing a 4He UCN source in a beam of thermal or cold neutrons and surrounding the source with a moderator-reflector, which plays the role of a source of cold neutrons (CNs) feeding the UCN source. The CN flux in the source can be several times larger than the incident flux, due to multiple neutron reflections from the moderator-reflector. We show that such a source at the PIK reactor would provide an order of magnitude larger density and production rate than an analogous source at the ILL reactor. We estimate parameters of a 4He source with solid methane (CH4) or/and liquid deuterium (D2) moderator-reflector. We show that such a source with CH4 moderator-reflector at the PIK reactor would provide the UCN density of ~1x10^5 1/cm^3, and the UCN production rate of ~2x10^7 1/s. These values are resp...

  7. UCN sources at external beams of thermal neutrons. An example of PIK reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lychagin, E.V., E-mail: lychag@nf.jinr.ru [Joint Institute for Nuclear Research, 6 Joliot-Curie, Dubna 141980 (Russian Federation); Mityukhlyaev, V.A., E-mail: victim@pnpi.spb.ru [Petersburg Nuclear Physics Institute, Orlova Roscha, Gatchina 188300 (Russian Federation); Muzychka, A.Yu., E-mail: muz@nf.jinr.ru [Joint Institute for Nuclear Research, 6 Joliot-Curie, Dubna 141980 (Russian Federation); Nekhaev, G.V., E-mail: grigorijnekhaev@yandex.ru [Joint Institute for Nuclear Research, 6 Joliot-Curie, Dubna 141980 (Russian Federation); Nesvizhevsky, V.V., E-mail: nesvizhevsky@ill.eu [Institut Max von Laue – Paul Langevin, 71 Avenue des Martyrs, Grenoble 38042 (France); Onegin, M.S., E-mail: oneginm@gmail.com [Petersburg Nuclear Physics Institute, Orlova Roscha, Gatchina 188300 (Russian Federation); Sharapov, E.I., E-mail: sharapov@nf.jinr.ru [Joint Institute for Nuclear Research, 6 Joliot-Curie, Dubna 141980 (Russian Federation); Strelkov, A.V., E-mail: str@jinr.ru [Joint Institute for Nuclear Research, 6 Joliot-Curie, Dubna 141980 (Russian Federation)

    2016-07-01

    We consider ultracold neutron (UCN) sources based on a new method of UCN production in superfluid helium ({sup 4}He). The PIK reactor is chosen as a perspective example of application of this idea, which consists of installing {sup 4}He UCN source in the beam of thermal or cold neutrons and surrounding the source with moderator-reflector, which plays the role of cold neutron (CN) source feeding the UCN source. CN flux in the source can be several times larger than the incident flux, due to multiple neutron reflections from the moderator–reflector. We show that such a source at the PIK reactor would provide an order of magnitude larger density and production rate than an analogous source at the ILL reactor. We estimate parameters of {sup 4}He source with solid methane (CH{sub 4}) or/and liquid deuterium (D{sub 2}) moderator–reflector. We show that such a source with CH{sub 4} moderator–reflector at the PIK reactor would provide the UCN density of ~1·10{sup 5} cm{sup −3}, and the UCN production rate of ~2·10{sup 7} s{sup −1}. These values are respectively 1000 and 20 times larger than those for the most intense UCN user source. The UCN density in a source with D{sub 2} moderator-reflector would reach the value of ~2·10{sup 5} cm{sup −3}, and the UCN production rate would be equal ~8·10{sup 7} s{sup −1}. Installation of such a source in a beam of CNs would slightly increase the density and production rate.

  8. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors

    Science.gov (United States)

    Recktenwald, Geoff; Deinert, Mark

    2010-03-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks.

  9. Analytical complex at the PIK reactor for studying the supra-atomic structure and dynamics of materials by neutron scattering

    Energy Technology Data Exchange (ETDEWEB)

    Lebedev, V. M., E-mail: lebedev@pnpi.spb.ru; Lebedev, V. T.; Ivanova, I. N.; Orlova, D. N. [Russian Academy of Sciences, St. Petersburg Nuclear Physics Institute (Russian Federation)

    2011-12-15

    A project of the center for studying reactor materials and solving problems of materials science is presented which will be equipped with the following neutron instruments: a small-angle Membrana diffractometer, a spin-echo spectrometer, and a time-of-flight spectrometer. It is proposed to irradiate materials in the PIK reactor core and use neutron-scattering tools to analyze the structure and dynamics of these materials and investigate radiative defects in the complete experimental cycle (initial material-irradiation-strength tests, thermal loads, and other effects) using materials science techniques.

  10. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  11. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  12. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

    2008-07-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  13. Reduced physics models in SOLPS for reactor scoping studies

    Energy Technology Data Exchange (ETDEWEB)

    Coster, D.P. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    2016-08-15

    Heat exhaust is a challenge for ITER and becomes even more of an issue for devices beyond ITER. The main reason for this is that the power produced in the core scales as R{sup 3} while relying on standard exhaust physics results in the heat exhaust scaling as R{sup 1} (R is the major radius). ITER has used SOLPS (B2-EIRENE) to design the ITER divertor, as well as to provide a database that supports the calculations of the ITER operational parameter space. The typical run time for such SOLPS runs is of the order 3 months (for D+C+He using EIRENE to treat the neutrals kinetically with an extensive choice of atomic and molecular physics). Future devices will be expected to radiate much of the power before it crosses the separatrix, and this requires treating extrinsic impurities such as Ne, Ar, Kr and Xe - the large number of charge states puts additional pressure on SOLPS, further slowing down the code. For design work of future machines, fast models have been implemented in system codes but these are usually unavoidably restricted in the included physics. As a bridge between system studies and detailed SOLPS runs, SOLPS offers a number of possibilities to speed up the code considerably at the cost of reducing the fidelity of the physics. By employing a fluid neutral model, aggressive bundling of the charge state of impurities, and reducing the size of the grids used, the run time for one second of physics time (which is often enough for the divertor to come to a steady state) can be reduced to approximately one day. This work looks at the impact of these trade-offs in the physics by comparing key parameters for different simulation assumptions. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  14. PIK3C2B inhibition improves function and prolongs survival in myotubular myopathy animal models

    Science.gov (United States)

    Sabha, Nesrin; Volpatti, Jonathan R.; Gonorazky, Hernan; Davidson, Ann E.; Li, Xingli; Eltayeb, Nadine M.; Dall’Armi, Claudia; Di Paolo, Gilbert; Brooks, Susan V.; Buj-Bello, Ana; Feldman, Eva L.; Dowling, James J.

    2016-01-01

    Myotubular myopathy (MTM) is a devastating pediatric neuromuscular disorder of phosphoinositide (PIP) metabolism resulting from mutations of the PIP phosphatase MTM1 for which there are no treatments. We have previously shown phosphatidylinositol-3-phosphate (PI3P) accumulation in animal models of MTM. Here, we tested the hypothesis that lowering PI3P levels may prevent or reverse the MTM disease process. To test this, we targeted class II and III PI3 kinases (PI3Ks) in an MTM1-deficient mouse model. Muscle-specific ablation of Pik3c2b, but not Pik3c3, resulted in complete prevention of the MTM phenotype, and postsymptomatic targeting promoted a striking rescue of disease. We confirmed this genetic interaction in zebrafish, and additionally showed that certain PI3K inhibitors prevented development of the zebrafish mtm phenotype. Finally, the PI3K inhibitor wortmannin improved motor function and prolonged lifespan of the Mtm1-deficient mice. In all, we have identified Pik3c2b as a genetic modifier of Mtm1 mutation and demonstrated that PIK3C2B inhibition is a potential treatment strategy for MTM. In addition, we set the groundwork for similar reciprocal inhibition approaches for treating other PIP metabolic disorders and highlight the importance of modifier gene pathways as therapeutic targets. PMID:27548528

  15. Semiclassical Distorted Wave Model Analysis of the $(\\pi^-,K^+)$ $\\Sigma$ Formation Inclusive Spectrum

    CERN Document Server

    Kohno, M; Kawai, M; Ogata, K; Watanabe, Y

    2006-01-01

    $(\\pi^-,K^+)$ hyperon production inclusive spectra with $p_\\pi =1.2$ GeV/c measured at KEK on $^{12}$C and $^{28}$Si are analyzed by the semiclassical distorted wave model. Single-particle wave functions of the target nucleus are treated using Wigner transformation. This method is able to account for the energy and angular dependences of the elementary process in nuclear medium without introducing the factorization approximation frequently employed. Calculations of the $(\\pi^+,K^+)$ $\\Lambda$ formation process, for which there is no free parameter since the $\\Lambda$ s.p. potential is known, demonstrate that the present model is useful to describe inclusive spectra. It is shown that in order to account for the experimental data of the $\\Sigma^-$ formation spectra a repulsive $\\Sigma$-nucleus potential is necessary whose magnitude is not so strong as around 100 MeV previously suggested.

  16. Mouse models of human PIK3CA-related brain overgrowth have acutely treatable epilepsy

    Science.gov (United States)

    Roy, Achira; Skibo, Jonathan; Kalume, Franck; Ni, Jing; Rankin, Sherri; Lu, Yiling; Dobyns, William B; Mills, Gordon B; Zhao, Jean J; Baker, Suzanne J; Millen, Kathleen J

    2015-01-01

    Mutations in the catalytic subunit of phosphoinositide 3-kinase (PIK3CA) and other PI3K-AKT pathway components have been associated with cancer and a wide spectrum of brain and body overgrowth. In the brain, the phenotypic spectrum of PIK3CA-related segmental overgrowth includes bilateral dysplastic megalencephaly, hemimegalencephaly and focal cortical dysplasia, the most common cause of intractable pediatric epilepsy. We generated mouse models expressing the most common activating Pik3ca mutations (H1047R and E545K) in developing neural progenitors. These accurately recapitulate all the key human pathological features including brain enlargement, cortical malformation, hydrocephalus and epilepsy, with phenotypic severity dependent on the mutant allele and its time of activation. Underlying mechanisms include increased proliferation, cell size and altered white matter. Notably, we demonstrate that acute 1 hr-suppression of PI3K signaling despite the ongoing presence of dysplasia has dramatic anti-epileptic benefit. Thus PI3K inhibitors offer a promising new avenue for effective anti-epileptic therapy for intractable pediatric epilepsy patients. DOI: http://dx.doi.org/10.7554/eLife.12703.001 PMID:26633882

  17. Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes

    Directory of Open Access Journals (Sweden)

    Lindley Benjamin A.

    2016-01-01

    adopts an integral configuration and a fully passive decay heat removal system to provide indefinite cooling capability for a class of accidents. This paper presents the equilibrium cycle core design and reactor physics behaviour of the I2S-LWR with U3Si2 and the advanced steel cladding. The results were obtained using the traditional two-stage approach, in which homogenized macroscopic cross-section sets were generated by WIMS and applied in a full 3D core solution with PANTHER. The results obtained with WIMS/PANTHER were compared against the Monte Carlo Serpent code developed by VTT and previously reported results for the I2S-LWR. The results were found to be in a good agreement (e.g. <200 pcm in reactivity among the compared codes, giving confidence that the WIMS/PANTHER reactor physics package can be reliably used in modelling advanced LWR systems.

  18. Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, David W.; Nielsen, Joseph W.; Norman, Daren R.

    2017-07-01

    The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be well outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.

  19. Assessment of CANDU reactor physics effects using a simplified whole-core MCNP model

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    2002-07-01

    A whole-core Monte Carlo n-particle (MCNP) model of a simplified CANDU reactor was developed and used to study core configurations and reactor physics phenomena of interest in CANDU safety analysis. The resulting reactivity data were compared with values derived from corresponding WIMS-AECL/RFSP, two-neutron-energy-group diffusion theory core simulations, thereby extending the range of CANDU-related code-to-code benchmark comparisons to include whole-core representations. These comparisons show a systematic discrepancy of about 6 mk between the respective absolute k{sub eff} values, but very good agreement to within about -0.15 {+-} 0.06 mk for the reactivity perturbation induced by G-core checkerboard coolant voiding. These findings are generally consistent with the results of much simpler uniform-lattice comparisons involving only WIMS-AECL and MCNP. In addition, MCNP fission-energy tallies were used to evaluate other core-wide properties, such as fuel bundle and total-channel power distributions, as well as intra-bundle details, such as outer-fuel-ring relative power densities and outer-ring fuel element azimuthal power variations, which cannot be determined directly from WIMS-AECL/RFSP core calculations. The average MCNP values for the ratio of outer fuel element to average fuel element power density agreed well with corresponding values derived from WIMS-AECL lattice-cell cases, showing a small systematic discrepancy of about 0.5 %, independent of fuel bum-up. For fuel bundles containing the highest-power fuel elements, the maximum peak-to-average outer-element azimuthal power variation was about 2.5% for cases where a statistically significant trend was observed, while much larger peak-to-average outer-element azimuthal power variations of up to around 42% were observed in low-power fuel bundles at the core/radial-neutron-reflector interface. (author)

  20. Physics-Based Multi-State Models of Passive Component Degradation for the R7 Reactor Simulation Environment

    Energy Technology Data Exchange (ETDEWEB)

    Unwin, Stephen D.; Layton, Robert F.; Johnson, Kenneth I.; Lowry, Peter P.

    2012-06-25

    Abstract: The Next Generation Systems Analysis Code - referred to as R7 - is reactor systems simulation software being developed to support the Risk-Informed Safety Margin Characterization Pathway of the U.S. Department of Energy's Light Water Reactor Sustainability Program. It will provide an integrated multi-physics environment, implemented in an uncertainty quantification (UQ) framework that can produce risk and other performance insights on long-term reactor operations. An element of this simulation environment will be the performance of passive components and materials. Conventional models of component reliability are largely parametric, relying on plant service data to estimate component lifetimes and failure rates. This type of model has limited usefulness in the R7 environment where the intent is to explicitly determine the influence of physical stressors on component degradation. In this paper, we describe a new class of multi-state physics-based component models designed to be R7-compatible. These models capture the physics of materials degradation while also incorporating the effects of interventions and component rejuvenation. The models are implemented in a cumulative damage framework that allows the impact of an evolving physical environment to be addressed without recourse to resampling within the Monte Carlo-based UQ framework. The paper describes an application to stress corrosion cracking in dissimilar metal welds - a principal contributor to potential loss of coolant accidents. So while R7 will have the more conventional capability of reactor simulation codes to model the impact of degraded components and systems on plant performance, the methodology described here allows R7 to model the inverse effect; the impact of the physical environment on component degradation and performance.

  1. Reactor antineutrinos and nuclear physics

    Science.gov (United States)

    Balantekin, A. B.

    2016-11-01

    Short-baseline reactor neutrino experiments successfully measured the neutrino parameters they set out to measure, but they also identified a shape distortion in the 5-7 MeV range as well as a reduction from the predicted value of the flux. Nuclear physics input into the calculations of reactor antineutrino spectra needs to be better refined if this anomaly is to be interpreted as due to sterile neutrino states.

  2. Fundamentals of Nuclear Reactor Physics

    CERN Document Server

    Lewis, E E

    2008-01-01

    This new streamlined text offers a one-semester treatment of the essentials of how the fission nuclear reactor works, the various approaches to the design of reactors, and their safe and efficient operation. The book includes numerous worked-out examples and end-of-chapter questions to help reinforce the knowledge presented. This textbook offers an engineering-oriented introduction to nuclear physics, with a particular focus on how those physics are put to work in the service of generating nuclear-based power, particularly the importance of neutron reactions and neutron behavior. Engin

  3. The research reactors their contribution to the reactors physics; Les reacteurs de recherche leur apport sur la physique des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Zaetta, A. [CEA/Cadarache, Direction des Reacteurs Nucleaires, DRN, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint Paul lez Durance (France); Mathoniere, G. [CEA/Saclay, DEN, 91 - Gif sur Yvette (France)] [and others

    2000-07-01

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  4. Reactor physics modelling of accident tolerant fuel for LWRs using answers codes

    OpenAIRE

    Lindley Benjamin A.; Kotlyar Dan; Parks Geoffrey T.; Lillington John N.; Petrovic Bojan

    2016-01-01

    This is the final version of the article. It first appeared from Springer via http://dx.doi.org/10.1051/epjn/2016012 The majority of nuclear reactors operating in the world today and similarly the majority of near-term new build reactors will be LWRs. These currently accommodate traditional Zr clad UO2/ PuO2 fuel designs which have an excellent performance record for normal operation and most transients. However, the events at Fukushima culminated in significant hydrogen production and hyd...

  5. Trypanosoma cruzi: Genome characterization of phosphatidylinositol kinase gene family (PIK and PIK-related) and identification of a novel PIK gene.

    Science.gov (United States)

    Oliveira, Priscila; Lima, Fabio Mitsuo; Cruz, Mario Costa; Ferreira, Renata Carmona; Sanchez-Flores, Alejandro; Cordero, Esteban Maurício; Cortez, Danielle Rodrigues; Ferreira, Éden Ramalho; Briones, Marcelo Ribeiro da Silva; Mortara, Renato Arruda; da Silveira, José Franco; Bahia, Diana

    2014-07-01

    Chagas disease is caused by the protozoan Trypanosoma cruzi which affects 10 million people worldwide. Very few kinases have been characterized in this parasite, including the phosphatidylinositol kinases (PIKs) that are at the heart of one of the major pathways of intracellular signal transduction. Recently, we have classified the PIK family in T. cruzi using five different models based on the presence of PIK conserved domains. In this study, we have mapped PIK genes to the chromosomes of two different T. cruzi lineages (G and CL Brener) and determined the cellular localization of two PIK members. The kinases have crucial roles in metabolism and are assumed to be conserved throughout evolution. For this reason, they should display a conserved localization within the same eukaryotic species. In spite of this, there is an extensive polymorphism regarding PIK localization at both genomic and cellular levels, among different T. cruzi isolates and between T. cruzi and Trypanosomabrucei, respectively. We showed in this study that the cellular localization of two PIK-related proteins (TOR1 and 2) in the T. cruzi lineage is distinct from that previously observed in T. brucei. In addition, we identified a new PIK gene with peculiar feature, that is, it codes for a FYVE domain at N-terminal position. FYVE-PIK genes are phylogenetically distant from the groups containing exclusively the FYVE or PIK domain. The FYVE-PIK architecture is only present in trypanosomatids and in virus such as Acanthamoeba mimivirus, suggesting a horizontal acquisition. Our Bayesian phylogenetic inference supports this hypothesis. The exact functions of this FYVE-PIK gene are unknown, but the presence of FYVE domain suggests a role in membranous compartments, such as endosome. Taken together, the data presented here strengthen the possibility that trypanosomatids are characterized by extensive genomic plasticity that may be considered in designing drugs and vaccines for prevention of Chagas disease.

  6. Radioactive target needs for nuclear reactor physics and nuclear astrophysics

    OpenAIRE

    Jurado, B.; Barreau, G.; Bacri, C. O.

    2010-01-01

    Nuclear Instruments and Methods in Physics Research Section A - In press.; Nuclear reaction cross sections of short-lived nuclei are key inputs for new generation nuclear reactor simulations and for models describing the nucleosynthesis of elements. After discussing various topics of nuclear astrophysics and reactor physics where the demand of nuclear data on unstable nuclei is strong, we describe the general characteristics of the targets needed to measure the requested data. In some cases t...

  7. Physics modeling support for the International Thermonuclear Experimental Reactor: Final report

    Energy Technology Data Exchange (ETDEWEB)

    1988-09-30

    There are two major sections to this report. The first section of the report is an executive summary of the work done this year. For each task, the major results are condensed for the reader's convenience. The major result of each memo, report or presentation is summarized briefly in this section. The second section of the report is a collection of appendices containing reports, memos, and presentations written this year. Here, the interested reader can investigate any topic discussed in the summary in more detail. The documentation is presented in chronological order, and we would like to note that the content of later documents may supercede that of earlier ones. The summaries are divided into sections, corresponding to the tasks outlined in the original proposal for the work. These sections are: MUMAK code development and application; Alfven wave stability problem; TETRA systems code development and application; lower hybrid heating and current drive; and advanced blanket modeling.

  8. Naval reactors physics handbook. Volume 3: The physics of intermediate spectrum reactors

    Energy Technology Data Exchange (ETDEWEB)

    Stehn, J.R. [ed.] [Knolls Atomic Power Lab., Schenectady, NY (United States)

    1958-09-01

    The present volume has been prepared for persons with some knowledge of the physics of nuclear reactors. It is intended to make available the accumulated physics experience of the Knolls Atomic Power Laboratory in its work on intermediate spectrum reactors. Only those portions have been selected which were deemed to be most useful and significant to other physicists concerned with the problems of reactor design. The volume is divided into four parts which are more or less independent of one another. Part 1 (Chaps. 2--9), Investigation of Reactor Characteristics by Critical Assemblies, reflects the importance of the properties of critical assemblies and of the techniques for obtaining experimental information about such assemblies. Part 2 (Chaps. 10--20), Reactivity Effects Associated with Reactor Operation, details the use of both critical assemblies and reactor theory to make and test predictions of the manner in which the reactivity of an intermediate power reactor will vary during operation. Part 3 (Chaps. 21--26), Heat Generation and Nuclear Materials Problems, considers how reactor heat generation is spread out in space and time, and what nuclear effects result from the presence of beryllium or sodium in the reactor. Part 4 (Chaps. 27--38), Reactor Kinetics and Temperature Coefficients, relates to the transient or near-transient behavior of intermediate reactors.

  9. Reactor physics methods, models, and applications used to support the conceptual design of the Advanced Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    Gehin, J.C.; Worley, B.A.; Renier, J.P. [Oak Ridge National Lab., TN (United States); Wemple, C.A.; Jahshan, S.N.; Ryskammp, J.M. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1995-08-01

    This report summarizes the neutronics analysis performed during 1991 and 1992 in support of characterization of the conceptual design of the Advanced Neutron Source (ANS). The methods used in the analysis, parametric studies, and key results supporting the design and safety evaluations of the conceptual design are presented. The analysis approach used during the conceptual design phase followed the same approach used in early ANS evaluations: (1) a strong reliance on Monte Carlo theory for beginning-of-cycle reactor performance calculations and (2) a reliance on few-group diffusion theory for reactor fuel cycle analysis and for evaluation of reactor performance at specific time steps over the fuel cycle. The Monte Carlo analysis was carried out using the MCNP continuous-energy code, and the few- group diffusion theory calculations were performed using the VENTURE and PDQ code systems. The MCNP code was used primarily for its capability to model the reflector components in realistic geometries as well as the inherent circumvention of cross-section processing requirements and use of energy-collapsed cross sections. The MCNP code was used for evaluations of reflector component reactivity effects and of heat loads in these components. The code was also used as a benchmark comparison against the diffusion-theory estimates of key reactor parameters such as region fluxes, control rod worths, reactivity coefficients, and material worths. The VENTURE and PDQ codes were used to provide independent evaluations of burnup effects, power distributions, and small perturbation worths. The performance and safety calculations performed over the subject time period are summarized, and key results are provided. The key results include flux and power distributions over the fuel cycle, silicon production rates, fuel burnup rates, component reactivities, control rod worths, component heat loads, shutdown reactivity margins, reactivity coefficients, and isotope production rates.

  10. Application of invariant embedding to reactor physics

    CERN Document Server

    Shimizu, Akinao; Parsegian, V L

    1972-01-01

    Application of Invariant Embedding to Reactor Physics describes the application of the method of invariant embedding to radiation shielding and to criticality calculations of atomic reactors. The authors intend to show how this method has been applied to realistic problems, together with the results of applications which will be useful to shielding design. The book is organized into two parts. Part A deals with the reflection and transmission of gamma rays by slabs. The chapters in this section cover topics such as the reflection and transmission problem of gamma rays; formulation of the probl

  11. A model of reactor kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, A.S.; Thompson, B.R.

    1988-09-01

    The analytical model of nuclear reactor transients, incorporating both mechanical and nuclear effects, simulates reactor kinetics. Linear analysis shows the stability borderline for small power perturbations. In a stable system, initial power disturbances die out with time. With an unstable combination of nuclear and mechanical characteristics, initial disturbances persist and may increase with time. With large instability, oscillations of great magnitude occur. Stability requirements set limits on the power density at which particular reactors can operate. The limiting power density depends largely on the product of two terms: the fraction of delayed neutrons and the frictional damping of vibratory motion in reactor core components. As the fraction of delayed neutrons is essentially fixed, mechanical damping largely determines the maximum power density. A computer program, based on the analytical model, calculates and plots reactor power as a nonlinear function of time in response to assigned values of mechanical and nuclear characteristics.

  12. Physics of reactor safety. Volume II. Quarterly report, April-June 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.

  13. Final Physics Report for the Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfe, I. B. [Savannah River Site (SRS), Aiken, SC (United States)

    1956-06-25

    This report is a summary of the physics design work performed on the Engineering Test Reactor. The ETR presents computational difficulties not found in other reactors because of the large number of experimental holes in the core. The physics of the ETR depends strongly upon the contents of the in-core experimental facilities. In order to properly evaluate the reactor; taking into account the experiments in the core, multi-region, two-dimensional calculations are required. These calculations require .the use of a large computer such as the Remington Rand Univac and are complex and expensive enough to warrant a five-stage program: 1. In the early stages of design, only preliminary two-dimensional calculations were performed .in order to obtain a rough idea of the general behavior of the reactor and its critical mass with tentative experiments in place. 2. A large amount of work was carried out in which the reactor was approximated as one with a uniform homogeneous core. With this model, detailed studies were carried out to investigate the feasibility and to obtain general design data on such points as the design and properties of the gray and black-control rods, the design of the beryllium reflector, gamma and neutron heating, the use of burnable poisons, etc. In performing these calculations, use was made of the IBM 650 PROD code obtained from KAPL. 3. With stages 1 and 2 carried out, two-dimensional calculations of the core at start-up conditions were performed on the Univac computer. 4. Detailed two-dimensional calculations of the properties of the ETR with a proposed first set of experiments in place were carried out. 5. A series of nuclear tests were performed at the reactivity measurements facility at the MTR site in order to confirm the validity of the analytical techniques in physics analysis. In performing the two-dimensional Univac calculations, the MUG code developed by KAPL and the Cuthill code developed at the David Taylor Model Basin were utilized. In

  14. Micro reactor physics of MOX fueled LWR

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ. (Japan)

    2001-09-01

    Upon the background that the LWR fuels become complicated in recent years because of the introduction of high burnup fuels, high density Gd fuels, MOX fuels, the author proposes the Micro Reactor Physics. He intends to investigate the behaviors of neutrons and reactions in a pin rod that have not yet been paid attention. Conventionally the resonance absorption has been evaluated by assuming the uniform effective cross sections in a pin rod. However, due to the self-shielding, the neutron spectrum near the surface of the rod is quite different with that of the center of rod. This fact affects the spatial distributions of Pu isotopes produced during burnup. The spatial distribution of temperature in a rod affects the Doppler coefficient. He solved this problem by the multi-band method. In the case where MOX rods are adjacent with U rods, the spectrum of the current from MOX rods to U rods is different with that of U to MOX. That makes the spatial distribution of azimuthal direction together with that of the infinite lattice. He solved this problem by a cell calculation based on the characteristic method. This report introduces several numerical results of his Micro Reactor Physics. One of the important results is the indication that the conventional Doppler coefficient gives 20% higher (not conservative) value. (K. Tsuchihashi)

  15. Micro reactor physics of MOX fueled LWR

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ. (Japan)

    2001-09-01

    Upon the background that the LWR fuels become complicated in recent years because of the introduction of high burnup fuels, high density Gd fuels, MOX fuels, the author proposes the Micro Reactor Physics. He intends to investigate the behaviors of neutrons and reactions in a pin rod that have not yet been paid attention. Conventionally the resonance absorption has been evaluated by assuming the uniform effective cross sections in a pin rod. However, due to the self-shielding, the neutron spectrum near the surface of the rod is quite different with that of the center of rod. This fact affects the spatial distributions of Pu isotopes produced during burnup. The spatial distribution of temperature in a rod affects the Doppler coefficient. He solved this problem by the multi-band method. In the case where MOX rods are adjacent with U rods, the spectrum of the current from MOX rods to U rods is different with that of U to MOX. That makes the spatial distribution of azimuthal direction together with that of the infinite lattice. He solved this problem by a cell calculation based on the characteristic method. This report introduces several numerical results of his Micro Reactor Physics. One of the important results is the indication that the conventional Doppler coefficient gives 20% higher (not conservative) value. (K. Tsuchihashi)

  16. Search for New Physics in reactor and accelerator experiments

    Science.gov (United States)

    Di Iura, A.; Girardi, I.; Meloni, D.

    2016-01-01

    We consider two scenarios of New Physics: the Large Extra Dimensions (LED), where sterile neutrinos can propagate in a (4+d) -dimensional space-time, and the Non Standard Interactions (NSI), where the neutrino interactions with ordinary matter are parametrized at low energy in terms of effective flavour-dependent complex couplings \\varepsilon_{αβ} . We study how these models have an impact on oscillation parameters in reactor and accelerator experiments.

  17. Modeling Chemical Reactors I: Quiescent Reactors

    CERN Document Server

    Michoski, C E; Schmitz, P G

    2010-01-01

    We introduce a fully generalized quiescent chemical reactor system in arbitrary space $\\vdim =1,2$ or 3, with $n\\in\\mathbb{N}$ chemical constituents $\\alpha_{i}$, where the character of the numerical solution is strongly determined by the relative scaling between the local reactivity of species $\\alpha_{i}$ and the local functional diffusivity $\\mathscr{D}_{ij}(\\alpha)$ of the reaction mixture. We develop an operator time-splitting predictor multi-corrector RK--LDG scheme, and utilize $hp$-adaptivity relying only on the entropy $\\mathscr{S}_{\\mathfrak{R}}$ of the reactive system $\\mathfrak{R}$. This condition preserves these bounded nonlinear entropy functionals as a necessarily enforced stability condition on the coupled system. We apply this scheme to a number of application problems in chemical kinetics; including a difficult classical problem arising in nonequilibrium thermodynamics known as the Belousov-Zhabotinskii reaction where we utilize a concentration-dependent diffusivity tensor $\\mathscr{D}_{ij}(...

  18. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  19. Sensitivity and Uncertainty Analysis of Coupled Reactor Physics Problems: Method Development for Multi-Physics in Reactors

    NARCIS (Netherlands)

    Perkó, Z.

    2015-01-01

    This thesis presents novel adjoint and spectral methods for the sensitivity and uncertainty (S&U) analysis of multi-physics problems encountered in the field of reactor physics. The first part focuses on the steady state of reactors and extends the adjoint sensitivity analysis methods well establish

  20. Sensitivity and Uncertainty Analysis of Coupled Reactor Physics Problems: Method Development for Multi-Physics in Reactors

    NARCIS (Netherlands)

    Perkó, Z.

    2015-01-01

    This thesis presents novel adjoint and spectral methods for the sensitivity and uncertainty (S&U) analysis of multi-physics problems encountered in the field of reactor physics. The first part focuses on the steady state of reactors and extends the adjoint sensitivity analysis methods well

  1. Advanced reactor physics methods for heterogeneous reactor cores

    Science.gov (United States)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  2. New applications of neutron noise theory in power reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, Vasiliy

    2000-04-01

    methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. By numerical modelling it was found that the applicability of the reactor physics approximations is better than in critical systems. Another interesting problem in neutron noise theory, which recently attracts more and more attention, is the treatment of moving boundaries. In this case one needs to redefine such common methods in reactor physics as point kinetic and adiabatic approximations because, generally speaking, the various functions involved have different regions of definition. The thesis as well presents one possible line of developing the general theory of linear kinetics as applied to systems with varying size. It also generalises the flux factorisation and develops further the Green's function technique.

  3. Cold fusion reactors and new modern physics

    Directory of Open Access Journals (Sweden)

    Huang Zhenqiang Huang Yuxiang

    2013-10-01

    Full Text Available The author of the "modern physics classical particle quantization orbital motion model general solution", referred to as the “new modern physics” a book. “The nuclear force constraint inertial guidance cold nuclear fusion collides” patent of invention referred to as the “cold nuclear fusion reactor” detailed technical data. Now provide to you, hope you help spread and the mainstream of modern physics of academic and fusion engineering academic communication. We work together to promote the cause of science and technology progress of mankind to contribute

  4. Two modelling approaches to water-quality simulation in a flooded iron-ore mine (Saizerais, Lorraine, France): a semi-distributed chemical reactor model and a physically based distributed reactive transport pipe network model.

    Science.gov (United States)

    Hamm, V; Collon-Drouaillet, P; Fabriol, R

    2008-02-19

    The flooding of abandoned mines in the Lorraine Iron Basin (LIB) over the past 25 years has degraded the quality of the groundwater tapped for drinking water. High concentrations of dissolved sulphate have made the water unsuitable for human consumption. This problematic issue has led to the development of numerical tools to support water-resource management in mining contexts. Here we examine two modelling approaches using different numerical tools that we tested on the Saizerais flooded iron-ore mine (Lorraine, France). A first approach considers the Saizerais Mine as a network of two chemical reactors (NCR). The second approach is based on a physically distributed pipe network model (PNM) built with EPANET 2 software. This approach considers the mine as a network of pipes defined by their geometric and chemical parameters. Each reactor in the NCR model includes a detailed chemical model built to simulate quality evolution in the flooded mine water. However, in order to obtain a robust PNM, we simplified the detailed chemical model into a specific sulphate dissolution-precipitation model that is included as sulphate source/sink in both a NCR model and a pipe network model. Both the NCR model and the PNM, based on different numerical techniques, give good post-calibration agreement between the simulated and measured sulphate concentrations in the drinking-water well and overflow drift. The NCR model incorporating the detailed chemical model is useful when a detailed chemical behaviour at the overflow is needed. The PNM incorporating the simplified sulphate dissolution-precipitation model provides better information of the physics controlling the effect of flow and low flow zones, and the time of solid sulphate removal whereas the NCR model will underestimate clean-up time due to the complete mixing assumption. In conclusion, the detailed NCR model will give a first assessment of chemical processes at overflow, and in a second time, the PNM model will provide more

  5. Neutrino Mixing Discriminates Geo-reactor Models

    CERN Document Server

    Dye, S T

    2009-01-01

    Geo-reactor models suggest the existence of natural nuclear reactors at different deep-earth locations with loosely defined output power. Reactor fission products undergo beta decay with the emission of electron antineutrinos, which routinely escape the earth. Neutrino mixing distorts the energy spectrum of the electron antineutrinos. Characteristics of the distorted spectrum observed at the earth's surface could specify the location of a geo-reactor, discriminating the models and facilitating more precise power measurement. The existence of a geo-reactor with known position could enable a precision measurement of the neutrino oscillation parameter delta-mass-squared.

  6. The prevalence of PIK3CA mutations in gastric and colon cancer

    NARCIS (Netherlands)

    Velho, S; Oliveira, C; Ferreira, A; Ferreira, AC; Suriano, G; Schwartz, S; Duval, A; Carneiro, F; Machado, JC; Hamelin, R; Seruca, R

    2005-01-01

    A wide variety of tumours show PIK3CA mutations leading to increased phosphatidylinositol-3 kinase (PI3K) activity. We have determined the frequency of PIK3CA mutations in exons 9 and 20 that has previously been reported as mutational hotspot regions in distinct tumour models. One hundred and fifty

  7. Current status of fast reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, H.H.

    1979-01-01

    The subject of calculation of reactivity coefficients for fast reactors is developed, starting with a discussion of the status of relevant nuclear data and proceeding to the subjects of group cross section generation and of methods of obtaining reactivity coefficients from group cross sections. Reactivity coefficients measured in critical experiments are compared with calculated values. Dependence of reactivity coefficients on reactor design is discussed. Finally, results of the recent international comparison of calculated reactivity coefficients are presented.

  8. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    Science.gov (United States)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  9. RSMASS-D models: An improved method for estimating reactor and shield mass for space reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C.

    1997-10-01

    Three relatively simple mathematical models have been developed to estimate minimum reactor and radiation shield masses for liquid-metal-cooled reactors (LMRs), in-core thermionic fuel element (TFE) reactors, and out-of-core thermionic reactors (OTRs). The approach was based on much of the methodology developed for the Reactor/Shield Mass (RSMASS) model. Like the original RSMASS models, the new RSMASS-derivative (RSMASS-D) models use a combination of simple equations derived from reactor physics and other fundamental considerations, along with tabulations of data from more detailed neutron and gamma transport theory computations. All three models vary basic design parameters within a range specified by the user to achieve a parameter choice that yields a minimum mass for the power level and operational time of interest. The impact of critical mass, fuel damage, and thermal limitations are accounted for to determine the required fuel mass. The effect of thermionic limitations are also taken into account for the thermionic reactor models. All major reactor component masses are estimated, as well as instrumentation and control (I&C), boom, and safety system masses. A new shield model was developed and incorporated into all three reactor concept models. The new shield model is more accurate and simpler to use than the approach used in the original RSMASS model. The estimated reactor and shield masses agree with the mass predictions from separate detailed calculations within 15 percent for all three models.

  10. SANA - project results and PIK contributions

    Energy Technology Data Exchange (ETDEWEB)

    Bellmann, K.; Erhard, M.; Flechsig, M.; Grote, R.; Suckow, F.

    1998-03-01

    This report includes the final project results of the two groups at PIK, involved in the project: Firstly, the newly developed physiologically-based forest growth model FORSANA was applied for the first time to three pine stands, which differed largely in their air pollution and deposition history. (The evaluation of the model is presented in PIK Report 32). The model was able to explain the growth during the last decades of at least two of the three stands from the climatic and deposition conditions at the sites. The third site was shown to be exceptional with respect to its relation between dimension and age, and was supposed to be exposed to major disturbances in the past, which could not be accounted for by the model. To extrapolate from the stand level to the regional level, FORSANA was initialised with spatially explicit data from forestry inventory and soil maps. Simulations were executed with measured weather records and regional distributions of deposition and air pollution, which were estimated on the basis of emission inventories and wind directions. Different assumptions about the development of air pollution had been applied to investigate different pollution abatement strategies. The results showed that a positive effect can be expected from the actual emission reductions close the main centres of emission, but showed also that this effect is decreasing with increasing distance from the emission source. (orig./KW)

  11. Neutrino Physics with Accelerator Driven Subcritical Reactors

    CERN Document Server

    Ciuffoli, Emilio; Zhao, Fengyi

    2015-01-01

    Accelerator driven system (ADS) subcritical nuclear reactors are under development around the world. They will be intense sources of free, 30-50 MeV antimuon decay at rest antimuon neutrinos. These ADS reactor neutrinos can provide a robust test of the LSND anomaly and a precise measurement of the leptonic CP-violating phase delta, including sign(cos(delta)). The first phase of many ADS programs includes the construction of a low energy, high intensity proton or deuteron accelerator, which can yield competitive bounds on sterile neutrinos.

  12. Once-through CANDU reactor models for the ORIGEN2 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % /sup 235/U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given.

  13. Monte Carlo modelling of TRIGA research reactor

    Science.gov (United States)

    El Bakkari, B.; Nacir, B.; El Bardouni, T.; El Younoussi, C.; Merroun, O.; Htet, A.; Boulaich, Y.; Zoubair, M.; Boukhal, H.; Chakir, M.

    2010-10-01

    The Moroccan 2 MW TRIGA MARK II research reactor at Centre des Etudes Nucléaires de la Maâmora (CENM) achieved initial criticality on May 2, 2007. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the neutronic analysis of the 2-MW TRIGA MARK II research reactor at CENM and validation of the results by comparisons with the experimental, operational, and available final safety analysis report (FSAR) values. The study was prepared in collaboration between the Laboratory of Radiation and Nuclear Systems (ERSN-LMR) from Faculty of Sciences of Tetuan (Morocco) and CENM. The 3-D continuous energy Monte Carlo code MCNP (version 5) was used to develop a versatile and accurate full model of the TRIGA core. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S( α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross-section libraries were generated by using the NJOY99 system updated to its more recent patch file "up259". The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. Core excess reactivity, total and integral control rods worth as well as power peaking factors were used in the validation process. Results of calculations are analysed and discussed.

  14. Physics of nuclear reactors; La physique des reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Marguet, S. [Ecole Nationale Superieure de Risques Industriels de Bourges, 18 (France); Institut de Transfert de Technologie d' EDF, 92 - Clamart (France)

    2011-07-01

    This manual covers all the aspects of the science of neutron transport in nuclear reactors and can be used with great advantage by students, engineers or even reactor experts. It is composed of 18 chapters: 1) basis of nuclear physics, 2) the interactions of neutrons with matter, 3) the interactions of electromagnetic radiations and charged-particles with matter, 4) neutron slowing-down, 5) resonant absorption, 6) Doppler effect, 7) neutron thermalization, 8) Boltzmann equation, 9) calculation methods in neutron transport theory, 10) neutron scattering, 11) reactor reactivity, 12) theory of the critical homogenous pile, 13) the neutron reflector, 14) the heterogeneous reactor, 15) the equations of the fuel cycle, 16) neutron counter-reactions, 17) reactor kinetics, and 18) calculation methods in neutron scattering

  15. The dual PI3K/mTOR inhibitor NVP-BEZ235 induces tumor regression in a genetically engineered mouse model of PIK3CA wild-type colorectal cancer.

    Directory of Open Access Journals (Sweden)

    Jatin Roper

    Full Text Available To examine the in vitro and in vivo efficacy of the dual PI3K/mTOR inhibitor NVP-BEZ235 in treatment of PIK3CA wild-type colorectal cancer (CRC.PIK3CA mutant and wild-type human CRC cell lines were treated in vitro with NVP-BEZ235, and the resulting effects on proliferation, apoptosis, and signaling were assessed. Colonic tumors from a genetically engineered mouse (GEM model for sporadic wild-type PIK3CA CRC were treated in vivo with NVP-BEZ235. The resulting effects on macroscopic tumor growth/regression, proliferation, apoptosis, angiogenesis, and signaling were examined.In vitro treatment of CRC cell lines with NVP-BEZ235 resulted in transient PI3K blockade, sustained decreases in mTORC1/mTORC2 signaling, and a corresponding decrease in cell viability (median IC(50 = 9.0-14.3 nM. Similar effects were seen in paired isogenic CRC cell lines that differed only in the presence or absence of an activating PIK3CA mutant allele. In vivo treatment of colonic tumor-bearing mice with NVP-BEZ235 resulted in transient PI3K inhibition and sustained blockade of mTORC1/mTORC2 signaling. Longitudinal tumor surveillance by optical colonoscopy demonstrated a 97% increase in tumor size in control mice (p = 0.01 vs. a 43% decrease (p = 0.008 in treated mice. Ex vivo analysis of the NVP-BEZ235-treated tumors demonstrated a 56% decrease in proliferation (p = 0.003, no effects on apoptosis, and a 75% reduction in angiogenesis (p = 0.013.These studies provide the preclinical rationale for studies examining the efficacy of the dual PI3K/mTOR inhibitor NVP-BEZ235 in treatment of PIK3CA wild-type CRC.

  16. Operating manual for the Health Physics Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1985-11-01

    This manual is intended to serve as a guide in the operation and maintenance of the Health Physics Researh Reactor (HPRR) of the Health Physics Dosimetry Applications Research (DOSAR) Facility. It includes descriptions of the HPRR and of associated equipment such as the reactor positioning devises and the derrick. Procedures for routine operation of the HPRR are given in detail, and checklists for the various steps are provided where applicable. Emergency procedures are similarly covered, and maintenance schedules are outlined. Also, a bibliography of references giving more detailed information on the DOSAR Facility is included. Changes to this manual will be approved by at least two of the following senior staff members: (1) the Operations Division Director, (2) the Reactor Operations Department Head, (3) the Supervisor of Reactor Operations TSF-HPRR Areas. The master copy and the copy of the manual issued to the HPRR Operations Supervisor will always reflect the latest revision. 22 figs.

  17. Fundamental interactions involving neutrons and neutrinos: reactor-based studies led by Petersburg Nuclear Physics Institute (National Research Centre 'Kurchatov Institute') [PNPI (NRC KI)

    Science.gov (United States)

    Serebrov, A. P.

    2015-11-01

    Neutrons of very low energy ( ˜ 10-7 eV), commonly known as ultracold, are unique in that they can be stored in material and magnetic traps, thus enhancing methodical opportunities to conduct precision experiments and to probe the fundamentals of physics. One of the central problems of physics, of direct relevance to the formation of the Universe, is the violation of time invariance. Experiments searching for the nonzero neutron electric dipole moment serve as a time invariance test, and the use of ultracold neutrons provides very high measurement precision. Precision neutron lifetime measurements using ultracold neutrons are extremely important for checking ideas on the early formation of the Universe. This paper discusses problems that arise in studies using ultracold neutrons. Also discussed are the currently highly topical problem of sterile neutrinos and the search for reactor antineutrino oscillations at distances of 6-12 meters from the reactor core. The field reviewed is being investigated at multiple facilities globally. The present paper mainly concentrates on the results of PNPI-led studies at WWR-M PNPI (Gatchina), ILL (Grenoble), and SM-3 (Dimitrovgrad) reactors, and also covers the results obtained during preparation for research at the PIK reactor which is under construction.

  18. Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)

    2013-07-01

    A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

  19. The physics of accelerator driven sub-critical reactors

    Indian Academy of Sciences (India)

    S B Degweker; Biplab Ghosh; Anil Bajpal; S D Pranjape

    2007-02-01

    In recent years, there has been an increasing worldwide interest in accelerator driven systems (ADS) due to their perceived superior safety characteristics and their potential for burning actinides and long-lived fission products. Indian interest in ADS has an additional dimension, which is related to our planned large-scale thorium utilization for future nuclear energy generation. The physics of ADS is quite different from that of critical reactors. As such, physics studies on ADS reactors are necessary for gaining an understanding of these systems. Development of theoretical tools and experimental facilities for studying the physics of ADS reactors constitute important aspect of the ADS development program at BARC. This includes computer codes for burnup studies based on transport theory and Monte Carlo methods, codes for studying the kinetics of ADS and sub-critical facilities driven by 14 MeV neutron generators for ADS experiments and development of sub-criticality measurement methods. The paper discusses the physics issues specific to ADS reactors and presents the status of the reactor physics program and some of the ADS concepts under study.

  20. Nuclear data and reactor physics activities in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Liem, P.H. [National Atomic Energy Agency, Tangerang (Indonesia). Center for Multipurpose Reactor

    1998-03-01

    The nuclear data and reactor physics activities in Indonesia, especially, in the National Atomic Energy Agency are presented. In the nuclear data field, the Agency is now taking the position of a user of the main nuclear data libraries such as JENDL and ENDF/B. These nuclear data libraries become the main sources for producing problem dependent cross section sets that are needed by cell calculation codes or transport codes for design, analysis and safety evaluation of research reactors. In the reactor physics field, besides utilising the existing core analysis codes obtained from bilateral and international co-operation, the Agency is putting much effort to self-develop Batan`s codes for reactor physics calculations, in particular, for research reactor and high temperature reactor design, analysis and fuel management. Under the collaboration with JAERI, Monte Carlo criticality calculations on the first criticality of RSG GAS (MPR-30) first core were done using JAERI continuous energy, vectorized Monte Carlo code, MVP, with JENDL-3.1 and JENDL-3.2 nuclear data libraries. The results were then compared with the experiment data collected during the commissioning phase. Monte Carlo calculations with both JENDL-3.1 and -3.2 libraries produced k{sub eff} values with excellent agreement with experiment data, however, systematically, JENDL-3.2 library showed slightly higher k{sub eff} values than JENDL-3.1 library. (author)

  1. Optimization of an ionized metal physical vapor deposition reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lu, J.; Kushner, M.J. [Univ. of Illinois, Urbana, IL (United States)

    1998-12-31

    Conventional sputtering for microelectronic fabrication produces poorly collimated neutral atom fluxes. Ion fluxes, however, can be accelerated and collimated by using a conventional dc or rf substrate bias. Hence, magnetron ionized metal physical vapor deposition (IMPVD) can produce highly ionized metal fluxes that can be used to fill high-aspect-ratio vias and trenches in microelectronic devices. Hopwood and Qian have examined design issues in IMPVD systems. In this study, a Design of Experiment (DOE) has been numerically performed for an IMPVD reactor using an inductively coupled plasma and a capacitively biased substrate. Gas pressure, reactor geometry, ICP power, and number of inductive coils are the design variables. Uniformity, magnitude, and ionization fraction of the depositing fluxes are the response variables. The influence of the design variables on the response variables is examined, with the goals of obtaining high uniformity, high magnitude, and high ionization fraction of the depositing metal fluxes. The computational tool used in this study is the two-dimensional Hybrid Plasma Equipment Model (HPEM). The aspect ratio of the reactor (height/radius) ranges from 0.5 to 1.0, the gas pressure ranges from 10 to 40 mTorr, the ICP power ranges from 0.5 to 2.0 kW, and the number of ICP coils ranges from 2 to 6. It was found that: (a) uniformity maximizes at high aspect ratio, low power, and high pressure; (b) flux magnitude maximizes at low aspect ratio, high power, and low pressure; (c) ionization fraction maximizes at high aspect ratio, high power, and high pressure.

  2. Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Bess, J. D.; Briggs, J. B.; Gulliford, J.; Ivanova, T.; Rozhikhin, E. V.; Semenov, M. Yu.; Tsibulya, A. M.; Koscheev, V. N.

    2017-07-01

    Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energy Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning

  3. Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem

    Energy Technology Data Exchange (ETDEWEB)

    Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Grudzinski, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-12-21

    This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.

  4. Dynamic model of Fast Breeder Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vaidyanathan, G., E-mail: vaidya@igcar.gov.i [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Kasinathan, N.; Velusamy, K. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)

    2010-04-15

    Fast Breeder Test Reactor (FBTR) is a 40 M Wt/13.2 MWe sodium cooled reactor operating since 1985. It is a loop type reactor. As part of the safety analysis the response of the plant to various transients is needed. In this connection a computer code named DYNAM was developed to model the reactor core, the intermediate heat exchanger, steam generator, piping, etc. This paper deals with the mathematical model of the various components of FBTR, the numerical techniques to solve the model, and comparison of the predictions of the code with plant measurements. Also presented is the benign response of the plant to a station blackout condition, which brings out the role of the various reactivity feedback mechanisms combined with a gradual coast down of reactor sodium flow.

  5. Chemical reactor modeling multiphase reactive flows

    CERN Document Server

    Jakobsen, Hugo A

    2014-01-01

    Chemical Reactor Modeling closes the gap between Chemical Reaction Engineering and Fluid Mechanics.  The second edition consists of two volumes: Volume 1: Fundamentals. Volume 2: Chemical Engineering Applications In volume 1 most of the fundamental theory is presented. A few numerical model simulation application examples are given to elucidate the link between theory and applications. In volume 2 the chemical reactor equipment to be modeled are described. Several engineering models are introduced and discussed. A survey of the frequently used numerical methods, algorithms and schemes is provided. A few practical engineering applications of the modeling tools are presented and discussed. The working principles of several experimental techniques employed in order to get data for model validation are outlined. The monograph is based on lectures regularly taught in the fourth and fifth years graduate courses in transport phenomena and chemical reactor modeling, and in a post graduate course in modern reactor m...

  6. PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, L.; Diamond, D.; Xu, J.; Carew, J.; Rorer, D.

    2004-03-31

    Detailed reactor physics and safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron and photon transport calculations were performed with the MCNP code to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim safety arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim safety arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the

  7. Mutations in PIK3CA are infrequent in neuroblastoma

    Directory of Open Access Journals (Sweden)

    Mazanek Pavel

    2006-07-01

    Full Text Available Abstract Background Neuroblastoma is a frequently lethal pediatric cancer in which MYCN genomic amplification is highly correlated with aggressive disease. Deregulated MYC genes require co-operative lesions to foster tumourigenesis and both direct and indirect evidence support activated Ras signaling for this purpose in many cancers. Yet Ras genes and Braf, while often activated in cancer cells, are infrequent targets for activation in neuroblastoma. Recently, the Ras effector PIK3CA was shown to be activated in diverse human cancers. We therefore assessed PIK3CA for mutation in human neuroblastomas, as well as in neuroblastomas arising in transgenic mice with MYCN overexpressed in neural-crest tissues. In this murine model we additionally surveyed for Ras family and Braf mutations as these have not been previously reported. Methods Sixty-nine human neuroblastomas (42 primary tumors and 27 cell lines were sequenced for PIK3CA activating mutations within the C2, helical and kinase domain "hot spots" where 80% of mutations cluster. Constitutional DNA was sequenced in cases with confirmed alterations to assess for germline or somatic acquisition. Additionally, Ras family members (Hras1, Kras2 and Nras and the downstream effectors Pik3ca and Braf, were sequenced from twenty-five neuroblastomas arising in neuroblastoma-prone transgenic mice. Results We identified mutations in the PIK3CA gene in 2 of 69 human neuroblastomas (2.9%. Neither mutation (R524M and E982D has been studied to date for effects on lipid kinase activity. Though both occurred in tumors with MYCN amplification the overall rate of PIK3CA mutations in MYCN amplified and single-copy tumors did not differ appreciably (2 of 31 versus 0 of 38, respectively. Further, no activating mutations were identified in a survey of Ras signal transduction genes (including Hras1, Kras2, Nras, Pik3ca, or Braf genes in twenty-five neuroblastic tumors arising in the MYCN-initiated transgenic mouse model

  8. Current Reactor Physics Benchmark Activities at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Margaret A. Marshall; Mackenzie L. Gorham; Joseph Christensen; James C. Turnbull; Kim Clark

    2011-11-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) [1] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) [2] were established to preserve integral reactor physics and criticality experiment data for present and future research. These valuable assets provide the basis for recording, developing, and validating our integral nuclear data, and experimental and computational methods. These projects are managed through the Idaho National Laboratory (INL) and the Organisation for Economic Co-operation and Development Nuclear Energy Agency (OECD-NEA). Staff and students at the Department of Energy - Idaho (DOE-ID) and INL are engaged in the development of benchmarks to support ongoing research activities. These benchmarks include reactors or assemblies that support Next Generation Nuclear Plant (NGNP) research, space nuclear Fission Surface Power System (FSPS) design validation, and currently operational facilities in Southeastern Idaho.

  9. Recent improvements of reactor physics codes in MHI

    Energy Technology Data Exchange (ETDEWEB)

    Kosaka, Shinya, E-mail: shinya-kosaka@mhi.co.jp; Yamaji, Kazuya; Kirimura, Kazuki; Kamiyama, Yohei; Matsumoto, Hideki [Mitsubishi Heavy Industries, Ltd. (Japan)

    2015-12-31

    This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO’s Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipated transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented.

  10. Recent improvements of reactor physics codes in MHI

    Science.gov (United States)

    Kosaka, Shinya; Yamaji, Kazuya; Kirimura, Kazuki; Kamiyama, Yohei; Matsumoto, Hideki

    2015-12-01

    This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO's Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipated transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented.

  11. An activating Pik3ca mutation coupled with Pten loss is sufficient to initiate ovarian tumorigenesis in mice.

    Science.gov (United States)

    Kinross, Kathryn M; Montgomery, Karen G; Kleinschmidt, Margarete; Waring, Paul; Ivetac, Ivan; Tikoo, Anjali; Saad, Mirette; Hare, Lauren; Roh, Vincent; Mantamadiotis, Theo; Sheppard, Karen E; Ryland, Georgina L; Campbell, Ian G; Gorringe, Kylie L; Christensen, James G; Cullinane, Carleen; Hicks, Rodney J; Pearson, Richard B; Johnstone, Ricky W; McArthur, Grant A; Phillips, Wayne A

    2012-02-01

    Mutations in the gene encoding the p110α subunit of PI3K (PIK3CA) that result in enhanced PI3K activity are frequently observed in human cancers. To better understand the role of mutant PIK3CA in the initiation or progression of tumorigenesis, we generated mice in which a PIK3CA mutation commonly detected in human cancers (the H1047R mutation) could be conditionally knocked into the endogenous Pik3ca locus. Activation of this mutation in the mouse ovary revealed that alone, Pik3caH1047R induced premalignant hyperplasia of the ovarian surface epithelium but no tumors. Concomitantly, we analyzed several human ovarian cancers and found PIK3CA mutations coexistent with KRAS and/or PTEN mutations, raising the possibility that a secondary defect in a co-regulator of PI3K activity may be required for mutant PIK3CA to promote transformation. Consistent with this notion, we found that Pik3caH1047R mutation plus Pten deletion in the mouse ovary led to the development of ovarian serous adenocarcinomas and granulosa cell tumors. Both mutational events were required for early, robust Akt activation. Pharmacological inhibition of PI3K/mTOR in these mice delayed tumor growth and prolonged survival. These results demonstrate that the Pik3caH1047R mutation with loss of Pten is enough to promote ovarian cell transformation and that we have developed a model system for studying possible therapies.

  12. Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Merzari, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Obabko, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Jain, Rajeev [Argonne National Lab. (ANL), Argonne, IL (United States); Mahadevan, Vijay [Argonne National Lab. (ANL), Argonne, IL (United States); Tautges, Timothy [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Solberg, Jerome [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ferencz, Robert Mark [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Whitesides, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-12-21

    This report describes to employ SHARP to perform a first-of-a-kind analysis of the core radial expansion phenomenon in an SFR. This effort required significant advances in the framework Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit used to drive the coupled simulations, manipulate the mesh in response to the deformation of the geometry, and generate the necessary modified mesh files. Furthermore, the model geometry is fairly complex, and consistent mesh generation for the three physics modules required significant effort. Fully-integrated simulations of a 7-assembly mini-core test problem have been performed, and the results are presented here. Physics models of a full-core model of the Advanced Burner Test Reactor have also been developed for each of the three physics modules. Standalone results of each of the three physics modules for the ABTR are presented here, which provides a demonstration of the feasibility of the fully-integrated simulation.

  13. Antitumor Efficacy of the Dual PI3K/mTOR Inhibitor PF-04691502 in a Human Xenograft Tumor Model Derived from Colorectal Cancer Stem Cells Harboring a PIK3CA Mutation.

    Directory of Open Access Journals (Sweden)

    Douglas D Fang

    Full Text Available PIK3CA (phosphoinositide-3-kinase, catalytic, alpha polypeptide mutations can help predict the antitumor activity of phosphatidylinositol-3-kinase (PI3K/mammalian target of rapamycin (mTOR pathway inhibitors in both preclinical and clinical settings. In light of the recent discovery of tumor-initiating cancer stem cells (CSCs in various tumor types, we developed an in vitro CSC model from xenograft tumors established in mice from a colorectal cancer patient tumor in which the CD133+/EpCAM+ population represented tumor-initiating cells. CD133+/EpCAM+ CSCs were enriched under stem cell culture conditions and formed 3-dimensional tumor spheroids. Tumor spheroid cells exhibited CSC properties, including the capability for differentiation and self-renewal, higher tumorigenic potential and chemo-resistance. Genetic analysis using an OncoCarta™ panel revealed a PIK3CA (H1047R mutation in these cells. Using a dual PI3K/mTOR inhibitor, PF-04691502, we then showed that blockage of the PI3K/mTOR pathway inhibited the in vitro proliferation of CSCs and in vivo xenograft tumor growth with manageable toxicity. Tumor growth inhibition in mice was accompanied by a significant reduction of phosphorylated Akt (pAKT (S473, a well-established surrogate biomarker of PI3K/mTOR signaling pathway inhibition. Collectively, our data suggest that PF-04691502 exhibits potent anticancer activity in colorectal cancer by targeting both PIK3CA (H1047R mutant CSCs and their derivatives. These results may assist in the clinical development of PF-04691502 for the treatment of a subpopulation of colorectal cancer patients with poor outcomes.

  14. Nuclear reactor core modelling in multifunctional simulators

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

    1999-06-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  15. The International Reactor Physics Experiment Evaluation Project (IRPHEP)

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Enrico Sartori; Lori Scott

    2006-09-01

    Since the beginning of the Nuclear Power industry, numerous experiments concerned with nuclear energy and technology have been performed at different research laboratories, worldwide. These experiments required a large investment in terms of infrastructure, expertise, and cost; however, many were performed without a high degree of attention to archival of results for future use. The degree and quality of documentation varies greatly. There is an urgent need to preserve integral reactor physics experimental data, including measurement methods, techniques, and separate or special effects data for nuclear energy and technology applications and the knowledge and competence contained therein. If the data are compromised, it is unlikely that any of these experiments will be repeated again in the future. The International Reactor Physics Evaluation Project (IRPhEP) was initiated, as a pilot activity in 1999 by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June of 2003. The purpose of the IRPhEP is to provide an extensively peer reviewed set of reactor physics related integral benchmark data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next generation reactors and establish the safety basis for operation of these reactors. A short history of the IRPhEP is presented and its purposes are discussed in this paper. Accomplishments of the IRPhEP, including the first publication of the IRPhEP Handbook, are highlighted and the future of the project outlined.

  16. Effects of Spatial Variations in Packing Fraction on Reactor Physics Parameters in Pebble-Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    William K. Terry; A. M. Ougouag; Farzad Rahnema; Michael Scott McKinley

    2003-04-01

    The well-known spatial variation of packing fraction near the outer boundary of a pebble-bed reactor core is cited. The ramifications of this variation are explored with the MCNP computer code. It is found that the variation has negligible effects on the global reactor physics parameters extracted from the MCNP calculations for use in analysis by diffusion-theory codes, but for local reaction rates the effects of the variation are naturally important. Included is some preliminary work in using first-order perturbation theory for estimating the effect of the spatial variation of packing fraction on the core eigenvalue and the fision density distribution.

  17. Modeling of Flow in Nuclear Reactor Fuel Cell Outlet

    Directory of Open Access Journals (Sweden)

    František URBAN

    2010-12-01

    Full Text Available Safe and effective load of nuclear reactor fuel cells demands qualitative and quantitative analysis of relations between coolant temperature in fuel cell outlet temperature measured by thermocouple and middle temperature of coolant in thermocouple plane position. In laboratory at Insitute of thermal power engineering of the Slovak University of Technology in Bratislava was installed an experimental physical fuel cell model of VVER 440 nuclear power plant with V 213 nuclear reactors. Objective of measurements on physical model was temperature and velocity profiles analysis in the fuel cell outlet. In this paper the measured temperature and velocity profiles are compared with the results of CFD simulation of fuel cell physical model coolant flow.

  18. Mathematical Modeling for Simulation of Nuclear Reactor Analysis

    OpenAIRE

    Salah Ud-Din Khan; Shahab Ud-Din Khan

    2013-01-01

    In this paper, we have developed a mathematical model for the nuclear reactor analysis to be implemented in the nuclear reactor code. THEATRe is nuclear reactor analysis code which can only work for the cylindrical type fuel reactor and cannot applicable for the plate type fuel nuclear reactor. Therefore, the current studies encompasses on the modification of THEATRe code for the plate type fuel element. This mathematical model is applicable to the thermal analysis of the reactor which is ver...

  19. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  20. Neutron physics for nuclear reactors unpublished writings by Enrico Fermi

    CERN Document Server

    Fermi, Enrico; Pisanti, O

    2010-01-01

    This unique volume gives an accurate and very detailed description of the functioning and operation of basic nuclear reactors, as emerging from yet unpublished papers by Nobel Laureate Enrico Fermi. In the first part, the entire course of lectures on Neutron Physics delivered by Fermi at Los Alamos is reported, according to the version made by Anthony P French. Here, the fundamental physical phenomena are described very clearly and comprehensively, giving the appropriate physics grounds for the functioning of nuclear piles. In the second part, all the patents issued by Fermi (and coworkers) on

  1. Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

    Energy Technology Data Exchange (ETDEWEB)

    Heeger, Karsten M. [Yale Univ., New Haven, CT (United States)

    2014-09-13

    This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zero$\\theta_{13}$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.

  2. Modeling of Reactor Kinetics and Dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Matthew Johnson; Scott Lucas; Pavel Tsvetkov

    2010-09-01

    In order to model a full fuel cycle in a nuclear reactor, it is necessary to simulate the short time-scale kinetic behavior of the reactor as well as the long time-scale dynamics that occur with fuel burnup. The former is modeled using the point kinetics equations, while the latter is modeled by coupling fuel burnup equations with the kinetics equations. When the equations are solved simultaneously with a nonlinear equation solver, the end result is a code with the unique capability of modeling transients at any time during a fuel cycle.

  3. Hydrodynamic models for slurry bubble column reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gidaspow, D. [IIT Center, Chicago, IL (United States)

    1995-12-31

    The objective of this investigation is to convert a {open_quotes}learning gas-solid-liquid{close_quotes} fluidization model into a predictive design model. This model is capable of predicting local gas, liquid and solids hold-ups and the basic flow regimes: the uniform bubbling, the industrially practical churn-turbulent (bubble coalescence) and the slugging regimes. Current reactor models incorrectly assume that the gas and the particle hold-ups (volume fractions) are uniform in the reactor. They must be given in terms of empirical correlations determined under conditions that radically differ from reactor operation. In the proposed hydrodynamic approach these hold-ups are computed from separate phase momentum balances. Furthermore, the kinetic theory approach computes the high slurry viscosities from collisions of the catalyst particles. Thus particle rheology is not an input into the model.

  4. Summary of ORSphere Critical and Reactor Physics Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Margaret A.; Bess, John D.

    2016-09-01

    In the early 1970s Dr. John T. Mihalczo (team leader), J. J. Lynn, and J. R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is summary summarize all the critical and reactor physics measurements evaluations and, when possible, to compare them to GODIVA experiment results.

  5. Plasma Reactors and Plasma Thrusters Modeling by Ar Complete Global Models

    Directory of Open Access Journals (Sweden)

    Chloe Berenguer

    2012-01-01

    Full Text Available A complete global model for argon was developed and adapted to plasma reactor and plasma thruster modeling. It takes into consideration ground level and excited Ar and Ar+ species and the reactor and thruster form factors. The electronic temperature, the species densities, and the ionization percentage, depending mainly on the pressure and the absorbed power, have been obtained and commented for various physical conditions.

  6. PIK3CA in colorectal cancer

    Directory of Open Access Journals (Sweden)

    Gieri eCathomas

    2014-03-01

    Full Text Available PIK3CA, the catalytic subunit of PI3K, is mutated in many different tumours, including colorectal cancer. Mutations of PIK3CA have been reported in 10 – 20% of colorectal cancer, about 80% of mutations found in two hot spots in exon 9 and exon 20. In RAS wild-type colorectal cancers, PIK3CA mutations have been associated with a worse clinical outcome and with a negative prediction of a response to targeted therapy by anti-EGFR monoclonal antibodies. However, these findings have not been confirmed in all studies and subsequent more detailed analysis has revealed that these effects may be restricted to mutations in Exon 20. Finally, mutations in PIK3CA may be the long sought biomarker for successful adjuvant therapy with aspirin in patients with colorectal cancer. Therefore, PIK3CA mutations appear to be a promising predictive biomarker; however, further data are needed to conclusively define the impact of somatic mutations in the PIK3CA gene for the management of patients with colorectal cancer.

  7. On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

    Energy Technology Data Exchange (ETDEWEB)

    Chauvin, J.P. [CEA,DEN, DER, SPEX, Experimental Physics Service, Cadarache, F-13108 St-Paul-Lez-Durance (France); Blaise, P. [CEA, DEN, DER, SPEX Experimental Programs Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France)

    2015-07-01

    The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physics calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)

  8. Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

    Science.gov (United States)

    Butler, C.; Albright, D.

    2007-01-01

    Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

  9. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  10. Engineering and Physics Optimization of Breed and Burn Fast Reactor Systems: Annual and Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Kevan D. Weaver; Theron Marshall; James Parry

    2005-10-01

    The Idaho National Laboratory (INL) contribution to the Nuclear Energy Research Initiative (NERI) project number 2002-005 was divided into reactor physics, and thermal-hydraulics and plant design. The research targeted credible physics and thermal-hydraulics models for a gas-cooled fast reactor, analyzing various fuel and in-core fuel cycle options to achieve a true breed and burn core, and performing a design basis Loss of Coolant Accident (LOCA) analysis on that design. For the physics analysis, a 1/8 core model was created using different enrichments and simulated equilibrium fuel loadings. The model was used to locate the hot spot of the reactor, and the peak to average energy deposition at that location. The model was also used to create contour plots of the flux and energy deposition over the volume of the reactor. The eigenvalue over time was evaluated using three different fuel configurations with the same core geometry. The breeding capabilities of this configuration were excellent for a 7% U-235 model and good in both a plutonium model and a 14% U-235 model. Changing the fuel composition from the Pu fuel which provided about 78% U-238 for breeding to the 14% U-235 fuel with about 86% U-238 slowed the rate of decrease in the eigenvalue a noticeable amount. Switching to the 7% U-235 fuel with about 93% U-238 showed an increase in the eigenvalue over time. For the thermal-hydraulic analysis, the reactor design used was the one forwarded by the MIT team. This reactor design uses helium coolant, a Brayton cycle, and has a thermal power of 600 MW. The core design parameters were supplied by MIT; however, the other key reactor components that were necessary for a plausible simulation of a LOCA were not defined. The thermal-hydraulic and plant design research concentrated on determining reasonable values for those undefined components. The LOCA simulation was intended to provide insights on the influence of the Reactor Cavity Cooling System (RCCS), the

  11. CFD Modeling of Melt Spreading on the Reactor Cavity Floor

    Energy Technology Data Exchange (ETDEWEB)

    Yeon, Wan Sik; Bang, Kwang Hyun [Korea Maritime University, Busan (Korea, Republic of); Cho, Young Jo; Lee, Jae Gon [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2010-05-15

    In the very unlikely event of a severe reactor accident involving core melt and reactor pressure vessel failure, it is important to provide an accident management strategy that would allow the molten core material to cool down, resolidify and bring the core debris to a stable coolable state for Light Water Reactors (LWRs). One approach to achieve a stable coolable state is to quench the core melt after its relocation from the reactor pressure vessel into the reactor cavity. This approach typically requires a large cavity floor area on which a large amount of core melt spreads well and forms a shallow melt thickness for small thermal resistance across the melt pool. Spreading of high temperature (approx3000 K), low superheat (approx200 K) core melt over a wide cavity floor has been a key question to the success of the ex-vessel core coolability and it has brought a number of experimental work (CORINE, ECOKATS, VULCANO) and analytical work (CORFLOW, MELTSPREAD, THEMA). These computational models are currently able to predict well the spreading of stimulant materials but yet have shown a limitation for prototypic core melt of UO{sub 2}+ZrO{sub 2} mixture. A computational model for the melt spreading requires a multiphase treatment of liquid melt, solidified melt, and air. Also solidification and thermal radiation physics should be included. The present work uses ANSYS-CFX code to simulate core melt spreading on the reactor cavity. The CFX code is a general-purpose multiphase code and the present work is focused on exploring the code's capability to model melt spreading problem in a step by step approach

  12. Reactor Physics Characterization of the HTR Module with UCO Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard Strydom

    2011-01-01

    ABSTRACT The HTR Module [1] is a graphite-moderated, helium cooled pebble bed High Temperature Reactor (HTR) design that has been extensively used as a reference template for the former South African and current Chinese HTR [2] programs. This design utilized spherical fuel elements packed into a dynamic pebble bed, consisting of TRISO coated uranium oxide (UO2) fuel kernels with a U-235 enrichment of 7.8% and a Heavy Metal loading of 7 grams per pebble. The main objective of this study is to compare several important reactor physics and core design parameters for the HTR Module and an identical design utilizing UCO fuel kernels. Fuel kernels of this type are currently being tested in the Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR) as part of the larger Next Generation Nuclear Plant (NGNP) project. The PEBBED-THERMIX [3] code, which was developed specifically for the analysis of pebble bed HTRs, was used to compare the coupled neutronic and thermal fluid performance of the two designs.

  13. A model for a countercurrent gas—solid—solid trickle flow reactor for equilibrium reactions. The methanol synthesis

    NARCIS (Netherlands)

    Westerterp, K.R.; Kuczynski, M.

    1987-01-01

    The theoretical background for a novel, countercurrent gas—solid—solid trickle flow reactor for equilibrium gas reactions is presented. A one-dimensional, steady-state reactor model is developed. The influence of the various process parameters on the reactor performance is discussed. The physical

  14. Models of iodine behavior in reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Weber, C.F.; Beahm, E.C.; Kress, T.S.

    1992-10-01

    Models are developed for many phenomena of interest concerning iodine behavior in reactor containments during severe accidents. Processes include speciation in both gas and liquid phases, reactions with surfaces, airborne aerosols, and other materials, and gas-liquid interface behavior. Although some models are largely empirical formulations, every effort has been made to construct mechanistic and rigorous descriptions of relevant chemical processes. All are based on actual experimental data generated at the Oak Ridge National Laboratory (ORNL) or elsewhere, and, hence, considerable data evaluation and parameter estimation are contained in this study. No application or encoding is attempted, but each model is stated in terms of rate processes, with the intention of allowing mechanistic simulation. Taken together, this collection of models represents a best estimate iodine behavior and transport in reactor accidents.

  15. Plasma flow reactor for steady state monitoring of physical and chemical processes at high temperatures

    Science.gov (United States)

    Koroglu, Batikan; Mehl, Marco; Armstrong, Michael R.; Crowhurst, Jonathan C.; Weisz, David G.; Zaug, Joseph M.; Dai, Zurong; Radousky, Harry B.; Chernov, Alex; Ramon, Erick; Stavrou, Elissaios; Knight, Kim; Fabris, Andrea L.; Cappelli, Mark A.; Rose, Timothy P.

    2017-09-01

    We present the development of a steady state plasma flow reactor to investigate gas phase physical and chemical processes that occur at high temperature (1000 reactor consists of a glass tube that is attached to an inductively coupled argon plasma generator via an adaptor (ring flow injector). We have modeled the system using computational fluid dynamics simulations that are bounded by measured temperatures. In situ line-of-sight optical emission and absorption spectroscopy have been used to determine the structures and concentrations of molecules formed during rapid cooling of reactants after they pass through the plasma. Emission spectroscopy also enables us to determine the temperatures at which these dynamic processes occur. A sample collection probe inserted from the open end of the reactor is used to collect condensed materials and analyze them ex situ using electron microscopy. The preliminary results of two separate investigations involving the condensation of metal oxides and chemical kinetics of high-temperature gas reactions are discussed.

  16. Development of analysis system and analysis on reactor physics for CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Gi; Bae, Chang Joon; Kwon, Oh Sun [Korea Power Electric Corporation, Taejon (Korea, Republic of)

    1997-07-01

    characteristics of reactor physics for CANFLEX-NU fuel core were calculated using final fuel design data. The results of analysis showed that there was no impact on reactor operations and safety. The above results of calculations and analysis were described in the physics design for CANFLEX-NU= fuel core. Various fuel models were evaluated for selecting high burnup fuel using recovered uranium. It is judged to be worse effects for reactor safety Hence, the use of graphite within fuel was proposed and its results showed to be better. The analysis system of reactor physics for design and analysis of high burnup fuel was evaluated. Lattice codes and core code were reviewed. From the results, the probability of WIMS-AECL and HELIOS is known to be high for analysis of high burnup fuel. For the core code, RFSP, it was evaluated that the simplified 2 group equation should be replaced by explicit 2 group equation. (Author) 32 refs., 25 tabs., 79 figs.

  17. Pebble Bed Reactor Dust Production Model

    Energy Technology Data Exchange (ETDEWEB)

    Abderrafi M. Ougouag; Joshua J. Cogliati

    2008-09-01

    The operation of pebble bed reactors, including fuel circulation, can generate graphite dust, which in turn could be a concern for internal components; and to the near field in the remote event of a break in the coolant circuits. The design of the reactor system must, therefore, take the dust into account and the operation must include contingencies for dust removal and for mitigation of potential releases. Such planning requires a proper assessment of the dust inventory. This paper presents a predictive model of dust generation in an operating pebble bed with recirculating fuel. In this preliminary work the production model is based on the use of the assumption of proportionality between the dust production and the normal force and distance traveled. The model developed in this work uses the slip distances and the inter-pebble forces computed by the authors’ PEBBLES. The code, based on the discrete element method, simulates the relevant static and kinetic friction interactions between the pebbles as well as the recirculation of the pebbles through the reactor vessel. The interaction between pebbles and walls of the reactor vat is treated using the same approach. The amount of dust produced is proportional to the wear coefficient for adhesive wear (taken from literature) and to the slip volume, the product of the contact area and the slip distance. The paper will compare the predicted volume with the measured production rates. The simulation tallies the dust production based on the location of creation. Two peak production zones from intra pebble forces are predicted within the bed. The first zone is located near the pebble inlet chute due to the speed of the dropping pebbles. The second peak zone occurs lower in the reactor with increased pebble contact force due to the weight of supported pebbles. This paper presents the first use of a Discrete Element Method simulation of pebble bed dust production.

  18. BENCHMARK EVALUATION OF THE START-UP CORE REACTOR PHYSICS MEASUREMENTS OF THE HIGH TEMPERATURE ENGINEERING TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    John Darrell Bess

    2010-05-01

    The benchmark evaluation of the start-up core reactor physics measurements performed with Japan’s High Temperature Engineering Test Reactor, in support of the Next Generation Nuclear Plant Project and Very High Temperature Reactor Program activities at the Idaho National Laboratory, has been completed. The evaluation was performed using MCNP5 with ENDF/B-VII.0 nuclear data libraries and according to guidelines provided for inclusion in the International Reactor Physics Experiment Evaluation Project Handbook. Results provided include updated evaluation of the initial six critical core configurations (five annular and one fully-loaded). The calculated keff eigenvalues agree within 1s of the benchmark values. Reactor physics measurements that were evaluated include reactivity effects measurements such as excess reactivity during the core loading process and shutdown margins for the fully-loaded core, four isothermal temperature reactivity coefficient measurements for the fully-loaded core, and axial reaction rate measurements in the instrumentation columns of three core configurations. The calculated values agree well with the benchmark experiment measurements. Fully subcritical and warm critical configurations of the fully-loaded core were also assessed. The calculated keff eigenvalues for these two configurations also agree within 1s of the benchmark values. The reactor physics measurement data can be used in the validation and design development of future High Temperature Gas-cooled Reactor systems.

  19. Fischer-Tropsch Slurry Reactor modeling

    Energy Technology Data Exchange (ETDEWEB)

    Soong, Y.; Gamwo, I.K.; Harke, F.W. [Pittsburgh Energy Technology Center, PA (United States)] [and others

    1995-12-31

    This paper reports experimental and theoretical results on hydrodynamic studies. The experiments were conducted in a hot-pressurized Slurry-Bubble Column Reactor (SBCR). It includes experimental results of Drakeol-10 oil/nitrogen/glass beads hydrodynamic study and the development of an ultrasonic technique for measuring solids concentration. A model to describe the flow behavior in reactors was developed. The hydrodynamic properties in a 10.16 cm diameter bubble column with a perforated-plate gas distributor were studied at pressures ranging from 0.1 to 1.36 MPa, and at temperatures from 20 to 200{degrees}C, using a dual hot-wire probe with nitrogen, glass beads, and Drakeol-10 oil as the gas, solid, and liquid phase, respectively. It was found that the addition of 20 oil wt% glass beads in the system has a slight effect on the average gas holdup and bubble size. A well-posed three-dimensional model for bed dynamics was developed from an ill-posed model. The new model has computed solid holdup distributions consistent with experimental observations with no artificial {open_quotes}fountain{close_quotes} as predicted by the earlier model. The model can be applied to a variety of multiphase flows of practical interest. An ultrasonic technique is being developed to measure solids concentration in a three-phase slurry reactor. Preliminary measurements have been made on slurries consisting of molten paraffin wax, glass beads, and nitrogen bubbles at 180 {degrees}C and 0.1 MPa. The data show that both the sound speed and attenuation are well-defined functions of both the solid and gas concentrations in the slurries. The results suggest possibilities to directly measure solids concentration during the operation of an autoclave reactor containing molten wax.

  20. Multi-physics approach to the analysis of molten salt reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nicolino, Claudio; Dulla, Sandra; Ravetto, Piero [Politecnico di Torino, DENER, Corso Duca degli Abruzzi, 24 - 10129 Torino (Italy); Lapenta, Giovanni [Departement Wiskunde, Katholieke Universiteit Leuven, Celestijnenlaan 200B, Heverlee, B-3001 (Belgium)

    2008-07-01

    A fully-coupled multi-physics non-linear model for the dynamic description of liquid-fuel multiplying fast systems is presented. Different physical variables are considered in the model, such as neutron flux, precursor concentration, velocity field and temperature. The corresponding set of equations is implemented in a code solved by a Jacobian-Free Newton-Krylov algorithm. The results presented enlighten the presence of unusual connections among the different aspects of the problem treated. A particular accent is posed on the description of those special features that need to be taken into account when dealing with a realistic analysis of a molten salt fast reactor. (authors)

  1. Pressure tube creep impact on the physics parameters for CANDU-6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, W. Y.; Min, B. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kam, S. C.; Kim, M. E. [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2004-07-01

    The lattice cell calculations are performed to assess the sensitivity of the reactor physics parameters to pressure tube creep resulting from radiation aging. The physics parameters of the lattice cell are calculated by using WIMSD-5B code, WIMS- AECL code, and MCNP code. The reference model(normal state) and two perturbed models accounting for the pressure tube creep are developed on the basis of CANDU-6 lattice cell. The 2.5% and 5% values of pressure tube diameter creep are considered. Also, The effects of the analyzed lattice parameters which are the coolant void reactivity, the fuel fission density and the atom density of Pu isotopes on the lattice.

  2. Isospin odd {pi}K scattering length

    Energy Technology Data Exchange (ETDEWEB)

    Schweizer, J. [Institut fuer Theoretische Physik, University of Vienna, A-1090 Vienna (Austria)]. E-mail: julia.schweizer@univie.ac.at

    2005-10-13

    We make use of the chiral two-loop representation of the {pi}K scattering amplitude [J. Bijnens, P. Dhonte, P. Talavera, JHEP 0405 (2004) 036] to investigate the isospin odd scattering length at next-to-next-to-leading order in the SU(3) expansion. This scattering length is protected against contributions of m{sub s} in the chiral expansion, in the sense that the corrections to the current algebra result are of order M{sub {pi}}{sup 2}. In view of the planned lifetime measurement on {pi}K atoms at CERN it is important to understand the size of these corrections.

  3. State space modeling of reactor core in a pressurized water reactor

    Science.gov (United States)

    Ashaari, A.; Ahmad, T.; Shamsuddin, Mustaffa; M, Wan Munirah W.; Abdullah, M. Adib

    2014-07-01

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  4. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  5. RNAi knockdown of PIK3CA preferentially inhibits invasion of mutant PIK3CA cells

    Institute of Scientific and Technical Information of China (English)

    Xin-Ke Zhou; Sheng-Song Tang; Gao Yi; Min Hou; Jin-Hui Chen; Bo Yang; Ji-Fang Liu; Zhi-Min He

    2011-01-01

    AIM: To explore the effects of siRNA silencing of PIK3CA on proliferation, migration and invasion of gastric cancer cells and to investigate the underlying mechanisms.METHODS: The mutation of PIK3CA in exons 9 and 20 of gastric cancer cell lines HGC-27, SGC-7901, BGC-823, MGC-803 and MKN-45 was screened by poly-merase chain reaction (PCR) followed by sequencing. BGC-823 cells harboring no mutations in either of the exons, and HGC-27 cells containing PIK3CA mutations were employed in the current study. siRNA targeting PIK3CA was chemically synthesized and was transfect-ed into these two cell lines in vitro. mRNA and protein expression of PIK3CA were detected by real-time PCR and Western blotting, respectively. We also measured phosphorylation of a serine/threonine protein kinase (Akt) using Western blotting. The proliferation, migra-tion and invasion of these cells were examined sepa-rately by 3-(4,5-dimethylthiazol-2-yl)-2,5-diphenyltet-razolium bromide (MTT), wound healing and Transwell chambers assay.RESULTS: The siRNA directed against PIK3CA effec-tively led to inhibition of both endogenous mRNA and protein expression of PIK3CA, and thus significantly down-regulated phosphorylation of Akt (P < 0.05). Furthermore, simultaneous silencing of PIK3CA result-ed in an obvious reduction in tumor cell proliferation activity, migration and invasion potential (P < 0.01). Intriguing, mutant HGC-27 cells exhibited stronger invasion ability than that shown by wild-type BGC-823 cells. Knockdown of PIK3CA in mutant HGC-27 cells contributed to a reduction in cell invasion to a greater extent than in non-mutant BGC-823 cells.CONCLUSION: siRNA mediated targeting of PIK3CA may specifically knockdown the expression of PIK3CA in gastric cancer cells, providing a potential implication for therapy of gastric cancer.

  6. Development of PIK-75 nanosuspension formulation with enhanced delivery efficiency and cytotoxicity for targeted anti-cancer therapy.

    Science.gov (United States)

    Talekar, Meghna; Ganta, Srinivas; Amiji, Mansoor; Jamieson, Stephen; Kendall, Jackie; Denny, William A; Garg, Sanjay

    2013-06-25

    PIK-75 is a phosphatidylinositol 3-kinase (PI3K) inhibitor that shows selectivity toward p110-α over the other PI3K class Ia isoforms p110-β and p110-δ, but it lacks solubility, stability and other kinase selectivity. The purpose of this study was to develop folate-targeted PIK-75 nanosuspension for tumor targeted delivery and to improve therapeutic efficacy in human ovarian cancer model. High pressure homogenization was used to prepare the non-targeted and targeted PIK-75 nanosuspensions which were characterized for size, zeta potential, entrapment efficiency, morphology, saturation solubility and dissolution velocity. In vitro analysis of drug uptake, cell viability and cell survival was conducted in SKOV-3 cells. Drug pharmacokinetics and pAkt expression were determined in SKOV-3 tumor bearing mice. PIK-75 nanosuspensions showed an improvement in dissolution velocity and an 11-fold increase in saturation solubility over pre-milled PIK-75. In vitro studies in SKOV-3 cells indicated a 2-fold improvement in drug uptake and 0.4-fold decrease in IC50 value of PIK-75 following treatment with targeted nanosuspension compared to non-targeted nanosuspension. The improvement in cytotoxicity was attributed to an increase in caspase 3/7 and hROS activity. In vivo studies indicated a 5-10-fold increased PIK-75 accumulation in the tumor with both the nanosuspension formulations compared to PIK-75 suspension. The targeted nanosuspension showed an enhanced downregulation of pAkt compared to non-targeted formulation system. These results illustrate the opportunity to formulate PIK-75 as a targeted nanosuspension to enhance uptake and cytotoxicity of the drug in tumor.

  7. Development of a system model for advanced small modular reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  8. Development of an automated core model for nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mosteller, R.D.

    1998-12-31

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input.

  9. ORIGEN2 model and results for the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A G; Bjerke, M A

    1982-06-01

    Reactor physics calculations and literature information acquisition have led to the development of a Clinch River Breeder Reactor (CRBR) model for the ORIGEN2 computer code. The model is based on cross sections taken directly from physics codes. Details are presented concerning the physical description of the fuel assemblies, the fuel management scheme, irradiation parameters, and initial material compositions. The ORIGEN2 model for the CRBR has been implemented, resulting in the production of graphical and tabular characteristics (radioactivity, thermal power, and toxicity) of CRBR spent fuel, high-level waste, and fuel-assembly structural material waste as a function of decay time. Characteristics for pressurized water reactors (PWRs), commercial liquid-metal fast breeder reactors (LMFBRs), and the Fast Flux Test Facility (FFTF) have also been included in this report for comparison with the CRBR data.

  10. Thermal-hydraulic modeling of reactivity accidents in MTR reactors

    Directory of Open Access Journals (Sweden)

    Khater Hany

    2006-01-01

    Full Text Available This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactors during a reactivity insertion accident. The model is formulated for coupling reactor kinetics with feedback reactivity and reactor core thermal-hydraulics. To represent the reactor core, two types of channels are considered, average and hot channels. The developed computer program is compiled and executed on a personal computer, using the FORTRAN language. The model is validated by safety-related benchmark calculations for MTR-TYPE reactors of IAEA 10 MW generic reactor for both slow and fast reactivity insertion transients. A good agreement is shown between the present model and the benchmark calculations. Then, the model is used for simulating the uncontrolled withdrawal of a control rod of an ETRR-2 reactor in transient with over power scram trip. The model results for ETRR-2 are analyzed and discussed.

  11. Educational reactor-physics experiments with the critical assemble TCA

    Energy Technology Data Exchange (ETDEWEB)

    Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki [Tokyo Inst. of Tech. (Japan); Horiki, Oichiro; Suzaki, Takenori

    1997-10-01

    The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for (1) Critical approach and Exponential experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, and (5) Measurement of safety plate worth by the rod drop method. (author)

  12. Modeling for Anaerobic Fixed-Bed Biofilm Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, B. Y. M.; Pfeffer, J. T.

    1989-06-01

    The specific objectives of this research were: 1. to develop an equilibrium model for chemical aspects of anaerobic reactors; 2. to modify the equilibrium model for non-equilibrium conditions; 3. to incorporate the existing biofilm models into the models above to study the biological and chemical behavior of the fixed-film anaerobic reactors; 4. to experimentally verify the validity of these models; 5. to investigate the biomass-holding ability of difference packing materials for establishing reactor design criteria.

  13. Modelization of physical phenomena in research reactors with the help of new developments in transport methods, and methodology validation with experimental data; Modelisation des phenomenes physiques dans les reacteurs de recherche a l'aide de developpements realises dans les methodes de transport et qualification

    Energy Technology Data Exchange (ETDEWEB)

    Rauck, St

    2000-10-01

    The aim of this work is to develop a scheme for experimental reactors, based on transport equations. This type of reactors is characterized by a small core, a complex, very heterogeneous geometry and a large leakage. The possible insertion of neutron beams in the reflector and the presence of absorbers in the core increase the difficulty of the 3D-geometrical description and the physical modeling of the component parameters of the reactor. The Orphee reactor has been chosen for our study. Physical models (homogenization, collapsing cross section in few groups, albedo multigroup condition) have been developed in the APOLLO2 and CRONOS2 codes to calculate flux and power maps in a 3D-geometry, with different burnup and through transport equations. Comparisons with experimental measurements have shown the interest of taking into account anisotropy, steep flux gradients by using Sn methods, and on the other hand using a 12-group cross section library. The modeling of neutron beams has been done outside the core modeling through Monte Carlo calculations and with the total geometry, including a large thickness of heavy water. Thanks to this calculations, one can evaluate the neutron beams anti-reactivity and determinate the core cycle. We assure these methods more accurate than usual transport-diffusion calculations will be used for the conception of new research reactors. (author)

  14. PIK3CAH1047R and Her2 initiated mammary tumors escape PI3K dependency by compensatory activation of MEK-ERK signaling

    Science.gov (United States)

    Cheng, Hailing; Liu, Pixu; Ohlson, Carolynn; Xu, Erbo; Symonds, Lynn; Isabella, Adam; Muller, William J.; Lin, Nancy U.; Krop, Ian E.; Roberts, Thomas M.; Winer, Eric P.; Arteaga, Carlos L.; Zhao, Jean J.

    2015-01-01

    Human breast cancers that have HER2 amplification/overexpression frequently carry PIK3CA mutations, and are often associated with a worse prognosis. However, the role of PIK3CA mutations in the initiation and maintenance of these breast cancers remains elusive. In the present study, we generated a compound mouse model that genetically mimics HER2 positive breast cancer with coexisting PIK3CAH1047R. Induction of PIK3CAH1047R expression in mouse mammary glands with constitutive expression of activated Her2/Neu resulted in accelerated mammary tumorigenesis with enhanced metastatic potential. Interestingly, inducible expression of mutant PIK3CA resulted in a robust activation of PI3K/AKT signaling but attenuation of Her2/Her3 signaling, and this can be reversed by deinduction of PIK3CAH1047R expression. Strikingly, while these Her2+ PIK3CAH1047R initiated primary mammary tumors are refractory to HER2-targeted therapy, all tumors responded to inactivation of the oncogenic PIK3CAH1047R, a situation closely mimicking the use of a highly effective inhibitor specifically targeting the mutant PIK3CA/p110a. Notably, these tumors eventually resumed growth, and a fraction of them escaped PI3K dependence by compensatory ERK activation, which can be blocked by combined inhibition of Her2 and MEK. Together, these results suggest that PIK3CA-specific inhibition as a monotherapy followed by combination therapy targeting MAPK and HER2 in a timely manner may be an effective treatment approach against HER2 positive cancers with coexisting PIK3CA-activating mutations. PMID:26640141

  15. Physical aspects of liquid-impelled loop reactors.

    NARCIS (Netherlands)

    Sonsbeek, van H.M.

    1992-01-01

    The liquid-impelled loop reactor (LLR) is a reactor that consists of two parts : the main tube and the circulation tube. Both parts are in open connection at the bottom and at the top. The reactor is filled with a liquid phase: the continuous phase. Another liquid phase is injected in the main tube

  16. The mechanics in the reactors physics; La mecanique dans la physique des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Zaetta, A. [CEA/Cadarache, Dept. d' Etudes des Reacteurs, DER/SPRC, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint-Paul-lez-Durance (France)] [and others

    1998-12-22

    This meeting of the 24 november 1998, took place in Paris and was organized by the SFEN. After three plenary sessions a technical meeting dealt on the mechanics in reactors physics. The plenary papers presented the state of the art in the PWR type reactors and fast neutron reactors systems and in the thermonuclear reactors system. Five more technical papers presented the seismic behavior of the reactors cores, the fuel-cladding interactions, the defects harmfulness in the fracture mechanics and the fuel rods control system wear. (A.L.B.)

  17. The mechanics in the reactors physics; La mecanique dans la physique des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Zaetta, A. [CEA/Cadarache, Dept. d' Etudes des Reacteurs, DER/SPRC, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint-Paul-lez-Durance (France)] [and others

    1998-12-22

    This meeting of the 24 november 1998, took place in Paris and was organized by the SFEN. After three plenary sessions a technical meeting dealt on the mechanics in reactors physics. The plenary papers presented the state of the art in the PWR type reactors and fast neutron reactors systems and in the thermonuclear reactors system. Five more technical papers presented the seismic behavior of the reactors cores, the fuel-cladding interactions, the defects harmfulness in the fracture mechanics and the fuel rods control system wear. (A.L.B.)

  18. Exploring Stochastic Sampling in Nuclear Data Uncertainties Assessment for Reactor Physics Applications and Validation Studies

    Directory of Open Access Journals (Sweden)

    Alexander Vasiliev

    2016-12-01

    Full Text Available The quantification of uncertainties of various calculation results, caused by the uncertainties associated with the input nuclear data, is a common task in nuclear reactor physics applications. Modern computation resources and improved knowledge on nuclear data allow nowadays to significantly advance the capabilities for practical investigations. Stochastic sampling is the method which has received recently a high momentum for its use and exploration in the domain of reactor design and safety analysis. An application of a stochastic sampling based tool towards nuclear reactor dosimetry studies is considered in the given paper with certain exemplary test evaluations. The stochastic sampling not only allows the input nuclear data uncertainties propagation through the calculations, but also an associated correlation analysis performance with no additional computation costs and for any parameters of interest can be done. Thus, an example of assessment of the Pearson correlation coefficients for several models, used in practical validation studies, is shown here. As a next step, the analysis of the obtained information is proposed for discussion, with focus on the systems similarities assessment. The benefits of the employed method and tools with respect to practical reactor dosimetry studies are consequently outlined.

  19. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-08-31

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

  20. Integrated physics analysis of plasma start-up scenario of helical reactor FFHR-d1

    Science.gov (United States)

    Goto, T.; Miyazawa, J.; Sakamoto, R.; Seki, R.; Suzuki, C.; Yokoyama, M.; Satake, S.; Sagara, A.; The FFHR Design Group

    2015-06-01

    1D physics analysis of the plasma start-up scenario of the large helical device (LHD)-type helical reactor FFHR-d1 was conducted. The time evolution of the plasma profile is calculated using a simple model based on the LHD experimental observations. A detailed assessment of the magnetohydrodynamic equilibrium and neo-classical energy loss was conducted using the integrated transport analysis code TASK3D. The robust controllability of the fusion power was confirmed by feedback control of the pellet fuelling and a simple staged variation of the external heating power with a small number of simple diagnostics (line-averaged electron density, edge electron density and fusion power). A baseline operation control scenario (plasma start-up and steady-state sustainment) of the FFHR-d1 reactor for both self-ignition and sub-ignition operation modes was demonstrated.

  1. Global gene expression profiling of a mouse model of ovarian clear cell carcinoma caused by ARID1A and PIK3CA mutations implicates a role for inflammatory cytokine signaling

    Directory of Open Access Journals (Sweden)

    Ronald L. Chandler

    2015-09-01

    Full Text Available Ovarian clear-cell carcinoma (OCCC is an aggressive form of epithelial ovarian cancer (EOC. OCCC represents 5–25% of all EOC incidences and is the second leading cause of death from ovarian cancer (Glasspool and McNeish, 2013 [1]. A recent publication by Chandler et al. reported the first mouse model of OCCC that resembles human OCCC both genetically and histologically by inducing a localized deletion of ARID1A and the expression of the PIK3CAH1047R substitution mutation (Chandler et al., 2015 [2]. We utilized Affymetrix Mouse Gene 2.1 ST arrays for the global gene expression profiling of mouse primary OCCC tumor samples and animal-matched normal ovaries to identify cancer-dependent gene expression. We describe the approach used to generate the differentially expressed genes from the publicly available data deposited at the Gene Expression Omnibus (GEO database under the accession number GSE57380. These data were used in cross-species comparisons to publically available human OCCC gene expression data and allowed the identification of coordinately regulated genes in both mouse and human OCCC and supportive of a role for inflammatory cytokine signaling in OCCC pathogenesis (Chandler et al., 2015 [2].

  2. 3D CAD model of the subcritical nuclear reactor of IPN; Modelo CAD 3D del reactor nuclear subcritico del IPN

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

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

  3. Identification of physical models

    DEFF Research Database (Denmark)

    Melgaard, Henrik

    1994-01-01

    The problem of identification of physical models is considered within the frame of stochastic differential equations. Methods for estimation of parameters of these continuous time models based on descrete time measurements are discussed. The important algorithms of a computer program for ML or MAP...... design of experiments, which is for instance the design of an input signal that are optimal according to a criterion based on the information provided by the experiment. Also model validation is discussed. An important verification of a physical model is to compare the physical characteristics...... of the model with the available prior knowledge. The methods for identification of physical models have been applied in two different case studies. One case is the identification of thermal dynamics of building components. The work is related to a CEC research project called PASSYS (Passive Solar Components...

  4. IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bostelmann, F. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on

  5. Physical Characteristics of Coffee Beans from Steaming Processin Single Column Reactor

    OpenAIRE

    Sukrisno Widyotomo; Sri Mulato; Hadi K. Purwadaria; A.M Syarief

    2010-01-01

    One of important steps in decaffeination process is steaming. The aim of steaming is to expand coffee beans porosity in order to obtain optimal condition for decaffeination process. Steaming can be done in single column reactor using saturated water vapour as media. The objective of this research is to study physical characteristics of coffee beans after steaming process using single column reactor. Material tested was Robusta coffee with 13—14% moisture content after dry processing. Reactor ...

  6. Core Physics of Pebble Bed High Temperature Nuclear Reactors

    NARCIS (Netherlands)

    Auwerda, G.J.

    2014-01-01

    To more accurately predict the temperature distribution inside the reactor core of pebble bed type high temperature reactors, in this thesis we investigated the stochastic properties of randomly stacked beds and the effects of the non-homogeneity of these beds on the neutronics and thermal-hydraulic

  7. Mutant PIK3CA accelerates HER2-driven transgenic mammary tumors and induces resistance to combinations of anti-HER2 therapies.

    Science.gov (United States)

    Hanker, Ariella B; Pfefferle, Adam D; Balko, Justin M; Kuba, María Gabriela; Young, Christian D; Sánchez, Violeta; Sutton, Cammie R; Cheng, Hailing; Perou, Charles M; Zhao, Jean J; Cook, Rebecca S; Arteaga, Carlos L

    2013-08-27

    Human epidermal growth factor receptor 2 (HER2; ERBB2) amplification and phosphatidylinositol-4,5-bisphosphate 3-kinase, catalytic subunit alpha (PIK3CA) mutations often co-occur in breast cancer. Aberrant activation of the phosphatidylinositol 3-kinase (PI3K) pathway has been shown to correlate with a diminished response to HER2-directed therapies. We generated a mouse model of HER2-overexpressing (HER2(+)), PIK3CA(H1047R)-mutant breast cancer. Mice expressing both human HER2 and mutant PIK3CA in the mammary epithelium developed tumors with shorter latencies compared with mice expressing either oncogene alone. HER2 and mutant PIK3CA also cooperated to promote lung metastases. By microarray analysis, HER2-driven tumors clustered with luminal breast cancers, whereas mutant PIK3CA tumors were associated with claudin-low breast cancers. PIK3CA and HER2(+)/PIK3CA tumors expressed elevated transcripts encoding markers of epithelial-to-mesenchymal transition and stem cells. Cells from HER2(+)/PIK3CA tumors more efficiently formed mammospheres and lung metastases. Finally, HER2(+)/PIK3CA tumors were resistant to trastuzumab alone and in combination with lapatinib or pertuzumab. Both drug resistance and enhanced mammosphere formation were reversed by treatment with a PI3K inhibitor. In sum, PIK3CA(H1047R) accelerates HER2-mediated breast epithelial transformation and metastatic progression, alters the intrinsic phenotype of HER2-overexpressing cancers, and generates resistance to approved combinations of anti-HER2 therapies.

  8. Adapting Dynamic Mathematical Models to a Pilot Anaerobic Digestion Reactor

    Directory of Open Access Journals (Sweden)

    F. Haugen, R. Bakke, and B. Lie

    2013-04-01

    Full Text Available A dynamic model has been adapted to a pilot anaerobic reactor fed diarymanure. Both steady-state data from online sensors and laboratory analysis anddynamic operational data from online sensors are used in the model adaptation.The model is based on material balances, and comprises four state variables,namely biodegradable volatile solids, volatile fatty acids, acid generatingmicrobes (acidogens, and methane generating microbes (methanogens. The modelcan predict the methane gas flow produced in the reactor. The model may beused for optimal reactor design and operation, state-estimation and control.Also, a dynamic model for the reactor temperature based on energy balance ofthe liquid in the reactor is adapted. This model may be used for optimizationand control when energy and economy are taken into account.

  9. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  10. Advanced Small Modular Reactor Economics Model Development

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Thomas J [ORNL

    2014-10-01

    The US Department of Energy Office of Nuclear Energy’s Advanced Small Modular Reactor (SMR) research and development activities focus on four key areas: Developing assessment methods for evaluating advanced SMR technologies and characteristics; and Developing and testing of materials, fuels and fabrication techniques; and Resolving key regulatory issues identified by US Nuclear Regulatory Commission and industry; and Developing advanced instrumentation and controls and human-machine interfaces. This report focuses on development of assessment methods to evaluate advanced SMR technologies and characteristics. Specifically, this report describes the expansion and application of the economic modeling effort at Oak Ridge National Laboratory. Analysis of the current modeling methods shows that one of the primary concerns for the modeling effort is the handling of uncertainty in cost estimates. Monte Carlo–based methods are commonly used to handle uncertainty, especially when implemented by a stand-alone script within a program such as Python or MATLAB. However, a script-based model requires each potential user to have access to a compiler and an executable capable of handling the script. Making the model accessible to multiple independent analysts is best accomplished by implementing the model in a common computing tool such as Microsoft Excel. Excel is readily available and accessible to most system analysts, but it is not designed for straightforward implementation of a Monte Carlo–based method. Using a Monte Carlo algorithm requires in-spreadsheet scripting and statistical analyses or the use of add-ons such as Crystal Ball. An alternative method uses propagation of error calculations in the existing Excel-based system to estimate system cost uncertainty. This method has the advantage of using Microsoft Excel as is, but it requires the use of simplifying assumptions. These assumptions do not necessarily bring into question the analytical results. In fact, the

  11. Deterministic Modeling of the High Temperature Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, J.; Cogliati, J. J.; Pope, M. A.; Ferrer, R. M.; Ougouag, A. M.

    2010-06-01

    Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL’s current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green’s Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2–3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the

  12. PEBBLES: A COMPUTER CODE FOR MODELING PACKING, FLOW AND RECIRCULATIONOF PEBBLES IN A PEBBLE BED REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2006-10-01

    A comprehensive, high fidelity model for pebble flow has been developed and embodied in the PEBBLES computer code. In this paper, a description of the physical artifacts included in the model is presented and some results from using the computer code for predicting the features of pebble flow and packing in a realistic pebble bed reactor design are shown. The sensitivity of models to various physical parameters is also discussed.

  13. Mathematical modeling of upflow anaerobic sludge blanket (UASB) reactor treating domestic wastewater.

    Science.gov (United States)

    Elmitwalli, Tarek

    2013-01-01

    Although the upflow anaerobic sludge blanket (UASB) reactor has been widely applied for domestic wastewater treatment in many developing countries, there is no sufficient mathematical model for proper design and operation of the reactor. An empirical model based on non-linear regression was developed to represent the physical and chemical removal of suspended solids (SS) in the reactor. Moreover, a simplified dynamic model based on ADM1 and the empirical model for SS removal was developed for anaerobic digestion of the entrapped SS and dissolved matter in the wastewater. The empirical model showed that effluent suspended chemical oxygen demand (COD(ss)) concentration is directly proportional to the influent COD(ss) concentration and inversely proportional to both the hydraulic retention time (HRT) of the reactor and wastewater temperature. For obtaining sufficient COD(ss) removal, the HRT of the UASB reactor must be higher than 4 h, and higher HRT than 12 h slightly improved COD(ss) removal. The dynamic model results showed that the required time for filling the reactor with sludge mainly depends on influent total chemical oxygen demand (COD(t)) concentration and HRT. The influent COD(t) concentration, HRT and temperature play a crucial role on the performance of the reactor. The results indicated that shorter HRT is needed for optimization of COD(t) removal, as compared with optimization of COD(t) conversion to methane. Based on the model results, the design HRT of the UASB reactor should be selected based on the optimization of wastewater conversion and minimization of biodegradable SS accumulation in the sludge bed, not only based on COD removal, to guarantee a stable reactor performance.

  14. Modeling of a Reverse Flow Reactor for Methanol Synthesis

    Institute of Scientific and Technical Information of China (English)

    陈晓春; P.L.Silveston; 等

    2003-01-01

    An accurate one-dimensional,heterogeneous model taking account of axial dispersion and heat transfer to the reactor wall,and heat conduction through the reactor wall for methanol synthesis in a bench scale reactor under periodic reversal of flow direction is presented.Adjustable parameters in this model are the effectiveness factors for each of the three reactions occurring in the synthesis and a factor for the bed to wall heat transfer coefficient correlation.Experimental data were used to evaluate these parameters and reasonable values of these parameters were obtained.The model was found to closely predict the reactor performance under a wide range of parameters were obtained.The model was found to closely predict the reactor preformance under a wide range of operating conditions,such as carbon oxide concentrations,volumetric flow rate,and cyclic period.

  15. Benchmark Data Through The International Reactor Physics Experiment Evaluation Project (IRPHEP)

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Dr. Enrico Sartori

    2005-09-01

    The International Reactor Physics Experiments Evaluation Project (IRPhEP) was initiated by the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency’s (NEA) Nuclear Science Committee (NSC) in June of 2002. The IRPhEP focus is on the derivation of internationally peer reviewed benchmark models for several types of integral measurements, in addition to the critical configuration. While the benchmarks produced by the IRPhEP are of primary interest to the Reactor Physics Community, many of the benchmarks can be of significant value to the Criticality Safety and Nuclear Data Communities. Benchmarks that support the Next Generation Nuclear Plant (NGNP), for example, also support fuel manufacture, handling, transportation, and storage activities and could challenge current analytical methods. The IRPhEP is patterned after the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and is closely coordinated with the ICSBEP. This paper highlights the benchmarks that are currently being prepared by the IRPhEP that are also of interest to the Criticality Safety Community. The different types of measurements and associated benchmarks that can be expected in the first publication and beyond are described. The protocol for inclusion of IRPhEP benchmarks as ICSBEP benchmarks and for inclusion of ICSBEP benchmarks as IRPhEP benchmarks is detailed. The format for IRPhEP benchmark evaluations is described as an extension of the ICSBEP format. Benchmarks produced by the IRPhEP add new dimension to criticality safety benchmarking efforts and expand the collection of available integral benchmarks for nuclear data testing. The first publication of the "International Handbook of Evaluated Reactor Physics Benchmark Experiments" is scheduled for January of 2006.

  16. Identification of physical models

    DEFF Research Database (Denmark)

    Melgaard, Henrik

    1994-01-01

    design of experiments, which is for instance the design of an input signal that are optimal according to a criterion based on the information provided by the experiment. Also model validation is discussed. An important verification of a physical model is to compare the physical characteristics...... and Systems Testing), on testing of building components related to passive solar energy conservation, tested under outdoor climate conditions. The second case study is related to the performance of a spark ignition car engine. A phenomenological model of the fuel flow is identified under various operating...

  17. Models in physics teaching

    DEFF Research Database (Denmark)

    Kneubil, Fabiana Botelho

    2016-01-01

    In this work we show an approach based on models, for an usual subject in an introductory physics course, in order to foster discussions on the nature of physical knowledge. The introduction of elements of the nature of knowledge in physics lessons has been emphasised by many educators and one uses...... the case of metals to show the theoretical and phenomenological dimensions of physics. The discussion is made by means of four questions whose answers cannot be reached neither for theoretical elements nor experimental measurements. Between these two dimensions it is necessary to realise a series...... of reasoning steps to deepen the comprehension of microscopic concepts, such as electrical resistivity, drift velocity and free electrons. When this approach is highlighted, beyond the physical content, aspects of its nature become explicit and may improve the structuring of knowledge for learners...

  18. A reference worldwide model for antineutrinos from reactors

    CERN Document Server

    Baldoncini, Marica; Fiorentini, Giovanni; Mantovani, Fabio; Ricci, Barbara; Strati, Virginia; Xhixha, Gerti

    2014-01-01

    Antineutrinos produced at nuclear reactors constitute a severe source of background for the detection of geoneutrinos, which bring to the Earth's surface information about natural radioactivity in the whole planet. In this framework we provide a reference worldwide model for antineutrinos from reactors, in view of reactors operational records yearly published by the International Atomic Energy Agency (IAEA). We evaluate the expected signal from commercial reactors for ongoing (KamLAND and Borexino), planned (SNO+) and proposed (Juno, RENO-50, LENA and Hanohano) experimental sites. Uncertainties related to reactor antineutrino production, propagation and detection processes are estimated using a Monte Carlo based approach, which provides an overall site dependent uncertainty on the signal in the geoneutrino energy window on the order of 3%. We also implement the off-equilibrium correction to the reference reactor spectra associated with the long-lived isotopes and we estimate a 2.4% increase of the unoscillate...

  19. Analytical model of plasma-chemical etching in planar reactor

    Science.gov (United States)

    Veselov, D. S.; Bakun, A. D.; Voronov, Yu A.; Kireev, V. Yu; Vasileva, O. V.

    2016-09-01

    The paper discusses an analytical model of plasma-chemical etching in planar diode- type reactor. Analytical expressions of etch rate and etch anisotropy were obtained. It is shown that etch anisotropy increases with increasing the ion current and ion energy. At the same time, etch selectivity of processed material decreases as compared with the mask. Etch rate decreases with the distance from the centre axis of the reactor. To decrease the loading effect, it is necessary to reduce the wafer temperature and pressure in the reactor, as well as increase the gas flow rate through the reactor.

  20. New Reactor Physics Benchmark Data in the March 2012 Edition of the IRPhEP Handbook

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; J. Blair Briggs; Jim Gulliford

    2012-11-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications. Numerous experiments that have been performed worldwide, represent a large investment of infrastructure, expertise, and cost, and are valuable resources of data for present and future research. These valuable assets provide the basis for recording, development, and validation of methods. If the experimental data are lost, the high cost to repeat many of these measurements may be prohibitive. The purpose of the IRPhEP is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. Contributors from around the world collaborate in the evaluation and review of selected benchmark experiments for inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [1]. Several new evaluations have been prepared for inclusion in the March 2012 edition of the IRPhEP Handbook.

  1. A physical description of fission product behavior fuels for advanced power reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

  2. Multiphysics modeling of porous CRUD deposits in nuclear reactors

    Science.gov (United States)

    Short, M. P.; Hussey, D.; Kendrick, B. K.; Besmann, T. M.; Stanek, C. R.; Yip, S.

    2013-11-01

    The formation of porous CRUD deposits on nuclear reactor fuel rods, a longstanding problem in the operation of pressurized water reactors (PWRs), is a significant challenge to science-based multiscale modeling and simulation. While existing, published studies have focused on individual or loosely coupled processes, such as heat transfer, fluid flow, and compound dissolution/precipitation, none have addressed their coupled effects sufficiently to enable a comprehensive, scientific understanding of CRUD. Here we present the formulation and results of a model, MAMBA-BDM, which begins to incorporate mechanistic details in describing CRUD in PWRs. CRUD is treated as a chemical deposition process in an environment of variable concentration, an arbitrary level of heating, and a complex fractal-based flow geometry. We present results on spatial distributions of temperature, pressure, velocity, and concentration that give insight into the interplay between these physical properties and geometrical parameters. We show the role of heat convection which has not been discussed previously. Furthermore, we suggest that the assumption of liquid saturation in the CRUD deserves scrutiny, as a result of our attempt to determine an effective CRUD thermal conductivity.

  3. Discrete element modelling of pebble packing in pebble bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Suikkanen, Heikki, E-mail: heikki.suikkanen@lut.fi; Ritvanen, Jouni, E-mail: jouni.ritvanen@lut.fi; Jalali, Payman, E-mail: payman.jalali@lut.fi; Kyrki-Rajamäki, Riitta, E-mail: riitta.kyrki-rajamaki@lut.fi

    2014-07-01

    Highlights: • A discrete element method code is developed for pebble bed reactor analyses. • Methods are established to extract packing information at various spatial scales. • Packing simulations inside annular core geometry are done varying input parameters. • The restitution coefficient has the strongest effect on the resulting packing density. • Detailed analyses reveal local densification especially near the walls. - Abstract: It is important to understand the packing characteristics and behaviour of the randomly packed pebble bed to further analyse the reactor physical and thermal-hydraulic behaviour and to design a safe and economically feasible pebble bed reactor. The objective of this work was to establish methods to model and analyse the pebble packing in detail to provide useful tools and data for further analyses. Discrete element method (DEM) is a well acknowledged method for analysing granular materials, such as the fuel pebbles in a pebble bed reactor. In this work, a DEM computer code was written specifically for pebble bed analyses. Analysis methods were established to extract data at various spatial scales from the pebble beds resulting from the DEM simulations. A comparison with available experimental data was performed to validate the DEM implementation. To test the code implementation in full-scale reactor calculations, DEM packing simulations were done in annular geometry with 450,000 pebbles. Effects of the initial packing configuration, friction and restitution coefficients and pebble size distribution to the resulting pebble bed were investigated. The packing simulations revealed that from the investigated parameters the restitution coefficient had the largest effect on the resulting average packing density while other parameters had smaller effects. Detailed local packing density analysis of pebble beds with different average densities revealed local variations especially strong in the regions near the walls. The implemented DEM

  4. Physics Characterization of a Heterogeneous Sodium Fast Reactor Transmutation System

    Energy Technology Data Exchange (ETDEWEB)

    Samuel E. Bays

    2007-09-01

    The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even mass number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active “fast reactor” driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both a non-flattened and a pancake core geometry. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of the same size.

  5. Probing new physics scenarios in accelerator and reactor neutrino experiments

    Science.gov (United States)

    Di Iura, A.; Girardi, I.; Meloni, D.

    2015-06-01

    We perform a detailed combined fit to the {{\\bar{ν }}e}\\to {{\\bar{ν }}e} disappearence data of the Daya Bay experiment and the appearance {{ν }μ }\\to {{ν }e} and disappearance {{ν }μ }\\to {{ν }μ } data of the Tokai to Kamioka (T2K) one in the presence of two models of new physics affecting neutrino oscillations, namely a model where sterile neutrinos can propagate in a large compactified extra dimension and a model where non-standard interactions (NSI) affect the neutrino production and detection. We find that the Daya Bay ⨁ T2K data combination constrains the largest radius of the compactified extra dimensions to be R≲ 0.17 μm at 2σ C.L. (for the inverted ordering of the neutrino mass spectrum) and the relevant NSI parameters in the range O({{10}-3})-O({{10}-2}), for particular choices of the charge parity violating phases.

  6. INTERVAL OBSERVER FOR A BIOLOGICAL REACTOR MODEL

    Directory of Open Access Journals (Sweden)

    T. A. Kharkovskaia

    2014-05-01

    Full Text Available The method of an interval observer design for nonlinear systems with parametric uncertainties is considered. The interval observer synthesis problem for systems with varying parameters consists in the following. If there is the uncertainty restraint for the state values of the system, limiting the initial conditions of the system and the set of admissible values for the vector of unknown parameters and inputs, the interval existence condition for the estimations of the system state variables, containing the actual state at a given time, needs to be held valid over the whole considered time segment as well. Conditions of the interval observers design for the considered class of systems are shown. They are: limitation of the input and state, the existence of a majorizing function defining the uncertainty vector for the system, Lipschitz continuity or finiteness of this function, the existence of an observer gain with the suitable Lyapunov matrix. The main condition for design of such a device is cooperativity of the interval estimation error dynamics. An individual observer gain matrix selection problem is considered. In order to ensure the property of cooperativity for interval estimation error dynamics, a static transformation of coordinates is proposed. The proposed algorithm is demonstrated by computer modeling of the biological reactor. Possible applications of these interval estimation systems are the spheres of robust control, where the presence of various types of uncertainties in the system dynamics is assumed, biotechnology and environmental systems and processes, mechatronics and robotics, etc.

  7. Fast reactors fuel cycle core physics results from the CAPRA-CADRA programme

    Energy Technology Data Exchange (ETDEWEB)

    Vasile, A.; Rimpault, G.; Tommasi, J.; Saint Jean, C. de; Delpech, M. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Hesketh, K. [BNFL, Inc., Denver, CO (United States); Beaumont, H.M.; Sunderland, R.E. [NNC Ltd. (United Kingdom); Newton, T.; Smith, P. [AEA Technology (United Kingdom); Raedt, Ch. de [SCK.CEN, Mol (Belgium); Vambenepe, G. [Electricite de France (EDF), 75 - Paris (France); Lefevre, J.C. [FRAMATOME, 92 - Paris-La-Defence (France); Maschek, W.; Haas, D

    2001-07-01

    This paper presents an overview of fast reactor core physics results obtained in the context of the CAPRA-CADRA European collaborative programme, whose aim is to investigate a broad range of possible options for plutonium and radioactive waste management. Different types of fast reactors have been studied to evaluate their potential capabilities with respect to the long term management of plutonium, minor actinides (MAs) and long- lived fission products (LLFPs). Among the several options aiming at reducing waste and consequently radio toxicity are: homogeneous recycling of Minor Actinides, heterogeneous recycling of Minor Actinides either without or with moderation, dedicated critical cores (fuelled mainly with Minor Actinides) and Accelerator Driven System (ADS) variants. In order to achieve a detailed understanding of the potential of the various options, advanced core physics methods have been implemented and tested and applied, for example, to improving control rod modeling and to studying safety aspects. There has also been code development and experimental work carried out to improve the understanding of fuel performance behaviors. (author)

  8. Validation of physics and thermalhydraulic computer codes for advanced Candu reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Wren, D.J.; Popov, N.; Snell, V.G. [Atomic Energy of Canada Ltd, (Canada)

    2004-07-01

    Atomic Energy of Canada Ltd. (AECL) is developing an Advanced Candu Reactor (ACR) that is an evolutionary advancement of the currently operating Candu 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular Candu concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the Candu 6, most of the key phenomena associated with the safety analyses of ACR are common, and the Candu industry standard tool-set of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design

  9. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  10. Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

    Science.gov (United States)

    2013-06-01

    Loss of Coolant Accident LWR Light Water Reactor MOX Mixed Oxide Fuel MTC Moderator Temperature Coefficient MWd/kgIHM Megawatt days per...working only with UO2 and UO2/PuO2 mixed oxide ( MOX ) fuels. 3.1 Studsvik Core Management Software CASMO-4E and SIMULATE-3 are the primary computational

  11. MicroRNA-375 inhibits colorectal cancer growth by targeting PIK3CA

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yihui [Department of Colorectal Surgery, The Third Affiliated Hospital of Harbin Medical University, 150 Haping Road, 150081 Harbin (China); Tang, Qingchao [Cancer Center, The Second Affiliated Hospital of Harbin Medical University, 246 Xuefu Road, 150086 Harbin (China); Li, Mingqi; Jiang, Shixiong [Department of Colorectal Surgery, The Third Affiliated Hospital of Harbin Medical University, 150 Haping Road, 150081 Harbin (China); Wang, Xishan, E-mail: wxshan12081@163.com [Cancer Center, The Second Affiliated Hospital of Harbin Medical University, 246 Xuefu Road, 150086 Harbin (China)

    2014-02-07

    Highlights: • miR-375 is downregulated in colorectal cancer cell lines and tissues. • miR-375 inhibits colorectal cancer cell growth by targeting PIK3CA. • miR-375 inhibits colorectal cancer cell growth in xenograft nude mice model. - Abstract: Colorectal cancer (CRC) is the second most common cause of death from cancer. MicroRNAs (miRNAs) represent a class of small non-coding RNAs that control gene expression by triggering RNA degradation or interfering with translation. Aberrant miRNA expression is involved in human disease including cancer. Herein, we showed that miR-375 was frequently down-regulated in human colorectal cancer cell lines and tissues when compared to normal human colon tissues. PIK3CA was identified as a potential miR-375 target by bioinformatics. Overexpression of miR-375 in SW480 and HCT15 cells reduced PIK3CA protein expression. Subsequently, using reporter constructs, we showed that the PIK3CA untranslated region (3′-UTR) carries the directly binding site of miR-375. Additionally, miR-375 suppressed CRC cell proliferation and colony formation and led to cell cycle arrest. Furthermore, miR-375 overexpression resulted in inhibition of phosphatidylinositol 3-kinase (PI3K)/Akt signaling pathway. SiRNA-mediated silencing of PIK3CA blocked the inhibitory effect of miR-375 on CRC cell growth. Lastly, we found overexpressed miR-375 effectively repressed tumor growth in xenograft animal experiments. Taken together, we propose that overexpression of miR-375 may provide a selective growth inhibition for CRC cells by targeting PI3K/Akt signaling pathway.

  12. Assessing reactor physics codes capabilities to simulate fast reactors on the example of the BN-600 benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, Vladimir [Scientific and Engineering Centre for Nuclear and Radiation Safety (SES NRS), Moscow (Russian Federation); Bousquet, Jeremy [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    This work aims to assess the capabilities of reactor physics codes (initially validated for thermal reactors) to simulate fast sodium cooled reactors. The BFS-62-3A critical experiment from the BN-600 Hybrid Core Benchmark Analyses was chosen for the investigation. Monte-Carlo codes (KENO from SCALE and SERPENT 2.1.23) and the deterministic diffusion code DYN3D-MG are applied to calculate the neutronic parameters. It was found that the multiplication factor and reactivity effects calculated by KENO and SERPENT using the ENDF/B-VII.0 continuous energy library are in a good agreement with each other and with the measured benchmark values. Few-groups macroscopic cross sections, required for DYN3D-MG, were prepared in applying different methods implemented in SCALE and SERPENT. The DYN3D-MG results of a simplified benchmark show reasonable agreement with results from Monte-Carlo calculations and measured values. The former results are used to justify DYN3D-MG implementation for sodium cooled fast reactors coupled deterministic analysis.

  13. Using thermal balance model to determine optimal reactor volume and insulation material needed in a laboratory-scale composting reactor.

    Science.gov (United States)

    Wang, Yongjiang; Pang, Li; Liu, Xinyu; Wang, Yuansheng; Zhou, Kexun; Luo, Fei

    2016-04-01

    A comprehensive model of thermal balance and degradation kinetics was developed to determine the optimal reactor volume and insulation material. Biological heat production and five channels of heat loss were considered in the thermal balance model for a representative reactor. Degradation kinetics was developed to make the model applicable to different types of substrates. Simulation of the model showed that the internal energy accumulation of compost was the significant heat loss channel, following by heat loss through reactor wall, and latent heat of water evaporation. Lower proportion of heat loss occurred through the reactor wall when the reactor volume was larger. Insulating materials with low densities and low conductive coefficients were more desirable for building small reactor systems. Model developed could be used to determine the optimal reactor volume and insulation material needed before the fabrication of a lab-scale composting system.

  14. APPLICATION OF MODEL PREDICTIVE CONTROL TO BATCH POLYMERIZATION REACTOR

    Directory of Open Access Journals (Sweden)

    N.M. Ghasem

    2006-06-01

    Full Text Available The absence of a stable operational state in polymerization reactors that operates in batches is factor that determine the need of a special control system. In this study, advanced control methodology is implemented for controlling the operation of a batch polymerization reactor for polystyrene production utilizingmodel predictive control. By utilizing a model of the polymerization process, the necessary operational conditions were determined for producing the polymer within the desired characteristics. The maincontrol objective is to bring the reactor temperature to its target temperature as rapidly as possible with minimal temperature overshoot. Control performance for the proposed method is encouraging. It has been observed that temperature overshoot can be minimized by the proposed method with the use of both reactor and jacket energy balance for reactor temperature control.

  15. Calcium phosphate precipitation modeling in a pellet reactor

    OpenAIRE

    Montastruc, Ludovic; Azzaro-Pantel, Catherine; Cabassud, Michel; Biscans, Béatrice

    2002-01-01

    The calcium phosphate precipitation in a pellet reactor can be evaluated by two main parameters: the phosphate conversion ratio and the phosphate removal efficiency. The conversion ratio depends mainly on the pH. The pellet reactor efficiency depends not only on pH but also on the hydrodynamical conditions. An efficiency model based on a thermochemical precipitation approach and an orthokinetic aggregation model is presented. In this paper, the results show that optimal conditions for pellet ...

  16. Research on physical and chemical parameters of coolant in Light-Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

  17. Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11

    The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

  18. Waste tyre pyrolysis: modelling of a moving bed reactor.

    Science.gov (United States)

    Aylón, E; Fernández-Colino, A; Murillo, R; Grasa, G; Navarro, M V; García, T; Mastral, A M

    2010-12-01

    This paper describes the development of a new model for waste tyre pyrolysis in a moving bed reactor. This model comprises three different sub-models: a kinetic sub-model that predicts solid conversion in terms of reaction time and temperature, a heat transfer sub-model that calculates the temperature profile inside the particle and the energy flux from the surroundings to the tyre particles and, finally, a hydrodynamic model that predicts the solid flow pattern inside the reactor. These three sub-models have been integrated in order to develop a comprehensive reactor model. Experimental results were obtained in a continuous moving bed reactor and used to validate model predictions, with good approximation achieved between the experimental and simulated results. In addition, a parametric study of the model was carried out, which showed that tyre particle heating is clearly faster than average particle residence time inside the reactor. Therefore, this fast particle heating together with fast reaction kinetics enables total solid conversion to be achieved in this system in accordance with the predictive model.

  19. Modelling and control design for SHARON/Anammox reactor sequence

    DEFF Research Database (Denmark)

    Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

    2012-01-01

    With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work presents a complete model of the SHARON/Anammox reactor sequence. The dynamics of the reactor were explored pointing out the different scales of the rates in the system: slow microbial...... metabolism against fast chemical reaction and mass transfer. Likewise, the analysis of the dynamics contributed to establish qualitatively the requirements for control of the reactors, both for regulation and for optimal operation. Work in progress on quantitatively analysing different control structure...

  20. An assessment of coupling algorithms for nuclear reactor core physics simulations

    Science.gov (United States)

    Hamilton, Steven; Berrill, Mark; Clarno, Kevin; Pawlowski, Roger; Toth, Alex; Kelley, C. T.; Evans, Thomas; Philip, Bobby

    2016-04-01

    This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss-Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton-Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.

  1. Physics parameter calculations for a Tandem Mirror Reactor with thermal barriers

    Energy Technology Data Exchange (ETDEWEB)

    Boghosian, B.M.; Lappa, D.A.; Logan, B.G.

    1979-11-06

    Thermal barriers are localized reductions in potential between the plugs and the central cell, which effectively insulate trapped plug electrons from the central cell electrons. By then applying electron heating in the plug, it is possible to obtain trapped electron temperatures that are much greater than those of the central cell electrons. This, in turn, effects an increase in the plug potential and central cell confinement with a concomitant decrease in plug density and injection power. Ions trapped in the barrier by collisions are removed by the injection of neutral beams directed inside the barrier cell loss cone; these beam neutrals convert trapped barrier ions to neutrals by charge exchange permitting their escape. We describe a zero-dimensional physics model for this type of reactor, and present some preliminary results for Q.

  2. Lessons learned for participation in recent OECD-NEA reactor physics and thermalhydraulic benchmarks

    Energy Technology Data Exchange (ETDEWEB)

    Novog, D.R.; Leung, K.H.; Ball, M. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)

    2013-07-01

    Over the last 6 years the OECD-NEA has initiated a series of computational benchmarks in the fields of reactor physics and thermalhydraulics. Within this context McMaster university has been a key contributor and applied several state of the art tools including TSUNAMI, DRAGON, ASSERT, STAR-CCM+, RELAP and TRACE. Considering the tremendous amount of international participation in these benchmarks, there were many lessons of both technical and non-technical that should be shared. This paper presents a summary of the benchmarks, the results and contributions from McMaster, and the authors opinion on the overall conclusions gained from these extensive benchmarks. The benchmarks discussed in this paper include the Uncertainty Analysis in Modelling (UAM), the BWR fine mesh bundle test (BFBT), the PWR Subchannel Boiling Test (PSBT), the MATiS mixing experiment and the IAEA super critical water benchmarks on heat transfer and stability. (author)

  3. Hybrid parallel code acceleration methods in full-core reactor physics calculations

    Energy Technology Data Exchange (ETDEWEB)

    Courau, T.; Plagne, L.; Ponicot, A. [EDF R and D, 1, Avenue du General de Gaulle, 92141 Clamart Cedex (France); Sjoden, G. [Nuclear and Radiological Engineering, Georgia Inst. of Technology, Atlanta, GA 30332 (United States)

    2012-07-01

    When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadrature required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)

  4. Development of multi-physics code systems based on the reactor dynamics code DYN3D

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Schuetze, Jochen [ANSYS Germany GmbH, Darmstadt (Germany); Frank, Thomas [ANSYS Germany GmbH, Otterfing (Germany); Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)

    2011-07-15

    The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

  5. Generomak: Fusion physics, engineering and costing model

    Energy Technology Data Exchange (ETDEWEB)

    Delene, J.G.; Krakowski, R.A.; Sheffield, J.; Dory, R.A.

    1988-06-01

    A generic fusion physics, engineering and economics model (Generomak) was developed as a means of performing consistent analysis of the economic viability of alternative magnetic fusion reactors. The original Generomak model developed at Oak Ridge by Sheffield was expanded for the analyses of the Senior Committee on Environmental Safety and Economics of Magnetic Fusion Energy (ESECOM). This report describes the Generomak code as used by ESECOM. The input data used for each of the ten ESECOM fusion plants and the Generomak code output for each case is given. 14 refs., 3 figs., 17 tabs.

  6. Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A L; Cetiner, M S; Wilson, Jr, T L

    2012-04-30

    system relate to flows within the reactor vessel during severe events and the resulting temperature profiles (temperature and duration) for major components. Critical components include the fuel, reactor vessel, primary piping, and the primary-to-intermediate heat exchangers (P-IHXs). The major AHTR power system loops are shown in Fig. 3. The intermediate heat transfer system is a group of three pumped salt loops that transports the energy produced in the primary system to the power conversion system. Two dynamic system models are used to analyze the AHTR. A Matlab/Simulink-based model initiated in 2011 has been updated to reflect the evolving design parameters related to the heat flows associated with the reactor vessel. The Matlab model utilizes simplified flow assumptions within the vessel and incorporates an empirical representation of the Direct Reactor Auxiliary Cooling System (DRACS). A Dymola/Modelica model incorporates a more sophisticated representation of primary coolant flow and a physics-based representation of the three-loop DRACS thermal hydraulics. This model is not currently operating in a fully integrated mode. The Matlab model serves as a prototype and provides verification for the Dymola model, and its use will be phased out as the Dymola model nears completion. The heat exchangers in the system are sized using spreadsheet-based, steady-state calculations. The detail features of the heat exchangers are programmed into the dynamic models, and the overall dimensions are used to generate realistic plant designs. For the modeling cases where the emphasis is on understanding responses within the intermediate and primary systems, the power conversion system may be modeled as a simple boundary condition at the intermediate-to-power conversion system heat exchangers.

  7. Identification of Chemical Reactor Plant’s Mathematical Model

    Directory of Open Access Journals (Sweden)

    Pyakillya Boris

    2015-01-01

    Full Text Available This work presents a solution of the identification problem of chemical reactor plant’s mathematical model. The main goal is to obtain a mathematical description of a chemical reactor plant from experimental data, which based on plant’s time response measurements. This data consists sequence of measurements for water jacket temperature and information about control input signal, which is used to govern plant’s behavior.

  8. Developments in Sensitivity Methodologies and the Validation of Reactor Physics Calculations

    Directory of Open Access Journals (Sweden)

    Giuseppe Palmiotti

    2012-01-01

    Full Text Available The sensitivity methodologies have been a remarkable story when adopted in the reactor physics field. Sensitivity coefficients can be used for different objectives like uncertainty estimates, design optimization, determination of target accuracy requirements, adjustment of input parameters, and evaluations of the representativity of an experiment with respect to a reference design configuration. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described.

  9. Fast burner reactor benchmark results from the NEA working party on physics of plutonium recycle

    Energy Technology Data Exchange (ETDEWEB)

    Hill, R.N.; Wade, D.C. [Argonne National Lab., IL (United States); Palmiotti, G. [CEA - Cadarache, Saint-Paul-Les-Durance (France)

    1995-12-01

    As part of a program proposed by the OECD/NEA Working Party on Physics of Plutonium Recycling (WPPR) to evaluate different scenarios for the use of plutonium, fast reactor physics benchmarks were developed; fuel cycle scenarios using either PUREX/TRUEX (oxide fuel) or pyrometallurgical (metal fuel) separation technologies were specified. These benchmarks were designed to evaluate the nuclear performance and radiotoxicity impact of a transuranic-burning fast reactor system. International benchmark results are summarized in this paper; and key conclusions are highlighted.

  10. Physics and Control Assessment of AN 850 Mw(e) - Leu-Candu Reactor.

    Science.gov (United States)

    Barbone, Michelangelo

    The physics and control assessment of an 850 MW(e) Low Enriched Uranium CANDU Pressurized Heavy Water (LEU -CANDU-PHW) reactor constitute the major objective of this thesis. The use of Low Enriched Uranium fuel in the present CANDU nuclear power generating stations is recognized as economically beneficial due to reduced fuelling costs. The LEU fuel cycle is also recognized as a stepping stone to transit from the present CANDU-PHW once-through natural Uranium cycle to advanced cycles such as those based on Plutonium recycle, once-through Th + U-235 cycle, Thorium with Uranium recycle and net U-235 feed, Thorium with Uranium recycle and Plutonium feed. However, although the use of Low Enriched Uranium in the present CANDU-PHW reactor has economic advantages, and it would act as a technical bridge between the present cycle and advanced cycles, technical problems in different areas of reactor physics and fuel management were anticipated. The present thesis research work adresses the areas of reactor physics, fuel management, and control (in particular, the spatial control of large CANDU-PHW reactors). The main conclusions that have been drawn following these studies are as follows: (1) The Low Enriched Uranium Cycle is feasible in a CANDU-PHW reactor of present design and provided that: (a) The enrichment is kept relatively low (that is, about 1% instead of 0.711%); (b) the number of bundles to be replaced at every refuelling operation is about one-half that of the natural Uranium fuel case; (c) The channels are refuelled in the same direction as the coolant. (2) The response of an LEU-CANDU-PHW reactor to reactivity perturbation such as single- and two-channel refuelling operation, shim transient, shutdown-start-up transient with enrichment levels of 0.9% and 1.2% is essentially very similar {provided that certain conditions in (1) are respected} to that of the natural uranium reactor core case without any reactor reoptimization. The general behaviour of the reactor

  11. Reactor physics studies in the GCFR phase-II critical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Pond, R B [ed.

    1976-09-01

    The reactor physics studies performed in the gas cooled fast reactor (GCFR) mockup on ZPR-9 are covered. This critical assembly, designated Phase II in the GCFR program, had a single zone PuO/sub 2/-UO/sub 2/ core composition and UO/sub 2/ radial and axial blankets. The assembly was built both with and without radial and axial stainless steel reflectors. The program included the following measurements: small-sample reactivity worths of reactor constituent materials (including helium); /sup 238/U Doppler effect; uranium and plutonium reaction rate distributions; thorium, uranium, and plutonium ..cap alpha.. and reactor kinetics. Analysis of the measurements used ENDF/B-IV nuclear data; anisotropic diffusion coefficients were used to account for neutron streaming effects. Comparison of measurements and calculations to GCFR Phase I are also made.

  12. System Thermal Model for the S-Prime Thermionic Reactor

    Science.gov (United States)

    Arx, Alan V. Von

    1994-07-01

    A model has been developed which numerically simulates heat transfer and flow characteristics of the thermal-hydraulic loop of the S-PRIME thermionic reactor. The components for which detailed models have been included are: the thermionic fuel elements (TFEs), heat pipe panels, flow loop and pumps. The reactor start-up operation was then modeled from zero to full power. It includes modelling of the melting of the heat pipe working fluid as well as correlations for the performance of the thermionic cells. The results show that there is stable operation during this period.

  13. Basic Model of a Control Assembly Drop in Nuclear Reactors

    Directory of Open Access Journals (Sweden)

    Radek BULÍN

    2013-06-01

    Full Text Available This paper is focused on the modelling and dynamic analysis of a nonlinear system representing a control assembly of the VVER 440/V213 nuclear reactor. A simple rigid body model intended for basic dynamic analyses is introduced. It contains the influences of the pressurized water and mainly the eects of possible control assembly contacts with guiding tubes inside the reactor. Another approach based on a complex multibody model is further described and the suitability of both modelling approaches is discussed.

  14. PIK3CA mutations define favorable prognostic biomarkers in operable breast cancer: a systematic review and meta-analysis

    Directory of Open Access Journals (Sweden)

    Liu YR

    2014-04-01

    Full Text Available Yi-Rong Liu,* Yi-Zhou Jiang,* Wen-Jia Zuo, Ke-Da Yu, Zhi-Ming ShaoDepartment of Breast Surgery, Cancer Center and Cancer Institute, Shanghai Medical College, Fudan University, Shanghai, People’s Republic of China, *These authors contributed equally to this publication Background: Mutations of the p110α catalytic subunit of phosphatidylinositol 3-kinase (PIK3CA are among the most common genetic aberrations in human breast cancer. At present, controversy exists concerning the prognostic value of the mutations. Methods: We performed a systematic review and meta-analysis to clarify the association between PIK3CA mutations and survival outcomes. A comprehensive, computerized literature search of PubMed, Web of Science databases, the Chinese Biomedical Literature Database, and Wangfang Data until August 27, 2013 was carried out. Eligible studies were included according to specific inclusion criteria. Pooled hazard ratio was estimated by using the fixed effects model or random effects model according to heterogeneity between studies. Results: Eight eligible studies were included in the analysis, all of which were retrospective cohort studies. The overall meta-analysis demonstrated that the PIK3CA mutations were associated with better clinical outcomes (hazard ratio 0.72; 95% confidence interval: 0.57–0.91; P=0.006. None of the single studies materially altered the original results and no evidence of publication bias was found. Further subgroup analysis of mutations in exons 9 and 20 did not show statistical significance. Conclusion: PIK3CA mutations in operable primary breast cancer indicate a good prognosis. Further studies should be conducted to investigate the effect of PIK3CA mutations on clinical outcomes in different histologic types, different molecular subtypes of breast cancer, and different exons of PIK3CA. Keywords: early breast cancer, p110g catalytic subunit of phosphatidylinositol 3-kinase, somatic mutations, prognosis

  15. Parametric study of the Incompletely Stirred Reactor modeling

    Energy Technology Data Exchange (ETDEWEB)

    Mobini, K. [Department of Mechanical Engineering, Shahid Rajaee University, Lavizan, Tehran (Iran); Bilger, R.W. [School of Aerospace, Mechanical and Mechatronic Engineering, University of Sydney, Sydney (Australia)

    2009-09-15

    The Incompletely Stirred Reactor (ISR) is a generalization of the widely-used Perfectly Stirred Reactor (PSR) model and allows for incomplete mixing within the reactor. Its formulation is based on the Conditional Moment Closure (CMC) method. This model is applicable to nonpremixed combustion with strong recirculation such as in a gas turbine combustor primary zone. The model uses the simplifying assumptions that the conditionally-averaged reactive-scalar concentrations are independent of position in the reactor: this results in ordinary differential equations in mixture fraction space. The simplicity of the model permits the use of very complex chemical mechanisms. The effects of the detailed chemistry can be found while still including the effects of micromixing. A parametric study is performed here on an ISR for combustion of methane at overall stoichiometric conditions to investigate the sensitivity of the model to different parameters. The focus here is on emissions of nitric oxide and carbon monoxide. It is shown that the most important parameters in the ISR model are reactor residence time, the chemical mechanism and the core-averaged Probability Density Function (PDF). Using several different shapes for the core-averaged PDF, it is shown that use of a bimodal PDF with a low minimum at stoichiometric mixture fraction and a large variance leads to lower nitric oxide formation. The 'rich-plus-lean' mixing or staged combustion strategy for combustion is thus supported. (author)

  16. Physiological Levels of Pik3caH1047R Mutation in the Mouse Mammary Gland Results in Ductal Hyperplasia and Formation of ERα-Positive Tumors

    Science.gov (United States)

    Tikoo, Anjali; Roh, Vincent; Montgomery, Karen G.; Ivetac, Ivan; Waring, Paul; Pelzer, Rebecca; Hare, Lauren; Shackleton, Mark; Humbert, Patrick; Phillips, Wayne A.

    2012-01-01

    PIK3CA, the gene coding for the p110α subunit of phosphoinositide 3-kinase, is frequently mutated in a variety of human tumors including breast cancers. To better understand the role of mutant PIK3CA in the initiation and/or progression of breast cancer, we have generated mice with a conditional knock-in of the common activating mutation, Pik3caH1047R, into one allele of the endogenous gene in the mammary gland. These mice developed a ductal anaplasia and hyperplasia by 6 weeks of age characterized by multi-layering of the epithelial lining of the mammary ducts and expansion of the luminal progenitor (Lin−; CD29lo; CD24+; CD61+) cell population. The Pik3caH1047R expressing mice eventually develop mammary tumors with 100% penetrance but with a long latency (>12 months). This is significantly longer than has been reported for transgenic models where expression of the mutant Pik3ca is driven by an exogenous promoter. Histological analysis of the tumors formed revealed predominantly ERα-positive fibroadenomas, carcinosarcomas and sarcomas. In vitro induction of Pik3caH1047R in immortalized mammary epithelial cells also resulted in tumor formation when injected into the mammary fat pad of immunodeficient recipient mice. This novel model, which reproduces the scenario of a heterozygous somatic mutation occurring in the endogenous PIK3CA gene, will thus be a valuable tool for investigating the role of Pik3caH1047R mutation in mammary tumorigenesis both in vivo and in vitro. PMID:22666336

  17. Physiological levels of Pik3ca(H1047R) mutation in the mouse mammary gland results in ductal hyperplasia and formation of ERα-positive tumors.

    Science.gov (United States)

    Tikoo, Anjali; Roh, Vincent; Montgomery, Karen G; Ivetac, Ivan; Waring, Paul; Pelzer, Rebecca; Hare, Lauren; Shackleton, Mark; Humbert, Patrick; Phillips, Wayne A

    2012-01-01

    PIK3CA, the gene coding for the p110α subunit of phosphoinositide 3-kinase, is frequently mutated in a variety of human tumors including breast cancers. To better understand the role of mutant PIK3CA in the initiation and/or progression of breast cancer, we have generated mice with a conditional knock-in of the common activating mutation, Pik3ca(H1047R), into one allele of the endogenous gene in the mammary gland. These mice developed a ductal anaplasia and hyperplasia by 6 weeks of age characterized by multi-layering of the epithelial lining of the mammary ducts and expansion of the luminal progenitor (Lin(-); CD29(lo); CD24(+); CD61(+)) cell population. The Pik3ca(H1047R) expressing mice eventually develop mammary tumors with 100% penetrance but with a long latency (>12 months). This is significantly longer than has been reported for transgenic models where expression of the mutant Pik3ca is driven by an exogenous promoter. Histological analysis of the tumors formed revealed predominantly ERα-positive fibroadenomas, carcinosarcomas and sarcomas. In vitro induction of Pik3ca(H1047R) in immortalized mammary epithelial cells also resulted in tumor formation when injected into the mammary fat pad of immunodeficient recipient mice. This novel model, which reproduces the scenario of a heterozygous somatic mutation occurring in the endogenous PIK3CA gene, will thus be a valuable tool for investigating the role of Pik3ca(H1047R) mutation in mammary tumorigenesis both in vivo and in vitro.

  18. Expression of PIK3CA, PTEN mRNA and PIK3CA mutations in primary breast cancer

    DEFF Research Database (Denmark)

    Palimaru, Irina; Brügmann, Anja; Wium-Andersen, Marie Kim;

    2013-01-01

    tissue samples of breast carcinoma and normal breast tissue were obtained from 175 breast cancer patients at the time of primary surgery, of these 105 patients were lymph node positive. Expression of PIK3CA and PTEN mRNA was quantified with Quantitative Real Time PCR. Somatic mutations in exon 9 and exon......PURPOSE: High activity of the intracellular phosphatidylinositol-3 kinase (PI3K) pathway is common in breast cancer. Here, we explore differences in expression of important PI3K pathway regulators: the activator, phosphatidylinositol-3-kinase catalytic subunit alpha (PIK3CA), and the tumour...... suppressor, phosphatase and tensin homolog (PTEN), in breast carcinoma tissue and normal breast tissue. Furthermore, we examine whether expression of PIK3CA and PTEN mRNA and occurrence of PIK3CA mutations are associated with lymph node metastases in patients with primary breast cancer. METHODS: Paired...

  19. Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

    Directory of Open Access Journals (Sweden)

    A. Rais

    2015-01-01

    Full Text Available In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

  20. Time variability of α from realistic models of Oklo reactors

    Science.gov (United States)

    Gould, C. R.; Sharapov, E. I.; Lamoreaux, S. K.

    2006-08-01

    We reanalyze Oklo Sm149 data using realistic models of the natural nuclear reactors. Disagreements among recent Oklo determinations of the time evolution of α, the electromagnetic fine structure constant, are shown to be due to different reactor models, which led to different neutron spectra used in the calculations. We use known Oklo reactor epithermal spectral indices as criteria for selecting realistic reactor models. Two Oklo reactors, RZ2 and RZ10, were modeled with MCNP. The resulting neutron spectra were used to calculate the change in the Sm149 effective neutron capture cross section as a function of a possible shift in the energy of the 97.3-meV resonance. We independently deduce ancient Sm149 effective cross sections and use these values to set limits on the time variation of α. Our study resolves a contradictory situation with previous Oklo α results. Our suggested 2σ bound on a possible time variation of α over 2 billion years is stringent: -0.11≤Δα/α≤0.24, in units of 10-7, but model dependent in that it assumes only α has varied over time.

  1. Aspects of the physics and chemistry of water radiolysis by fast neutrons and fast electrons in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    McCracken, D.R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Tsang, K.T. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Laughton, P.J

    1998-09-01

    Detailed radiation physics calculations of energy deposition have been done for the coolant of CANDU reactors and Pressurized Water Reactors (PWRs). The geometry of the CANDU fuel channel was modelled in detail. Fluxes and energy-deposition rates for neutrons, recoil ions, photons, and fast electrons have been calculated using MCNP4B, WIMS-AECL, and specifically derived energy-transfer factors. These factors generate the energy/flux spectra of recoil ions from fast-neutron energy/flux spectra. The energy spectrum was divided into 89 discrete ranges (energy bins).The production of oxidizing species and net coolant radiolysis can be suppressed by the addition of hydrogen to the coolant of nuclear reactors. It is argued that the net dissociation of coolant by gamma rays is suppressed by lower levels of excess hydrogen than when dissociation is by ion recoils. This has consequences for the modelling of coolant radiolysis by homogeneous kinetics. More added hydrogen is required to stop water radiolysis by recoil ions acting alone than if recoil ions and gamma rays acted concurrently in space and time. Homogeneous kinetic models and experimental data suggest that track overlap is very inefficient in providing radicals from gamma-ray tracks to recombine molecular products in ion-recoil tracks. An inhomogeneous chemical model is needed that incorporates ionizing-particle track structure and track overlap. Such a model does not yet exist, but a number of limiting cases using homogeneous kinetics are discussed. There are sufficient uncertainties and contradictions in the data relevant to the radiolysis of reactor coolant that the relatively high CHC's (critical hydrogen concentration) observed in NRU reactor experiments (compared to model predictions) may be explainable by errors in fundamental data and understanding of water radiolysis under reactor conditions. The radiation chemistry program at CRL has been focused to generate quantitative water-radiolysis data in a

  2. Physical models of cognition

    Science.gov (United States)

    Zak, Michail

    1994-05-01

    This paper presents and discusses physical models for simulating some aspects of neural intelligence, and, in particular, the process of cognition. The main departure from the classical approach here is in utilization of a terminal version of classical dynamics introduced by the author earlier. Based upon violations of the Lipschitz condition at equilibrium points, terminal dynamics attains two new fundamental properties: it is spontaneous and nondeterministic. Special attention is focused on terminal neurodynamics as a particular architecture of terminal dynamics which is suitable for modeling of information flows. Terminal neurodynamics possesses a well-organized probabilistic structure which can be analytically predicted, prescribed, and controlled, and therefore which presents a powerful tool for modeling real-life uncertainties. Two basic phenomena associated with random behavior of neurodynamic solutions are exploited. The first one is a stochastic attractor—a stable stationary stochastic process to which random solutions of a closed system converge. As a model of the cognition process, a stochastic attractor can be viewed as a universal tool for generalization and formation of classes of patterns. The concept of stochastic attractor is applied to model a collective brain paradigm explaining coordination between simple units of intelligence which perform a collective task without direct exchange of information. The second fundamental phenomenon discussed is terminal chaos which occurs in open systems. Applications of terminal chaos to information fusion as well as to explanation and modeling of coordination among neurons in biological systems are discussed. It should be emphasized that all the models of terminal neurodynamics are implementable in analog devices, which means that all the cognition processes discussed in the paper are reducible to the laws of Newtonian mechanics.

  3. Modified Mathematical Model For Neutralization System In Stirred Tank Reactor

    Directory of Open Access Journals (Sweden)

    Ahmmed Saadi Ibrehem

    2011-05-01

    Full Text Available A modified model for the neutralization process of Stirred Tank Reactors (CSTR reactor is presented in this study. The model accounts for the effect of strong acid [HCL] flowrate and strong base [NaOH] flowrate with the ionic concentrations of [Cl-] and [Na+] on the Ph of the system. In this work, the effect of important reactor parameters such as ionic concentrations and acid and base flowrates on the dynamic behavior of the CSTR is investigated and the behavior of mathematical model is compared with the reported models for the McAvoy model and Jutila model. Moreover, the results of the model are compared with the experimental data in terms of pH dynamic study. A good agreement is observed between our model prediction and the actual plant data. © 2011 BCREC UNDIP. All rights reserved(Received: 1st March 2011, Revised: 28th March 2011; Accepted: 7th April 2011[How to Cite: A.S. Ibrehem. (2011. Modified Mathematical Model For Neutralization System In Stirred Tank Reactor. Bulletin of Chemical Reaction Engineering & Catalysis, 6(1: 47-52. doi:10.9767/bcrec.6.1.825.47-52][How to Link / DOI: http://dx.doi.org/10.9767/bcrec.6.1.825.47-52 || or local:  http://ejournal.undip.ac.id/index.php/bcrec/article/view/825 ] | View in 

  4. Towards an efficient multiphysics model for nuclear reactor dynamics

    Directory of Open Access Journals (Sweden)

    Obaidurrahman K.

    2015-01-01

    Full Text Available Availability of fast computer resources nowadays has facilitated more in-depth modeling of complex engineering systems which involve strong multiphysics interactions. This multiphysics modeling is an important necessity in nuclear reactor safety studies where efforts are being made worldwide to combine the knowledge from all associated disciplines at one place to accomplish the most realistic simulation of involved phenomenon. On these lines coupled modeling of nuclear reactor neutron kinetics, fuel heat transfer and coolant transport is a regular practice nowadays for transient analysis of reactor core. However optimization between modeling accuracy and computational economy has always been a challenging task to ensure the adequate degree of reliability in such extensive numerical exercises. Complex reactor core modeling involves estimation of evolving 3-D core thermal state, which in turn demands an expensive multichannel based detailed core thermal hydraulics model. A novel approach of power weighted coupling between core neutronics and thermal hydraulics presented in this work aims to reduce the bulk of core thermal calculations in core dynamics modeling to a significant extent without compromising accuracy of computation. Coupled core model has been validated against a series of international benchmarks. Accuracy and computational efficiency of the proposed multiphysics model has been demonstrated by analyzing a reactivity initiated transient.

  5. Monte Carlo Modeling Electronuclear Processes in Cascade Subcritical Reactor

    CERN Document Server

    Bznuni, S A; Zhamkochyan, V M; Polyanskii, A A; Sosnin, A N; Khudaverdian, A G

    2000-01-01

    Accelerator driven subcritical cascade reactor composed of the main thermal neutron reactor constructed analogous to the core of the VVER-1000 reactor and a booster-reactor, which is constructed similar to the core of the BN-350 fast breeder reactor, is taken as a model example. It is shown by means of Monte Carlo calculations that such system is a safe energy source (k_{eff}=0.94-0.98) and it is capable of transmuting produced radioactive wastes (neutron flux density in the thermal zone is PHI^{max} (r,z)=10^{14} n/(cm^{-2} s^{-1}), neutron flux in the fast zone is respectively equal PHI^{max} (r,z)=2.25 cdot 10^{15} n/(cm^{-2} s^{-1}) if the beam current of the proton accelerator is k_{eff}=0.98 and I=5.3 mA). Suggested configuration of the "cascade" reactor system essentially reduces the requirements on the proton accelerator current.

  6. Computational fluid dynamic modeling of fluidized-bed polymerization reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rokkam, Ram [Iowa State Univ., Ames, IA (United States)

    2012-01-01

    Polyethylene is one of the most widely used plastics, and over 60 million tons are produced worldwide every year. Polyethylene is obtained by the catalytic polymerization of ethylene in gas and liquid phase reactors. The gas phase processes are more advantageous, and use fluidized-bed reactors for production of polyethylene. Since they operate so close to the melting point of the polymer, agglomeration is an operational concern in all slurry and gas polymerization processes. Electrostatics and hot spot formation are the main factors that contribute to agglomeration in gas-phase processes. Electrostatic charges in gas phase polymerization fluidized bed reactors are known to influence the bed hydrodynamics, particle elutriation, bubble size, bubble shape etc. Accumulation of electrostatic charges in the fluidized-bed can lead to operational issues. In this work a first-principles electrostatic model is developed and coupled with a multi-fluid computational fluid dynamic (CFD) model to understand the effect of electrostatics on the dynamics of a fluidized-bed. The multi-fluid CFD model for gas-particle flow is based on the kinetic theory of granular flows closures. The electrostatic model is developed based on a fixed, size-dependent charge for each type of particle (catalyst, polymer, polymer fines) phase. The combined CFD model is first verified using simple test cases, validated with experiments and applied to a pilot-scale polymerization fluidized-bed reactor. The CFD model reproduced qualitative trends in particle segregation and entrainment due to electrostatic charges observed in experiments. For the scale up of fluidized bed reactor, filtered models are developed and implemented on pilot scale reactor.

  7. Modelling of advanced structural materials for GEN IV reactors

    Science.gov (United States)

    Samaras, M.; Hoffelner, W.; Victoria, M.

    2007-09-01

    The choice of suitable materials and the assessment of long-term materials damage are key issues that need to be addressed for the safe and reliable performance of nuclear power plants. Operating conditions such as high temperatures, irradiation and a corrosive environment degrade materials properties, posing the risk of very expensive or even catastrophic plant damage. Materials scientists are faced with the scientific challenge to determine the long-term damage evolution of materials under service exposure in advanced plants. A higher confidence in life-time assessments of these materials requires an understanding of the related physical phenomena on a range of scales from the microscopic level of single defect damage effects all the way up to macroscopic effects. To overcome lengthy and expensive trial-and-error experiments, the multiscale modelling of materials behaviour is a promising tool, bringing new insights into the fundamental understanding of basic mechanisms. This paper presents the multiscale modelling methodology which is taking root internationally to address the issues of advanced structural materials for Gen IV reactors.

  8. Thermohydraulic and nuclear modeling of natural fission reactors

    Science.gov (United States)

    Viggato, Jason Charles

    Experimental verification of proposed nuclear waste storage schemes in geologic repositories is not possible, however, a natural analog exists in the form of ancient natural reactors that existed in uranium-rich ores. Two billion years ago, the enrichment of natural uranium was high enough to allow a sustained chain reaction in the presence of water as a moderator. Several natural reactors occurred in Gabon, Africa and were discovered in the early 1970's. These reactors operated at low power levels for hundreds of thousands of years. Heated water generated from the reactors also leached uranium from the surrounding rock strata and deposited it in the reactor cores. This increased the concentration of uranium in the core over time and served to "refuel" the reactor. This has strong implications in the design of modern geologic repositories for spent nuclear fuel. The possibility of accidental fission events in man-made repositories exists and the geologic evidence from Oklo suggests how those events may progress and enhance local concentrations of uranium. Based on a review of the literature, a comprehensive code was developed to model the thermohydraulic behavior and criticality conditions that may have existed in the Oklo reactor core. A two-dimensional numerical model that incorporates modeling of fluid flow, temperatures, and nuclear fission and subsequent heat generation was developed for the Oklo natural reactors. The operating temperatures ranged from about 456 K to about 721 K. Critical reactions were observed for a wide range of concentrations and porosity values (9 to 30 percent UO2 and 10 to 20 percent porosity). Periodic operation occurred in the computer model prediction with UO2 concentrations of 30 percent in the core and 5 percent in the surrounding material. For saturated conditions and 30 percent porosity, the model predicted temperature transients with a period of about 5 hours. Kuroda predicted 3 to 4 hour durations for temperature transients

  9. Students' Assessment of Interactive Distance Experimentation in Nuclear Reactor Physics Laboratory Education

    Science.gov (United States)

    Malkawi, Salaheddin; Al-Araidah, Omar

    2013-01-01

    Laboratory experiments develop students' skills in dealing with laboratory instruments and physical processes with the objective of reinforcing the understanding of the investigated subject. In nuclear engineering, where research reactors play a vital role in the practical education of students, the high cost and long construction time of research…

  10. 78 FR 31821 - Physical Protection of Shipments of Irradiated Reactor Fuel

    Science.gov (United States)

    2013-05-28

    ...: Nuclear Regulatory Commission. ACTION: NUREG; issuance. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is issuing Revision 2 of NUREG-0561, ``Physical Protection of Shipments of Irradiated Reactor... individuals granted unescorted access to SNF during transportation. DATES: Revision 2 of NUREG-0561...

  11. Mathematical modelling of methane steam reforming in a membrane reactor: an isothermal model

    Energy Technology Data Exchange (ETDEWEB)

    Assaf, E.M. [Sao Paulo Univ., Sao Carlos, SP (Brazil). Dept. de Fisico-Quimica; Jesus, C.D.F.; Assaf, J.M. [Sao Carlos Univ., SP (Brazil). Dept. de Engenharia Quimica

    1998-06-01

    A mathematical modelling of one-dimensional, stationary and isothermic membrane reactor for methane steam reforming was developed to compare the maximum yield for methane conversion in this reactor with that in a conventional fixed-bed reactor. Fick`s first law was used to describe the mechanism of hydrogen permeation. The variables studied include: reaction temperature, hydrogen feed flow rate and membrane thickness. The results show that the membrane reactor presents a higher methane conversion yield than the conventional fixed-bed reactor. (author) 16 refs., 5 figs., 1 tab.; e-mail: eassaf at iqsc.sc.usp.br; mansur at power.ufscar.br

  12. MATHEMATICAL MODELLING OF METHANE STEAM REFORMING IN A MEMBRANE REACTOR: AN ISOTHERMIC MODEL

    Directory of Open Access Journals (Sweden)

    E.M. ASSAF

    1998-06-01

    Full Text Available A mathematical modelling of one-dimensional, stationary and isothermic membrane reactor for methane steam reforming was developed to compare the maximum yield for methane conversion in this reactor with that in a conventional fixed-bed reactor. Fick's first law was used to describe the mechanism of hydrogen permeation. The variables studied include: reaction temperature, hydrogen feed flow rate and membrane thickness. The results show that the membrane reactor presents a higher methane conversion yield than the conventional fixed-bed reactor.

  13. BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Williamson, Richard L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Novascone, Stephen Rhead [Idaho National Lab. (INL), Idaho Falls, ID (United States); Medvedev, Pavel G. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on the formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.

  14. Reactor modeling in heterogeneous photocatalysis: toxicity and biodegradability assessment.

    Science.gov (United States)

    Satuf, M L; José, S; Paggi, J C; Brandi, R J; Cassano, A E; Alfano, O M

    2010-01-01

    Photocatalysis employing titanium dioxide is a useful method to degrade a wide variety of organic and inorganic pollutants from water and air. However, the application of this advanced oxidation process at industrial scale requires the development of mathematical models to design and scale-up photocatalytic reactors. In the present work, intrinsic kinetic expressions previously obtained in a laboratory reactor are employed to predict the performance of a bench scale reactor of different configuration and operating conditions. 4-Chlorophenol was chosen as the model pollutant. The toxicity and biodegradability of the irradiated mixture in the bench photoreactor was also assessed. Good agreement was found between simulation and experimental data. The root mean square error of the estimations was 9.9%. The photocatalytic process clearly enhances the biodegradability of the reacting mixture, and the initial toxicity of the pollutant was significantly reduced by the treatment.

  15. Precision Neutrino Oscillation Physics with an Intermediate Baseline Reactor Neutrino Experiment

    CERN Document Server

    Choubey, S; Piai, M; Choubey, Sandhya

    2003-01-01

    We discuss the physics potential of intermediate $L \\sim 20 \\div 30$ km baseline experiments at reactor facilities, assuming that the solar neutrino oscillation parameters $\\Delta m^2_{\\odot}$ and $\\theta_{\\odot}$ lie in the high-LMA solution region. We show that such an intermediate baseline reactor experiment can determine both $\\Delta m^2_{\\odot}$ and $\\theta_{\\odot}$ with a remarkably high precision. We perform also a detailed study of the sensitivity of the indicated experiment to $\\Delta m^2_{\\rm atm}$, which drives the dominant atmospheric $\

  16. Evolutionary Industrial Physical Model Generation

    Science.gov (United States)

    Carrascal, Alberto; Alberdi, Amaia

    Both complexity and lack of knowledge associated to physical processes makes physical models design an arduous task. Frequently, the only available information about the physical processes are the heuristic data obtained from experiments or at best a rough idea on what are the physical principles and laws that underlie considered physical processes. Then the problem is converted to find a mathematical expression which fits data. There exist traditional approaches to tackle the inductive model search process from data, such as regression, interpolation, finite element method, etc. Nevertheless, these methods either are only able to solve a reduced number of simple model typologies, or the given black-box solution does not contribute to clarify the analyzed physical process. In this paper a hybrid evolutionary approach to search complex physical models is proposed. Tests carried out on a real-world industrial physical process (abrasive water jet machining) demonstrate the validity of this approach.

  17. Landslides as weathering reactors; links between physical erosion and weathering in rapidly eroding mountain belts

    Science.gov (United States)

    Emberson, R.; Hovius, N.; Galy, A.

    2014-12-01

    The link between physical erosion and chemical weathering is generally modelled with a surface-blanketing weathering zone, where the supply of fresh minerals is tied to the average rate of denudation. In very fast eroding environments, however, sediment production is dominated by landsliding, which acts in a stochastic fashion across the landscape, contrasting strongly with more uniform denudation models. If physical erosion is a driver of weathering at the highest erosion rates, then an alternative weathering model is required. Here we show that landslides can be effective 'weathering reactors'. Previous work modelling the effect of landslides on chemical weathering (Gabet 2007) considered the fresh bedrock surfaces exposed in landslide scars. However, fracturing during the landslide motion generates fresh surfaces, the total surface area of which exceeds that of the exposed scar by many orders of magnitude. Moreover, landslides introduce concavity into hillslopes, which acts to catch precipitation. This is funnelled into a deposit of highly fragmented rock mass with large reactive surface area and limited hydraulic conductivity (Lo et al. 2007). This allows percolating water reaction time for chemical weathering; any admixture of macerated organic debris could yield organic acid to further accelerate weathering. In the South island of New Zealand, seepage from recent landslide deposits has systematically high solute concentrations, far outstripping concentration in runoff from locations where soils are present. River total dissolved load in the western Southern Alps is highly correlated with the rate of recent (erosion; this contrasts with persistent and ubiquitous weathering associated with soil production. Solute fluxes from fast eroding landscapes therefore likely depend on climatic or tectonic forcing of mass wasting; greater precipitation would drive increased weathering, while earthquakes, in generating landslides (Dadson et al. 2003; Chen & Hawkins 2009

  18. Two-dimensional model for circulating fluidized-bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schoenfelder, H.; Kruse, M.; Werther, J. [Technical Univ. Hamburg-Harburg, Hamburg (Germany). Dept. of Chemical Engineering

    1996-07-01

    Circulating fluidized bed reactors are widely used for the combustion of coal in power stations as well as for the cracking of heavy oil in the petroleum industry. A two-dimensional reactor model for circulating fluidized beds (CFB) was studied based on the assumption that at every location within the riser, a descending dense phase and a rising lean phase coexist. Fluid mechanical variables may be calculated from one measured radial solids flux profile (upward and downward). The internal mass-transfer behavior is described on the basis of tracer gas experiments. The CFB reactor model was tested against data from ozone decomposition experiments in a CFB cold flow model (15.6-m height, 0.4-m ID) operated in the ranges 2.5--4.5 m/s and 9--45 kg/(m{sup 2}{center_dot}s) of superficial gas velocity and solids mass flux, respectively. Based on effective reaction rate constants determined from the ozone exit concentration, the model was used to predict the spatial reactant distribution within the reactor. Model predictions agreed well with measurements.

  19. Technical Basis for Physical Fidelity of NRC Control Room Training Simulators for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Minsk, Brian S.; Branch, Kristi M.; Bates, Edward K.; Mitchell, Mark R.; Gore, Bryan F.; Faris, Drury K.

    2009-10-09

    The objective of this study is to determine how simulator physical fidelity influences the effectiveness of training the regulatory personnel responsible for examination and oversight of operating personnel and inspection of technical systems at nuclear power reactors. It seeks to contribute to the U.S. Nuclear Regulatory Commission’s (NRC’s) understanding of the physical fidelity requirements of training simulators. The goal of the study is to provide an analytic framework, data, and analyses that inform NRC decisions about the physical fidelity requirements of the simulators it will need to train its staff for assignment at advanced reactors. These staff are expected to come from increasingly diverse educational and experiential backgrounds.

  20. Reactor physics studies for the Advanced Fuel Cycle Initiative (AFCI) Reactor-Accelerator Coupling Experiments (RACE) Project

    Science.gov (United States)

    Stankovskiy, Evgeny Yuryevich

    In the recently completed RACE Project of the AFCI, accelerator-driven subcritical systems (ADS) experiments were conducted to develop technology of coupling accelerators to nuclear reactors. In these experiments electron accelerators induced photon-neutron reactions in heavy-metal targets to initiate fission reactions in ADS. Although the Idaho State University (ISU) RACE ADS was constructed only to develop measurement techniques for advanced experiments, many reactor kinetics experiments were conducted there. In the research reported in this dissertation, a method was developed to calculate kinetics parameters for measurement and calculation of the reactivity of ADS, a safety parameter that is necessary for control and monitoring of power production. Reactivity is measured in units of fraction of delayed versus prompt neutron from fission, a quantity that cannot be directly measured in far-subcritical reactors such as the ISU RACE configuration. A new technique is reported herein to calculate it accurately and to predict kinetic behavior of a far-subcritical ADS. Experiments conducted at ISU are first described and experimental data are presented before development of the kinetic theory used in the new computational method. Because of the complexity of the ISU ADS, the Monte-Carlo method as applied in the MCNP code is most suitable for modeling reactor kinetics. However, the standard method of calculating the delayed neutron fraction produces inaccurate values. A new method was developed and used herein to evaluate actual experiments. An advantage of this method is that its efficiency is independent of the fission yield of delayed neutrons, which makes it suitable for fuel with a minor actinide component (e.g. transmutation fuels). The implementation of this method is based on a correlated sampling technique which allows the accurate evaluation of delayed and prompt neutrons. The validity of the obtained results is indicated by good agreement between experimental

  1. Multi-physics design and analyses of long life reactors for lunar outposts

    Science.gov (United States)

    Schriener, Timothy M.

    event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor's neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete

  2. INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

    2008-09-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR’06 are highlighted, and the future of the two projects is discussed.

  3. Model biases in high-burnup fast reactor simulations

    Energy Technology Data Exchange (ETDEWEB)

    Touran, N.; Cheatham, J.; Petroski, R. [TerraPower LLC, 11235 S.E. 6th St, Bellevue, WA 98004 (United States)

    2012-07-01

    A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)

  4. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  5. Study on Modeling Technology in Digital Reactor System

    Institute of Scientific and Technical Information of China (English)

    刘晓平; 罗月童; 童莉莉

    2004-01-01

    Modeling is the kernel part of a digital reactor system. As an extensible platform for reactor conceptual design, it is very important to study modeling technology and develop some kind of tools to speed up preparation of all classical computing models. This paper introduces the background of the project and basic conception of digital reactor. MCAM is taken as an example for modeling and its related technologies used are given. It is an interface program for MCNP geometry model developed by FDS team (ASIPP & HUT), and designed to run on windows system. MCAM aims at utilizing CAD technology to facilitate creation of MCNP geometry model. There have been two ways for MCAM to utilize CAD technology:(1) Making use of user interface technology in aid of generation of MCNP geometry model;(2) Making use of existing 3D CAD model to accelerate creation of MCNP geometry model. This paper gives an overview of MCAM's major function. At last, several examples are given to demonstrate MCAM's various capabilities.

  6. Process Model of the Gas Recovery System in an IFE reactor

    Science.gov (United States)

    Gentile, Charles; Aristova, Maria

    2007-11-01

    It is necessary to develop a detailed representative model for the fuel recovery system (FRS) in the prospective direct drive inertial fusion energy (IFE) reactor. In order to observe the interaction of all components, a chemical process model is developed as part of the conceptual design phase of the project. Initially, the reactants, system structure, and processes are defined using the known contents of the vacuum vessel exhaust. The output, which will include physical properties and chemical content of the products, is analyzed to determine the most efficient and productive system parameters. The results of the modeling will be presented in this paper. This modeling exercise will be instrumental in optimizing and closing the fusion fuel cycle in the IFE power reactor.

  7. Advanced Mesh-Enabled Monte carlo capability for Multi-Physics Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Paul; Evans, Thomas; Tautges, Tim

    2012-12-24

    This project will accumulate high-precision fluxes throughout reactor geometry on a non- orthogonal grid of cells to support multi-physics coupling, in order to more accurately calculate parameters such as reactivity coefficients and to generate multi-group cross sections. This work will be based upon recent developments to incorporate advanced geometry and mesh capability in a modular Monte Carlo toolkit with computational science technology that is in use in related reactor simulation software development. Coupling this capability with production-scale Monte Carlo radiation transport codes can provide advanced and extensible test-beds for these developments. Continuous energy Monte Carlo methods are generally considered to be the most accurate computational tool for simulating radiation transport in complex geometries, particularly neutron transport in reactors. Nevertheless, there are several limitations for their use in reactor analysis. Most significantly, there is a trade-off between the fidelity of results in phase space, statistical accuracy, and the amount of computer time required for simulation. Consequently, to achieve an acceptable level of statistical convergence in high-fidelity results required for modern coupled multi-physics analysis, the required computer time makes Monte Carlo methods prohibitive for design iterations and detailed whole-core analysis. More subtly, the statistical uncertainty is typically not uniform throughout the domain, and the simulation quality is limited by the regions with the largest statistical uncertainty. In addition, the formulation of neutron scattering laws in continuous energy Monte Carlo methods makes it difficult to calculate adjoint neutron fluxes required to properly determine important reactivity parameters. Finally, most Monte Carlo codes available for reactor analysis have relied on orthogonal hexahedral grids for tallies that do not conform to the geometric boundaries and are thus generally not well

  8. Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin Allen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, Sebastian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gleicher, Frederick Nathan [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, Mark David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition, this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.

  9. V.S.O.P. (99) for WINDOWS and UNIX : computer code system for reactor physics and fuel cycle simulation

    OpenAIRE

    Rütten, H. J.; Haas, K. A.; Brockmann, H.; Ohlig, U.; Scherer, W.

    2000-01-01

    V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The...

  10. V.S.O.P.(97) Computer Code System for Reactor Physics and Fuel Cycle Simulation : Input Manual and Comments

    OpenAIRE

    Rütten, H. J.; Haas, K. A.; Brockmann, H.; Ohlig, U.; Scherer, W.

    1998-01-01

    V.S.O.P. (97) is a computer code system for the comprehensive numerical simulation ofthe physics of thermal reactors. It implies processing ofcross sections, the setup ofthe reactor and ofthe fuel element, repeated neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to EM and to two spatial dimensi...

  11. MODELING THE ELECTROLYTIC DECHLORINATION OF TRICHLOROETHYLENE IN A GRANULAR GRAPHITE-PACKED REACTOR

    Science.gov (United States)

    A comprehensive reactor model was developed for the electrolytic dechlorination of trichloroethylene (TCE) at a granular-graphite cathode. The reactor model describes the dynamic processes of TCE dechlorination and adsorption, and the formation and dechlorination of all the major...

  12. Building Mental Models by Dissecting Physical Models

    Science.gov (United States)

    Srivastava, Anveshna

    2016-01-01

    When students build physical models from prefabricated components to learn about model systems, there is an implicit trade-off between the physical degrees of freedom in building the model and the intensity of instructor supervision needed. Models that are too flexible, permitting multiple possible constructions require greater supervision to…

  13. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  14. KRASness and PIK3CAness in patients with advanced colorectal cancer: outcome after treatment with early-phase trials with targeted pathway inhibitors.

    Directory of Open Access Journals (Sweden)

    Ignacio Garrido-Laguna

    Full Text Available PURPOSE: To evaluate clinicopathologic and molecular features of patients with metastatic colorectal cancer (mCRC and their outcomes in early-phase trials using pathway-targeting agents. PATIENTS AND METHODS: We analyzed characteristics of 238 patients with mCRC referred to the phase 1 trials unit at MD Anderson Cancer Center. KRAS, PIK3CA and BRAF status were tested using PCR-based DNA sequencing. RESULTS: Fifty-one percent of patients harbored KRAS mutations; 15% had PIK3CA mutations. In the multivariate regression model for clinical characteristics KRAS mutations were associated with an increased incidence of lung and bone metastases and decreased incidence of adrenal metastases; PIK3CA mutations were marginally correlated with mucinous tumors (p = 0.05. In the univariate analysis, KRAS and PIK3CA mutations were strongly associated. Advanced Duke's stage (p<0.0001 and KRAS mutations (p = 0.01 were the only significant independent predictors of poor survival (Cox proportional hazards model. Patients with PIK3CA mutations had a trend toward shorter progression-free survival when treated with anti-EGFR therapies (p = 0.07. Eighteen of 78 assessable patients (23% treated with PI3K/Akt/mTOR axis inhibitors achieved stable disease [SD] ≥6 months or complete response/partial response (CR/PR, only one of whom were in the subgroup (N = 15 with PIK3CA mutations, perhaps because 10 of these 15 patients (67% had coexisting KRAS mutations. No SD ≥6 months/CR/PR was observed in the 10 patients treated with mitogen-activating protein kinase (MAPK pathway targeting drugs. CONCLUSIONS: KRAS and PIK3CA mutations frequently coexist in patients with colorectal cancer, and are associated with clinical characteristics and outcome. Overcoming resistance may require targeting both pathways.

  15. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, D.W.; Schwint, A.E.; Hartwell, J.K.; Heber, E.M.; Trivillin, V.; Castillo, J.; Wentzeis, L.; Sloan, P.; Wemple, C.A.

    2004-10-04

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  16. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Amanda E. Schwint; John K. Hartwell; Elisa M. Heber; Veronica Trivillin; Jorge Castillo; Luis Wentzeis; Patrick Sloan; Charles A. Wemple

    2004-10-01

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  17. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    Energy Technology Data Exchange (ETDEWEB)

    Chakraborty, Pritam [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Biner, Suleyman Bulent [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Spencer, Benjamin Whiting [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-07-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  18. Simulation of MILD combustion using Perfectly Stirred Reactor model

    KAUST Repository

    Chen, Z.

    2016-07-06

    A simple model based on a Perfectly Stirred Reactor (PSR) is proposed for moderate or intense low-oxygen dilution (MILD) combustion. The PSR calculation is performed covering the entire flammability range and the tabulated chemistry approach is used with a presumed joint probability density function (PDF). The jet, in hot and diluted coflow experimental set-up under MILD conditions, is simulated using this reactor model for two oxygen dilution levels. The computed results for mean temperature, major and minor species mass fractions are compared with the experimental data and simulation results obtained recently using a multi-environment transported PDF approach. Overall, a good agreement is observed at three different axial locations for these comparisons despite the over-predicted peak value of CO formation. This suggests that MILD combustion can be effectively modelled by the proposed PSR model with lower computational cost.

  19. Developments in Sensitivity Methodologies and the Validation of Reactor Physics Calculations

    OpenAIRE

    Giuseppe Palmiotti; Massimo Salvatores

    2012-01-01

    The sensitivity methodologies have been a remarkable story when adopted in the reactor physics field. Sensitivity coefficients can be used for different objectives like uncertainty estimates, design optimization, determination of target accuracy requirements, adjustment of input parameters, and evaluations of the representativity of an experiment with respect to a reference design configuration. A review of the methods used is provided, and several examples illustrate the success of the metho...

  20. Thermal and neutron-physical features of the nuclear reactor for a power pulsation plant for space applications

    Science.gov (United States)

    Gordeev, É. G.; Kaminskii, A. S.; Konyukhov, G. V.; Pavshuk, V. A.; Turbina, T. A.

    2012-05-01

    We have explored the possibility of creating small-size reactors with a high power output with the provision of thermal stability and nuclear safety under standard operating conditions and in emergency situations. The neutron-physical features of such a reactor have been considered and variants of its designs preserving the main principles and approaches of nuclear rocket engine technology are presented.

  1. Studies on modelling of bubble driven flows in chemical reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grevskott, Sverre

    1997-12-31

    Multiphase reactors are widely used in the process industry, especially in the petrochemical industry. They very often are characterized by very good thermal control and high heat transfer coefficients against heating and cooling surfaces. This thesis first reviews recent advances in bubble column modelling, focusing on the fundamental flow equations, drag forces, transversal forces and added mass forces. The mathematical equations for the bubble column reactor are developed, using an Eulerian description for the continuous and dispersed phase in tensor notation. Conservation equations for mass, momentum, energy and chemical species are given, and the k-{epsilon} and Rice-Geary models for turbulence are described. The different algebraic solvers used in the model are described, as are relaxation procedures. Simulation results are presented and compared with experimental values. Attention is focused on the modelling of void fractions and gas velocities in the column. The energy conservation equation has been included in the bubble column model in order to model temperature distributions in a heated reactor. The conservation equation of chemical species has been included to simulate absorption of CO{sub 2}. Simulated axial and radial mass fraction profiles for CO{sub 2} in the gas phase are compared with measured values. Simulations of the dynamic behaviour of the column are also presented. 189 refs., 124 figs., 1 tab.

  2. A bibliography on finite element and related methods analysis in reactor physics computations (1971--1997)

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, D.C.

    1998-01-01

    This bibliography provides a list of references on finite element and related methods analysis in reactor physics computations. These references have been published in scientific journals, conference proceedings, technical reports, thesis/dissertations and as chapters in reference books from 1971 to the present. Both English and non-English references are included. All references contained in the bibliography are sorted alphabetically by the first author`s name and a subsort by date of publication. The majority of the references relate to reactor physics analysis using the finite element method. Related topics include the boundary element method, the boundary integral method, and the global element method. All aspects of reactor physics computations relating to these methods are included: diffusion theory, deterministic radiation and neutron transport theory, kinetics, fusion research, particle tracking in finite element grids, and applications. For user convenience, many of the listed references have been categorized. The list of references is not all inclusive. In general, nodal methods were purposely excluded, although a few references do demonstrate characteristics of finite element methodology using nodal methods (usually as a non-conforming element basis). This area could be expanded. The author is aware of several other references (conferences, thesis/dissertations, etc.) that were not able to be independently tracked using available resources and thus were not included in this listing.

  3. Modelling of an ASR countercurrent pyrolysis reactor with nonlinear kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Chiarioni, A.; Reverberi, A.P.; Dovi, V.G. [Universita degli Studi di Genova (Italy). Dipartimento di Ingegneria Chimica e di Processo ' G.B. Bonino' ; El-Shaarawi, A.H. [National Water Research Institute, Burlington, Ont. (Canada)

    2003-10-01

    The main objective of this work is focused on the modelling of a steady-state reactor where an automotive shredder residue (ASR) is subject to pyrolysis. The gas and solid temperature inside the reactor and the relevant density profiles of both phases are simulated for fixed values of the geometry of the apparatus and a lumped kinetic model is adopted to take into account the high heterogeneity of the ASR material. The key elements for the simulation are the inlet solid temperature and the outlet gas temperature. The problem is modelled by a system of first-order boundary-value ordinary differential equations and it is solved by means of a relaxation technique owing to the nonlinearities contained in the chemical kinetic expression. (author)

  4. A CFD model for biomass fast pyrolysis in fluidized-bed reactors

    Science.gov (United States)

    Xue, Qingluan; Heindel, T. J.; Fox, R. O.

    2010-11-01

    A numerical study is conducted to evaluate the performance and optimal operating conditions of fluidized-bed reactors for fast pyrolysis of biomass to bio-oil. A comprehensive CFD model, coupling a pyrolysis kinetic model with a detailed hydrodynamics model, is developed. A lumped kinetic model is applied to describe the pyrolysis of biomass particles. Variable particle porosity is used to account for the evolution of particle physical properties. The kinetic scheme includes primary decomposition and secondary cracking of tar. Biomass is composed of reference components: cellulose, hemicellulose, and lignin. Products are categorized into groups: gaseous, tar vapor, and solid char. The particle kinetic processes and their interaction with the reactive gas phase are modeled with a multi-fluid model derived from the kinetic theory of granular flow. The gas, sand and biomass constitute three continuum phases coupled by the interphase source terms. The model is applied to investigate the effect of operating conditions on the tar yield in a fluidized-bed reactor. The influence of various parameters on tar yield, including operating temperature and others are investigated. Predicted optimal conditions for tar yield and scale-up of the reactor are discussed.

  5. Progress Report for Period Ending December 1961. Department of Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Tell, B. (ed.)

    1962-08-15

    This is the second Progress Report from the Department for Reactor Physics of Aktiebolaget Atomenergi, which is issued for the information of institutions and persons interested in the progress of the work. In this report the activities of the General Physics Section have been included, since this section nowadays belongs to the department. This is merely an informal progress report, and the results and data presented must be taken as preliminary. Final results will be submitted for publication either in the regular technical journals or as monographs in the series AE-reports.

  6. Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

    KAUST Repository

    Zhao, Zhenlong

    2013-01-17

    Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

  7. Numerical modeling of disperse material evaporation in axisymmetric thermal plasma reactor

    Directory of Open Access Journals (Sweden)

    Stefanović Predrag Lj.

    2003-01-01

    Full Text Available A numerical 3D Euler-Lagrangian stochastic-deterministic (LSD model of two-phase flow laden with solid particles was developed. The model includes the relevant physical effects, namely phase interaction, panicle dispersion by turbulence, lift forces, particle-particle collisions, particle-wall collisions, heat and mass transfer between phases, melting and evaporation of particles, vapour diffusion in the gas flow. It was applied to simulate the processes in thermal plasma reactors, designed for the production of the ceramic powders. Paper presents results of extensive numerical simulation provided (a to determine critical mechanism of interphase heat and mass transfer in plasma flows, (b to show relative influence of some plasma reactor parameters on solid precursor evaporation efficiency: 1 - inlet plasma temperature, 2 - inlet plasma velocity, 3 - particle initial diameter, 4 - particle injection angle a, and 5 - reactor wall temperature, (c to analyze the possibilities for high evaporation efficiency of different starting solid precursors (Si, Al, Ti, and B2O3 powder, and (d to compare different plasma reactor configurations in conjunction with disperse material evaporation efficiency.

  8. REACTOR AND SHIELD PHYSICS. Comprehensive Technical Report, General Electric Direct-Air-Cycle, Aircraft Nuclear Propulsion Program.

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.E.; Simpson, J.D.

    1962-01-01

    This volume is one of twenty-one summarizing the Aircraft Nuclear Propulsion Program of the General Electric Company. This volume describes the experimental and theoretical work accomplished in the areas of reactor and shield physics.

  9. VIPRE modeling of VVER-1000 reactor core for DNB analyses

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Y.; Nguyen, Q. [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Cizek, J. [Nuclear Research Institute, Prague, (Czech Republic)

    1995-09-01

    Based on the one-pass modeling approach, the hot channels and the VVER-1000 reactor core can be modeled in 30 channels for DNB analyses using the VIPRE-01/MOD02 (VIPRE) code (VIPRE is owned by Electric Power Research Institute, Palo Alto, California). The VIPRE one-pass model does not compromise any accuracy in the hot channel local fluid conditions. Extensive qualifications include sensitivity studies of radial noding and crossflow parameters and comparisons with the results from THINC and CALOPEA subchannel codes. The qualifications confirm that the VIPRE code with the Westinghouse modeling method provides good computational performance and accuracy for VVER-1000 DNB analyses.

  10. Scalable Methods for Uncertainty Quantification, Data Assimilation and Target Accuracy Assessment for Multi-Physics Advanced Simulation of Light Water Reactors

    Science.gov (United States)

    Khuwaileh, Bassam

    High fidelity simulation of nuclear reactors entails large scale applications characterized with high dimensionality and tremendous complexity where various physics models are integrated in the form of coupled models (e.g. neutronic with thermal-hydraulic feedback). Each of the coupled modules represents a high fidelity formulation of the first principles governing the physics of interest. Therefore, new developments in high fidelity multi-physics simulation and the corresponding sensitivity/uncertainty quantification analysis are paramount to the development and competitiveness of reactors achieved through enhanced understanding of the design and safety margins. Accordingly, this dissertation introduces efficient and scalable algorithms for performing efficient Uncertainty Quantification (UQ), Data Assimilation (DA) and Target Accuracy Assessment (TAA) for large scale, multi-physics reactor design and safety problems. This dissertation builds upon previous efforts for adaptive core simulation and reduced order modeling algorithms and extends these efforts towards coupled multi-physics models with feedback. The core idea is to recast the reactor physics analysis in terms of reduced order models. This can be achieved via identifying the important/influential degrees of freedom (DoF) via the subspace analysis, such that the required analysis can be recast by considering the important DoF only. In this dissertation, efficient algorithms for lower dimensional subspace construction have been developed for single physics and multi-physics applications with feedback. Then the reduced subspace is used to solve realistic, large scale forward (UQ) and inverse problems (DA and TAA). Once the elite set of DoF is determined, the uncertainty/sensitivity/target accuracy assessment and data assimilation analysis can be performed accurately and efficiently for large scale, high dimensional multi-physics nuclear engineering applications. Hence, in this work a Karhunen-Loeve (KL

  11. Designing visual displays and system models for safe reactor operations

    Energy Technology Data Exchange (ETDEWEB)

    Brown-VanHoozer, S.A.

    1995-12-31

    The material presented in this paper is based on two studies involving the design of visual displays and the user`s prospective model of a system. The studies involve a methodology known as Neuro-Linguistic Programming and its use in expanding design choices from the operator`s perspective image. The contents of this paper focuses on the studies and how they are applicable to the safety of operating reactors.

  12. Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; J. Blair Briggs; Jim Gulliford; Ian Hill

    2014-10-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

  13. p55PIK Transcriptionally Activated by MZF1 Promotes Colorectal Cancer Cell Proliferation

    Directory of Open Access Journals (Sweden)

    Yu Deng

    2013-01-01

    Full Text Available p55PIK, regulatory subunit of class IA phosphatidylinositol 3-kinase (PI3K, plays a crucial role in cell cycle progression by interaction with tumor repressor retinoblastoma (Rb protein. A recent study showed that Rb protein can localize to the mitochondria in proliferative cells. Aberrant p55PIK expression may contribute to mitochondrial dysfunction in cancer progression. To reveal the mechanisms of p55PIK transcriptional regulation, the p55PIK promoter characteristics were analyzed. The data show that myeloid zinc finger 1, MZF1, is necessary for p55PIK gene transcription activation. ChIP (Chromatin immuno-precipitation assay shows that MZF1 binds to the cis-element “TGGGGA” in p55PIK promoter. In MZF1 overexpressed cells, the promoter activity, expression of p55PIK, and cell proliferation rate were observed to be significantly enhanced. Whereas in MZF1-silenced cells, the promoter activity and expression of p55PIK and cell proliferation level was statistically decreased. In CRC tissues, MZF1 and p55PIK mRNA expression were increased (P=0.046, P=0.047, resp.. A strong positive correlation (Rs=0.94 between MZF1 and p55PIK mRNA expression was observed. Taken together, we concluded that p55PIK is transcriptionally activated by MZF1, resulting in increased proliferation of colorectal cancer cells.

  14. GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; John D. Bess; Jim Gulliford

    2011-09-01

    Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the

  15. Physical Modeling Modular Boxes: PHOXES

    DEFF Research Database (Denmark)

    Gelineck, Steven; Serafin, Stefania

    2010-01-01

    This paper presents the development of a set of musical instruments, which are based on known physical modeling sound synthesis techniques. The instruments are modular, meaning that they can be combined in various ways. This makes it possible to experiment with physical interaction and sonic expl...

  16. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    Energy Technology Data Exchange (ETDEWEB)

    Adams, S.R.

    1985-10-01

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.

  17. Causal diagrams for physical models

    CERN Document Server

    Kinsler, Paul

    2015-01-01

    I present a scheme of drawing causal diagrams based on physically motivated mathematical models expressed in terms of temporal differential equations. They provide a means of better understanding the processes and causal relationships contained within such systems.

  18. Mathematical modeling of a three-phase trickle bed reactor

    Directory of Open Access Journals (Sweden)

    J. D. Silva

    2012-09-01

    Full Text Available The transient behavior in a three-phase trickle bed reactor system (N2/H2O-KCl/activated carbon, 298 K, 1.01 bar was evaluated using a dynamic tracer method. The system operated with liquid and gas phases flowing downward with constant gas flow Q G = 2.50 x 10-6 m³ s-1 and the liquid phase flow (Q L varying in the range from 4.25x10-6 m³ s-1 to 0.50x10-6 m³ s-1. The evolution of the KCl concentration in the aqueous liquid phase was measured at the outlet of the reactor in response to the concentration increase at reactor inlet. A mathematical model was formulated and the solutions of the equations fitted to the measured tracer concentrations. The order of magnitude of the axial dispersion, liquid-solid mass transfer and partial wetting efficiency coefficients were estimated based on a numerical optimization procedure where the initial values of these coefficients, obtained by empirical correlations, were modified by comparing experimental and calculated tracer concentrations. The final optimized values of the coefficients were calculated by the minimization of a quadratic objective function. Three correlations were proposed to estimate the parameters values under the conditions employed. By comparing experimental and predicted tracer concentration step evolutions under different operating conditions the model was validated.

  19. An Axial Dispersion Model for Evaporating Bubble Column Reactor

    Institute of Scientific and Technical Information of China (English)

    谢刚; 李希

    2004-01-01

    Evaporating bubble column reactor (EBCR) is a kind of aerated reactor in which the reaction heat is removed by the evaporation of volatile reaction mixture. In this paper, a mathematical model that accounts for the gas-liquid exothermic reaction and axial dispersions of both gas and liquid phase is employed to study the performance of EBCR for the process of p-xylene(PX) oxidation. The computational results show that there are remarkable concentration and temperature gradients in EBCR for high ratio of height to diameter (H/DT). The temperature is lower at the bottom of column and higher at the top, due to rapid evaporation induced by the feed gas near the bottom. The concentration profiles in the gas phase are more nonuniform than those (except PX) in the liquid phase, which causes more solvent burning consumption at high H/DT ratio. For p-xylene oxidation, theo ptimal H/DT is around 5.

  20. Advanced Multiphysics Thermal-Hydraulics Models for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Prashant K [ORNL; Freels, James D [ORNL

    2015-01-01

    Engineering design studies to determine the feasibility of converting the High Flux Isotope Reactor (HFIR) from using highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL). This work is part of an effort sponsored by the US Department of Energy (DOE) Reactor Conversion Program. HFIR is a very high flux pressurized light-water-cooled and moderated flux-trap type research reactor. HFIR s current missions are to support neutron scattering experiments, isotope production, and materials irradiation, including neutron activation analysis. Advanced three-dimensional multiphysics models of HFIR fuel were developed in COMSOL software for safety basis (worst case) operating conditions. Several types of physics including multilayer heat conduction, conjugate heat transfer, turbulent flows (RANS model) and structural mechanics were combined and solved for HFIR s inner and outer fuel elements. Alternate design features of the new LEU fuel were evaluated using these multiphysics models. This work led to a new, preliminary reference LEU design that combines a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone. Preliminary results of estimated thermal safety margins are presented. Fuel design studies and model enhancement continue.

  1. Modelling of gas-liquid reactors - stability and dynamic behaviour of a hydroformylation reactor, influence of mass transfer in the kinetics controlled regime

    NARCIS (Netherlands)

    Elk, E.P. van; Borman, P.C.; Kuipers, J.A.M.; Versteeg, G.F.

    2001-01-01

    On behalf of the development of new hydroformylation reactors, a research project was initiated to examine the dynamics of hydroformylation processes. The current paper presents the results of applying the rigorous reactor model and the approximate reactor model on a new, to be developed, hydroformy

  2. Hdr reactor containment fire modeling with Br12

    Energy Technology Data Exchange (ETDEWEB)

    Rockett, J.A.; Keski-Rahkonen, O.; Heikkilae, L.

    1992-01-01

    Fire tests at the German test reactor, HDR, were simulated using a Japanese zone model code, BRI2. Eight and ten room models of the containment building were developed. Critical phenomena occurring during simulation were explored. BRI2 can be used for this type of work but care must be exercised where a side wind increases entrainment by the fire plume. Horizontal vents were described by effective vertical vents. The effect of location of the vent to the ambient was found critical during severely oxygen limited burning. (Copyright (c) Valtion teknillinen tutkimuskeskus (VTT) 1992.)

  3. High Flux Isotope Reactor system RELAP5 input model

    Energy Technology Data Exchange (ETDEWEB)

    Morris, D.G.; Wendel, M.W.

    1993-01-01

    A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model.

  4. Initial Testing of the Microscopic Depletion Implementation in the MAMMOTH Reactor Physics Application

    Energy Technology Data Exchange (ETDEWEB)

    J. Ortensi; Y. Wang; S. Schunert; B.D. Ganapol; F.N. Gleicher; B. Baker; M.D. DeHart

    2016-09-01

    Present and new nuclear fuels that will be tested at the Transient Reactor Test (TREAT) facility will be analyzed with the MAMMOTH reactor physics application, currently under development, at Idaho National Laboratory. MAMMOTH natively couples the BISON, RELAP-7, and Rattlesnake applications within the MOOSE framework. This system allows the irradiation of fuel from beginning of life in a nuclear reactor until it is placed in TREAT for fuel testing within the same analysis mesh and, thus, retaining a very high level of resolution and fidelity. The calculation of the isotopic distribution in fuel requires the solution to the decay and transmutation equations coupled to the neutron transport equation. The Chebyshev Rational Approximation Method (CRAM) is the current state-of-the-art in the field, as was chosen to be the solver for the decay and transmutation equations. This report shows that the implementation of the CRAM solver within MAMMOTH is correct with various analytic benchmarks for decay and transmutation of nuclides. The results indicate that the solutions with CRAM order 16 achieve the level of precision of the benchmark. The CRAM solutions show little sensitivity to the time step size and consistently produce a high level of accuracy for isotopic decay for time steps of 1x10^11 years. Comparisons to DRAGON5 with 297 isotopes yield comparable results, but some differences need to be further analyzed.

  5. Seismic Physical Modeling Technology and Its Applications

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    This paper introduces the seismic physical modeling technology in the CNPC Key Lab of Geophysical Exploration. It includes the seismic physical model positioning system, the data acquisition system, sources, transducers,model materials, model building techniques, precision measurements of model geometry, the basic principles of the seismic physical modeling and experimental methods, and two physical model examples.

  6. Phenomenological and mechanistic modeling of melt-structure-water interactions in a light water reactor severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Bui, V.A

    1998-10-01

    The objective of this work is to address the modeling of the thermal hydrodynamic phenomena and interactions occurring during the progression of reactor severe accidents. Integrated phenomenological models are developed to describe the accident scenarios, which consist of many processes, while mechanistic modeling, including direct numerical simulation, is carried out to describe separate effects and selected physical phenomena of particular importance 88 refs, 54 figs, 7 tabs

  7. Physical Characteristics of Coffee Beans from Steaming Processin Single Column Reactor

    Directory of Open Access Journals (Sweden)

    Sukrisno Widyotomo

    2010-05-01

    Full Text Available One of important steps in decaffeination process is steaming. The aim of steaming is to expand coffee beans porosity in order to obtain optimal condition for decaffeination process. Steaming can be done in single column reactor using saturated water vapour as media. The objective of this research is to study physical characteristics of coffee beans after steaming process using single column reactor. Material tested was Robusta coffee with 13—14% moisture content after dry processing. Reactor capacity is 6 kg dried coffee beans and 30 l water to produce water vapour. Dried coffee beans classified in 4 grades, i.e. diameter size (d d>7,5 mm; 6,5reactor, decaffeination.

  8. Simulation of styrene polymerization reactors: kinetic and thermodynamic modeling

    Directory of Open Access Journals (Sweden)

    A. S. Almeida

    2008-06-01

    Full Text Available A mathematical model for the free radical polymerization of styrene is developed to predict the steady-state and dynamic behavior of a continuous process. Special emphasis is given for the kinetic and thermodynamic models, where the most sensitive parameters were estimated using data from an industrial plant. The thermodynamic model is based on a cubic equation of state and a mixing rule applied to the low-pressure vapor-liquid equilibrium of polymeric solutions, suitable for modeling the auto-refrigerated polymerization reactors, which use the vaporization rate to remove the reaction heat from the exothermic reactions. The simulation results show the high predictive capability of the proposed model when compared with plant data for conversion, average molecular weights, polydispersity, melt flow index, and thermal properties for different polymer grades.

  9. An overview of modeling methods for thermal mixing and stratification in large enclosures for reactor safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Per F. Peterson

    2010-10-01

    Thermal mixing and stratification phenomena play major roles in the safety of reactor systems with large enclosures, such as containment safety in current fleet of LWRs, long-term passive containment cooling in Gen III+ plants including AP-1000 and ESBWR, the cold and hot pool mixing in pool type sodium cooled fast reactor systems (SFR), and reactor cavity cooling system behavior in high temperature gas cooled reactors (HTGR), etc. Depending on the fidelity requirement and computational resources, 0-D steady state models (heat transfer correlations), 0-D lumped parameter based transient models, 1-D physical-based coarse grain models, and 3-D CFD models are available. Current major system analysis codes either have no models or only 0-D models for thermal stratification and mixing, which can only give highly approximate results for simple cases. While 3-D CFD methods can be used to analyze simple configurations, these methods require very fine grid resolution to resolve thin substructures such as jets and wall boundaries. Due to prohibitive computational expenses for long transients in very large volumes, 3-D CFD simulations remain impractical for system analyses. For mixing in stably stratified large enclosures, UC Berkeley developed 1-D models basing on Zuber’s hierarchical two-tiered scaling analysis (HTTSA) method where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. This paper will present an overview on important thermal mixing and stratification phenomena in large enclosures for different reactors, major modeling methods and their advantages and limits, potential paths to improve simulation capability and reduce analysis uncertainty in this area for advanced reactor system analysis tools.

  10. MELCOR Model Development of High Temperature Gas-cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Changyong; Huh, Changwook [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    The High Temperature Gas-cooled Reactor is one of the major challenging issues on the development of licensing technology for HTGR. The safety evaluation tools of HTGR can be developed in two ways - development of new HTGR-specific codes or revision of existing codes. The KINS is considering using existing analytic tools to the extent feasible, with appropriate modifications for the intended purpose. The system-level MELCOR code is traditionally used for LWR safety analysis, which is capable of performing thermal-fluid and accident analysis, including fission-product transport and release. Recently, this code is being modified for the NGNP HTGR by the NRC. In this study, the MELCOR input model for HTGR with Reactor Cavity Cooling System (RCCS) was developed and the steady state performance was analyzed to evaluate the applicability in HTGR. HTGR model with design characteristics of GT-MHR was developed using MELCOR 2.1 code to validate the applicability of MELCOR code to HTGR. In addition, the steady state of GT-MHR was analyzed with the developed model. It was evaluated to predict well the design parameters of GT-MHR. The developed model can be used as the basis for accident analysis of HTGR with further update of packages such as Radio Nuclide (RN) package.

  11. Meso-scale modeling of irradiated concrete in test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giorla, A. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Vaitová, M. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic); Le Pape, Y., E-mail: lepapeym@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Štemberk, P. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic)

    2015-12-15

    Highlights: • A meso-scale finite element model for irradiated concrete is developed. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • Confrontation with expansion and damage obtained from experiments is successful. • Effects of paste shrinkage, creep and ductility are discussed. - Abstract: A numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale is detailed in this paper. Irradiation experiments in test reactor (Elleuch et al., 1972), i.e., in accelerated conditions, are simulated. Concrete is considered as a two-phase material made of elastic inclusions (aggregate) subjected to thermal and irradiation-induced swelling and embedded in a cementitious matrix subjected to shrinkage and thermal expansion. The role of the hardened cement paste in the post-peak regime (brittle-ductile transition with decreasing loading rate), and creep effects are investigated. Radiation-induced volumetric expansion (RIVE) of the aggregate cause the development and propagation of damage around the aggregate which further develops in bridging cracks across the hardened cement paste between the individual aggregate particles. The development of damage is aggravated when shrinkage occurs simultaneously with RIVE during the irradiation experiment. The post-irradiation expansion derived from the simulation is well correlated with the experimental data and, the obtained damage levels are fully consistent with previous estimations based on a micromechanical interpretation of the experimental post-irradiation elastic properties (Le Pape et al., 2015). The proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-01-15

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

  13. piK Scattering in Three Flavour ChPT

    Science.gov (United States)

    Bijnens, Johan; Dhonte, Pierre; Talavera, Pere

    2004-05-01

    We present the scattering lengths for the piK processes in the three flavour Chiral Perturbation Theory (ChPT) framework at next-to-next-to-leading order (NNLO). The calculation has been performed analytically but we only include analytical results for the dependence on the low-energy constants (LECs) at NNLO due to the size of the expressions. These results, together with resonance estimates of the NNLO LECs are used to obtain constraints on the Zweig rule suppressed LECs at NLO, L4r and L6r. Contrary to expectations from NLO order calculations we find them to be compatible with zero. We do a preliminary study of combining the results from pipi scattering, piK scattering and the scalar form-factors and find only a marginal compatibility with all experimental/dispersive input data.

  14. A Roadmap and Discussion of Issues for Physics Analyses Required to Support Plutonium Disposition in VVER-1000 Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Primm, R.T.; Drischler, J.D.; Pavlovichev, A.M. Styrine, Y.A.

    2000-06-01

    The purpose of this report is to document the physics analyses that must be performed to successfully disposition weapons-usable plutonium in VVER-1000 reactors in the Russian Federation. The report is a document to support programmatic and financial planning. It does not include documentation of the technical procedures by which physics analyses are performed, nor are the results of any analyses included.

  15. An integration scheme for stiff solid-gas reactor models

    Directory of Open Access Journals (Sweden)

    Bjarne A. Foss

    2001-04-01

    Full Text Available Many dynamic models encounter numerical integration problems because of a large span in the dynamic modes. In this paper we develop a numerical integration scheme for systems that include a gas phase, and solid and liquid phases, such as a gas-solid reactor. The method is based on neglecting fast dynamic modes and exploiting the structure of the algebraic equations. The integration method is suitable for a large class of industrially relevant systems. The methodology has proven remarkably efficient. It has in practice performed excellent and been a key factor for the success of the industrial simulator for electrochemical furnaces for ferro-alloy production.

  16. Assessment of sensitivity of neutron-physical parameters of fast neutron reactor to purification of reprocessed fuel from minor actinides

    Science.gov (United States)

    Cherny, V. A.; Kochetkov, L. A.; Nevinitsa, A. I.

    2013-12-01

    The work is devoted to computational investigation of the dependence of basic physical parameters of fast neutron reactors on the degree of purification of plutonium from minor actinides obtained as a result of pyroelectrochemical reprocessing of spent nuclear fuel and used for manufacturing MOX fuel to be reloaded into the reactors mentioned. The investigations have shown that, in order to preserve such important parameters of a BN-800 type reactor as the criticality, the sodium void reactivity effect, the Doppler effect, and the efficiency of safety rods, it is possible to use the reprocessed fuel without separation of minor actinides for refueling (recharging) the core.

  17. Nanostructure evolution of neutron-irradiated reactor pressure vessel steels: Revised Object kinetic Monte Carlo model

    Science.gov (United States)

    Chiapetto, M.; Messina, L.; Becquart, C. S.; Olsson, P.; Malerba, L.

    2017-02-01

    This work presents a revised set of parameters to be used in an Object kinetic Monte Carlo model to simulate the microstructure evolution under neutron irradiation of reactor pressure vessel steels at the operational temperature of light water reactors (∼300 °C). Within a "grey-alloy" approach, a more physical description than in a previous work is used to translate the effect of Mn and Ni solute atoms on the defect cluster diffusivity reduction. The slowing down of self-interstitial clusters, due to the interaction between solutes and crowdions in Fe is now parameterized using binding energies from the latest DFT calculations and the solute concentration in the matrix from atom-probe experiments. The mobility of vacancy clusters in the presence of Mn and Ni solute atoms was also modified on the basis of recent DFT results, thereby removing some previous approximations. The same set of parameters was seen to predict the correct microstructure evolution for two different types of alloys, under very different irradiation conditions: an Fe-C-MnNi model alloy, neutron irradiated at a relatively high flux, and a high-Mn, high-Ni RPV steel from the Swedish Ringhals reactor surveillance program. In both cases, the predicted self-interstitial loop density matches the experimental solute cluster density, further corroborating the surmise that the MnNi-rich nanofeatures form by solute enrichment of immobilized small interstitial loops, which are invisible to the electron microscope.

  18. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  19. Mutational profiling reveals PIK3CA mutations in gallbladder carcinoma

    Directory of Open Access Journals (Sweden)

    Bardeesy Nabeel

    2011-02-01

    Full Text Available Abstract Background The genetics of advanced biliary tract cancers (BTC, which encompass intra- and extra-hepatic cholangiocarcinomas as well as gallbladder carcinomas, are heterogeneous and remain to be fully defined. Methods To better characterize mutations in established known oncogenes and tumor suppressor genes we tested a mass spectrometric based platform to interrogate common cancer associated mutations across a panel of 77 formalin fixed paraffin embedded archived BTC cases. Results Mutations among three genes, KRAS, NRAS and PIK3CA were confirmed in this cohort. Activating mutations in PIK3CA were identified exclusively in GBC (4/32, 12.5%. KRAS mutations were identified in 3 (13% intra-hepatic cholangiocarcinomas and 1 (33% perihillar cholangiocarcinoma but were not identified in gallbladder carcinomas and extra-hepatic cholangiocarcinoma. Conclusions The presence of activating mutations in PIK3CA specifically in GBC has clinical implications in both the diagnosis of this cancer type, as well as the potential utility of targeted therapies such as PI3 kinase inhibitors.

  20. COUNTERCURRENT FLOW LIMITATION EXPERIMENTS AND MODELING FOR IMPROVED REACTOR SAFETY

    Energy Technology Data Exchange (ETDEWEB)

    Vierow, Karen

    2008-09-26

    This project is investigating countercurrent flow and “flooding” phenomena in light water reactor systems to improve reactor safety of current and future reactors. To better understand the occurrence of flooding in the surge line geometry of a PWR, two experimental programs were performed. In the first, a test facility with an acrylic test section provided visual data on flooding for air-water systems in large diameter tubes. This test section also allowed for development of techniques to form an annular liquid film along the inner surface of the “surge line” and other techniques which would be difficult to verify in an opaque test section. Based on experiences in the air-water testing and the improved understanding of flooding phenomena, two series of tests were conducted in a large-diameter, stainless steel test section. Air-water test results and steam-water test results were directly compared to note the effect of condensation. Results indicate that, as for smaller diameter tubes, the flooding phenomena is predominantly driven by the hydrodynamics. Tests with the test sections inclined were attempted but the annular film was easily disrupted. A theoretical model for steam venting from inclined tubes is proposed herein and validated against air-water data. Empirical correlations were proposed for air-water and steam-water data. Methods for developing analytical models of the air-water and steam-water systems are discussed, as is the applicability of the current data to the surge line conditions. This report documents the project results from July 1, 2005 through June 30, 2008.

  1. Parameter estimation for LLDPE gas-phase reactor models

    Directory of Open Access Journals (Sweden)

    G. A. Neumann

    2007-06-01

    Full Text Available Product development and advanced control applications require models with good predictive capability. However, in some cases it is not possible to obtain good quality phenomenological models due to the lack of data or the presence of important unmeasured effects. The use of empirical models requires less investment in modeling, but implies the need for larger amounts of experimental data to generate models with good predictive capability. In this work, nonlinear phenomenological and empirical models were compared with respect to their capability to predict the melt index and polymer yield of a low-density polyethylene production process consisting of two fluidized bed reactors connected in series. To adjust the phenomenological model, the optimization algorithms based on the flexible polyhedron method of Nelder and Mead showed the best efficiency. To adjust the empirical model, the PLS model was more appropriate for polymer yield, and the melt index needed more nonlinearity like the QPLS models. In the comparison between these two types of models better results were obtained for the empirical models.

  2. Physical model of Nernst element

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Hiroaki [Venture Business Lab., Nagoya Univ., Nagoya (Japan); Ikeda, Kazuaki; Yamaguchi, Satarou

    1998-08-01

    Generation of electric power by the Nernst effect is a new application of a semiconductor. A key point of this proposal is to find materials with a high thermomagnetic figure-of-merit, which are called Nernst elements. In order to find candidates of the Nernst element, a physical model to describe its transport phenomena is needed. As the first model, we began with a parabolic two-band model in classical statistics. According to this model, we selected InSb as candidates of the Nernst element and measured their transport coefficients in magnetic fields up to 4 Tesla within a temperature region from 270 K to 330 K. In this region, we calculated transport coefficients numerically by our physical model. For InSb, experimental data are coincident with theoretical values in strong magnetic field. (author)

  3. Developing Fully Coupled Dynamical Reactor Core Isolation System Models in RELAP-7 for Extended Station Black-Out Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang; David Andrs; Richard Martineau

    2014-04-01

    The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup water to the reactor vessel for core cooling when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. It was one of the very few safety systems still available during the Fukushima Daiichi accidents after the tsunamis hit the plants and the system successfully delayed the core meltdown for a few days for unit 2 & 3. Therefore, detailed models for RCIC system components are indispensable to understand extended station black-out accidents (SBO) for BWRs. As part of the effort to develop the new generation reactor system safety analysis code RELAP-7, major components to simulate the RCIC system have been developed. This paper describes the models for those components such as turbine, pump, and wet well. Selected individual component test simulations and a simplified SBO simulation up to but before core damage is presented. The successful implementation of the simplified RCIC and wet well models paves the way to further improve the models for safety analysis by including more detailed physical processes in the near future.

  4. Model based design of biochemical micro-reactors

    Directory of Open Access Journals (Sweden)

    Tobias eElbinger

    2016-02-01

    Full Text Available Mathematical modelling of biochemical pathways is an important resource in Synthetic Biology, as the predictive power of simulating synthetic pathways represents an important step in the design of synthetic metabolons. In this paper, we are concerned with the mathematical modeling, simulation and optimization of metabolic processes in biochemical micro-reactors able to carry out enzymatic reactions and to exchange metabolites with their surrounding medium. The results of the reported modeling approach are incorporated in the design of the first micro-reactor prototypes that are under construction. These microreactors consist of compartments separated by membranes carrying specific transporters for the input of substrates and export of products. Inside the compartments multi-enzyme complexes assembled on nano-beads by peptide adapters are used to carry out metabolic reactions.The spatially resolved mathematical model describing the ongoing processes consists of a system of diffusion equations together with boundary and initial conditions. The boundary conditions model the exchange of metabolites with the neighboring compartments and the reactions at the surface of the nano-beads carrying the multi-enzyme complexes. Efficient and accurate approaches for numerical simulation of the mathematical model and for optimal design of the micro-reactor are developed. As a proof-of-concept scenario, a synthetic pathway for the conversion of sucrose to glucose-6-phosphate (G6P was chosen. In this context, the mathematical model is employed to compute the spatio-temporal distributions of the metabolite concentrations, as well as application relevant quantities like the outflow rate of G6P. These computations are performed for different scenarios, where the number of beads as well as their loading capacity are varied. The computed metabolite distributions show spatial patterns which differ for different experimental arrangements. Furthermore, the total output

  5. An earthquake transient method for pebble-bed reactors and a fuel temperature model for TRISO fueled reactors

    Science.gov (United States)

    Ortensi, Javier

    This investigation is divided into two general topics: (1) a new method for analyzing the safe shutdown earthquake event in a pebble bed reactor core, and (2) the development of an explicit tristructural-isotropic fuel model for high temperature reactors. The safe shutdown earthquake event is one of the design basis accidents for the pebble bed reactor. The new method captures the dynamic geometric compaction of the pebble bed core. The neutronic and thermal-fluids grids are dynamically re-meshed to simulate the re-arrangement of the pebbles in the reactor during the earthquake. Results are shown for the PBMR-400 assuming it is subjected to the Idaho National Laboratory's design basis earthquake. The study concludes that the PBMR-400 can safely withstand the reactivity insertions induced by the slumping of the core and the resulting relative withdrawal of the control rods. This characteristic stems from the large negative Doppler feedback of the fuel. This Doppler feedback mechanism is a major contributor to the passive safety of gas-cooled, graphite-moderated, high-temperature reactors that use fuel based on TRISO particles. The correct prediction of the magnitude and time-dependence of this feedback effect is essential to the conduct of safety analyses for these reactors. An explicit TRISO fuel temperature model named THETRIS has been developed in this work and incorporated in the CYNOD-THERMIX-KONVEK suite of coupled codes. The new model yields similar results to those obtained with more complex methods, requiring multi-TRISO calculations within one control volume. The performance of the code during fast and moderately-slow transients is verified. These analyses show how explicit TRISO models improve the predictions of the fuel temperature, and consequently, of the power escalation. In addition, a brief study of the potential effects on the transient behavior of high-temperature reactors due to the presence of a gap inside the TRISO particles is included

  6. Standard Model of Particle Physics--a health physics perspective.

    Science.gov (United States)

    Bevelacqua, J J

    2010-11-01

    The Standard Model of Particle Physics is reviewed with an emphasis on its relationship to the physics supporting the health physics profession. Concepts important to health physics are emphasized and specific applications are presented. The capability of the Standard Model to provide health physics relevant information is illustrated with application of conservation laws to neutron and muon decay and in the calculation of the neutron mean lifetime.

  7. MODELLING AND CONTROL OF CONTINUOUS STIRRED TANK REACTOR WITH PID CONTROLLER

    Directory of Open Access Journals (Sweden)

    Artur Wodołażski

    2016-09-01

    Full Text Available This paper presents a model of dynamics control for continuous stirred tank reactor (CSTR in methanol synthesis in a three-phase system. The reactor simulation was carried out for steady and transient state. Efficiency ratio to achieve maximum performance of the product per reactor unit volume was calculated. Reactor dynamics simulation in closed loop allowed to received data for tuning PID controller (proportional-integral-derivative. The results of the regulation process allow to receive data for optimum reactor production capacity, along with local hot spots eliminations or temperature runaway.

  8. Hydrodynamical modelling of upflow anaerobic sludge blanket reactors; Modelagem hidrodinamica de reatores anaerobios de escoamento ascendente e manta de lodo (UASB)

    Energy Technology Data Exchange (ETDEWEB)

    Hanisch, Werner Siegfried

    1995-12-31

    The increasing need to treat wastewater consuming a minimum amount of energy is a clear indication of the appropriateness of anaerobic processes. One of them, the upflow anaerobic sludge blanket reactor (UASB), has shown to be a feasible option to treat industrial wastewater and domestic sewage. To improve this treatment system the knowledge if of its hydrodynamic behaviour is fundamental. In this work a mathematical model is proposed to describe physical simulations that were performed in bench scale UASB reactors. The results allow to conclude that the proposed mathematical model is adequate to describe the hydrodynamical behaviour of the above mentioned reactors 27 refs., 78 figs., 12 tabs.

  9. Physics beyond the Standard Model

    CERN Document Server

    Valle, José W F

    1991-01-01

    We discuss some of the signatures associated with extensions of the Standard Model related to the neutrino and electroweak symmetry breaking sectors, with and without supersymmetry. The topics include a basic discussion of the theory of neutrino mass and the corresponding extensions of the Standard Model that incorporate massive neutrinos; an overview of the present observational status of neutrino mass searches, with emphasis on solar neutrinos, as well the as cosmological data on the amplitude of primordial density fluctuations; the implications of neutrino mass in cosmological nucleosynthesis, non-accelerator, as well as in high energy particle collider experiments. Turning to the electroweak breaking sector, we discuss the physics potential for Higgs boson searches at LEP200, including Majoron extensions of the Standard Model, and the physics of invisibly decaying Higgs bosons. We discuss the minimal supersymmetric Standard Model phenomenology, as well as some of the laboratory signatures that would be as...

  10. Wave Generation in Physical Models

    DEFF Research Database (Denmark)

    Andersen, Thomas Lykke; Frigaard, Peter

    The present book describes the most important aspects of wave generation techniques in physical models. Moreover, the book serves as technical documentation for the wave generation software AwaSys 6, cf. Aalborg University (2012). In addition to the two main authors also Tue Hald and Michael...

  11. A physical model for dementia

    Science.gov (United States)

    Sotolongo-Costa, O.; Gaggero-Sager, L. M.; Becker, J. T.; Maestu, F.; Sotolongo-Grau, O.

    2017-04-01

    Aging associated brain decline often result in some kind of dementia. Even when this is a complex brain disorder a physical model can be used in order to describe its general behavior. A probabilistic model for the development of dementia is obtained and fitted to some experimental data obtained from the Alzheimer's Disease Neuroimaging Initiative. It is explained how dementia appears as a consequence of aging and why it is irreversible.

  12. Process Model of A Fusion Fuel Recovery System for a Direct Drive IFE Power Reactor

    Science.gov (United States)

    Natta, Saswathi; Aristova, Maria; Gentile, Charles

    2008-11-01

    A task has been initiated to develop a detailed representative model for the fuel recovery system (FRS) in the prospective direct drive inertial fusion energy (IFE) reactor. As part of the conceptual design phase of the project, a chemical process model is developed in order to observe the interaction of system components. This process model is developed using FEMLAB Multiphysics software with the corresponding chemical engineering module (CEM). Initially, the reactants, system structure, and processes are defined using known chemical species of the target chamber exhaust. Each step within the Fuel recovery system is modeled compartmentally and then merged to form the closed loop fuel recovery system. The output, which includes physical properties and chemical content of the products, is analyzed after each step of the system to determine the most efficient and productive system parameters. This will serve to attenuate possible bottlenecks in the system. This modeling evaluation is instrumental in optimizing and closing the fusion fuel cycle in a direct drive IFE power reactor. The results of the modeling are presented in this paper.

  13. PIK3CA mutations may be discordant between primary and corresponding metastatic disease in Breast Cancer

    DEFF Research Database (Denmark)

    Dupont Jensen, Jeanette; Laenkholm, Anne-Vibeke; Knoop, Ann;

    2011-01-01

    PURPOSE: PIK3CA mutations are frequent in breast cancer and activate the PI3K/Akt pathway. Unexpectedly, PIK3CA mutation appears in general to be associated with better outcome. In a cohort of patients where both primary and metastatic lesions were available the objective was to assess changes...... recurrence than wild type cases (p=0.03). CONCLUSIONS: PIK3CA mutations occur at high frequency in primary and metastatic breast cancer; these may not necessarily confer increased aggressiveness as mutants had a longer time to recurrence. Because PIK3CA status quite frequently changes between primary...

  14. Tritium distribution modeling in a Light Water New Production Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jaeckle, J.W.

    1989-05-01

    The tritium distribution and tritium release pathways in a new light water production reactor were examined. A computer model was developed to track the tritium as it makes its way through the various plant systems and ends up either as a release to the atmosphere, the cooling tower blowdown or to the solid waste system. The model was designed to predict the integrated yearly tritium releases and provide estimated airborne tritium concentrations in various locations within the plant. WNP-1 was used as a representative model for a Light Water New Production Reactor (LWNPR). The Tritium Distribution Model solves for the time dependent tritium concentration in a system of nodes. These nodes are connected to one another via a set of internodal flow paths and to various sources and sinks. For example, plant systems such as the primary system are the nodes, piping and leaks are the internodal flow paths, make-up water is a source, and release to the atmosphere is a sink. The expected water mass of each node; the flow rates between nodes, sources, and sinks; and tritium source rates are provided as input. The code will solve for the time dependent tritium concentration in each node and the amount of tritium ''released'' to the sinks. Preliminary calculations have been performed using WNP-1 plant specific information obtained primarily from the WNP-1 FSAR. Further work is currently in progress to refine the model and provide a more realistic set of input values which will better represent an operating LWNPR. 1 ref., 1 fig., 1 tab.

  15. Resonances in pi-K scattering

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, David J. [Old Dominion University, Norfolk, VA

    2014-06-23

    We have obtained clear signals of resonances in coupled-channel pi K - eta K scattering. Using distillation and a large basis of operators we are able to extract a precise spectrum of energy levels using the variational method. These energies are analysed using inelastic extensions of the Luescher method to obtain scattering amplitudes that clearly describe S, P and D wave resonances, corresponding to the physical K_0^*(1430), the K^*(892) and the K_2^*(1430).

  16. Dispersed plug flow model for upflow anaerobic sludge bed reactors with focus on granular sludge dynamics

    NARCIS (Netherlands)

    Kalyuzhnyi, S.V.; Fedorovich, V.V.; Lens, P.N.L.

    2006-01-01

    A new approach to model upflow anaerobic sludge bed (UASB)-reactors, referred to as a one-dimensional dispersed plug flow model, was developed. This model focusses on the granular sludge dynamics along the reactor height, based on the balance between dispersion, sedimentation and convection using on

  17. Effective lifetime measurements in the $B_{s}^{0} \\rightarrow K^{+}K^{-}$, $B^{0} \\rightarrow K^{+}\\pi^{-}$ and $B_{s}^{0} \\rightarrow \\pi^{+}K^{-}$ decays

    CERN Document Server

    Aaij, R.; Adinolfi, M.; Affolder, A.; Ajaltouni, Z.; Albrecht, J.; Alessio, F.; Alexander, M.; Ali, S.; Alkhazov, G.; Cartelle, P. Alvarez; Alves, A.A.; Amato, S.; Amerio, S.; Amhis, Y.; An, L.; Anderlini, L.; Anderson, J.; Andreassen, R.; Andreotti, M.; Andrews, J.E.; Appleby, R.B.; Gutierrez, O. Aquines; Archilli, F.; Artamonov, A.; Artuso, M.; Aslanides, E.; Auriemma, G.; Baalouch, M.; Bachmann, S.; Back, J.J.; Badalov, A.; Balagura, V.; Baldini, W.; Barlow, R.J.; Barschel, C.; Barsuk, S.; Barter, W.; Batozskaya, V.; Bauer, Th.; Bay, A.; Beddow, J.; Bedeschi, F.; Bediaga, I.; Belogurov, S.; Belous, K.; Belyaev, I.; Ben-Haim, E.; Bencivenni, G.; Benson, S.; Benton, J.; Berezhnoy, A.; Bernet, R.; Bettler, M. -O.; van Beuzekom, M.; Bien, A.; Bifani, S.; Bird, T.; Bizzeti, A.; Bjørnstad, P.M.; Blake, T.; Blanc, F.; Blouw, J.; Blusk, S.; Bocci, V.; Bondar, A.; Bondar, N.; Bonivento, W.; Borghi, S.; Borgia, A.; Borsato, M.; Bowcock, T.J.V.; Bowen, E.; Bozzi, C.; Brambach, T.; Brand, J. van den; Bressieux, J.; Brett, D.; Britsch, M.; Britton, T.; Brook, N.H.; Brown, H.; Bursche, A.; Busetto, G.; Buytaert, J.; Cadeddu, S.; Calabrese, R.; Calvi, M.; Gomez, M. Calvo; Camboni, A.; Campana, P.; Perez, D. Campora; Carbone, A.; Carboni, G.; Cardinale, R.; Cardini, A.; Carranza-Mejia, H.; Carson, L.; Akiba, K. Carvalho; Casse, G.; Cassina, L.; Garcia, L. Castillo; Cattaneo, M.; Cauet, Ch.; Cenci, R.; Charles, M.; Charpentier, Ph.; Cheung, S. -F.; Chiapolini, N.; Chrzaszcz, M.; Ciba, K.; Vidal, X. Cid; Ciezarek, G.; Clarke, P.E.L.; Clemencic, M.; Cliff, H.V.; Closier, J.; Coco, V.; Cogan, J.; Cogneras, E.; Collins, P.; Comerma-Montells, A.; Contu, A.; Cook, A.; Coombes, M.; Coquereau, S.; Corti, G.; Corvo, M.; Counts, I.; Couturier, B.; Cowan, G.A.; Craik, D.C.; Torres, M. Cruz; Cunliffe, S.; Currie, R.; D'Ambrosio, C.; Dalseno, J.; David, P.; David, P.N.Y.; Davis, A.; De Bruyn, K.; De Capua, S.; De Cian, M.; De Miranda, J.M.; De Paula, L.; De Silva, W.; De Simone, P.; Decamp, D.; Deckenhoff, M.; Del Buono, L.; Déléage, N.; Derkach, D.; Deschamps, O.; Dettori, F.; Di Canto, A.; Dijkstra, H.; Donleavy, S.; Dordei, F.; Dorigo, M.; Suárez, A. Dosil; Dossett, D.; Dovbnya, A.; Dupertuis, F.; Durante, P.; Dzhelyadin, R.; Dziurda, A.; Dzyuba, A.; Easo, S.; Egede, U.; Egorychev, V.; Eidelman, S.; Eisenhardt, S.; Eitschberger, U.; Ekelhof, R.; Eklund, L.; Rifai, I. El; Elsasser, Ch.; Esen, S.; Falabella, A.; Färber, C.; Farinelli, C.; Farley, N.; Farry, S.; Fay, RF; Ferguson, D.; Albor, V. Fernandez; Rodrigues, F. Ferreira; Ferro-Luzzi, M.; Filippov, S.; Fiore, M.; Fiorini, M.; Firlej, M.; Fitzpatrick, C.; Fiutowski, T.; Fontana, M.; Fontanelli, F.; Forty, R.; Francisco, O.; Frank, M.; Frei, C.; Frosini, M.; Fu, J.; Furfaro, E.; Torreira, A. Gallas; Galli, D.; Gallorini, S.; Gambetta, S.; Gandelman, M.; Gandini, P.; Gao, Y.; Garofoli, J.; Tico, J. Garra; Garrido, L.; Gaspar, C.; Gauld, R.; Gavardi, L.; Geraci, A.; Gersabeck, E.; Gersabeck, M.; Gershon, T.; Ghez, Ph.; Gianelle, A.; Giani', S.; Gibson, V.; Giubega, L.; Gligorov, V.V.; Göbel, C.; Golubkov, D.; Golutvin, A.; Gomes, A.; Gordon, H.; Gotti, C.; Gándara, M. Grabalosa; Diaz, R. Graciani; Cardoso, L. A. Granado; Graugés, E.; Graziani, G.; Grecu, A.; Greening, E.; Gregson, S.; Griffith, P.; Grillo, L.; Grünberg, O.; Gui, B.; Gushchin, E.; Guz, Yu.; Gys, T.; Hadjivasiliou, C.; Haefeli, G.; Haen, C.; Haines, S.C.; Hall, S.; Hamilton, B.; Hampson, T.; Han, X.; Hansmann-Menzemer, S.; Harnew, N.; Harnew, S.T.; Harrison, J.; Hartmann, T.; He, J.; Head, T.; Heijne, V.; Hennessy, K.; Henrard, P.; Henry, L.; Morata, J. A. Hernando; van Herwijnen, E.; Heß, M.; Hicheur, A.; Hill, D.; Hoballah, M.; Hombach, C.; Hulsbergen, W.; Hunt, P.; Hussain, N.; Hutchcroft, D.; Hynds, D.; Idzik, M.; Ilten, P.; Jacobsson, R.; Jaeger, A.; Jalocha, J.; Jans, E.; Jaton, P.; Jawahery, A.; Jing, F.; John, M.; Johnson, D.; Jones, C.R.; Joram, C.; Jost, B.; Jurik, N.; Kaballo, M.; Kandybei, S.; Kanso, W.; Karacson, M.; Karbach, T.M.; Kelsey, M.; Kenyon, I.R.; Ketel, T.; Khanji, B.; Khurewathanakul, C.; Klaver, S.; Kochebina, O.; Kolpin, M.; Komarov, I.; Koopman, R.F.; Koppenburg, P.; Korolev, M.; Kozlinskiy, A.; Kravchuk, L.; Kreplin, K.; Kreps, M.; Krocker, G.; Krokovny, P.; Kruse, F.; Kucharczyk, M.; Kudryavtsev, V.; Kurek, K.; Kvaratskheliya, T.; La Thi, V.N.; Lacarrere, D.; Lafferty, G.; Lai, A.; Lambert, D.; Lambert, R.W.; Lanciotti, E.; Lanfranchi, G.; Langenbruch, C.; Langhans, B.; Latham, T.; Lazzeroni, C.; Gac, R. Le; van Leerdam, J.; Lees, J. -P.; Lefèvre, R.; Leflat, A.; Lefrançois, J.; Leo, S.; Leroy, O.; Lesiak, T.; Leverington, B.; Li, Y.; Liles, M.; Lindner, R.; Linn, C.; Lionetto, F.; Liu, B.; Liu, G.; Lohn, S.; Longstaff, I.; Lopes, J.H.; Lopez-March, N.; Lowdon, P.; Lu, H.; Lucchesi, D.; Luo, H.; Lupato, A.; Luppi, E.; Lupton, O.; Machefert, F.; Machikhiliyan, I.V.; Maciuc, F.; Maev, O.; Malde, S.; Manca, G.; Mancinelli, G.; Mapelli, A.; Maratas, J.; Marchand, J.F.; Marconi, U.; Benito, C. Marin; Marino, P.; Märki, R.; Marks, J.; Martellotti, G.; Martens, A.; Sánchez, A. Martín; Martinelli, M.; Santos, D. Martinez; Vidal, F. Martinez; Tostes, D. Martins; Massafferri, A.; Matev, R.; Mathe, Z.; Matteuzzi, C.; Mazurov, A.; McCann, M.; McCarthy, J.; McNab, A.; McNulty, R.; McSkelly, B.; Meadows, B.; Meier, F.; Meissner, M.; Merk, M.; Milanes, D.A.; Minard, M. -N.; Moggi, N.; Rodriguez, J. Molina; Monteil, S.; Moran, D.; Morandin, M.; Morawski, P.; Mordà, A.; Morello, M.J.; Moron, J.; Mountain, R.; Muheim, F.; Müller, K.; Muresan, R.; Mussini, M.; Muster, B.; Naik, P.; Nakada, T.; Nandakumar, R.; Nasteva, I.; Needham, M.; Neri, N.; Neubert, S.; Neufeld, N.; Neuner, M.; Nguyen, A.D.; Nguyen, T.D.; Nguyen-Mau, C.; Nicol, M.; Niess, V.; Niet, R.; Nikitin, N.; Nikodem, T.; Novoselov, A.; Oblakowska-Mucha, A.; Obraztsov, V.; Oggero, S.; Ogilvy, S.; Okhrimenko, O.; Oldeman, R.; Onderwater, G.; Orlandea, M.; Goicochea, J. M. Otalora; Owen, P.; Oyanguren, A.; Pal, B.K.; Palano, A.; Palombo, F.; Palutan, M.; Panman, J.; Papanestis, A.; Pappagallo, M.; Parkes, C.; Parkinson, C.J.; Passaleva, G.; Patel, G.D.; Patel, M.; Patrignani, C.; Alvarez, A. Pazos; Pearce, A.; Pellegrino, A.; Altarelli, M. Pepe; Perazzini, S.; Trigo, E. Perez; Perret, P.; Perrin-Terrin, M.; Pescatore, L.; Pesen, E.; Petridis, K.; Petrolini, A.; Olloqui, E. Picatoste; Pietrzyk, B.; Pilař, T.; Pinci, D.; Pistone, A.; Playfer, S.; Casasus, M. Plo; Polci, F.; Poluektov, A.; Polycarpo, E.; Popov, A.; Popov, D.; Popovici, B.; Potterat, C.; Powell, A.; Prisciandaro, J.; Pritchard, A.; Prouve, C.; Pugatch, V.; Navarro, A. Puig; Punzi, G.; Qian, W.; Rachwal, B.; Rademacker, J.H.; Rakotomiaramanana, B.; Rama, M.; Rangel, M.S.; Raniuk, I.; Rauschmayr, N.; Raven, G.; Reichert, S.; Reid, M.M.; Reis, A. C. dos; Ricciardi, S.; Richards, A.; Rihl, M.; Rinnert, K.; Molina, V. Rives; Romero, D. A. Roa; Robbe, P.; Rodrigues, A.B.; Rodrigues, E.; Perez, P. Rodriguez; Roiser, S.; Romanovsky, V.; Vidal, A. Romero; Rotondo, M.; Rouvinet, J.; Ruf, T.; Ruffini, F.; Ruiz, H.; Valls, P. Ruiz; Sabatino, G.; Silva, J. J. Saborido; Sagidova, N.; Sail, P.; Saitta, B.; Guimaraes, V. Salustino; Mayordomo, C. Sanchez; Sedes, B. Sanmartin; Santacesaria, R.; Rios, C. Santamarina; Santovetti, E.; Sapunov, M.; Sarti, A.; Satriano, C.; Satta, A.; Savrie, M.; Savrina, D.; Schiller, M.; Schindler, H.; Schlupp, M.; Schmelling, M.; Schmidt, B.; Schneider, O.; Schopper, A.; Schune, M. -H.; Schwemmer, R.; Sciascia, B.; Sciubba, A.; Seco, M.; Semennikov, A.; Senderowska, K.; Sepp, I.; Serra, N.; Serrano, J.; Sestini, L.; Seyfert, P.; Shapkin, M.; Shapoval, I.; Shcheglov, Y.; Shears, T.; Shekhtman, L.; Shevchenko, V.; Shires, A.; Coutinho, R. Silva; Simi, G.; Sirendi, M.; Skidmore, N.; Skwarnicki, T.; Smith, N.A.; Smith, E.; Smith, J.; Smith, M.; Snoek, H.; Sokoloff, M.D.; Soler, F.J.P.; Soomro, F.; Souza, D.; De Paula, B. Souza; Spaan, B.; Sparkes, A.; Spradlin, P.; Stagni, F.; Stahl, S.; Steinkamp, O.; Stenyakin, O.; Stevenson, S.; Stoica, S.; Stone, S.; Storaci, B.; Stracka, S.; Straticiuc, M.; Straumann, U.; Stroili, R.; Subbiah, V.K.; Sun, L.; Sutcliffe, W.; Swientek, K.; Swientek, S.; Syropoulos, V.; Szczekowski, M.; Szczypka, P.; Szilard, D.; Szumlak, T.; T'Jampens, S.; Teklishyn, M.; Tellarini, G.; Teubert, F.; Thomas, C.; Thomas, E.; van Tilburg, J.; Tisserand, V.; Tobin, M.; Tolk, S.; Tomassetti, L.; Tonelli, D.; Topp-Joergensen, S.; Torr, N.; Tournefier, E.; Tourneur, S.; Tran, M.T.; Tresch, M.; Tsaregorodtsev, A.; Tsopelas, P.; Tuning, N.; Garcia, M. Ubeda; Ukleja, A.; Ustyuzhanin, A.; Uwer, U.; Vagnoni, V.; Valenti, G.; Vallier, A.; Gomez, R. Vazquez; Regueiro, P. Vazquez; Sierra, C. Vázquez; Vecchi, S.; Velthuis, J.J.; Veltri, M.; Veneziano, G.; Vesterinen, M.; Viaud, B.; Vieira, D.; Diaz, M. Vieites; Vilasis-Cardona, X.; Vollhardt, A.; Volyanskyy, D.; Voong, D.; Vorobyev, A.; Vorobyev, V.; Voß, C.; Voss, H.; de Vries, J.A.; Waldi, R.; Wallace, C.; Wallace, R.; Walsh, J.; Wandernoth, S.; Wang, J.; Ward, D.R.; Watson, N.K.; Websdale, D.; Whitehead, M.; Wicht, J.; Wiedner, D.; Wilkinson, G.; Williams, M.P.; Williams, M.; Wilson, F.F.; Wimberley, J.; Wishahi, J.; Wislicki, W.; Witek, M.; Wormser, G.; Wotton, S.A.; Wright, S.; Wu, S.; Wyllie, K.; Xie, Y.; Xing, Z.; Xu, Z.; Yang, Z.; Yuan, X.; Yushchenko, O.; Zangoli, M.; Zavertyaev, M.; Zhang, F.; Zhang, L.; Zhang, W.C.; Zhang, Y.; Zhelezov, A.; Zhokhov, A.; Zhong, L.; Zvyagin, A.

    2014-01-01

    Measurements of the effective lifetimes in the $B_{s}^{0} \\rightarrow K^{+}K^{-}$, $B^{0} \\rightarrow K^{+}\\pi^{-}$ and $B_{s}^{0} \\rightarrow \\pi^{+}K^{-}$ decays are presented using $1.0~\\mathrm{fb^{-1}}$ of $pp$ collision data collected at a centre-of-mass energy of 7 TeV by the LHCb experiment. The analysis uses a data-driven approach to correct for the decay time acceptance. The measured effective lifetimes are $\\tau_{B_{s}^{0} \\rightarrow K^{+}K^{-}}$ = $1.407~\\pm~0.016~\\pm~0.007~\\mathrm{ps}$, $\\tau_{B^{0} \\rightarrow K^{+}\\pi^{-}}$ = $1.524~\\pm~0.011~\\pm~0.004~\\mathrm{ps}$, $\\tau_{B_{s}^{0} \\rightarrow \\pi^{+}K^{-}}$ = $1.60~\\pm~0.06~\\pm~0.01~\\mathrm{ps}$. This is the most precise determination to date of the effective lifetime in the $B_{s}^{0} \\rightarrow K^{+}K^{-}$ decay and provides constraints on contributions from physics beyond the Standard Model to the $B_{s}^{0}$ mixing phase and the width difference $\\Delta\\Gamma_{s}$.

  18. Effective lifetime measurements in the $B_{s}^{0} \\rightarrow K^{+}K^{-}$, $B^{0} \\rightarrow K^{+}\\pi^{-}$ and $B_{s}^{0} \\rightarrow \\pi^{+}K^{-}$ decays

    CERN Document Server

    INSPIRE-00258707; Adeva, B.; Adinolfi, M.; Affolder, A.; Ajaltouni, Z.; Albrecht, J.; Alessio, F.; Alexander, M.; Ali, S.; Alkhazov, G.; Alvarez Cartelle, P.; Alves Jr, A.A.; Amato, S.; Amerio, S.; Amhis, Y.; An, L.; Anderlini, L.; Anderson, J.; Andreassen, R.; Andreotti, M.; Andrews, J.E.; Appleby, R.B.; Aquines Gutierrez, O.; Archilli, F.; Artamonov, A.; Artuso, M.; Aslanides, E.; Auriemma, G.; Baalouch, M.; Bachmann, S.; Back, J.J.; Badalov, A.; Balagura, V.; Baldini, W.; Barlow, R.J.; Barschel, C.; Barsuk, S.; Barter, W.; Batozskaya, V.; Bauer, Th.; Bay, A.; Beddow, J.; Bedeschi, F.; Bediaga, I.; Belogurov, S.; Belous, K.; Belyaev, I.; Ben-Haim, E.; Bencivenni, G.; Benson, S.; Benton, J.; Berezhnoy, A.; Bernet, R.; Bettler, M.O.; van Beuzekom, M.; Bien, A.; Bifani, S.; Bird, T.; Bizzeti, A.; Bjornstad, P.M.; Blake, T.; Blanc, F.; Blouw, J.; Blusk, S.; Bocci, V.; Bondar, A.; Bondar, N.; Bonivento, W.; Borghi, S.; Borgia, A.; Borsato, M.; Bowcock, T.J.V.; Bowen, E.; Bozzi, C.; Brambach, T.; van den Brand, J.; Bressieux, J.; Brett, D.; Britsch, M.; Britton, T.; Brook, N.H.; Brown, H.; Bursche, A.; Busetto, G.; Buytaert, J.; Cadeddu, S.; Calabrese, R.; Calvi, M.; Calvo Gomez, M.; Camboni, A.; Campana, P.; Campora Perez, D.; Carbone, A.; Carboni, G.; Cardinale, R.; Cardini, A.; Carranza-Mejia, H.; Carson, L.; Carvalho Akiba, K.; Casse, G.; Cassina, L.; Garcia, L.Castillo; Cattaneo, M.; Cauet, Ch.; Cenci, R.; Charles, M.; Charpentier, Ph.; Cheung, S.F.; Chiapolini, N.; Chrzaszcz, M.; Ciba, K.; Cid Vidal, X.; Ciezarek, G.; Clarke, P.E.L.; Clemencic, M.; Cliff, H.V.; Closier, J.; Coco, V.; Cogan, J.; Cogneras, E.; Collins, P.; Comerma-Montells, A.; Contu, A.; Cook, A.; Coombes, M.; Coquereau, S.; Corti, G.; Corvo, M.; Counts, I.; Couturier, B.; Cowan, G.A.; Craik, D.C.; Cruz Torres, M.; Cunliffe, S.; Currie, R.; D'Ambrosio, C.; Dalseno, J.; David, P.; David, P.N.Y.; Davis, A.; De Bruyn, K.; De Capua, S.; De Cian, M.; de Miranda, J.M.; De Paula, L.; De Silva, W.; De Simone, P.; Decamp, D.; Deckenhoff, M.; Del Buono, L.; Deleage, N.; Derkach, D.; Deschamps, O.; Dettori, F.; Di Canto, A.; Dijkstra, H.; Donleavy, S.; Dordei, F.; Dorigo, M.; Dosil Suarez, A.; Dossett, D.; Dovbnya, A.; Dupertuis, F.; Durante, P.; Dzhelyadin, R.; Dziurda, A.; Dzyuba, A.; Easo, S.; Egede, U.; Egorychev, V.; Eidelman, S.; Eisenhardt, S.; Eitschberger, U.; Ekelhof, R.; Eklund, L.; El Rifai, I.; Elsasser, Ch.; Esen, S.; Evans, T.; Falabella, A.; Farber, C.; Farinelli, C.; Farley, N.; Farry, S.; Ferguson, D.; Fernandez Albor, V.; Ferreira Rodrigues, F.; Ferro-Luzzi, M.; Filippov, S.; Fiore, M.; Fiorini, M.; Firlej, M.; Fitzpatrick, C.; Fiutowski, T.; Fontana, M.; Fontanelli, F.; Forty, R.; Francisco, O.; Frank, M.; Frei, C.; Frosini, M.; Fu, J.; Furfaro, E.; Gallas Torreira, A.; Galli, D.; Gallorini, S.; Gambetta, S.; Gandelman, M.; Gandini, P.; Gao, Y.; Garofoli, J.; Garra Tico, J.; Garrido, L.; Gaspar, C.; Gauld, R.; Gavardi, L.; Gersabeck, E.; Gersabeck, M.; Gershon, T.; Ghez, Ph.; Gianelle, A.; Giani', S.; Gibson, V.; Giubega, L.; Gligorov, V.V.; Gobel, C.; Golubkov, D.; Golutvin, A.; Gomes, A.; Gordon, H.; Gotti, C.; Grabalosa Gandara, M.; Graciani Diaz, R.; Granado Cardoso, L.A.; Grauges, E.; Graziani, G.; Grecu, A.; Greening, E.; Gregson, S.; Griffith, P.; Grillo, L.; Grunberg, O.; Gui, B.; Gushchin, E.; Guz, Yu.; Gys, T.; Hadjivasiliou, C.; Haefeli, G.; Haen, C.; Haines, S.C.; Hall, S.; Hamilton, B.; Hampson, T.; Han, X.; Hansmann-Menzemer, S.; Harnew, N.; Harnew, S.T.; Harrison, J.; Hartmann, T.; He, J.; Head, T.; Heijne, V.; Hennessy, K.; Henrard, P.; Henry, L.; Hernando Morata, J.A.; van Herwijnen, E.; Hess, M.; Hicheur, A.; Hill, D.; Hoballah, M.; Hombach, C.; Hulsbergen, W.; Hunt, P.; Hussain, N.; Hutchcroft, D.; Hynds, D.; Idzik, M.; Ilten, P.; Jacobsson, R.; Jaeger, A.; Jalocha, J.; Jans, E.; Jaton, P.; Jawahery, A.; Jezabek, M.; Jing, F.; John, M.; Johnson, D.; Jones, C.R.; Joram, C.; Jost, B.; Jurik, N.; Kaballo, M.; Kandybei, S.; Kanso, W.; Karacson, M.; Karbach, T.M.; Kelsey, M.; Kenyon, I.R.; Ketel, T.; Khanji, B.; Khurewathanakul, C.; Klaver, S.; Kochebina, O.; Kolpin, M.; Komarov, I.; Koopman, R.F.; Koppenburg, P.; Korolev, M.; Kozlinskiy, A.; Kravchuk, L.; Kreplin, K.; Kreps, M.; Krocker, G.; Krokovny, P.; Kruse, F.; Kucharczyk, M.; Kudryavtsev, V.; Kurek, K.; Kvaratskheliya, T.; La Thi, V.N.; Lacarrere, D.; Lafferty, G.; Lai, A.; Lambert, D.; Lambert, R.W.; Lanciotti, E.; Lanfranchi, G.; Langenbruch, C.; Langhans, B.; Latham, T.; Lazzeroni, C.; Le Gac, R.; van Leerdam, J.; Lees, J.P.; Lefevre, R.; Leflat, A.; Lefrancois, J.; Leo, S.; Leroy, O.; Lesiak, T.; Leverington, B.; Li, Y.; Liles, M.; Lindner, R.; Linn, C.; Lionetto, F.; Liu, B.; Liu, G.; Lohn, S.; Longstaff, I.; Lopes, J.H.; Lopez-March, N.; Lowdon, P.; Lu, H.

    2014-01-01

    Measurements of the effective lifetimes in the $B_{s}^{0} \\rightarrow K^{+}K^{-}$, $B^{0} \\rightarrow K^{+}\\pi^{-}$ and $B_{s}^{0} \\rightarrow \\pi^{+}K^{-}$ decays are presented using $1.0~\\mathrm{fb^{-1}}$ of $pp$ collision data collected at a centre-of-mass energy of 7 TeV by the LHCb experiment. The analysis uses a data-driven approach to correct for the decay time acceptance. The measured effective lifetimes are $\\tau_{B_{s}^{0} \\rightarrow K^{+}K^{-}}$ = $1.407~\\pm~0.016~\\pm~0.007~\\mathrm{ps}$, $\\tau_{B^{0} \\rightarrow K^{+}\\pi^{-}}$ = $1.524~\\pm~0.011~\\pm~0.004~\\mathrm{ps}$, $\\tau_{B_{s}^{0} \\rightarrow \\pi^{+}K^{-}}$ = $1.60~\\pm~0.06~\\pm~0.01~\\mathrm{ps}$. This is the most precise determination to date of the effective lifetime in the $B_{s}^{0} \\rightarrow K^{+}K^{-}$ decay and provides constraints on contributions from physics beyond the Standard Model to the $B_{s}^{0}$ mixing phase and the width difference $\\Delta\\Gamma_{s}$.

  19. Heterogeneous Nuclear Reactor Models for Optimal Xenon Control.

    Science.gov (United States)

    Gondal, Ishtiaq Ahmad

    Nuclear reactors are generally modeled as homogeneous mixtures of fuel, control, and other materials while in reality they are heterogeneous-homogeneous configurations comprised of fuel and control rods along with other materials. Similarly, for space-time studies of a nuclear reactor, homogeneous, usually one-group diffusion theory, models are used, and the system equations are solved by either nodal or modal expansion approximations. Study of xenon-induced problems has also been carried out using similar models and with the help of dynamic programming or classical calculus of variations or the minimum principle. In this study a thermal nuclear reactor is modeled as a two-dimensional lattice of fuel and control rods placed in an infinite-moderator in plane geometry. The two-group diffusion theory approximation is used for neutron transport. Space -time neutron balance equations are written for two groups and reduced to one space-time algebraic equation by using the two-dimensional Fourier transform. This equation is written at all fuel and control rod locations. Iodine -xenon and promethium-samarium dynamic equations are also written at fuel rod locations only. These equations are then linearized about an equilibrium point which is determined from the steady-state form of the original nonlinear system equations. After studying poisonless criticality, with and without control, and the stability of the open-loop system and after checking its controllability, a performance criterion is defined for the xenon-induced spatial flux oscillation problem in the form of a functional to be minimized. Linear -quadratic optimal control theory is then applied to solve the problem. To perform a variety of different additional useful studies, this formulation has potential for various extensions and variations; for example, different geometry of the problem, with possible extension to three dimensions, heterogeneous -homogeneous formulation to include, for example, homogeneously

  20. Prestressed concrete reactor vessel thermal cylinder model study

    Energy Technology Data Exchange (ETDEWEB)

    Callahan, J.P.; Canonico, D.A.; Richardson, M.; Corum, J.M.; Dodge, W.G.; Robinson, G.C.; Whitman, G.D.

    1977-05-04

    The thermal cylinder experiment was designed both to provide information for evaluating the capability of analytical methods to predict the time-dependent stress-strain behavior of a /sup 1///sub 6/-scale model of the barrel section of a single-cavity prestressed concrete reactor vessel and to demonstrate the structural behavior under design and off-design thermal conditions. The model was a thick-walled cylinder having a height of 1.22 m, a thickness of 0.46 m, and an outer diameter of 2.06 m. It was prestressed both axially and circumferentially and subjected to 4.83 MPa internal pressure together with a thermal crossfall imposed by heating the inner surface to 338.8 K and cooling the outer surface to 297.1 K. The initial 460 days of testing were divided into time periods that simulated prestressing, heatup, reactor operation, and shutdown. At the conclusion of the simulated operating period, the model was repressurized and subjected to localized heating at 505.4 K for 84 days to produce an off-design hot-spot condition. Comparisons of experimental data with calculated values obtained using the SAFE-CRACK finite-element computer program showed that the program was capable of predicting time-dependent behavior in a vessel subjected to normal operating conditions, but that it was unable to accurately predict the behavior during off-design hot-spot heating. Readings made using a neutron and gamma-ray backscattering moisture probe showed little, if any, migration of moisture in the concrete cross section. Destructive examination indicated that the model maintained its basic structural integrity during localized hot-spot heating.

  1. Reactor physics and safety aspects of various design options of a Russian light water reactor with rock-like fuels

    Science.gov (United States)

    Bondarenko, A. V.; Komissarov, O. V.; Kozmenkov, Ya. K.; Matveev, Yu. V.; Orekhov, Yu. I.; Pivovarov, V. A.; Sharapov, V. N.

    2003-06-01

    This paper presents results of analytical studies on weapons grade plutonium incineration in VVER (640) medium size light water reactors using a special composition of rock-like fuel (ROX-fuel) to assure spent fuel long-term storage without its reprocessing. The main goal is to achieve high degree of plutonium incineration in once-through cycle. In this paper we considered two fuel compositions. In both compositions weapons grade plutonium is used as fissile material. Spinel (MgAl 2O 4) is used as the 'preserving' material assuring safe storage of the spent fuel. Besides an inert matrix, the option of rock-like fuel with thorium dioxide was studied. One of principal problems in the realization of the proposed approach is the substantial change of properties of the light water reactor core when passing to the use of the ROX-fuel, in particular: (i) due to the absence of 238U the Doppler effect playing a crucial role in reactor's self-regulation and limiting the consequences of reactivity accidents, decreases significantly, (ii) no fuel breeding on one hand, and the quest to attain the maximum plutonium burnup on the other hand, would result in a drastical change of the fuel assembly power during the lifetime and, as a consequence, the rise in irregularity of the power density of fuel assemblies, (iii) both the control rods worth and dissolved boron worth decrease in view of neutron spectrum hardening brought on by the larger absorption cross-section of plutonium as compared to uranium, (iv) βeff is markedly reduced. All these distinctive features are potentially detrimental to the reactor nuclear safety. The principal objective of this work is that to identify a variant of the fuel composition and the reactor layout, which would permit neutralize the negative effect of the above-mentioned distinctive features.

  2. Modelling solid-convective flash pyrolysis of straw and wood in the Pyrolysis Centrifuge Reactor

    DEFF Research Database (Denmark)

    Bech, Niels; Larsen, Morten Boberg; Jensen, Peter Arendt

    2009-01-01

    Less than a handful of solid-convective pyrolysis reactors for the production of liquid fuel from biomass have been presented and for only a single reactor a detailed mathematical model has been presented. In this article we present a predictive mathematical model of the pyrolysis process...... in the Pyrolysis Centrifuge Reactor, a novel solid-convective flash pyrolysis reactor. The model relies on the original concept for ablative pyrolysis of particles being pyrolysed through the formation of an intermediate liquid compound which is further degraded to form liquid organics, char, and gas. To describe...... that the reacting particle continuously shed the formed char layer....

  3. Parameter estimation and accuracy matching strategies for 2-D reactor models

    Science.gov (United States)

    Nowak, U.; Grah, A.; Schreier, M.

    2005-11-01

    The mathematical modelling of a special modular catalytic reactor kit leads to a system of partial differential equation in two space dimensions. As customary, this model contains uncertain physical parameters, which may be adapted to fit experimental data. To solve this nonlinear least-squares problem we apply a damped Gauss-Newton method. A method of lines approach is used to evaluate the associated model equations. By an a priori spatial discretization, a large DAE system is derived and integrated with an adaptive, linearly implicit extrapolation method. For sensitivity evaluation we apply an internal numerical differentiation technique, which reuses linear algebra information from the model integration. In order not to interfere with the control of the Gauss-Newton iteration these computations are done usually very accurately and, therefore, with substantial cost. To overcome this difficulty, we discuss several accuracy adaptation strategies, e.g., a master-slave mode. Finally, we present some numerical experiments.

  4. Mathematical modeling of methyl ester concentration distribution in a continuous membrane tubular reactor and comparison with conventional tubular reactor

    Science.gov (United States)

    Talaghat, M. R.; Jokar, S. M.; Modarres, E.

    2017-04-01

    The reduction of fossil fuel resources and environmental issues made researchers find alternative fuels include biodiesels. One of the most widely used methods for production of biodiesel on a commercial scale is transesterification method. In this work, the biodiesel production by a transesterification method was modeled. Sodium hydroxide was considered as a catalyst to produce biodiesel from canola oil and methanol in a continuous tubular ceramic membranes reactor. As the Biodiesel production reaction from triglycerides is an equilibrium reaction, the reaction rate constants depend on temperature and related linearly to catalyst concentration. By using the mass balance for a membrane tubular reactor and considering the variation of raw materials and products concentration with time, the set of governing equations were solved by numerical methods. The results clearly show the superiority of membrane reactor than conventional tubular reactors. Afterward, the influences of molar ratio of alcohol to oil, weight percentage of the catalyst, and residence time on the performance of biodiesel production reactor were investigated.

  5. Physical and mechanical characteristics and chemical compatibility of aluminum nitride insulator coatings for fusion reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Rink, D.L. [Argonne National Lab., IL (United States). Energy Technology Div.

    1996-04-01

    The blanket system is one of the most important components in a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The Blanket Comparison and Selection Study, conducted earlier, described the overall comparative performance of various concepts, including liquid metal, molten salt, water, and helium. Based on the requirements for an electrically insulating coating on the first-wall structural material to minimize the MHD pressure drop during the flow of liquid metal in a magnetic field, AlN was selected as a candidate coating material for the Li self-cooled blanket concept. This report discusses the results from an ongoing study of physical and mechanical characteristics and chemical compatibility of AlN electrical insulator coatings in a liquid Li environment. Details are presented on the AlN coating fabrication methods, and experimental data are reported for microstructures, chemistry of coatings, pretreatment of substrate, heat treatment of coatings, hardness data for coatings, coating/lithium interactions, and electrical resistance before and after exposure to lithium. Thermodynamic calculations are presented to establish regions of stability for AlN coatings in an Li environment as a function of O concentration and temperature, which can aid in-situ development of AlN coatings in Li.

  6. Identification of the reduced order models of a BWR reactor; Identificacion de modelos de orden reducido de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez S, A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: augusto@correo.unam.mx

    2004-07-01

    The present work has as objective to analyze the relative stability of a BWR type reactor. It is analyzed that so adaptive it turns out to identify the parameters of a model of reduced order so that this it reproduces a condition of given uncertainty. This will take of a real fact happened in the La Salle plant under certain operation conditions of power and flow of coolant. The parametric identification is carried out by means of an algorithm of recursive least square and an Output Error model (Output Error), measuring the output power of the reactor when the instability is present, and considering that it is produced by a change in the reactivity of the system in the same way that a sign of type step. Also it is carried out an analytic comparison of the relative stability, analyzing two types of answers: the original answer of the uncertainty of the reactor vs. the obtained response identifying the parameters of the model of reduced order, reaching the conclusion that it is very viable to adapt a model of reduced order to study the stability of a reactor, under the only condition to consider that the dynamics of the reactivity is of step type. (Author)

  7. Physical models of cell motility

    CERN Document Server

    2016-01-01

    This book surveys the most recent advances in physics-inspired cell movement models. This synergetic, cross-disciplinary effort to increase the fidelity of computational algorithms will lead to a better understanding of the complex biomechanics of cell movement, and stimulate progress in research on related active matter systems, from suspensions of bacteria and synthetic swimmers to cell tissues and cytoskeleton.Cell motility and collective motion are among the most important themes in biology and statistical physics of out-of-equilibrium systems, and crucial for morphogenesis, wound healing, and immune response in eukaryotic organisms. It is also relevant for the development of effective treatment strategies for diseases such as cancer, and for the design of bioactive surfaces for cell sorting and manipulation. Substrate-based cell motility is, however, a very complex process as regulatory pathways and physical force generation mechanisms are intertwined. To understand the interplay between adhesion, force ...

  8. Dietary restriction-resistant human tumors harboring the PIK3CA-activating mutation H1047R are sensitive to metformin

    Science.gov (United States)

    Cufí, Sílvia; Corominas-Faja, Bruna; Lopez-Bonet, Eugeni; Bonavia, Rosa; Pernas, Sonia; López, Isabel álvarez; Dorca, Joan; Martínez, Susana; López, Norberto Batista; Fernández, Severina Domínguez; Cuyàs, Elisabet; Visa, Joana; Rodríguez-Gallego, Esther; Quirantes-Piné, Rosa; Segura-Carretero, Antonio; Joven, Jorge; Martin-Castillo, Begoña; Menendez, Javier A.

    2013-01-01

    Cancer cells expressing constitutively active phosphatidylinositol-3 kinase (PI3K) are proliferative regardless of the absence of insulin, and they form dietary restriction (DR)-resistant tumors in vivo. Because the binding of insulin to its receptors activates the PI3K/AKT/mammalian target of rapamycin (mTOR) signaling cascade, activating mutations in the PIK3CA oncogene may determine tumor response to DR-like pharmacological strategies targeting the insulin and mTOR pathways. The anti-diabetic drug metformin is a stereotypical DR mimetic that exerts its anti-cancer activity through a dual mechanism involving insulin-related (systemic) and mTOR-related (cell-autonomous) effects. However, it remains unclear whether PIK3CA-activating mutations might preclude the anti-cancer activity of metformin in vivo. To model the oncogenic PIK3CA-driven early stages of cancer, we used the clonal breast cancer cell line MCF10DCIS.com, which harbors the gain-of-function H1047R hot-spot mutation in the catalytic domain of the PI3KCA gene and has been shown to form DR-refractory xenotumors. To model PIK3CA-activating mutations in late stages of cancer, we took advantage of the isogenic conversion of a PIK3CA-wild-type tumor into a PIK3CA H1047R-mutated tumor using the highly metastatic colorectal cancer cell line SW48. MCF10DCIS.com xenotumors, although only modestly affected by treatment with oral metformin (approximately 40% tumor growth inhibition), were highly sensitive to the intraperitoneal (i.p.) administration of metformin, the anti-cancer activity of which increased in a time-dependent manner and reached >80% tumor growth inhibition by the end of the treatment. Metformin treatment via the i.p. route significantly reduced the proliferation factor mitotic activity index (MAI) and decreased tumor cellularity in MCF10DCIS.com cancer tissues. Whereas SW48-wild-type (PIK3CA+/+) cells rapidly formed metformin-refractory xenotumors in mice, ad libitum access to water containing

  9. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  10. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    Energy Technology Data Exchange (ETDEWEB)

    Hamann, S., E-mail: hamann@inp-greifswald.de; Röpcke, J. [INP-Greifswald, Felix-Hausdorff-Str. 2, 17489 Greifswald (Germany); Börner, K.; Burlacov, I.; Spies, H.-J. [TU Bergakademie Freiberg, Institute of Materials Engineering, Gustav-Zeuner-Str. 5, 09599 Freiberg (Germany); Strämke, M.; Strämke, S. [ELTRO GmbH, Arnold-Sommerfeld-Ring 3, 52499 Baesweiler (Germany)

    2015-12-15

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH{sub 4}, C{sub 2}H{sub 2}, HCN, and NH{sub 3}). With the help of OES, the rotational temperature of the screen plasma could be determined.

  11. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    Science.gov (United States)

    Hamann, S.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.; Röpcke, J.

    2015-12-01

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH4, C2H2, HCN, and NH3). With the help of OES, the rotational temperature of the screen plasma could be determined.

  12. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen.

    Science.gov (United States)

    Hamann, S; Börner, K; Burlacov, I; Spies, H-J; Strämke, M; Strämke, S; Röpcke, J

    2015-12-01

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH4, C2H2, HCN, and NH3). With the help of OES, the rotational temperature of the screen plasma could be determined.

  13. A physical model of sprinting.

    Science.gov (United States)

    Gaudet, S

    2014-09-22

    A new physical model of all-out sprinting is presented. The first models for the applied forces in the block, drive and maintenance phases, as well as for braking forces, are proposed and are based on experimental observations. The applied forces and the aerodynamic drag forces along with the speed and position of the sprinter are calculated by the model as functions of time. The model's unknown parameters are physically relevant and are quantitatively comparable to quantities measured experimentally. A novel mathematical method, not based on curve fitting, is proposed along with the model which requires two observable quantities, time of first step and start of maintenance phase, and four time splits. The model was validated by modeling several elite sprints from available split data, as well as measured splits for non-elite sprinters, over 100 m and 200 m distances. Excellent agreement between the split times and the simulated times was obtained and the model was shown to accurately predict 100 m times from 60 m splits for non-elite runners and 200 m times from 100 m splits for elite sprinters. The model was also applied to the study of wind and altitude effects for elite sprinters in 100 and 200 m sprints. The model presented in this paper may also be useful as a coaching tool for non-elite sprinters by enabling comparisons with elite sprinters, the identification of weaknesses (comparing phases, braking coefficient) and by allowing predictions of 100 m times based on 60 m (indoor) performances and 200 m times based on 100 m splits.

  14. Reactor physics analysis for the design of nuclear fuel lattices with burnable poisons

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, G. [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, D.F. (Mexico); Guzman, Juan R., E-mail: maestro_juan_rafael@hotmail.com [Departamento de Fisica y Matematicas, Instituto Politecnico Nacional, Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Mexico, D.F. (Mexico)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer A fuel rod optimization for the coupled bundle-core design in a BWR is developed. Black-Right-Pointing-Pointer An algorithm to minimize the rod power peaking factor is used. Black-Right-Pointing-Pointer The fissile content is divided in two factors. Black-Right-Pointing-Pointer A reactor physics analysis of these factors is performed. Black-Right-Pointing-Pointer The algorithm is applied to a typical BWR fuel lattice. - Abstract: The main goals in nuclear fuel lattice design are: (1) minimizing the rod power peaking factor (PPF) in order that the power level distribution is the most uniform; (2) obtaining a prescribed target value for the multiplication factor (k) at the end of the irradiation in order that the fuel lattice reaches the desired reactivity; and (3) obtaining a prescribed target value for the k at the beginning of the irradiation in order that the reactivity excess is neither a high value (to ease the maneuvering of the control systems) nor a low value (to avoid the penalization of the high cost of the burnable poison content). In this work a simple algorithm to design the burnable poison bearing nuclear fuel lattice is presented. This algorithm is based on a reactor physics analysis. The algorithm is focused on finding the radial distribution of the fuel rods having different fissile and burnable poison contents in order to obtain: (1) an adequate minimum PPF; (2) a prescribed target value of the k at the end of the irradiation; and (3) a prescribed target value of the k at the beginning of the irradiation. This algorithm is based on the factorization of the fissile and burnable poison contents of each fuel rod and on the application of the first-order perturbation theory. The performance of the algorithm is demonstrated with the design of a fuel lattice composed of uranium dioxide (UO{sub 2}) and gadolinium dioxide (Gd{sub 2}O{sub 3}) for boiling water reactors (BWR). This algorithm has been accomplished

  15. Model description and kinetic parameter analysis of MTBE biodegradation in a packed bed reactor

    DEFF Research Database (Denmark)

    Waul, Christopher Kevin; Arvin, Erik; Schmidt, Jens Ejbye

    2008-01-01

    A dynamic modeling approach was used to estimate in-situ model parameters, which describe the degradation of methyl tert-butyl ether (MTBE) in a laboratory packed bed reactor. The measured dynamic response of MTBE pulses injected at the reactor's inlet was analyzed by least squares and parameter...

  16. The modelling of counter-rotating twin screw extruders as reactors for single-component reactions

    NARCIS (Netherlands)

    Ganzeveld, K.J.; Capel, J.E.; Wal, D.J. van der; Janssen, L.P.B.M.

    1994-01-01

    Numerical models are useful to study the behaviour of the extruder as a polymerization reactor. With a correct numerical model a theoretical analysis of the influence of several reaction and extruder parameters can be made, the limitations of the use of the extruder reactor can be determined and the

  17. The modelling of counter-rotating twin screw extruders as reactors for single-component reactions

    NARCIS (Netherlands)

    Ganzeveld, K.J.; Capel, J.E.; Wal, D.J. van der; Janssen, L.P.B.M.

    1994-01-01

    Numerical models are useful to study the behaviour of the extruder as a polymerization reactor. With a correct numerical model a theoretical analysis of the influence of several reaction and extruder parameters can be made, the limitations of the use of the extruder reactor can be determined and the

  18. Modeling and simulation of high-pressure industrial autoclave polyethylene reactor

    Directory of Open Access Journals (Sweden)

    2008-01-01

    Full Text Available High-pressure technology for polyethylene production has been widely used by industries around the world. A good model for the reactor fluid dynamics is essential to set the operating conditions of an autoclave reactor. The high-pressure autoclave reactor model developed in this work was based on a non-isothermal dynamic model, where PID control equations are used to maintain the operation at the unstable steady state. The kinetic mechanism to describe the polymerization rate and molecular weight averages are presented. The model is capable of computing temperature, concentration gradients and polymer characteristics. The model was validated for an existing industrial reactor and data for production of homopolymer polyethylene and has represented well the behavior of the autoclave reactor used in ethylene homopolymerization.

  19. A reduced fidelity model for the rotary chemical looping combustion reactor

    KAUST Repository

    Iloeje, Chukwunwike O.

    2017-01-11

    The rotary chemical looping combustion reactor has great potential for efficient integration with CO capture-enabled energy conversion systems. In earlier studies, we described a one-dimensional rotary reactor model, and used it to demonstrate the feasibility of continuous reactor operation. Though this detailed model provides a high resolution representation of the rotary reactor performance, it is too computationally expensive for studies that require multiple model evaluations. Specifically, it is not ideal for system-level studies where the reactor is a single component in an energy conversion system. In this study, we present a reduced fidelity model (RFM) of the rotary reactor that reduces computational cost and determines an optimal combination of variables that satisfy reactor design requirements. Simulation results for copper, nickel and iron-based oxygen carriers show a four-order of magnitude reduction in simulation time, and reasonable prediction accuracy. Deviations from the detailed reference model predictions range from 3% to 20%, depending on oxygen carrier type and operating conditions. This study also demonstrates how the reduced model can be modified to deal with both optimization and design oriented problems. A parametric study using the reduced model is then applied to analyze the sensitivity of the optimal reactor design to changes in selected operating and kinetic parameters. These studies show that temperature and activation energy have a greater impact on optimal geometry than parameters like pressure or feed fuel fraction for the selected oxygen carrier materials.

  20. Modeling of a Flooding Induced Station Blackout for a Pressurized Water Reactor Using the RISMC Toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Mandelli, Diego; Prescott, Steven R; Smith, Curtis L; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua J; Kinoshita, Robert A

    2011-07-01

    In the Risk Informed Safety Margin Characterization (RISMC) approach we want to understand not just the frequency of an event like core damage, but how close we are (or are not) to key safety-related events and how might we increase our safety margins. The RISMC Pathway uses the probabilistic margin approach to quantify impacts to reliability and safety by coupling both probabilistic (via stochastic simulation) and mechanistic (via physics models) approaches. This coupling takes place through the interchange of physical parameters and operational or accident scenarios. In this paper we apply the RISMC approach to evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., system activation) and to perform statistical analyses (e.g., run multiple RELAP-7 simulations where sequencing/timing of events have been changed according to a set of stochastic distributions). By using the RISMC toolkit, we can evaluate how power uprate affects the system recovery measures needed to avoid core damage after the PWR lost all available AC power by a tsunami induced flooding. The simulation of the actual flooding is performed by using a smooth particle hydrodynamics code: NEUTRINO.

  1. Modeling-based optimization of a fixed-bed industrial reactor for oxidative dehydrogenation of propane

    Institute of Scientific and Technical Information of China (English)

    Ali Darvishi; Razieh Davand; Farhad Khorasheh; Moslem Fattahi

    2016-01-01

    An industrial scale propylene production via oxidative dehydrogenation of propane (ODHP) in multi-tubular re-actors was modeled. Multi-tubular fixed-bed reactor used for ODHP process, employing 10000 of smal diameter tubes immersed in a shel through a proper coolant flows. Herein, a theory-based pseudo-homogeneous model to describe the operation of a fixed bed reactor for the ODHP to correspondence olefin over V2O5/γ-Al2O3 catalyst was presented. Steady state one dimensional model has been developed to identify the operation parameters and to describe the propane and oxygen conversions, gas process and coolant temperatures, as well as other pa-rameters affecting the reactor performance such as pressure. Furthermore, the applied model showed that a double-bed multitubular reactor with intermediate air injection scheme was superior to a single-bed design due to the increasing of propylene selectivity while operating under lower oxygen partial pressures resulting in propane conversion of about 37.3%. The optimized length of the reactor needed to reach 100%conversion of the oxygen was theoretically determined. For the single-bed reactor the optimized length of 11.96 m including 0.5 m of inert section at the entrance region and for the double-bed reactor design the optimized lengths of 5.72 m for the first and 7.32 m for the second reactor were calculated. Ultimately, the use of a distributed oxygen feed with limited number of injection points indicated a significant improvement on the reactor performance in terms of propane conversion and propylene selectivity. Besides, this concept could overcome the reactor run-away temperature problem and enabled operations at the wider range of conditions to obtain enhanced propyl-ene production in an industrial scale reactor.

  2. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Science.gov (United States)

    2010-01-01

    ... necessary based upon site-specific analyses, assessments, or other performance indicators. (iii) By... 10 Energy 2 2010-01-01 2010-01-01 false Requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage. 73.55 Section 73.55 Energy NUCLEAR...

  3. Physical Design of Critical Experiment Facility for Verifying Characteristics and Effects of Coupling Between Reactor and Spallation Target of ADS

    Institute of Scientific and Technical Information of China (English)

    YIN; Sheng-gui; ZHOU; Qi; LI; Yan

    2013-01-01

    For the purpose of studying and verifying characteristics and effects of coupling between reactor and spallation target of ADS,based on the critical experimental facility design criteria and the availableexperiment condition,physical design of a critical experiment facility with lead coolant is completed,using critical calculation code MONK-9A.The contents of physical designs mainly include nuclear fuel,array of fuel rods,neutron source

  4. Modeling of Complex Wear Behavior Associated with Grid-to-Rod Fretting in Light Water Nuclear Reactors

    Science.gov (United States)

    Blau, P. J.; Qu, J.; Lu, R.

    2016-09-01

    Fretting wear damage to fuel cladding from flow-induced vibrations can be a significant concern in the operation of light water nuclear reactors. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. The multi-stage model accounts for oxide layers and wear rate transitions. This paper describes the basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.

  5. Modeling of Complex Wear Behavior Associated with Grid-to-Rod Fretting in Light Water Nuclear Reactors

    Science.gov (United States)

    Blau, P. J.; Qu, J.; Lu, R.

    2016-11-01

    Fretting wear damage to fuel cladding from flow-induced vibrations can be a significant concern in the operation of light water nuclear reactors. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. The multi-stage model accounts for oxide layers and wear rate transitions. This paper describes the basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.

  6. Comparison of sodium and lead-cooled fast reactors regarding reactor physics aspects, severe safety and economical issues

    Energy Technology Data Exchange (ETDEWEB)

    Tucek, Kamil [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands)]. E-mail: kamil.tucek@jrc.nl; Carlsson, Johan [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands); Wider, Hartmut [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands)

    2006-08-15

    A large number of new fast reactors may be needed earlier than foreseen in the Generation IV plans. According to the median forecast of the Special Report on Emission Scenarios commissioned by the Intergovernmental Panel on Climate Control nuclear power will increase by a factor of four by 2050. The drivers for this expected boost are the increasing energy demand in developing countries, energy security, but also climate concerns. However, staying with a once-through cycle will lead to both a substantially increased amount of high-level nuclear waste and an upward pressure on the price of uranium and even concerns about its availability in the coming decades. Therefore, it appears wise to accelerate the development of fast reactors and efficient re-processing technologies. In this paper, two fast reactor systems are discussed-the sodium-cooled fast reactor, which has already been built and can be further improved, and the lead-cooled fast reactor that could be developed relatively soon. An accelerated development of the latter is possible due to the sizeable experience on lead/bismuth eutectic coolant in Russian Alpha-class submarine reactors and the research efforts on accelerator-driven systems in the EU and other countries. First, comparative calculations on critical masses, fissile enrichments and burn-up swings of mid-sized SFRs and LFRs (600 MW{sub e}) are presented. Monte Carlo transport and burn-up codes were used in the analyses. Moreover, Doppler and coolant temperature and axial fuel expansion reactivity coefficients were also evaluated with MCNP and subsequently used in the European Accident Code-2 to calculate reactivity transients and unprotected Loss-of-Flow (ULOF) and Loss-of-Heat Sink (ULOHS) accidents. Further, ULOFs as well as decay heat removal (protected Total Loss-of-Power, TLOP) were calculated with the STAR-CD CFD code for both systems. We show that LFRs and SFRs can be used both as burners and as self-breeders, homogeneously incinerating

  7. Evaluation of model parameters for simulating TiO(2) coated UV reactors.

    Science.gov (United States)

    Duran, J E; Taghipour, F; Mohseni, M

    2011-01-01

    A CFD-based model for simulating TiO(2) coated photocatalytic reactors used in drinking water treatment applications was preliminarily evaluated. The model includes aspects of hydrodynamics, mass transfer, UV-radiation field, and surface chemical reactions. Appropriate models for each of the associated physicochemical phenomena were experimentally or analytically examined. Once defined and evaluated, the individual models were integrated into a CFD-based model for simulating photocatalytic reactor performance, which was experimentally evaluated.

  8. Modeling of operating history of the research nuclear reactor

    Science.gov (United States)

    Naymushin, A.; Chertkov, Yu; Shchurovskaya, M.; Anikin, M.; Lebedev, I.

    2016-06-01

    The results of simulation of the IRT-T reactor operation history from 2012 to 2014 are presented. Calculations are performed using continuous energy Monte Carlo code MCU-PTR. Comparison is made between calculation and experimental data for the critical reactor.

  9. CFD analysis of PWR core top and reactor vessel upper plenum internal subdomain models

    Energy Technology Data Exchange (ETDEWEB)

    Kao, Min-Tsung; Wu, Chung-Yun [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Chieng, Ching-Chang, E-mail: cchieng@ess.nthu.edu.tw [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Xu Yiban; Yuan Kun; Dzodzo, Milorad; Conner, Michael; Beltz, Steven; Ray, Sumit; Bissett, Teresa [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States)

    2011-10-15

    Highlights: > The paper develops a CFD flow model for upper portion of AP1000 and determines how lateral flow in the top core and upper plenum. > Mesh sensitivities and geometrical modification strategies give the guidelines to reduce the size of overall computation mesh. > Pressure drop measurement data act as a guideline for the mesh selection. > Lateral flows are mainly exiting through upper and lower windows of guide tubes ({approx}81%) and 18% flow through small side gaps. > The interactions between guide tubes and neighboring support column as well as flow characteristic are revealed. - Abstract: One aspect of the Westinghouse AP1000{sup TM} reactor design is the reduction in the number of major components and simplification in manufacturing. One design change relative to current Westinghouse reactors of similar size is that AP1000 reactor vessel has two nozzles/hot legs instead of three. With regard to fuel performance, this design difference creates a different flow field in the reactor vessel upper plenum. The flow exiting from the core and entering the upper plenum must turn toward one of the two outlet nozzles and flow laterally around numerous control rod guide tubes and support columns. Also, below the upper plenum are the upper core plate and the top core region of the 157 fuel assemblies and 69 guidetube assemblies. To determine how the lateral flow in the top of the core and upper plenum compares to the current reactors a CFD model of the flow in the upper portion of the AP1000 reactor vessel was created. Before detailed CFD simulations of the flow in the entire upper plenum and top core regions were performed, conducting local simulations for smaller sections of the domain provided crucial and detailed physical aspects of the flow. These sub-domain models were used to perform mesh sensitivities and to assess what geometrical details may be eliminated from the larger model in order to reduce mesh size and computational requirements. In this paper

  10. Physical modeling of the piano

    Science.gov (United States)

    Giordano, N.; Jiang, M.

    2003-10-01

    Over the past several years, this project has been aimed at constructing a physical model of the piano. The goal is to use Newton's laws to describe the motion of the hammers, strings, soundboard, and surrounding air, and thereby calculate the sound produced by the instrument entirely from first principles. The structure of the model is described, along with experiments that have provided essential tests and guidance to the calculations. The state of the model and, especially, how this work can lead to new insights and understanding into the piano are discussed. In many cases the work and the specific questions addressed along the way have followed paths initially inspired and developed by Gabriel Weinreich. [Work supported by NSF.

  11. Models and Stability Analysis of Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    John Dorning

    2002-04-15

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is

  12. Modeling of a fluidized bed reactor for the ethylene-propylene copolymerization

    Directory of Open Access Journals (Sweden)

    Juan Guillermo Cadavid Estrada

    2010-04-01

    Full Text Available A mathematical model for the ethylene - propylene copolymerization with a Ziegler - Natta catalyst in a gas phase fludized bed reactor is presented. The model includes a two active site kinetic model with spontaneous transfer reactions and site deactivation. Also, it is studied and simulated the growth of a polymeric particle which is exposed to an outside atmosphere (monomers concentrations and temperature that represent the emulsion phase conditions of the reactor. Particle growth model is the basis for the study of the sizes distribution into the reactor. Two phase model of Kunii-Levenspiel is the basis for the modelling and simulation of the fluid bed reactor, the models developed consider two extreme cases for the gas mixed grade in emulsion phase (perfectly mixed and plug flow. The solution of the models includes mass (for the two monomers and energy balances, coupled with the particle growth and residence time distribution models.

  13. Excellence in Physics Education Award: Modeling Theory for Physics Instruction

    Science.gov (United States)

    Hestenes, David

    2014-03-01

    All humans create mental models to plan and guide their interactions with the physical world. Science has greatly refined and extended this ability by creating and validating formal scientific models of physical things and processes. Research in physics education has found that mental models created from everyday experience are largely incompatible with scientific models. This suggests that the fundamental problem in learning and understanding science is coordinating mental models with scientific models. Modeling Theory has drawn on resources of cognitive science to work out extensive implications of this suggestion and guide development of an approach to science pedagogy and curriculum design called Modeling Instruction. Modeling Instruction has been widely applied to high school physics and, more recently, to chemistry and biology, with noteworthy results.

  14. Physics design for the Brookhaven Medical Research Reactor epithermal neutron source.

    Science.gov (United States)

    Wheeler, F J; Parsons, D K; Nigg, D W; Wessol, D E; Miller, L G; Fairchild, R G

    1990-01-01

    A collaborative effort by researchers at the Idaho National Engineering Laboratory and the Brookhaven National Laboratory has resulted in the design and implementation of an epithermal-neutron source at the Brookhaven Medical Research Reactor (BMRR). Large aluminum containers, filled with aluminum oxide tiles and aluminum spacers, were tailored to pre-existing compartments on the animal side of the reactor facility. A layer of cadmium was used to minimize the thermal-neutron component. Additional bismuth was added to the pre-existing bismuth shield to minimize the gamma component of the beam. Lead was also added to reduce gamma streaming around the bismuth. The physics design methods are outlined in this paper. Information available to date shows close agreement between calculated and measured beam parameters. The neutron spectrum is predominantly in the intermediate energy range (0.5 eV - 10 keV). The peak flux intensity is 6.4E + 12 n/(m2.s.MW) at the center of the beam on the outer surface of the final gamma shield. The corresponding neutron current is 3.8E + 12 n/(m2.s.MW). Presently, the core operates at a maximum of 3 MW. The fast-neutron KERMA is 3.6E-15 cGy/(n/m2) and the gamma KERMA is 5.0E-16 cGY/(n/m2) for the unperturbed beam. The neutron intensity falls off rapidly with distance from the outer shield and the thermal flux realized in phantom or tissue is strongly dependent on the beam-delimiter and target geometry.

  15. Models and analyses for inertial-confinement fusion-reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Bohachevsky, I.O.

    1981-05-01

    This report describes models and analyses devised at Los Alamos National Laboratory to determine the technical characteristics of different inertial confinement fusion (ICF) reactor elements required for component integration into a functional unit. We emphasize the generic properties of the different elements rather than specific designs. The topics discussed are general ICF reactor design considerations; reactor cavity phenomena, including the restoration of interpulse ambient conditions; first-wall temperature increases and material losses; reactor neutronics and hydrodynamic blanket response to neutron energy deposition; and analyses of loads and stresses in the reactor vessel walls, including remarks about the generation and propagation of very short wavelength stress waves. A discussion of analytic approaches useful in integrations and optimizations of ICF reactor systems concludes the report.

  16. Cabin Environment Physics Risk Model

    Science.gov (United States)

    Mattenberger, Christopher J.; Mathias, Donovan Leigh

    2014-01-01

    This paper presents a Cabin Environment Physics Risk (CEPR) model that predicts the time for an initial failure of Environmental Control and Life Support System (ECLSS) functionality to propagate into a hazardous environment and trigger a loss-of-crew (LOC) event. This physics-of failure model allows a probabilistic risk assessment of a crewed spacecraft to account for the cabin environment, which can serve as a buffer to protect the crew during an abort from orbit and ultimately enable a safe return. The results of the CEPR model replace the assumption that failure of the crew critical ECLSS functionality causes LOC instantly, and provide a more accurate representation of the spacecraft's risk posture. The instant-LOC assumption is shown to be excessively conservative and, moreover, can impact the relative risk drivers identified for the spacecraft. This, in turn, could lead the design team to allocate mass for equipment to reduce overly conservative risk estimates in a suboptimal configuration, which inherently increases the overall risk to the crew. For example, available mass could be poorly used to add redundant ECLSS components that have a negligible benefit but appear to make the vehicle safer due to poor assumptions about the propagation time of ECLSS failures.

  17. PIK3CA Mutation in Colorectal Cancer: Relationship with Genetic and Epigenetic Alterations

    Directory of Open Access Journals (Sweden)

    Katsuhiko Nosho

    2008-06-01

    Full Text Available Somatic PIK3CA mutations are often present in colorectal cancer. Mutant PIK3CA activates AKT signaling, which up-regulates fatty acid synthase (FASN. Microsatellite instability (MSI and CpG island methylator phenotype (CIMP are important molecular classifiers in colorectal cancer. However, the relationship between PIK3CA mutation, MSI and CIMP remains uncertain. Using Pyrosequencing technology, we detected PIK3CA mutations in 91 (15% of 590 population-based colorectal cancers. To determine CIMP status, we quantified DNA methylation in eight CIMP-specific promoters [CACNA1G, CDKN2A (p16, CRABP1, IGF2, MLH1, NEUROG1, RUNX3, and SOCS1] by real-time polymerase chain reaction (MethyLight. PIK3CA mutation was significantly associated with mucinous tumors [P = .0002; odds ratio (OR = 2.44], KRAS mutation (P < .0001; OR = 2.68, CIMP-high (P = .03; OR = 2.08, phospho–ribosomal protein S6 expression (P = .002; OR = 2.19, and FASN expression (P = .02; OR = 1.85 and inversely with p53 expression (P = .01; OR = 0.54 and β-catenin (CTNNB1 alteration (P = .004; OR = 0.43. In addition, PIK3CA G-to-A mutations were associated with MGMT loss (P = .001; OR = 3.24 but not with MGMT promoter methylation. In conclusion, PIK3CA mutation is significantly associated with other key molecular events in colorectal cancer, and MGMT loss likely contributes to the development of PIK3CA G>A mutation. In addition, Pyrosequencing is useful in detecting PIK3CA mutation in archival paraffin tumor tissue. PIK3CA mutational data further emphasize heterogeneity of colorectal cancer at the molecular level.

  18. Modeling of single tube Fischer-Tropsch reactor for model biosyngas

    Energy Technology Data Exchange (ETDEWEB)

    Rafiq, Muhammad Hamid; Hustad, Johan Einar

    2010-07-01

    Fischer-Tropsch Synthesis is an important chemical process for the production of liquid fuels. The present study addresses the modeling of low temperature single tube Fischer- Tropsch reactor for a model biosyngas (33%H2, 17%CO and 50%N2). Cobalt based catalyst is used for synthesis due to its high activity and selectivity for linear hydrocarbons and lower price compared with other noble metals. The chemistry taking place in a FT reactor is complex but can be simplified by the following reaction (see original paper). For cobalt catalyst methanation reaction and shift reaction is neglected. Yates and Satterfield[1] determined the intrinsic rate constant of H2 consumption on a commercial cobalt catalyst. According to Steynberg et al.[2], the intrinsic activity of modern industrial cobalt catalyst is by a factor of three times higher then those reported by the above mentioned author. So, the equation of hydrogen consumption on a commercial cobalt catalyst is estimated (using the threefold value) and is given below: (see original paper). Modeling of Single tube fixed bed Fischer-Tropsch reactor is done with one or two dimensional pseudo homogeneous model. Among many thing the influence of cooling temperature effects are studied on the axial molar composition profiles, molar flow of reactant and product and reactant conversion. In addition effect of cooling temperature on the axial temperature profiles in a single tube Fischer-Tropsch reactor is also studied. (AG)

  19. Tandem Mirror Reactor Systems Code (Version I)

    Energy Technology Data Exchange (ETDEWEB)

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.; Barrett, R.J.; Gorker, G.E.; Spampinaton, P.T.; Bulmer, R.H.; Dorn, D.W.; Perkins, L.J.; Ghose, S.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost.

  20. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010

    Energy Technology Data Exchange (ETDEWEB)

    Rahmat Aryaeinejad; Douglas S. Crawford; Mark D. DeHart; George W. Griffith; D. Scott Lucas; Joseph W. Nielsen; David W. Nigg; James R. Parry; Jorge Navarro

    2010-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or “Core Modeling Update”) Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  1. Reactor modeling and process analysis for partial oxidation of natural gas

    NARCIS (Netherlands)

    Albrecht, Bogdan Alexandru

    2004-01-01

    This thesis analyses a novel process of partial oxidation of natural gas and develops a numerical tool for the partial oxidation reactor modeling. The proposed process generates syngas in an integrated plant of a partial oxidation reactor, a syngas turbine and an air separation unit. This is called

  2. Fast pyrolysis in a novel wire-mesh reactor: decomposition of pine wood and model compounds

    NARCIS (Netherlands)

    Hoekstra, E.; Swaaij, van W.P.M.; Kersten, S.R.A.; Hogendoorn, J.A.

    2012-01-01

    In fast pyrolysis, biomass decomposition processes are followed by vapor phase reactions. Experimental results were obtained in a unique wire-mesh reactor using pine wood, KCl impregnated pine wood and several model compounds (cellulose, xylan, lignin, levoglucosan, glucose). The wire-mesh reactor w

  3. Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1986-04-01

    Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated.

  4. Air purification in a reverse-flow reactor: Model simulations vs. experiments

    NARCIS (Netherlands)

    Beld, van de L.; Westerterp, K.R.

    1996-01-01

    The behavior of a reverse-flow reactor was studied for the purification of polluted air by catalytic combustion. A heterogeneous one-dimensional model was extended with a heat balance for the reactor wall. An overall heat transport term is included to account for the small heat losses in radial dire

  5. The Thermal-Hydraulic model for the pebble bed modular reactor (PBMR) plant operator training simulator system

    Energy Technology Data Exchange (ETDEWEB)

    Dudley, Trevor [Pebble Bed Modular Reactor (Proprietary) Limited, Die Anker Building, Centurion 0046 (South Africa)], E-mail: trevor.dudley@pbmr.co.za; Bouwer, Werner; Villiers, Piet de [Pebble Bed Modular Reactor (Proprietary) Limited, Die Anker Building, Centurion 0046 (South Africa); Wang Zen [GSE Systems, Inc., 7133 Rutherford Suite 200, Baltimore, MD 21244 (United States)

    2008-11-15

    This paper provides a discussion of the model development status and verification efforts for the Reactor Core Thermal-Hydraulic model developed for the full-scope plant Operator Training Simulator System of the Pebble Bed Modular Reactor (PBMR). Due to the First of a Kind Engineering nature and lack of reference plant data, model verification has mainly been focused on benchmarking the model configurations against test cases performed by PBMR design analysis codes, i.e. TINTE, VSOP and FLOWNEX. As a first step, due to the symmetrical physical nature of the PBMR core, a two-dimensional (2D) model configuration in radial and axial directions (axial-symmetry) was developed. The design was subsequently extended to a three-dimensional (3D) configuration. Through the use of cross-flow and cross-conduction links, three nearly identical 2D configurations were glued together to form this 3D model configuration. To date, the 3D configuration represents the most comprehensive model to simulate the PBMR core thermo-hydraulics. This paper concludes with the verification of thermodynamic and heat-transfer properties of two steady state (100% and 40% power) conditions between the 3D Reactor Core Thermal-Hydraulic model and the available FLOWNEX and TINTE design code analysis. The transient operations between these two power levels are also discussed.

  6. Models and structures: mathematical physics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-07-01

    This document gathers research activities along 5 main directions. 1) Quantum chaos and dynamical systems. Recent results concern the extension of the exact WKB method that has led to a host of new results on the spectrum and wave functions. Progress have also been made in the description of the wave functions of chaotic quantum systems. Renormalization has been applied to the analysis of dynamical systems. 2) Combinatorial statistical physics. We see the emergence of new techniques applied to various such combinatorial problems, from random walks to random lattices. 3) Integrability: from structures to applications. Techniques of conformal field theory and integrable model systems have been developed. Progress is still made in particular for open systems with boundary conditions, in connection to strings and branes physics. Noticeable links between integrability and exact WKB quantization to 2-dimensional disordered systems have been highlighted. New correlations of eigenvalues and better connections to integrability have been formulated for random matrices. 4) Gravities and string theories. We have developed aspects of 2-dimensional string theory with a particular emphasis on its connection to matrix models as well as non-perturbative properties of M-theory. We have also followed an alternative path known as loop quantum gravity. 5) Quantum field theory. The results obtained lately concern its foundations, in flat or curved spaces, but also applications to second-order phase transitions in statistical systems.

  7. Decision model for evaluating reactor disposition of excess plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Edmunds, T.

    1995-02-01

    The US Department of Energy is currently considering a range of technologies for disposition of excess weapon plutonium. Use of plutonium fuel in fission reactors to generate spent fuel is one class of technology options. This report describes the inputs and results of decision analyses conducted to evaluate four evolutionary/advanced and three existing fission reactor designs for plutonium disposition. The evaluation incorporates multiple objectives or decision criteria, and accounts for uncertainty. The purpose of the study is to identify important and discriminating decision criteria, and to identify combinations of value judgments and assumptions that tend to favor one reactor design over another.

  8. Modeling the Ductile Brittle Fracture Transition in Reactor Pressure Vessel Steels using a Cohesive Zone Model based approach

    Energy Technology Data Exchange (ETDEWEB)

    Pritam Chakraborty; S. Bulent Biner

    2013-10-01

    Fracture properties of Reactor Pressure Vessel (RPV) steels show large variations with changes in temperature and irradiation levels. Brittle behavior is observed at lower temperatures and/or higher irradiation levels whereas ductile mode of failure is predominant at higher temperatures and/or lower irradiation levels. In addition to such temperature and radiation dependent fracture behavior, significant scatter in fracture toughness has also been observed. As a consequence of such variability in fracture behavior, accurate estimates of fracture properties of RPV steels are of utmost importance for safe and reliable operation of reactor pressure vessels. A cohesive zone based approach is being pursued in the present study where an attempt is made to obtain a unified law capturing both stable crack growth (ductile fracture) and unstable failure (cleavage fracture). The parameters of the constitutive model are dependent on both temperature and failure probability. The effect of irradiation has not been considered in the present study. The use of such a cohesive zone based approach would allow the modeling of explicit crack growth at both stable and unstable regimes of fracture. Also it would provide the possibility to incorporate more physical lower length scale models to predict DBT. Such a multi-scale approach would significantly improve the predictive capabilities of the model, which is still largely empirical.

  9. Modeling a multivariable reactor and on-line model predictive control.

    Science.gov (United States)

    Yu, D W; Yu, D L

    2005-10-01

    A nonlinear first principle model is developed for a laboratory-scaled multivariable chemical reactor rig in this paper and the on-line model predictive control (MPC) is implemented to the rig. The reactor has three variables-temperature, pH, and dissolved oxygen with nonlinear dynamics-and is therefore used as a pilot system for the biochemical industry. A nonlinear discrete-time model is derived for each of the three output variables and their model parameters are estimated from the real data using an adaptive optimization method. The developed model is used in a nonlinear MPC scheme. An accurate multistep-ahead prediction is obtained for MPC, where the extended Kalman filter is used to estimate system unknown states. The on-line control is implemented and a satisfactory tracking performance is achieved. The MPC is compared with three decentralized PID controllers and the advantage of the nonlinear MPC over the PID is clearly shown.

  10. Development and analysis of some versions of the fractional-order point reactor kinetics model for a nuclear reactor with slab geometry

    Science.gov (United States)

    Vyawahare, Vishwesh A.; Nataraj, P. S. V.

    2013-07-01

    In this paper, we report the development and analysis of some novel versions and approximations of the fractional-order (FO) point reactor kinetics model for a nuclear reactor with slab geometry. A systematic development of the FO Inhour equation, Inverse FO point reactor kinetics model, and fractional-order versions of the constant delayed neutron rate approximation model and prompt jump approximation model is presented for the first time (for both one delayed group and six delayed groups). These models evolve from the FO point reactor kinetics model, which has been derived from the FO Neutron Telegraph Equation for the neutron transport considering the subdiffusive neutron transport. Various observations and the analysis results are reported and the corresponding justifications are addressed using the subdiffusive framework for the neutron transport. The FO Inhour equation is found out to be a pseudo-polynomial with its degree depending on the order of the fractional derivative in the FO model. The inverse FO point reactor kinetics model is derived and used to find the reactivity variation required to achieve exponential and sinusoidal power variation in the core. The situation of sudden insertion of negative reactivity is analyzed using the FO constant delayed neutron rate approximation. Use of FO model for representing the prompt jump in reactor power is advocated on the basis of subdiffusion. Comparison with the respective integer-order models is carried out for the practical data. Also, it has been shown analytically that integer-order models are a special case of FO models when the order of time-derivative is one. Development of these FO models plays a crucial role in reactor theory and operation as it is the first step towards achieving the FO control-oriented model for a nuclear reactor. The results presented here form an important step in the efforts to establish a step-by-step and systematic theory for the FO modeling of a nuclear reactor.

  11. Linear stability analysis of a nuclear reactor using the lumped model

    Directory of Open Access Journals (Sweden)

    Kale Vivek A.

    2016-01-01

    Full Text Available The stability analysis of a nuclear reactor is an important aspect in the design and operation of the reactor. A stable neutronic response to perturbations is essential from the safety point of view. In this paper, a general methodology has been developed for the linear stability analysis of nuclear reactors using the lumped reactor model. The reactor kinetics has been modelled using the point kinetics equations and the reactivity feedbacks from fuel, coolant and xenon have been modelled through the appropriate time dependent equations. These governing equations are linearized considering small perturbations in the reactor state around a steady operating point. The characteristic equation of the system is used to establish the stability zone of the reactor considering the reactivity coefficients as parameters. This methodology has been used to identify the stability region of a typical pressurized heavy water reactor. It is shown that the positive reactivity feedback from xenon narrows down the stability region. Further, it is observed that the neutron kinetics parameters (such as the number of delayed neutron precursor groups considered, the neutron generation time, the delayed neutron fractions, etc. do not have a significant influence on the location of the stability boundary. The stability boundary is largely influenced by the parameters governing the evolution of the fuel and coolant temperature and xenon concentration.

  12. V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

    Energy Technology Data Exchange (ETDEWEB)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

    2010-07-15

    V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

  13. V.S.O.P.(97) computer code system for reactor physics and fuel cycle simulation. Input manual and comments

    Energy Technology Data Exchange (ETDEWEB)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Scherer, W.

    1998-04-01

    V.S.O.P. (97) is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies processing of cross sections, the setup of the reactor and of the fuel element, repeated neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. V.S.O.P. (97) can simulate the reactor operation from the initial core towards the equilibrium core. V.S.O.P. (97) - on the basis of V.S.O.P. (94) - has been improved with regard to a more detailed treatment of the build-up and the depletion of the heavy metal isotopes. Their chains now include the minor actinides. Resonance cross sections of the lumped resonance absorbers are evaluated burnup-dependent. Beyond this, the code has been reviewed in many details, aiming at an improved precision in the computer simulation of the features of the reactors and of their fuel cycle. The code consists of about 65000 FORTRAN statements. A memory of 32 MB should be available for its use. (orig.)

  14. Model Based Cyber Security Analysis for Research Reactor Protection System

    Energy Technology Data Exchange (ETDEWEB)

    Sho, Jinsoo; Rahman, Khalil Ur; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Son, Hanseong [Joongbu Univ., Geumsan (Korea, Republic of)

    2013-07-01

    The study on the qualitative risk due to cyber-attacks into research reactors was performed using bayesian Network (BN). This was motivated to solve the issues of cyber security raised due to digitalization of instrumentation and control (I and C) system. As a demonstrative example, we chose the reactor protection system (RPS) of research reactors. Two scenarios of cyber-attacks on RPS were analyzed to develop mitigation measures against vulnerabilities. The one is the 'insertion of reactor trip' and the other is the 'scram halt'. The six mitigation measures are developed for five vulnerability for these scenarios by getting the risk information from BN.

  15. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Song, C. H.; Chung, M. K.; Park, C. K. and others

    2005-04-15

    The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.

  16. Modeling of Fischer-Tropsch Synthesis in a Slurry Reactor with Water Permeable Membrane

    Institute of Scientific and Technical Information of China (English)

    Fabiano A. N. Fernandes

    2007-01-01

    Fischer-Tropsch synthesis is an important chemical process for the production of liquid fuels and olefins. In recent years, the abundant availability of natural gas and the increasing demand of olefins, diesel, and waxes have led to a high interest to further develop this process. A mathematical model of a slurry membrane reactor used for syngas polymerization was developed to simulate and compare the maximum yields and operating conditions in the reactor with that in a conventional slurry reactor.The carbon polymerization was studied from a modeling point of view in a slurry reactor with a water permeable membrane and a conventional slurry reactor. Simulation results show that different parameters affect syngas conversion and carbon product distribution, such as the hydrogen to carbon monoxide ratio,and the membrane parameters such as membrane permeance.

  17. Physical model of aluminium refining process in URC-7000

    Directory of Open Access Journals (Sweden)

    M. Saternus

    2009-07-01

    Full Text Available The paper presents short characteristics of the most frequently used reactors for the continuous refining of aluminium and its alloys. Refining depends on the flow rate of refining gas. It influences the shape and size of gas bubbles. So the physical model was created to determine the level of gas bubble dispersion in liquids. Schemes of gas dispersion in liquid metal were presented taking into consideration that the gas flow rate is changing from 2 to 30 dm3/min. The range selection of the flow rate of refining gas value for five patterns of the dispersion (no dispersion, minimum, intimate, uniform and overdispersion was also done.

  18. Steam separator modeling for various nuclear reactor transients

    Energy Technology Data Exchange (ETDEWEB)

    Paik, C Y; Mullen, G; Knoess, C; Griffith, P

    1987-06-01

    In a pressurized water reactor steam generator, a moisture separator is used to separate steam and liquid and to insure that essentially dry steam is supplied to the turbine. During a steam line break or combined steam line break plus tube rupture, a number of phenomena can occur in the separator which have no counterparts during steady-state operation. How the separator will perform under these circumstances is important for two reasons, it affects the carry-over of radioactive iodine and the water inventory in the secondary side. This study has as its goal the development of a simple separator model which can be applied to a variety of steam generator for off-design conditions. Experiments were performed using air and water on three different types of centrifugal separators: a cyclone as a generic separator, a Combustion Engineering type stationary swirl vane separator, and a Westinghouse type separator. The cyclone separator system has three stages of separation: first the cyclone, then a gravity separator, and finally a chevron plate separator. The other systems have only a centrifugal separator to isolate the effect of the primary separator. Experiments were also done in MIT blowdown rig, with and without a separator, using steam and water. The separators appear to perform well at flow rates well above the design values as long as the downcomer water level is not high. High downcomer water level rather than high flow rates appear to be the primary cause of degraded performance. Appreciable carry-over from the separator section of a steam generator occurs when the drain lines from three stages of separation are unable to carry off the liquid flow. Failure scenarios of the separator for extreme range of conditions from the quasi-steady state transient to the fast transients are presented. A general model structure and simple separator models are provided.

  19. Dynamic Model of an Ammonia Synthesis Reactor Based on Open Information

    OpenAIRE

    Jinasena, Asanthi; Lie, Bernt; Glemmestad, Bjørn

    2016-01-01

    Ammonia is a widely used chemical, hence the ammonia manufacturing process has become a standard case study in the scientific community. In the field of mathematical modeling of the dynamics of ammonia synthesis reactors, there is a lack of complete and well documented models. Therefore, the main aim of this work is to develop a complete and well documented mathematical model for observing the dynamic behavior of an industrial ammonia synthesis reactor system. The mode...

  20. Evaluation of the physical protection system of the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vaz, Antonio C.A.; Conti, Thadeu das N., E-mail: acavaz@ipen.br, E-mail: tnconti@yahoo.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The '09/11' in New York and the accident at the Fukushima power plant are two events that served as worldwide reference to review some aspects of the Physical Protection System (PPS) in nuclear areas. The nuclear research reactor IEA-R1 has followed this new world order and improved the protection systems that are directly related to detection (CCTV, sensors, alarms, etc), delay (turnstile, gates, barriers, etc) and response (communication systems, response force, etc), for operation against malicious act, seeking always to avoid or minimize any possibility of threat, theft and sabotage. These actions were performed to prevent and to mitigate the consequence on the environment, economy and society from damages caused by natural hazard, as well. This study evaluates the PPS of the IEA-R1 regarding the weaknesses, strengths,and impacts of the changes resulting from the system implanted. The analyses were based on methodology developed by security experts from SANDIA National Laboratories in Texas - U.S.A, allowing the evaluation of the system through probabilistic and hypothetical analysis. (author)

  1. The use of active learning strategies in the instruction of Reactor Physics concepts

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Michael A.

    2000-01-01

    Each of the Active Learning strategies employed to teach Reactor Physics material has been or promises to be instructionally successful. The Cooperative Group strategy has demonstrated a statistically significant increase in student performance on the unit exam in teaching conceptually difficult, transport and diffusion theory material. However, this result was achieved at the expense of a modest increase in class time. The Tutorial CBI programs have enabled learning equally as well as classroom lectures without the direct intervention of an instructor. Thus, the Tutorials have been successful as homework assignments, releasing classroom time for other instruction. However, the time required for development of these tools was large, on the order of two hundred hours per hour of instruction. The initial introduction of the Case-Based strategy was roughly as effective as the traditional classroom instruction. Case-Based learning could well, after important modifications, perform better than traditional instruction. A larger percentage of the students prefer active learning strategies than prefer traditional lecture presentations. Student preferences for the active strategies were particularly strong when they believed that the strategies helped them learn the material better than they would have by using a lecture format. In some cases, students also preferred the active strategies because they were different from traditional instruction, a change of pace. Some students preferred lectures to CBI instruction, primarily because the CBI did not afford them the opportunity to question the instructor during the presentation.

  2. Mutations in PIK3CA sensitize breast cancer cells to physiologic levels of aspirin.

    Science.gov (United States)

    Turturro, Sanja B; Najor, Matthew S; Ruby, Carl E; Cobleigh, Melody A; Abukhdeir, Abde M

    2016-02-01

    A review of the literature finds that women diagnosed with breast cancer, who were on an aspirin regimen, experienced a decreased risk of distant metastases and death. Several recent studies have reported an improvement in overall survival in colorectal cancer patients who harbored mutations in the oncogene PIK3CA and received a daily aspirin regimen. Breast cancer patients on a daily aspirin regimen experienced decreased risk of distant metastases and death. PIK3CA is the most frequently mutated oncogene in breast cancer, occurring in up to 45 % of all breast cancers. In order to determine if mutations in PIK3CA sensitized breast cancers to aspirin treatment, we employed the use of isogenic cellular clones of the non-tumorigenic, breast epithelial cell line MCF-10A that harbored mutations in either PIK3CA or KRAS or both. We report that mutations in both PIK3CA and KRAS are required for the greatest aspirin sensitivity in breast cancer, and that the GSK3β protein was hyperphosphorylated in aspirin-treated double knockin cells, but not in other clones/treatments. A more modest effect was observed with single mutant PIK3CA, but not KRAS alone. These observations were further confirmed in a panel of breast cancer cell lines. Our findings provide the first evidence that mutations in PIK3CA sensitize breast cancer cells to aspirin.

  3. Activity report of working party on reactor physics of accelerator-driven system. July 1999 to March 2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-02-01

    Under the Research Committee on Reactor Physics, the Working Party on Reactor Physics of Accelerator-Driven System (ADS-WP) was set in July 1999 to review and investigate special subjects related to reactor physics research for the Accelerator-Driven Subcritical System (ADS). The ADS-WP, at the first meeting, discussed a guideline of its activity for two years and decided to concentrate upon three subjects: (1) neutron transport calculations in high energy range, (2) static and kinetic (safety-related) characteristics of subcritical system, and (3) system design including ADS concepts and elemental technology developments required. The activity of ADS-WP continued from July 1999 to March 2001. In this duration, the members of ADS-WP met together four times and discussed the above subjects. In addition, the ADS-WP conducted a questionnaire on requests and proposals for the plan of Transmutation Physics Experimental Facility in the High-Intensity Proton Accelerator Project, which is a joint project between JAERI and KEK (High Energy Accelerator Research Organization). This report summarizes the results obtained by the above ADS-WP activity. (author)

  4. Integral Reactor Containment Condensation Model and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Qiao [Oregon State Univ., Corvallis, OR (United States); Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States)

    2016-05-02

    This NEUP funded project, NEUP 12-3630, is for experimental, numerical and analytical studies on high-pressure steam condensation phenomena in a steel containment vessel connected to a water cooling tank, carried out at Oregon State University (OrSU) and the University of Wisconsin at Madison (UW-Madison). In the three years of investigation duration, following the original proposal, the planned tasks have been completed: (1) Performed a scaling study for the full pressure test facility applicable to the reference design for the condensation heat transfer process during design basis accidents (DBAs), modified the existing test facility to route the steady-state secondary steam flow into the high pressure containment for controllable condensation tests, and extended the operations at negative gage pressure conditions (OrSU). (2) Conducted a series of DBA and quasi-steady experiments using the full pressure test facility to provide a reliable high pressure condensation database (OrSU). (3) Analyzed experimental data and evaluated condensation model for the experimental conditions, and predicted the prototypic containment performance under accidental conditions (UW-Madison). A film flow model was developed for the scaling analysis, and the results suggest that the 1/3 scaled test facility covers large portion of laminar film flow, leading to a lower average heat transfer coefficient comparing to the prototypic value. Although it is conservative in reactor safety analysis, the significant reduction of heat transfer coefficient (50%) could under estimate the prototypic condensation heat transfer rate, resulting in inaccurate prediction of the decay heat removal capability. Further investigation is thus needed to quantify the scaling distortion for safety analysis code validation. Experimental investigations were performed in the existing MASLWR test facility at OrST with minor modifications. A total of 13 containment condensation tests were conducted for pressure

  5. V.S.O.P. (99/05) Computer Code System : computer code system for reactor physics and fuel cycle simulation

    OpenAIRE

    Scherer, W.; Brockmann, H.; Haas, K. A.; Rütten, H. J.

    2005-01-01

    V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The...

  6. Fusion reactor theory and conceptual design. (Latest citations from the INSPEC: Information Services for the Physics and Engineering Communities database). Published Search

    Energy Technology Data Exchange (ETDEWEB)

    1992-11-01

    The bibliography contains citations concerning theoretical and conceptual aspects of fusion reactor physics and designs. A variety of fusion reactors is discussed, including Tokamak, experimental, commercial, tandem mirror, and superconducting magnetic. Topics also include fusion reactor materials, Tokamak devices, blanket design, divertors, fusion plasma production, superconducting magnets, and cryogenic systems. (Contains a minimum of 159 citations and includes a subject term index and title list.)

  7. Development of a Reactor Model for Chemical Conversion of Lunar Regolith

    Science.gov (United States)

    Hegde, U.; Balasubramaniam, R.; Gokoglu, S.

    2009-01-01

    Lunar regolith will be used for a variety of purposes such as oxygen and propellant production and manufacture of various materials. The design and development of chemical conversion reactors for processing lunar regolith will require an understanding of the coupling among the chemical, mass and energy transport processes occurring at the length and time scales of the overall reactor with those occurring at the corresponding scales of the regolith particles. To this end, a coupled transport model is developed using, as an example, the reduction of ilmenite-containing regolith by a continuous flow of hydrogen in a flow-through reactor. The ilmenite conversion occurs on the surface and within the regolith particles. As the ilmenite reduction proceeds, the hydrogen in the reactor is consumed, and this, in turn, affects the conversion rate of the ilmenite in the particles. Several important quantities are identified as a result of the analysis. Reactor scale parameters include the void fraction (i.e., the fraction of the reactor volume not occupied by the regolith particles) and the residence time of hydrogen in the reactor. Particle scale quantities include the time for hydrogen to diffuse into the pores of the regolith particles and the chemical reaction time. The paper investigates the relationships between these quantities and their impact on the regolith conversion. Application of the model to various chemical reactor types, such as fluidized-bed, packed-bed, and rotary-bed configurations, are discussed.

  8. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2013

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg

    2013-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  9. Modeling and Control of a Large Nuclear Reactor A Three-Time-Scale Approach

    CERN Document Server

    Shimjith, S R; Bandyopadhyay, B

    2013-01-01

    Control analysis and design of large nuclear reactors requires a suitable mathematical model representing the steady state and dynamic behavior of the reactor with reasonable accuracy. This task is, however, quite challenging because of several complex dynamic phenomena existing in a reactor. Quite often, the models developed would be of prohibitively large order, non-linear and of complex structure not readily amenable for control studies. Moreover, the existence of simultaneously occurring dynamic variations at different speeds makes the mathematical model susceptible to numerical ill-conditioning, inhibiting direct application of standard control techniques. This monograph introduces a technique for mathematical modeling of large nuclear reactors in the framework of multi-point kinetics, to obtain a comparatively smaller order model in standard state space form thus overcoming these difficulties. It further brings in innovative methods for controller design for systems exhibiting multi-time-scale property,...

  10. Adaptive control using a hybrid-neural model: application to a polymerisation reactor

    Directory of Open Access Journals (Sweden)

    Cubillos F.

    2001-01-01

    Full Text Available This work presents the use of a hybrid-neural model for predictive control of a plug flow polymerisation reactor. The hybrid-neural model (HNM is based on fundamental conservation laws associated with a neural network (NN used to model the uncertain parameters. By simulations, the performance of this approach was studied for a peroxide-initiated styrene tubular reactor. The HNM was synthesised for a CSTR reactor with a radial basis function neural net (RBFN used to estimate the reaction rates recursively. The adaptive HNM was incorporated in two model predictive control strategies, a direct synthesis scheme and an optimum steady state scheme. Tests for servo and regulator control showed excellent behaviour following different setpoint variations, and rejecting perturbations. The good generalisation and training capacities of hybrid models, associated with the simplicity and robustness characteristics of the MPC formulations, make an attractive combination for the control of a polymerisation reactor.

  11. A Special Topic From Nuclear Reactor Dynamics for the Undergraduate Physics Curriculum

    Science.gov (United States)

    Sevenich, R. A.

    1977-01-01

    Presents an intuitive derivation of the point reactor equations followed by formulation of equations for inverse and direct kinetics which are readily programmed on a digital computer. Suggests several computer simulations involving the effect of control rod motion on reactor power. (MLH)

  12. Hydrogeological and Groundwater Flow Model for C, K, L, and P Reactor Areas, Savannah River Site, Aiken, South Carolina

    Energy Technology Data Exchange (ETDEWEB)

    Flach, G.P.

    1999-02-24

    A regional groundwater flow model encompassing approximately 100 mi{sup 2} surrounding the C, K. L. and P reactor areas has been developed. The Reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department.

  13. An experimental study of a VVER reactor's steam generator model operating in the condensing mode

    Science.gov (United States)

    Morozov, A. V.; Remizov, O. V.

    2012-05-01

    Results obtained from an experimental study of a VVER reactor's steam generator model operating in the condensing mode are presented. The obtained empirical dependence for calculating the power of heat exchangers operating in the steam condensation mode is presented.

  14. Modeling of a slurry bubble column reactor for Fischer-Tropsch synthesis

    Institute of Scientific and Technical Information of China (English)

    QIAN Wei-xin; MA Hong-fang; LI Tao; YING Wei-yong; FANG Ding-ye

    2012-01-01

    On the basis of the global CO consumption rate model,the lumped product distribution model and the sedimentation-dispersion model of a catalyst,a steady-state,one-dimensional mathematical model of the slurry bubble column reactor for Fischer-Tropsch synthesis were established.The mathematical simulation of the slurry bubble column reactor for Fischer-Tropsch synthesis was carried out under the following typical industrial operating conditions:temperature 230 ℃,pressure 3.0 MPa,gas flow 5× 105 m3/h,catalyst content in slurry phase 30%,reactor diameter 5.0 m and the composition of feed gas:y(H2)=0.60,y(CO)=0.30,y(N2)=0.10.The influences of operating pressure,temperature and m(H2)/m(CO) in feed gas on the reactor's reaction performance were simulated.

  15. Fluidized-bed reactor model with generalized particle balances. Part 1. Formulation and solution

    Energy Technology Data Exchange (ETDEWEB)

    Overturf, B.W.; Reklaitis, G.V.

    1983-09-01

    In this first part, a particle balance model is developed for a fluidized-bed gas-solid reactor which accommodates particle distributions dependent on both size and density, as well as populations consisting of multiple solids.

  16. Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

    Energy Technology Data Exchange (ETDEWEB)

    Lázaro, A., E-mail: aurelio.lazaro-chueca@ec.europa.eu [JRC-IET European Commission—Westerduinweg 3, PO Box-2, 1755 ZG Petten (Netherlands); UPV—Universidad Politecnica de Valencia, Cami de vera s/n-46002, Valencia (Spain); Ammirabile, L. [JRC-IET European Commission—Westerduinweg 3, PO Box-2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Massara, S. [EDF, 1 avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2-28006 Madrid (Spain); Mikityuk, K. [PSI—Paul Scherrer Institut, 5232 Villigen Switzerland (Switzerland); Schikorr, M.; Bubelis, E.; Ponomarev, A.; Kruessmann, R. [KIT—Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen Germany (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, PO Box 9034 6800 ES, Arnhem (Netherlands)

    2014-01-15

    Highlights: • Ten system-code models of the ESFR were developed in the frame of the CP-ESFR project. • Eight different thermohydraulic system codes adapted to sodium fast reactor's technology. • Benchmarking exercise settled to check the consistency of the calculations. • Upgraded system codes able to simulate the reactivity feedback and key safety parameters. -- Abstract: The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes.

  17. STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart

    2016-05-01

    With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

  18. Transient modeling of the thermohydraulic behavior of high temperature heat pipes for space reactor applications

    Science.gov (United States)

    Hall, Michael L.; Doster, Joseph M.

    1986-01-01

    Many proposed space reactor designs employ heat pipes as a means of conveying heat. Previous researchers have been concerned with steady state operation, but the transient operation is of interest in space reactor applications due to the necessity of remote startup and shutdown. A model is being developed to study the dynamic behavior of high temperature heat pipes during startup, shutdown and normal operation under space environments. Model development and preliminary results for a hypothetical design of the system are presented.

  19. Sensitivity Analysis on LOCCW of Westinghouse typed Reactors Considering WOG2000 RCP Seal Leakage Model

    Energy Technology Data Exchange (ETDEWEB)

    Na, Jang-Hwan; Jeon, Ho-Jun; Hwang, Seok-Won [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this paper, we focus on risk insights of Westinghouse typed reactors. We identified that Reactor Coolant Pump (RCP) seal integrity is the most important contributor to Core Damage Frequency (CDF). As we reflected the latest technical report; WCAP-15603(Rev. 1-A), 'WOG2000 RCP Seal Leakage Model for Westinghouse PWRs' instead of the old version, RCP seal integrity became more important to Westinghouse typed reactors. After Fukushima accidents, Korea Hydro and Nuclear Power (KHNP) decided to develop Low Power and Shutdown (LPSD) Probabilistic Safety Assessment (PSA) models and upgrade full power PSA models of all operating Nuclear Power Plants (NPPs). As for upgrading full power PSA models, we have tried to standardize the methodology of CCF (Common Cause Failure) and HRA (Human Reliability Analysis), which are the most influential factors to risk measures of NPPs. Also, we have reviewed and reflected the latest operating experiences, reliability data sources and technical methods to improve the quality of PSA models. KHNP has operating various types of reactors; Optimized Pressurized Reactor (OPR) 1000, CANDU, Framatome and Westinghouse. So, one of the most challengeable missions is to keep the balance of risk contributors of all types of reactors. This paper presents the method of new RCP seal leakage model and the sensitivity analysis results from applying the detailed method to PSA models of Westinghouse typed reference reactors. To perform the sensitivity analysis on LOCCW of the reference Westinghouse typed reactors, we reviewed WOG2000 RCP seal leakage model and developed the detailed event tree of LOCCW considering all scenarios of RCP seal failures. Also, we performed HRA based on the T/H analysis by using the leakage rates for each scenario. We could recognize that HRA was the sensitive contributor to CDF, and the RCP seal failure scenario of 182gpm leakage rate was estimated as the most important scenario.

  20. Lectures on Physics Beyond the Standard Model

    OpenAIRE

    Gripaios, Ben

    2015-01-01

    These four lectures, given at the British Universities Summer School in Theoretical Elementary Particle Physics (BUSSTEPP), held in 2014 in Southampton, are a brief introduction to a selection of current topics in physics Beyond the Standard Model.

  1. Lectures on Physics Beyond the Standard Model

    OpenAIRE

    Gripaios, Ben

    2015-01-01

    These four lectures, given at the British Universities Summer School in Theoretical Elementary Particle Physics (BUSSTEPP), held in 2014 in Southampton, are a brief introduction to a selection of current topics in physics Beyond the Standard Model.

  2. Modeling of a continuous pretreatment reactor using computational fluid dynamics.

    Science.gov (United States)

    Berson, R Eric; Dasari, Rajesh K; Hanley, Thomas R

    2006-01-01

    Computational fluid dynamic simulations are employed to predict flow characteristics in a continuous auger driven reactor designed for the dilute acid pretreatment of biomass. Slurry containing a high concentration of biomass solids exhibits a high viscosity, which poses unique mixing issues within the reactor. The viscosity increases significantly with a small increase in solids concentration and also varies with temperature. A well-mixed slurry is desirable to evenly distribute acid on biomass, prevent buildup on the walls of the reactor, and provides an uniform final product. Simulations provide flow patterns obtained over a wide range of viscosities and pressure distributions, which may affect reaction rates. Results provide a tool for analyzing sources of inconsistencies in product quality and insight into future design and operating parameters.

  3. Development of Subspace-based Hybrid Monte Carlo-Deterministric Algorithms for Reactor Physics Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Abdel-Khalik, Hany S. [North Carolina State Univ., Raleigh, NC (United States); Zhang, Qiong [North Carolina State Univ., Raleigh, NC (United States)

    2014-05-20

    The development of hybrid Monte-Carlo-Deterministic (MC-DT) approaches, taking place over the past few decades, have primarily focused on shielding and detection applications where the analysis requires a small number of responses, i.e. at the detector locations(s). This work further develops a recently introduced global variance reduction approach, denoted by the SUBSPACE approach is designed to allow the use of MC simulation, currently limited to benchmarking calculations, for routine engineering calculations. By way of demonstration, the SUBSPACE approach is applied to assembly level calculations used to generate the few-group homogenized cross-sections. These models are typically expensive and need to be executed in the order of 103 - 105 times to properly characterize the few-group cross-sections for downstream core-wide calculations. Applicability to k-eigenvalue core-wide models is also demonstrated in this work. Given the favorable results obtained in this work, we believe the applicability of the MC method for reactor analysis calculations could be realized in the near future.

  4. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

  5. Applications of Neutrino Physics

    OpenAIRE

    Christensen, Eric Kurt

    2014-01-01

    Neutrino physics has entered a precision era in which understanding backgrounds and systematic uncertainties is particularly important. With a precise understanding of neutrino physics, we can better understand neutrino sources. In this work, we demonstrate dependency of single detector oscillation experiments on reactor neutrino flux model. We fit the largest reactor neutrino flux model error, weak magnetism, using data from experiments. We use reactor burn-up simulations in combination with...

  6. Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

    Energy Technology Data Exchange (ETDEWEB)

    Turinsky, Paul J., E-mail: turinsky@ncsu.edu [North Carolina State University, PO Box 7926, Raleigh, NC 27695-7926 (United States); Kothe, Douglas B., E-mail: kothe@ornl.gov [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6164 (United States)

    2016-05-15

    The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics “core simulator” based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL

  7. Hydraulic characterization of an activated sludge reactor with recycling system by tracer experiment and analytical models.

    Science.gov (United States)

    Sánchez, F; Viedma, A; Kaiser, A S

    2016-09-15

    Fluid dynamic behaviour plays an important role in wastewater treatment. An efficient treatment requires the inexistence of certain hydraulic problems such as dead zones or short-circuiting flows. Residence time distribution (RTD) analysis is an excellent technique for detecting these inefficiencies. However, many wastewater treatment installations include water or sludge recycling systems, which prevent us from carrying out a conventional tracer pulse experiment to obtain the RTD curve of the installation. This paper develops an RTD analysis of an activated sludge reactor with recycling system. A tracer experiment in the reactor is carried out. Three analytical models, derived from the conventional pulse model, are proposed to obtain the RTD curve of the reactor. An analysis of the results is made, studying which model is the most suitable for each situation. This paper is useful to analyse the hydraulic efficiency of reactors with recycling systems.

  8. KINETIC MODELLING OF CONTINUOUS-MIX ANAEROBIC REACTORS OPERATING UNDER DIURNALLY CYCLIC TEMPERATURE ENVIRONMENT

    Directory of Open Access Journals (Sweden)

    E. A. Echiegu

    2014-01-01

    Full Text Available A two-culture dynamic model which incorporated the effects of diurnally cyclic temperature was developed and used to predict the dynamic response of anaerobic reactors operated on dairy manure under two diurnally cyclic temperature ranges of 20-40°C and 15-25°C which represent the summer and winter in Nigeria. The digesters were operated at various hydraulic retention times and solid concentrations and some useful kinetic parameters were determined. The model predicted biogas production, volatile solid reduction, methane yield and treatment efficiency with reasonable accuracy (R2 = 0.70 to 0.90. The model, however, under-predicted the cell mass concentration in the reactor probably because the Volatile Suspended Solid (VSS, which was used as the estimator of the actual cell mass concentration in the reactor, was not a good indicator of the active cell mass concentration in anaerobic reactors operating on dairy manure.

  9. KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Victor Hugo; Miassoedov, Alexei; Steinbrueck, M.; Tromm, W. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany)

    2016-05-15

    This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

  10. Supplemental Reactor Physics Calculations and Analysis of ELF Mk 1A Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Michael A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    These calculations supplement previous the reactor physics work evaluating the Enhanced Low Enriched Uranium (LEU) Fuel (ELF) Mk 1A element. This includes various additional comparisons between the current Highly Enriched Uranium (HEU) and LEU along with further characterization of the performance of the ELF fuel. The excess reactivity to be held down at BOC for ELF Mk 1A fuel is estimated to be approximately $2.75 greater than with HEU for a typical cycle. This is a combined effect of the absence of burnable poison in the ELF fuel and the reduced neck shim worth in LEU fuel compared to HEU. Burnable poison rods were conceptualized for use in the small B positions containing Gd2O3 absorber. These were shown to provide $2.37 of negative reactivity at BOC and to burn out in less than half of a cycle. The worth of OSCCs is approximately the same between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. This was evaluated by rotating all banks simultaneously. The safety rod worth is relatively unchanged between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. However, this should be reevaluated with different loadings. Neutron flux, both total and fast (>1 MeV), is either the same or reduced upon changing from HEU to ELF Mk 1A (LEU) fuels in the representative loading evaluated. This is consistent with the well-established trend of lower neutron fluxes for a given power in LEU than HEU.The IPT loop void reactivity is approximately the same or less positive with ELF Mk 1A (LEU) fuel than HEU in the representative loading evaluated.

  11. FN approximation of the solution to a singular integral equation of classical reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Ganapol, B.D. [Department of Aerospace and Mechanical Engineering, University of Arizona, AME Building, Tucson, AZ 85721 (United States)]. E-mail: ganapol@ame.arizona.edu

    2004-11-01

    The iterated FN method is applied to a singular integral equation arising from a classical problem of reactor physics to determine the distribution of fissile material giving a spatially uniform flux. The FN iterations are accelerated toward convergence through the Wynn-algorithm - but first - Happy Birthday 'Fast Eddie' Larsen Why do I refer to the well known, most proper and exquisitely accomplished Edward W. Larsen as 'Fast Eddie'. Well our story begins in a small back bar room in the lobby of one of Los Alamos' finest and most luxurious hotels. Two young men were having a transport theoretic discussion while they were engaged in a serious game of pool with monetary benefits going to the winner. In addition, the two were sipping their most favorite lavation in rather large quantities - one, a short stocky man with thinning hair, was sipping to forget the cost of his recent divorce, and the other, a shorter stockier man also with thinning hair, was drinking, well because he liked to drink and it just made him silly. As they continued their transport discussion, one stocky man turned to the other and said, 'I wonder what 'Fast Eddie' Larsen would say to our transport question'. The other stocky man immediately thought the 'Fast Eddie' reference was to Paul Newman who played 'Fast Eddie', an expert at applied particle transport theory (a pool player) in the movie the Hustler and asked if indeed this was the case. The first stocky man said 'No. I call everyone with the name Ed 'Fast Eddie' ' - and that's the story of how 'Fast Eddie' Larsen got his name. Happy 60th Ed and thanks for all the great transport theory - from one of your biggest fans.

  12. Update on Small Modular Reactors Dynamic System Modeling Tool: Web Application

    Energy Technology Data Exchange (ETDEWEB)

    Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cetiner, Sacit M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Batteh, John J [Modelon Corporation (Sweden); Tiller, Michael M. [Xogeny Corporation (United States)

    2015-01-01

    Previous reports focused on the development of component and system models as well as end-to-end system models using Modelica and Dymola for two advanced reactor architectures: (1) Advanced Liquid Metal Reactor and (2) fluoride high-temperature reactor (FHR). The focus of this report is the release of the first beta version of the web-based application for model use and collaboration, as well as an update on the FHR model. The web-based application allows novice users to configure end-to-end system models from preconfigured choices to investigate the instrumentation and controls implications of these designs and allows for the collaborative development of individual component models that can be benchmarked against test systems for potential inclusion in the model library. A description of this application is provided along with examples of its use and a listing and discussion of all the models that currently exist in the library.

  13. Modeling and temperature regulation of a thermally coupled reactor system via internal model control strategy

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.Y.; Coronella, C.J.; Bhadkamkar, A.S.; Seader, J.D. [Univ. of Utah, Salt Lake City, UT (United States). Dept. of Chemical and Fuels Engineering

    1993-12-01

    A two-stage, thermally coupled fluidized-bed reactor system has been developed for energy-efficient conversion of tar-sand bitumen to synthetic crude oil. Modeling and temperature control of a system are addressed in this study. A process model and transfer function are determined by a transient response technique and the reactor temperature are controlled by PI controllers with tuning settings determined by an internal model control (IMC) strategy. Using the IMC tuning method, sufficiently good control performance was experimentally observed without lengthy on-line tuning. It is shown that IMC strategy provides a means to directly use process knowledge to make a control decision. Although this control method allows for fine tuning by adjusting a single tuning parameter, it is not easy to determine the optimal value of this tuning parameter, which must be specified by the user. A novel method is presented to evaluate that parameter, which must be specified by the user. A novel method is presented to evaluate that parameter in this study. It was selected based on the magnitude of elements on the off-diagonal of the relative gain array to account for the effect of thermal coupling on control performance. It is shown that this method provides stable and fast control of reactor temperatures. By successfully decoupling the system, a simple method of extending the IMC tuning technique to multiinput/multioutput systems is obtained.

  14. V.S.O.P. (99) for WINDOWS and UNIX. Computer code system for reactor physics and fuel cycle simulation

    Energy Technology Data Exchange (ETDEWEB)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Scherer, W.

    2000-10-01

    V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. V.S.O.P.(99) represents the further development of V.S.O.P. (97). Compared to its precursor, the code system has been improved in many details. Major extensions have been included concerning the thermal hydraulic sections. Beyond that, the many modules of the code-system have been condensed to only 2 executables in the ''99''-release of V.S.O.P., to be comfortably handled on a WINDOWS-PC or a UNIX-computer. The necessary data input as well as the handling and book-keeping of intermediate data sets has been condensed and simplified. A 64 MB memory should be available for the execution of the code. The hard disk requirement for the executables and the basic libraries associated with the code amounts to about 7 MB. (orig.)

  15. Physical and economical aspects of Pu multiple recycling on the basis of REMIX reprocessing technology in thermal reactors

    Directory of Open Access Journals (Sweden)

    Teplov Pavel S.

    2016-01-01

    Full Text Available The basic strategy of Russian nuclear energy is propagation of a closed fuel cycle on the basis of fast breeder and thermal reactors, as well as the solution of the spent nuclear fuel accumulation and resource problems. The three variants of multiple Pu and U recycling in Russian pressurized water reactor concept reactors on the basis of REgenerated MIXture of U, Pu oxides (REMIX reprocessing technology are considered in this work. The REMIX fuel is fabricated from an unseparated mixture of uranium and plutonium obtained during spent fuel reprocessing with further makeup by enriched natural U or reactor grade Pu. This makes it possible to recycle several times the total amount of Pu obtained from the spent fuel. The main difference in Pu recycling is the concept of 100% or partial fuel loading of the core. The third variant is heterogeneous composition of enriched uranium and uranium–plutonium mixed oxide fuel pins in one fuel assembly. It should be noted that all fuel assemblies with Pu require the involvement of expensive technologies during manufacturing. These three variants of the full core loadings can be balanced on zero Pu accumulation in the cycle. The various physical and economical aspects of Pu and U multiple recycling in selected variants are observed in the given work.

  16. M and c'99 : Mathematics and computation, reactor physics and environmental analysis in nuclear applications, Madrid, September 27-30, 1999

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J. M.; Ahnert, C.; Cabellos, O.

    1999-07-01

    The international conference on mathematics and computation, reactor physics and environmental analysis in nuclear applications in the biennial topical meeting of the mathematics and computation division of the American Nuclear Society. (Author)

  17. Inverse modeling approach for evaluation of kinetic parameters of a biofilm reactor using tabu search.

    Science.gov (United States)

    Kumar, B Shiva; Venkateswarlu, Ch

    2014-08-01

    The complex nature of biological reactions in biofilm reactors often poses difficulties in analyzing such reactors experimentally. Mathematical models could be very useful for their design and analysis. However, application of biofilm reactor models to practical problems proves somewhat ineffective due to the lack of knowledge of accurate kinetic models and uncertainty in model parameters. In this work, we propose an inverse modeling approach based on tabu search (TS) to estimate the parameters of kinetic and film thickness models. TS is used to estimate these parameters as a consequence of the validation of the mathematical models of the process with the aid of measured data obtained from an experimental fixed-bed anaerobic biofilm reactor involving the treatment of pharmaceutical industry wastewater. The results evaluated for different modeling configurations of varying degrees of complexity illustrate the effectiveness of TS for accurate estimation of kinetic and film thickness model parameters of the biofilm process. The results show that the two-dimensional mathematical model with Edward kinetics (with its optimum parameters as mu(max)rho(s)/Y = 24.57, Ks = 1.352 and Ki = 102.36) and three-parameter film thickness expression (with its estimated parameters as a = 0.289 x 10(-5), b = 1.55 x 10(-4) and c = 15.2 x 10(-6)) better describes the biofilm reactor treating the industry wastewater.

  18. FBR for catalytic propylene polymerization: Controlled mixing and reactor modeling

    NARCIS (Netherlands)

    Meier, G.B.; Weickert, G.; Swaaij, van W.P.M.

    2002-01-01

    Particle mixing and segregation have been studied in a small-scale fluidized-bed reactor (FBR) under pressure. The solids mixing is relatively faster than the residence time of catalyst particles in the case of a polymerization process, but smaller particles accumulate in the upper zone. Semibatch p

  19. Computer-aided modeling framework – a generic modeling template for catalytic membrane fixed bed reactors

    DEFF Research Database (Denmark)

    Fedorova, Marina; Sin, Gürkan; Gani, Rafiqul

    2013-01-01

    This work focuses on development of computer-aided modeling framework. The framework is a knowledge-based system that is built on a generic modeling language and structured based on workflows for different general modeling tasks. The overall objective of this work is to support the model developers...... and users to generate and test models systematically, efficiently and reliably. In this way, development of products and processes can be faster, cheaper and very efficient. In this contribution, as part of the framework a generic modeling template for the systematic derivation of problem specific catalytic...... membrane fixed bed models is developed. The application of the modeling template is highlighted with a case study related to the modeling of a catalytic membrane reactor coupling dehydrogenation of ethylbenzene with hydrogenation of nitrobenzene....

  20. An investigation of sulfate production in clouds using a flow-through chemical reactor model approach

    Science.gov (United States)

    Hong, M. S.; Carmichael, G. R.

    1983-01-01

    A flow-through chemical reactor model is developed to describe the mass transfer and chemical processes that atmospheric gases undergo in clouds. The model includes the simultaneous absorption of SO2, NH3, O3, NO(x), HNO3, CO2 and H2O2, the accompanying dissociation and oxidation reactions in cloud water, considers electrical neutrality, and includes qualitative parameterization of cloud microphysics. The model is used to assess the importance of the oxidation reactions H2O2-S(IV), O3-S(IV), and S(IV)-Mn(2+) catalysis, and the effects of cloud parameters such as drop size, rain intensity, liquid water content, and updraft velocity. Both precipitating and nonprecipitating clouds are studied. Model results predict sulfate production rates varying from 3 percent/hr to 230 percent/hr. The actual rate is highly dependent on the chemical composition of the uptake air and the physical conditions of the cloud. Model results also show that both the H2O2 and the O3 oxidation reactions can be significant.

  1. COMPUTATIONAL AND EXPERIMENTAL MODELING OF SLURRY BUBBLE COLUMN REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Paul C.K. Lam; Isaac K. Gamwo; Dimitri Gidaspow

    2002-05-01

    The objective of this study was to develop a predictive experimentally verified computational fluid dynamics (CFD) model for gas-liquid-solid flow. A three dimensional transient computer code for the coupled Navier-Stokes equations for each phase was developed and is appended in this report. The principal input into the model is the viscosity of the particulate phase which was determined from a measurement of the random kinetic energy of the 800 micron glass beads and a Brookfield viscometer. The details are presented in the attached paper titled ''CFD Simulation of Flow and Turbulence in a Slurry Bubble Column''. This phase of the work is in press in a referred journal (AIChE Journal, 2002) and was presented at the Fourth International Conference on Multiphase Flow (ICMF 2001) in New Orleans, May 27-June 1, 2001 (Paper No. 909). The computed time averaged particle velocities and concentrations agree with Particle Image Velocimetry (PIV) measurements of velocities and concentrations, obtained using a combination of gamma-ray and X-ray densitometers, in a slurry bubble column, operated in the bubbly-coalesced fluidization regime with continuous flow of water. Both the experiment and the simulation show a down-flow of particles in the center of the column and up-flow near the walls and nearly uniform particle concentration. Normal and shear Reynolds stresses were constructed from the computed instantaneous particle velocities. The PIV measurement and the simulation produced instantaneous particle velocities. The PIV measurement and the simulation produced similar nearly flat horizontal profiles of turbulent kinetic energy of particles. To better understand turbulence we studied fluidization in a liquid-solid bed. This work was also presented at the Fourth International Conference on Multiphase Flow (ICMF 2001, Paper No. 910). To understand turbulence in risers, measurements were done in the IIT riser with 530 micron glass beads using a PIV

  2. PIK3CA mutations frequently coexist with RAS and BRAF mutations in patients with advanced cancers.

    Directory of Open Access Journals (Sweden)

    Filip Janku

    Full Text Available BACKGROUND: Oncogenic mutations of PIK3CA, RAS (KRAS, NRAS, and BRAF have been identified in various malignancies, and activate the PI3K/AKT/mTOR and RAS/RAF/MEK pathways, respectively. Both pathways are critical drivers of tumorigenesis. METHODS: Tumor tissues from 504 patients with diverse cancers referred to the Clinical Center for Targeted Therapy at MD Anderson Cancer Center starting in October 2008 were analyzed for PIK3CA, RAS (KRAS, NRAS, and BRAF mutations using polymerase chain reaction-based DNA sequencing. RESULTS: PIK3CA mutations were found in 54 (11% of 504 patients tested; KRAS in 69 (19% of 367; NRAS in 19 (8% of 225; and BRAF in 31 (9% of 361 patients. PIK3CA mutations were most frequent in squamous cervical (5/14, 36%, uterine (7/28, 25%, breast (6/29, 21%, and colorectal cancers (18/105, 17%; KRAS in pancreatic (5/9, 56%, colorectal (49/97, 51%, and uterine cancers (3/20, 15%; NRAS in melanoma (12/40, 30%, and uterine cancer (2/11, 18%; BRAF in melanoma (23/52, 44%, and colorectal cancer (5/88, 6%. Regardless of histology, KRAS mutations were found in 38% of patients with PIK3CA mutations compared to 16% of patients with wild-type (wtPIK3CA (p = 0.001. In total, RAS (KRAS, NRAS or BRAF mutations were found in 47% of patients with PIK3CA mutations vs. 24% of patients wtPIK3CA (p = 0.001. PIK3CA mutations were found in 28% of patients with KRAS mutations compared to 10% with wtKRAS (p = 0.001 and in 20% of patients with RAS (KRAS, NRAS or BRAF mutations compared to 8% with wtRAS (KRAS, NRAS or wtBRAF (p = 0.001. CONCLUSIONS: PIK3CA, RAS (KRAS, NRAS, and BRAF mutations are frequent in diverse tumors. In a wide variety of tumors, PIK3CA mutations coexist with RAS (KRAS, NRAS and BRAF mutations.

  3. Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C., E-mail: amir@cdtn.b, E-mail: souzarm@cdtn.b, E-mail: avf@cdtn.b, E-mail: ajp@cdtn.b, E-mail: aclc@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

  4. Report on Reactor Physics Assessment of Candidate Accident Tolerant Fuel Cladding Materials in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan [Univ. of Tennessee, Knoxville, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-28

    This work focuses on ATF concepts being researched at Oak Ridge National Laboratory (ORNL), expanding on previous studies of using alternate cladding materials in pressurized water reactors (PWRs). The neutronic performance of two leading alternate cladding materials were assessed in boiling water reactors (BWRs): iron-chromium-aluminum (FeCrAl) cladding, and silicon carbide (SiC)-based composite cladding. This report fulfills ORNL Milestone M3FT-15OR0202332 within the fiscal year 2015 (FY15)

  5. Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

    Energy Technology Data Exchange (ETDEWEB)

    Moses, David Lewis [ORNL

    2011-10-01

    This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be

  6. Integral Reactor Containment Condensation Model and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Qiao [Oregon State Univ., Corvallis, OR (United States); Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States)

    2016-05-02

    This NEUP funded project, NEUP 12-3630, is for experimental, numerical and analytical studies on high-pressure steam condensation phenomena in a steel containment vessel connected to a water cooling tank, carried out at Oregon State University (OrSU) and the University of Wisconsin at Madison (UW-Madison). In the three years of investigation duration, following the original proposal, the planned tasks have been completed: (1) Performed a scaling study for the full pressure test facility applicable to the reference design for the condensation heat transfer process during design basis accidents (DBAs), modified the existing test facility to route the steady-state secondary steam flow into the high pressure containment for controllable condensation tests, and extended the operations at negative gage pressure conditions (OrSU). (2) Conducted a series of DBA and quasi-steady experiments using the full pressure test facility to provide a reliable high pressure condensation database (OrSU). (3) Analyzed experimental data and evaluated condensation model for the experimental conditions, and predicted the prototypic containment performance under accidental conditions (UW-Madison). A film flow model was developed for the scaling analysis, and the results suggest that the 1/3 scaled test facility covers large portion of laminar film flow, leading to a lower average heat transfer coefficient comparing to the prototypic value. Although it is conservative in reactor safety analysis, the significant reduction of heat transfer coefficient (50%) could under estimate the prototypic condensation heat transfer rate, resulting in inaccurate prediction of the decay heat removal capability. Further investigation is thus needed to quantify the scaling distortion for safety analysis code validation. Experimental investigations were performed in the existing MASLWR test facility at OrST with minor modifications. A total of 13 containment condensation tests were conducted for pressure

  7. Liquid biopsy of PIK3CA mutations in cervical cancer in Hong Kong Chinese women.

    Science.gov (United States)

    Chung, Tony K H; Cheung, Tak Hong; Yim, So Fan; Yu, Mei Yun; Chiu, Rossa W K; Lo, Keith W K; Lee, Ida P C; Wong, Raymond R Y; Lau, Kitty K M; Wang, Vivian W; Worley, Michael J; Elias, Kevin M; Fiascone, Stephen J; Smith, David I; Berkowitz, Ross S; Wong, Yick Fu

    2017-08-01

    Cervical cancer is the fourth most common female cancer worldwide. The prognosis for women with advanced-stage or recurrent cervical cancer remains poor and response to treatment is variable. Standardized management protocols leave little room for individualization. We report on a novel blood-based liquid biopsy for specific PIK3CA mutations as a clinically useful biomarker in patients with invasive cervical cancer. One hundred seventeen Hong Kong Chinese women with primary invasive cervical cancer and their pre-treatment plasma samples were investigated. Two PIK3CA mutations, p.E542K and p.E545K were measured in cell free DNA (cfDNA) extracted from plasma using droplet digital PCR. This liquid biopsy of PIK3CA in cervical cancer was correlated to clinico-pathological features to verify the potential of PIK3CA as a clinically useful molecular biomarker for predicting disease prognosis and monitoring for progression. PIK3CA mutations, either p.E542K or p.E545K, were detected in plasma cfDNA from 22.2% of the patients. PIK3CA mutation status was significantly correlated to median tumor size (p<0.01). PIK3CA mutations detected in the plasma were significantly associated with decreased disease-free survival and overall survival (p<0.05). As a liquid molecular biopsy, analysis of circulating PIK3CA mutations shows promise as a way to refine risk stratification of individual patients with cervical cancer, and provides a platform for further research to offer individualized therapy with the purpose of improving outcomes. Copyright © 2017 Elsevier Inc. All rights reserved.

  8. Fluid modeling and design of gas channels of solar non-stoichiometric redox reactor

    Science.gov (United States)

    Kedlaya, Aditya

    The present numerical study in FLUENT analyzes the fluid flow field within a solar powered reactor designed for syngas production. The thermochemical reactor is based on continuous cycling of cerium oxide (ceria) in a non-stoichiometric reduction/oxidation cycle. The reactor uses a hollow cylinder of porous ceria which rotates through a high-temperature zone, by exposure to concentrated sunlight and partially reduced in an inert atmosphere due to flow of the sweep gas (N2), and then through a lower temperature zone where the reduced ceria is re-oxidized with a flow of CO2 and/or H2O, to produce CO and/or H2. In terms of fluid flow modeling, the issue of crossover of species (leakage) within the reactor is critical for proper functioning of the reactor. The first part of the work relates to the geometry and placement of the inlet/outlet gas channels for the reactor optimized to minimize crossover of the species. This is done by conducting a parametric study of geometric variables associated with the inlet/outlet geometry. A simplified 2D fluid flow reactor model which incorporates multi-species flow is used for this study. Further, 2D and 3D reactor models which capture the internal structure more accurately are used to refine the inlet/outlet design. The optimized reactor model is found to have an O2 crossover of 2%-6% and oxidizer crossover of 8%-21% at different flow rates of the sweep gas and the oxidizer studied. In the second part of the work, the reactor model is simulated under varying test conditions. Different working conditions include morphologies of the reactive material, rotational speed of the ceria ring and the recuperator, flow rates of sweep gas and the oxidizer, types of oxidizer (CO2, H2O). The 3D reactor model is also tested using one, two and three discrete inlet/outlet ports and compared with slot configuration.

  9. CO2 Absorption in a Lab-Scale Fixed Solid Bed Reactor: Modelling and Experimental Tests

    Directory of Open Access Journals (Sweden)

    Roberto Gabbrielli

    2004-09-01

    Full Text Available The CO2 absorption in a lab-scale fixed solid bed reactor filled with different solid sorbents has been studied under different operative conditions regarding temperature (20-200°C and input gas composition (N2, O2, CO2, H2O at 1bar pressure. The gas leaving the reactor has been analysed to measure the CO2 and O2 concentrations and, consequently, to evaluate the overall CO2 removal efficiency. In order to study the influence of solid sorbent type (i.e. CaO, coal bottom ash, limestone and blast furnace slag and of mass and heat transfer processes on CO2 removal efficiency, a one-dimensional time dependent mathematical model of the reactor, which may be considered a Plug Flow Reactor, has been developed. The quality of the model has been confirmed using the experimental results.

  10. Dynamic Modeling for the Design and Cyclic Operation of an Atomic Layer Deposition (ALD Reactor

    Directory of Open Access Journals (Sweden)

    Curtisha D. Travis

    2013-08-01

    Full Text Available A laboratory-scale atomic layer deposition (ALD reactor system model is derived for alumina deposition using trimethylaluminum and water as precursors. Model components describing the precursor thermophysical properties, reactor-scale gas-phase dynamics and surface reaction kinetics derived from absolute reaction rate theory are integrated to simulate the complete reactor system. Limit-cycle solutions defining continuous cyclic ALD reactor operation are computed with a fixed point algorithm based on collocation discretization in time, resulting in an unambiguous definition of film growth-per-cycle (gpc. A key finding of this study is that unintended chemical vapor deposition conditions can mask regions of operation that would otherwise correspond to ideal saturating ALD operation. The use of the simulator for assisting in process design decisions is presented.

  11. KRAS, BRAF and PIK3CA status in squamous cell anal carcinoma (SCAC.

    Directory of Open Access Journals (Sweden)

    Andrea Casadei Gardini

    Full Text Available Anti-EGFR therapy appears to be a potential treatment option for squamous cell anal carcinoma (SCAC. KRAS mutation is a rare event in SCAC, indicating the absence of the principal mechanism of resistance to this type of therapy. However, no information is available from the literature regarding the status of BRAF or PIK3CA in this cancer type. We analysed KRAS, BRAF and PIK3CA status in SCAC patients in relation to the clinical-pathological characteristics of patients and to the presence of the human papilloma virus (HPV. One hundred and three patients were treated with the Nigro scheme for anal cancer from March 2001 to August 2012. Fifty patients were considered for the study as there was insufficient paraffin-embedded tumour tissue to perform molecular analysis the remaining 53. DNA was extracted from paraffin-embedded sections. KRAS, BRAF and PIK3CA gene status and HPV genotype were evaluated by pyrosequencing. KRAS and BRAF genes were wild-type in all cases. Conversely, PIK3CA gene was found to be mutated in 11 (22% cases. In particular, 8 mutations occurred in exon 9 and 3 in exon 20 of the PIK3CA gene. These findings suggest that SCAC could potentially respond to an anti-EGFR drug. PIK3CA mutation may be involved in the process of carcinogenesis in some cases of SCAC.

  12. Evaluating a Model of Youth Physical Activity

    Science.gov (United States)

    Heitzler, Carrie D.; Lytle, Leslie A.; Erickson, Darin J.; Barr-Anderson, Daheia; Sirard, John R.; Story, Mary

    2010-01-01

    Objective: To explore the relationship between social influences, self-efficacy, enjoyment, and barriers and physical activity. Methods: Structural equation modeling examined relationships between parent and peer support, parent physical activity, individual perceptions, and objectively measured physical activity using accelerometers among a…

  13. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  14. A methodology for modeling photocatalytic reactors for indoor pollution control using previously estimated kinetic parameters

    Energy Technology Data Exchange (ETDEWEB)

    Passalia, Claudio; Alfano, Orlando M. [INTEC - Instituto de Desarrollo Tecnologico para la Industria Quimica, CONICET - UNL, Gueemes 3450, 3000 Santa Fe (Argentina); FICH - Departamento de Medio Ambiente, Facultad de Ingenieria y Ciencias Hidricas, Universidad Nacional del Litoral, Ciudad Universitaria, 3000 Santa Fe (Argentina); Brandi, Rodolfo J., E-mail: rbrandi@santafe-conicet.gov.ar [INTEC - Instituto de Desarrollo Tecnologico para la Industria Quimica, CONICET - UNL, Gueemes 3450, 3000 Santa Fe (Argentina); FICH - Departamento de Medio Ambiente, Facultad de Ingenieria y Ciencias Hidricas, Universidad Nacional del Litoral, Ciudad Universitaria, 3000 Santa Fe (Argentina)

    2012-04-15

    Highlights: Black-Right-Pointing-Pointer Indoor pollution control via photocatalytic reactors. Black-Right-Pointing-Pointer Scaling-up methodology based on previously determined mechanistic kinetics. Black-Right-Pointing-Pointer Radiation interchange model between catalytic walls using configuration factors. Black-Right-Pointing-Pointer Modeling and experimental validation of a complex geometry photocatalytic reactor. - Abstract: A methodology for modeling photocatalytic reactors for their application in indoor air pollution control is carried out. The methodology implies, firstly, the determination of intrinsic reaction kinetics for the removal of formaldehyde. This is achieved by means of a simple geometry, continuous reactor operating under kinetic control regime and steady state. The kinetic parameters were estimated from experimental data by means of a nonlinear optimization algorithm. The second step was the application of the obtained kinetic parameters to a very different photoreactor configuration. In this case, the reactor is a corrugated wall type using nanosize TiO{sub 2} as catalyst irradiated by UV lamps that provided a spatially uniform radiation field. The radiative transfer within the reactor was modeled through a superficial emission model for the lamps, the ray tracing method and the computation of view factors. The velocity and concentration fields were evaluated by means of a commercial CFD tool (Fluent 12) where the radiation model was introduced externally. The results of the model were compared experimentally in a corrugated wall, bench scale reactor constructed in the laboratory. The overall pollutant conversion showed good agreement between model predictions and experiments, with a root mean square error less than 4%.

  15. PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1979-10-01

    The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined.

  16. Air purification in a reverse-flow reactor: Model simulations vs. experiments

    OpenAIRE

    Beld, van de, L.; Westerterp, K.R.

    1996-01-01

    The behavior of a reverse-flow reactor was studied for the purification of polluted air by catalytic combustion. A heterogeneous one-dimensional model was extended with a heat balance for the reactor wall. An overall heat transport term is included to account for the small heat losses in radial direction. The calculations are compared to experimental data without using fit parameters. The agreement between simulations and experiments is generally good. Discrepancies can be explained mainly by...

  17. Modeling of natural circulation for the inherent safety analysis of sodium cooled fast reactors

    Directory of Open Access Journals (Sweden)

    A.S. Bochkarev

    2016-12-01

    Full Text Available The paper discusses a set of developed integrated one-dimensional models of thermal-hydraulic processes that contribute to the removal of decay heat in a BN-type reactor. The assumptions and constraints involved in one-dimensional equations of unsteady natural convection in closed circuits have been analyzed. It has been shown that the calculated values of the primary circuit sodium temperature and flow rate in conditions with a loss of heat sink and with a forced circulation of the primary coolant are in a reasonable agreement with the results of a benchmark experiment in the PHENIX reactor. The model makes it possible to assess the effects general thermophysical and geometrical parameters and the selected technology have on the efficiency of passive heat removal by the natural coolant convection in the reactor tank and in the emergency heat removal system's intermediate circuit and by the heat transfer through the reactor vessel. The model is a part of an integrated algorithm used to assess the inherent safety level of advanced fast neutron reactors and is intended primarily to develop, at the early conceptual design stage, the recommendations and requirements with respect to the reactor equipment parameters leading to an increase in the reactor inherent safety. The model will be used to identify the set of quantitative thermal-hydraulic criteria that have an effect on the dynamics of emergency transients leading to a potential loss of integrity by the reactor safety barriers, and to formulate such limits for the defined criteria as would cause, if observed, the requirement for the safety barrier integrity to be met under any combination of the accident initiating events.

  18. Reactor models for a series of continuous stirred tank reactors with a gas-liquid-solid leaching system: Part III. Model application

    Science.gov (United States)

    Papangelakis, V. G.; Demopoulos, G. P.

    1992-12-01

    A mathematical model developed to describe the steady-state performance of a three-phase leaching reactor is applied to the analysis and simulation of an industrial process: the high-temperature (180 °C to 200 °C) aqueous pressure oxidation (O2-H2SO4) of refractory pyrite-arsenopyrite (FeS2-FeAsS) gold concentrates. The simulation work reported here centers on the analysis of the autothermal operation of a continuous multistage horizontal autoclave. The focus is on the performance of the first autoclave compartment, since its autothermal “initialization” determines the rate of the whole process. The analysis of the whole autoclave is subsequently done on a stage-by-stage basis. The model considers both possible reaction control regimes, that is, reactor operation limited by the rate of the particle dissolution reaction (surface reaction control) or limited by the rate of O2 transfer at the g-1 interface (gas-transfer control). The decision whether the reactor operates under surface reaction control or gas transfer control is based on whether the gas-transfer capacity of the reactor can or cannot satisfy the oxygen demands of the leaching reactions. With the aid of the model, the effects of feed rate, feed preheating, cooling with water injection, slurry recycling, and autoclave configuration are critically evaluated from the standpoint of optimum autoclave performance.

  19. Development of an Integrated Performance Model for TRISO-Coated Gas Reactor Particle Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew; Miller, Gregory Kent; Martin, David George; Maki, John Thomas

    2005-05-01

    The success of gas reactors depends upon the safety and quality of the coated particle fuel. The understanding and evaluation of this fuel requires development of an integrated mechanistic fuel performance model that fully describes the mechanical and physico-chemical behavior of the fuel particle under irradiation. Such a model, called PARFUME (PARticle Fuel ModEl), is being developed at the Idaho National Engineering and Environmental Laboratory. PARFUME is based on multi-dimensional finite element modeling of TRISO-coated gas reactor fuel. The goal is to represent all potential failure mechanisms and to incorporate the statistical nature of the fuel. The model is currently focused on carbide, oxide nd oxycarbide uranium fuel kernels, while the coating layers are the classical IPyC/SiC/OPyC. This paper reviews the current status of the mechanical aspects of the model and presents results of calculations for irradiations from the New Production Modular High Temperature Gas Reactor program.

  20. Reactor based plutonium disposition - physics and fuel behaviour benchmark studies of an OECD/NEA experts group

    Energy Technology Data Exchange (ETDEWEB)

    D' Hondt, P. [SCK.CEN, Mol (Belgium); Gehin, J. [ORNL, Oak Ridge, TN (United States); Na, B.C.; Sartori, E. [Organisation for Economic Co-Operation and Development, Nuclear Energy Agency, 92 - Issy les Moulineaux (France); Wiesenack, W. [Organisation for Economic Co-Operation and Development/HRP, Halden (Norway)

    2001-07-01

    One of the options envisaged for disposing of weapons grade plutonium, declared surplus for national defence in the Russian Federation and Usa, is to burn it in nuclear power reactors. The scientific/technical know-how accumulated in the use of MOX as a fuel for electricity generation is of great relevance for the plutonium disposition programmes. An Expert Group of the OECD/Nea is carrying out a series of benchmarks with the aim of facilitating the use of this know-how for meeting this objective. This paper describes the background that led to establishing the Expert Group, and the present status of results from these benchmarks. The benchmark studies cover a theoretical reactor physics benchmark on a VVER-1000 core loaded with MOX, two experimental benchmarks on MOX lattices and a benchmark concerned with MOX fuel behaviour for both solid and hollow pellets. First conclusions are outlined as well as future work. (author)

  1. Nonlinear Dynamic Modeling and Simulation of a Passively Cooled Small Modular Reactor

    Science.gov (United States)

    Arda, Samet Egemen

    A nonlinear dynamic model for a passively cooled small modular reactor (SMR) is developed. The nuclear steam supply system (NSSS) model includes representations for reactor core, steam generator, pressurizer, hot leg riser and downcomer. The reactor core is modeled with the combination of: (1) neutronics, using point kinetics equations for reactor power and a single combined neutron group, and (2) thermal-hydraulics, describing the heat transfer from fuel to coolant by an overall heat transfer resistance and single-phase natural circulation. For the helical-coil once-through steam generator, a single tube depiction with time-varying boundaries and three regions, i.e., subcooled, boiling, and superheated, is adopted. The pressurizer model is developed based upon the conservation of fluid mass, volume, and energy. Hot leg riser and downcomer are treated as first-order lags. The NSSS model is incorporated with a turbine model which permits observing the power with given steam flow, pressure, and enthalpy as input. The overall nonlinear system is implemented in the Simulink dynamic environment. Simulations for typical perturbations, e.g., control rod withdrawal and increase in steam demand, are run. A detailed analysis of the results show that the steady-state values for full power are in good agreement with design data and the model is capable of predicting the dynamics of the SMR. Finally, steady-state control programs for reactor power and pressurizer pressure are also implemented and their effect on the important system variables are discussed.

  2. Regional groundwater flow model for C, K. L. and P reactor areas, Savannah River Site, Aiken, SC

    Energy Technology Data Exchange (ETDEWEB)

    Flach, G.P.

    2000-02-11

    A regional groundwater flow model encompassing approximately 100 mi2 surrounding the C, K, L, and P reactor areas has been developed. The reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department. The model provides a quantitative understanding of groundwater flow on a regional scale within the near surface aquifers and deeper semi-confined to confined aquifers. The model incorporates historical and current field characterization data up through Spring 1999. Model preprocessing is automated so that future updates and modifications can be performed quickly and efficiently. The CKLP regional reactor model can be used to guide characterization, perform scoping analyses of contaminant transport, and serve as a common base for subsequent finer-scale transport and remedial/feasibility models for each reactor area.

  3. Modelling Homogeneous Nucleation in Sodium Fast Reactors under BDBA Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, M.; Herranz, L. E.; Kissane, M.

    2014-07-01

    During postulated Beyond Design Basis Accidents (BDBAs) in Sodium-cooled Fast Reactors (SFRs), the contaminated coolant discharge at high temperature into the containment is considered as a potential scenario during the severe accident progression. In this scenario, the vaporization of sodium and its subsequent combustion (oxidation) would result in supersaturated sodium oxide vapours and formation of large quantities of contaminated aerosols by nucleation of these combustion products. (Author)

  4. An approach to model reactor core nodalization for deterministic safety analysis

    Science.gov (United States)

    Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

    2016-01-01

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  5. An approach to model reactor core nodalization for deterministic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my [Nuclear Energy Department, Regulatory Economics & Planning Division, Tenaga Nasional Berhad (Malaysia); Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my [Prototypes & Plant Development Center, Malaysian Nuclear Agency (Malaysia); Roslan, Ridha, E-mail: ridha@aelb.gov.my; Sadri, Abd Aziz [Nuclear Installation Divisions, Atomic Energy Licensing Board (Malaysia); Farid, Mohd Fairus Abd [Reactor Technology Center, Malaysian Nuclear Agency (Malaysia)

    2016-01-22

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  6. Modeling, simulation, and analysis of a reactor system for the generation of white liquor of a pulp and paper industry

    Directory of Open Access Journals (Sweden)

    Ricardo Andreola

    2011-02-01

    Full Text Available An industrial system for the production of white liquor of a pulp and paper industry, Klabin Paraná Papéis, formed by ten reactors was modeled, simulated, and analyzed. The developed model considered possible water losses by the evaporation and reaction, in addition to variations in the volumetric flow of lime mud across the reactors due to the composition variations. The model predictions agreed well with the process measurements at the plant and the results showed that the slaking reaction was nearly complete at the third causticizing reactor, while causticizing ends by the seventh reactor. Water loss due to slaking reaction and evaporation occurred more pronouncedly in the slaker reactor than in the final causticizing reactors; nevertheless, the lime mud flow remained nearly constant across the reactors.

  7. Modeling and Design Optimization of Multifunctional Membrane Reactors for Direct Methane Aromatization.

    Science.gov (United States)

    Fouty, Nicholas J; Carrasco, Juan C; Lima, Fernando V

    2017-08-29

    Due to the recent increase of natural gas production in the U.S., utilizing natural gas for higher-value chemicals has become imperative. Direct methane aromatization (DMA) is a promising process used to convert methane to benzene, but it is limited by low conversion of methane and rapid catalyst deactivation by coking. Past work has shown that membrane separation of the hydrogen produced in the DMA reactions can dramatically increase the methane conversion by shifting the equilibrium toward the products, but it also increases coke production. Oxygen introduction into the system has been shown to inhibit this coke production while not inhibiting the benzene production. This paper introduces a novel mathematical model and design to employ both methods in a multifunctional membrane reactor to push the DMA process into further viability. Multifunctional membrane reactors, in this case, are reactors where two different separations occur using two differently selective membranes, on which no systems studies have been found. The proposed multifunctional membrane design incorporates a hydrogen-selective membrane on the outer wall of the reaction zone, and an inner tube filled with airflow surrounded by an oxygen-selective membrane in the middle of the reactor. The design is shown to increase conversion via hydrogen removal by around 100%, and decrease coke production via oxygen addition by 10% when compared to a tubular reactor without any membranes. Optimization studies are performed to determine the best reactor design based on methane conversion, along with coke and benzene production. The obtained optimal design considers a small reactor (length = 25 cm, diameter of reaction tube = 0.7 cm) to subvert coke production and consumption of the product benzene as well as a high permeance (0.01 mol/s·m²·atm(1/4)) through the hydrogen-permeable membrane. This modeling and design approach sets the stage for guiding further development of multifunctional membrane reactor

  8. A Well-Posed Two Phase Flow Model and its Numerical Solutions for Reactor Thermal-Fluids Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kadioglu, Samet Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Berry, Ray [Idaho National Lab. (INL), Idaho Falls, ID (United States); Martineau, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    A 7-equation two-phase flow model and its numerical implementation is presented for reactor thermal-fluids applications. The equation system is well-posed and treats both phases as compressible flows. The numerical discretization of the equation system is based on the finite element formalism. The numerical algorithm is implemented in the next generation RELAP-7 code (Idaho National Laboratory (INL)’s thermal-fluids code) built on top of an other INL’s product, the massively parallel multi-implicit multi-physics object oriented code environment (MOOSE). Some preliminary thermal-fluids computations are presented.

  9. Critical review of the reactor-safety study radiological health effects model. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, D.W.; Evans, J.S.; Jacob, N.; Kase, K.R.; Maletskos, C.J.; Robertson, J.B.; Smith, D.G.

    1983-03-01

    This review of the radiological health effects models originally presented in the Reactor Safety Study (RSS) and currently used by the US Nuclear Regulatory Commission (NRC) was undertaken to assist the NRC in determining whether or not to revise the models and to aid in the revision, if undertaken. The models as presented in the RSS and as implemented in the CRAC (Calculations of Reactor Accident Consequences) Code are described and critiqued. The major elements analyzed are those concerning dosimetry, early effects, and late effects. The published comments on the models are summarized, as are the important findings since the publication of the RSS.

  10. A simple dynamic model and transient simulation of the nuclear power reactor on microcomputers

    Energy Technology Data Exchange (ETDEWEB)

    Han, Yang Gee; Park, Cheol [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    A simple dynamic model is developed for the transient simulation of the nuclear power reactor. The dynamic model includes the normalized neutron kinetics model with reactivity feedback effects and the core thermal-hydraulics model. The main objective of this paper demonstrates the capability of the developed dynamic model to simulate various important variables of interest for a nuclear power reactor transient. Some representative results of transient simulations show the expected trends in all cases, even though no available data for comparison. In this work transient simulations are performed on a microcomputer using the DESIRE/N96T continuous system simulation language which is applicable to nuclear power reactor transient analysis. 3 refs., 9 figs. (Author)

  11. Incorporation of Reaction Kinetics into a Multiphase, Hydrodynamic Model of a Fischer Tropsch Slurry Bubble Column Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna Guillen, PhD; Anastasia Gribik; Daniel Ginosar, PhD; Steven P. Antal, PhD

    2008-11-01

    This paper describes the development of a computational multiphase fluid dynamics (CMFD) model of the Fischer Tropsch (FT) process in a Slurry Bubble Column Reactor (SBCR). The CMFD model is fundamentally based which allows it to be applied to different industrial processes and reactor geometries. The NPHASE CMFD solver [1] is used as the robust computational platform. Results from the CMFD model include gas distribution, species concentration profiles, and local temperatures within the SBCR. This type of model can provide valuable information for process design, operations and troubleshooting of FT plants. An ensemble-averaged, turbulent, multi-fluid solution algorithm for the multiphase, reacting flow with heat transfer was employed. Mechanistic models applicable to churn turbulent flow have been developed to provide a fundamentally based closure set for the equations. In this four-field model formulation, two of the fields are used to track the gas phase (i.e., small spherical and large slug/cap bubbles), and the other two fields are used for the liquid and catalyst particles. Reaction kinetics for a cobalt catalyst is based upon values reported in the published literature. An initial, reaction kinetics model has been developed and exercised to demonstrate viability of the overall solution scheme. The model will continue to be developed with improved physics added in stages.

  12. Incorporation of Reaction Kinetics into a Multiphase, Hydrodynamic Model of a Fischer Tropsch Slurry Bubble Column Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna Guillen, PhD; Anastasia Gribik; Daniel Ginosar, PhD; Steven P. Antal, PhD

    2008-11-01

    This paper describes the development of a computational multiphase fluid dynamics (CMFD) model of the Fischer Tropsch (FT) process in a Slurry Bubble Column Reactor (SBCR). The CMFD model is fundamentally based which allows it to be applied to different industrial processes and reactor geometries. The NPHASE CMFD solver [1] is used as the robust computational platform. Results from the CMFD model include gas distribution, species concentration profiles, and local temperatures within the SBCR. This type of model can provide valuable information for process design, operations and troubleshooting of FT plants. An ensemble-averaged, turbulent, multi-fluid solution algorithm for the multiphase, reacting flow with heat transfer was employed. Mechanistic models applicable to churn turbulent flow have been developed to provide a fundamentally based closure set for the equations. In this four-field model formulation, two of the fields are used to track the gas phase (i.e., small spherical and large slug/cap bubbles), and the other two fields are used for the liquid and catalyst particles. Reaction kinetics for a cobalt catalyst is based upon values reported in the published literature. An initial, reaction kinetics model has been developed and exercised to demonstrate viability of the overall solution scheme. The model will continue to be developed with improved physics added in stages.

  13. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Devin A. Steuhm

    2011-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose. Furthermore

  14. Mathematical modeling of a slurry reactor for DME direct synthesis from syngas

    Institute of Scientific and Technical Information of China (English)

    Sadegh Papari; Mohammad Kazemeini; Moslem Fattahi

    2012-01-01

    In this paper,an axial dispersion mathematical model is developed to simulate a three-phase slurry bubble column reactor for direct synthesis of dimethyl ether (DME) from syngas.This large-scale reactor is modeled using mass and energy balances,catalyst sedimentation and single-bubble as well as two-bubbles class flow hydrodynamics.A comparison between the two hydrodynamic models through pilot plant experimental data from the literature shows that heterogeneous two-bubbles flow model is in better agreement with the experimental data than homogeneous single-bubble gas flow model.Also,by investigating the heterogeneous gas flow and axial dispersion model for small bubbles as well as the large bubbles and slurry (i.e.including paraffins and the catalyst) phase,the temperature profile along the reactor is obtained.A comparison between isothermal and non-isothermal reactors reveals no obvious performance difference between them.The optimum values of reactor diameter and height were obtained at 7 m and 50 m,respectively.The effects of operating variables on the axial catalyst distribution,DME productivity and CO conversion are also investigated in this research.

  15. Process development and modeling of fluidized-bed reactor with coimmobilized biocatalyst for fuel ethanol production

    Science.gov (United States)

    Sun, May Yongmei

    was 25--44g/L-hr (based on reactor volume), the average yield was 0.45 g ethanol/g starch, the biocatalyst retained physical integrity and contamination did not affect fermentation. For the Z. mobilis system the maximum volumetric productivity was 38 g ethanol/L-h, the average yield was 0.51 g ethanol/g starch and the FBR was successfully operated for almost one month. In order to develop, scale-up and economically evaluate this system more efficiently, a predictive mathematical model that is based on fundamental principles was developed and verified. This model includes kinetics of reactions, transport phenomena of the reactant and product by diffusion within the biocatalyst bead, and the hydrodynamics of the three phase fluidized bed. The co-immobilized biocatalyst involves a consecutive reaction mechanism The mathematical descriptions of the effectiveness factors of reactant and the intermediate product were developed. Hydrodynamic literature correlations were used to develop the dispersion coefficient and gas, liquid, and solid holdup. The solutions of coupled non-linear second order equations for biocatalyst bead and reactor together with the boundary conditions were solved numerically. This model gives considerable information about the system, such as concentration profiles inside both the beads and column, flow rate and feed concentration influences on productivity and phase hold up, and the influence of enzyme and cell mass loading in the catalyst. This model is generic in nature such that it can be easily applied to a diverse set of applications and operating conditions.

  16. CO2 conversion by plasma technology: insights from modeling the plasma chemistry and plasma reactor design

    Science.gov (United States)

    Bogaerts, A.; Berthelot, A.; Heijkers, S.; Kolev, St.; Snoeckx, R.; Sun, S.; Trenchev, G.; Van Laer, K.; Wang, W.

    2017-06-01

    In recent years there has been growing interest in the use of plasma technology for CO2 conversion. To improve this application, a good insight into the underlying mechanisms is of great importance. This can be obtained from modeling the detailed plasma chemistry in order to understand the chemical reaction pathways leading to CO2 conversion (either in pure form or mixed with another gas). Moreover, in practice, several plasma reactor types are being investigated for CO2 conversion, so in addition it is essential to be able to model these reactor geometries so that their design can be improved, and the most energy efficient CO2 conversion can be achieved. Modeling the detailed plasma chemistry of CO2 conversion in complex reactors is, however, very time-consuming. This problem can be overcome by using a combination of two different types of model: 0D chemical reaction kinetics models are very suitable for describing the detailed plasma chemistry, while the characteristic features of different reactor geometries can be studied by 2D or 3D fluid models. In the first instance the latter can be developed in argon or helium with a simple chemistry to limit the calculation time; however, the ultimate aim is to implement the more complex CO2 chemistry in these models. In the present paper, examples will be given of both the 0D plasma chemistry models and the 2D and 3D fluid models for the most common plasma reactors used for CO2 conversion in order to emphasize the complementarity of both approaches. Furthermore, based on the modeling insights, the paper discusses the possibilities and limitations of plasma-based CO2 conversion in different types of plasma reactors, as well as what is needed to make further progress in this field.

  17. A Parametric Sizing Model for Molten Regolith Electrolysis Reactors to Produce Oxygen from Lunar Regolith

    Science.gov (United States)

    Schreiner, Samuel S.; Dominguez, Jesus A.; Sibille, Laurent; Hoffman, Jeffrey A.

    2015-01-01

    We present a parametric sizing model for a Molten Electrolysis Reactor that produces oxygen and molten metals from lunar regolith. The model has a foundation of regolith material properties validated using data from Apollo samples and simulants. A multiphysics simulation of an MRE reactor is developed and leveraged to generate a vast database of reactor performance and design trends. A novel design methodology is created which utilizes this database to parametrically design an MRE reactor that 1) can sustain the required mass of molten regolith, current, and operating temperature to meet the desired oxygen production level, 2) can operate for long durations via joule heated, cold wall operation in which molten regolith does not touch the reactor side walls, 3) can support a range of electrode separations to enable operational flexibility. Mass, power, and performance estimates for an MRE reactor are presented for a range of oxygen production levels. The effects of several design variables are explored, including operating temperature, regolith type/composition, batch time, and the degree of operational flexibility.

  18. Modeling Cyber Physical War Gaming

    Science.gov (United States)

    2017-08-07

    Prepare physical facilities, means of communication , and paper or computer -based products to conduct the game. • Play: Assemble all cells and begin......estimated to average 1 hour per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the

  19. Time-variability of alpha from realistic models of Oklo reactors

    Science.gov (United States)

    Gould, Chris; Sharapov, Eduard; Lamoreaux, Steve

    2006-10-01

    We reanalyze Oklo ^149Sm data using realistic models of the natural nuclear reactors. Disagreements among recent Oklo determinations of the time evolution of α, the electromagnetic fine structure constant, are shown to be due to different reactor models, which led to different neutron spectra used in the calculations. We use known Oklo reactor epithermal spectral indices as criteria for selecting realistic reactor models. Two Oklo reactors, RZ2 and RZ10, were modeled with MCNP. The resulting neutron spectra were used to calculate the change in the ^149Sm effective neutron capture cross section as a function of a possible shift in the energy of the 97.3-meV resonance. We independently deduce ancient ^149Sm effective cross sections, and use these values to set limits on the time-variation of α. Our study resolves a contradictory situation with previous Oklo α-results. Our suggested 2 σ bound on a possible time variation of α over two billion years is stringent: -0.11 <=δαα <=0.24, in units of 10-7, but model dependent in that it assumes only α has varied over time.

  20. Modeling the Pyrochemical Reduction of Spent UO2 Fuel in a Pilot-Scale Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Steven D. Herrmann; Michael F. Simpson

    2006-08-01

    A kinetic model has been derived for the reduction of oxide spent nuclear fuel in a radial flow reactor. In this reaction, lithium dissolved in molten LiCl reacts with UO2 and fission product oxides to form a porous, metallic product. As the reaction proceeds, the depth of the porous layer around the exterior of each fuel particle increases. The observed rate of reaction has been found to be only dependent upon the rate of diffusion of lithium across this layer, consistent with a classic shrinking core kinetic model. This shrinking core model has been extended to predict the behavior of a hypothetical, pilot-scale reactor for oxide reduction. The design of the pilot-scale reactor includes forced flow through baskets that contain the fuel particles. The results of the modeling indicate that this is an essential feature in order to minimize the time needed to achieve full conversion of the fuel.

  1. Physical Analysis of the Initial Core and Running-In Phase for Pebble-Bed Reactor HTR-PM

    Directory of Open Access Journals (Sweden)

    Jingyu Zhang

    2017-01-01

    Full Text Available The pebble-bed reactor HTR-PM is being built in China and is planned to be critical in one or two years. At present, one emphasis of engineering design is to determine the fuel management scheme of the initial core and running-in phase. There are many possible schemes, and many factors need to be considered in the process of scheme evaluation and analysis. Based on the experience from the constructed or designed pebble-bed reactors, the fuel enrichment and the ratio of fuel spheres to graphite spheres are important. In this paper, some relevant physical considerations of the initial core and running-in phase of HTR-PM are given. Then a typical scheme of the initial core and running-in phase is proposed and simulated with VSOP code, and some key physical parameters, such as the maximum power per fuel sphere, the maximum fuel temperature, the refueling rate, and the discharge burnup, are calculated. Results of the physical parameters all satisfy the relevant design requirements, which means the proposed scheme is safe and reliable and can provide support for the fuel management of HTR-PM in the future.

  2. The IAEA Coordinated Research Program on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis: Description of the Benchmark Test Cases and Phases

    Energy Technology Data Exchange (ETDEWEB)

    Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

    2012-10-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The uncertainties in the HTR analysis tools are today typically assessed with sensitivity analysis and then a few important input uncertainties (typically based on a PIRT process) are varied in the analysis to find a spread in the parameter of importance. However, one wish to apply a more fundamental approach to determine the predictive capability and accuracies of coupled neutronics/thermal-hydraulics and depletion simulations used for reactor design and safety assessment. Today there is a broader acceptance of the use of uncertainty analysis even in safety studies and it has been accepted by regulators in some cases to replace the traditional conservative analysis. Finally, there is also a renewed focus in supplying reliable covariance data (nuclear data uncertainties) that can then be used in uncertainty methods. Uncertainty and sensitivity studies are therefore becoming an essential component of any significant effort in data and simulation improvement. In order to address uncertainty in analysis and methods in the HTGR community the IAEA launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modelling early in 2012. The project is built on the experience of the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity, but focuses specifically on the peculiarities of HTGR designs and its simulation requirements. Two benchmark problems were defined with the prismatic type design represented by the MHTGR-350 design from General Atomics (GA) while a 250 MW modular pebble bed design, similar to the INET (China) and indirect-cycle PBMR (South Africa) designs are also included. In the paper more detail on the benchmark cases, the different specific phases and tasks and the latest

  3. Preliminary assessment of Geant4 HP models and cross section libraries by reactor criticality benchmark calculations

    DEFF Research Database (Denmark)

    Cai, Xiao-Xiao; Llamas-Jansa, Isabel; Mullet, Steven

    2013-01-01

    to reactor modelling. Before version 9.5, Geant4 HP thermal scattering model (i.e. the S(α; β) model ) supports only three bounded isotopes, namely, H in water and polyethylene, and C in graphite. Newly supported materials include D in heavy water, O and Be in beryllium oxide, H and Zr in zirconium hydride......, U and O in uranium dioxide, Al metal, Be metal, and Fe metal. The native HP cross section library G4NDL does not include data for elements with atomic number larger than 92. Therefore, transuranic elements, which have impacts for a realistic reactor, can not be simulated by the combination of the HP...

  4. Dynamic behavior of the HTR-10 reactor: Dual temperature feedback model

    Directory of Open Access Journals (Sweden)

    Hosseini Seyed Ali

    2015-01-01

    Full Text Available The current work aims at presenting a simple model for PBM-type reactors' dynamic behavior analysis. The proposed model is based on point kinetics equations coupled with feedbacks from fuel and moderator temperatures. The temperature reactivity coefficients were obtained through MCNP code and via available experimental data. Parameters such as heat capacity and heat conductivity were carefully analyzed and the final system of equations was numerically solved. The obtained results, while in partial agreement with previously proposed models, suggest lower sensitivity to step reactivity insertion as compared to other reactor designs and inherent safety of the design.

  5. Capabilities of the ATHENA computer code for modeling the SP-100 space reactor concept

    Science.gov (United States)

    Fletcher, C. D.

    1985-09-01

    The capability to perform thermal-hydraulic analyses of an SP-100 space reactor was demonstrated using the ATHENA computer code. The preliminary General Electric SP-100 design was modeled using Athena. The model simulates the fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of this design. Two ATHENA demonstration calculations were performed simulating accident scenarios. A mask for the SP-100 model and an interface with the Nuclear Plant Analyzer (NPA) were developed, allowing a graphic display of the calculated results on the NPA.

  6. On the study of catalytic membrane reactor for water detritiation: Modeling approach

    Energy Technology Data Exchange (ETDEWEB)

    Liger, Karine, E-mail: karine.liger@cea.fr [CEA, DEN, DTN/SMTA/LIPC Cadarache, Saint Paul-lez-Durance F-13108 (France); Mascarade, Jérémy [CEA, DEN, DTN/SMTA/LIPC Cadarache, Saint Paul-lez-Durance F-13108 (France); Joulia, Xavier; Meyer, Xuan-Mi [Université de Toulouse, INPT, UPS, Laboratoire de Génie Chimique, 4, Allée Emile Monso, Toulouse F-31030 (France); CNRS, Laboratoire de Génie Chimique, Toulouse F-31030 (France); Troulay, Michèle; Perrais, Christophe [CEA, DEN, DTN/SMTA/LIPC Cadarache, Saint Paul-lez-Durance F-13108 (France)

    2016-11-01

    Highlights: • Experimental results for the conversion of tritiated water (using deuterium as a simulant of tritium) by means of a catalytic membrane reactor in view of tritium recovery. • Phenomenological 2D model to represent catalytic membrane reactor behavior including the determination of the compositions of gaseous effluents. • Good agreement between the simulation results and experimental measurements performed on the dedicated facility. • Explanation of the unexpected behavior of the catalytic membrane reactor by the modeling results and in particular the gas composition estimation. - Abstract: In the framework of tritium recovery from tritiated water, efficiency of packed bed membrane reactors have been successfully demonstrated. Thanks to protium isotope swamping, tritium bonded water can be recovered under the valuable Q{sub 2} form (Q = H, D or T) by means of isotope exchange reactions occurring on catalyst surface. The use of permselective Pd-based membrane allows withdrawal of reactions products all along the reactor, and thus limits reverse reaction rate to the benefit of the direct one (shift effect). The reactions kinetics, which are still little known or unknown, are generally assumed to be largely greater than the permeation ones so that thermodynamic equilibriums of isotope exchange reactions are generally assumed. This paper proposes a new phenomenological 2D model to represent catalytic membrane reactor behavior with the determination of gas effluents compositions. A good agreement was obtained between the simulation results and experimental measurements performed on a dedicated facility. Furthermore, the gas composition estimation permits to interpret unexpected behavior of the catalytic membrane reactor. In the next future, further sensitivity analysis will be performed to determine the limits of the model and a kinetics study will be conducted to assess the thermodynamic equilibrium of reactions.

  7. Analysis of PIK3CA Mutations and Activation Pathways in Triple Negative Breast Cancer.

    Directory of Open Access Journals (Sweden)

    Paolo Cossu-Rocca

    Full Text Available Triple Negative Breast Cancer (TNBC accounts for 12-24% of all breast carcinomas, and shows worse prognosis compared to other breast cancer subtypes. Molecular studies demonstrated that TNBCs are a heterogeneous group of tumors with different clinical and pathologic features, prognosis, genetic-molecular alterations and treatment responsivity. The PI3K/AKT is a major pathway involved in the regulation of cell survival and proliferation, and is the most frequently altered pathway in breast cancer, apparently with different biologic impact on specific cancer subtypes. The most common genetic abnormality is represented by PIK3CA gene activating mutations, with an overall frequency of 20-40%. The aims of our study were to investigate PIK3CA gene mutations on a large series of TNBC, to perform a wider analysis on genetic alterations involving PI3K/AKT and BRAF/RAS/MAPK pathways and to correlate the results with clinical-pathologic data.PIK3CA mutation analysis was performed by using cobas® PIK3CA Mutation Test. EGFR, AKT1, BRAF, and KRAS genes were analyzed by sequencing. Immunohistochemistry was carried out to identify PTEN loss and to investigate for PI3K/AKT pathways components.PIK3CA mutations were detected in 23.7% of TNBC, whereas no mutations were identified in EGFR, AKT1, BRAF, and KRAS genes. Moreover, we observed PTEN loss in 11.3% of tumors. Deregulation of PI3K/AKT pathways was revealed by consistent activation of pAKT and p-p44/42 MAPK in all PIK3CA mutated TNBC.Our data shows that PIK3CA mutations and PI3K/AKT pathway activation are common events in TNBC. A deeper investigation on specific TNBC genomic abnormalities might be helpful in order to select patients who would benefit from current targeted therapy strategies.

  8. Mutation and genomic amplification of the PIK3CA proto-oncogene in pituitary adenomas

    Energy Technology Data Exchange (ETDEWEB)

    Murat, C.B.; Braga, P.B.S.; Fortes, M.A.H.Z. [Laboratório de Endocrinologia Celular e Molecular (LIM-25), Faculdade de Medicina, Universidade de São Paulo, São Paulo, SP (Brazil); Bronstein, M.D. [Unidade de Neuroendocrinologia, Serviço de Endocrinologia, Hospital das Clínicas, Faculdade de Medicina, Universidade de São Paulo, São Paulo, SP (Brazil); Corrêa-Giannella, M.L.C.; Giorgi, R.R. [Laboratório de Endocrinologia Celular e Molecular (LIM-25), Faculdade de Medicina, Universidade de São Paulo, São Paulo, SP (Brazil)

    2012-07-13

    The tumorigenesis of pituitary adenomas is poorly understood. Mutations of the PIK3CA proto-oncogene, which encodes the p110-α catalytic subunit of PI3K, have been reported in various types of human cancers regarding the role of the gene in cell proliferation and survival through activation of the PI3K/Akt signaling pathway. Only one Chinese study described somatic mutations and amplification of the PIK3CA gene in a large series of pituitary adenomas. The aim of the present study was to determine genetic alterations of PIK3CA in a second series that consisted of 33 pituitary adenomas of different subtypes diagnosed by immunohistochemistry: 6 adrenocorticotropic hormone-secreting microadenomas, 5 growth hormone-secreting macroadenomas, 7 prolactin-secreting macroadenomas, and 15 nonfunctioning macroadenomas. Direct sequencing of exons 9 and 20 assessed by qPCR was employed to investigate the presence of mutations and genomic amplification defined as a copy number ≥4. Previously identified PIK3CA mutations (exon 20) were detected in four cases (12.1%). Interestingly, the Chinese study reported mutations only in invasive tumors, while we found a PIK3CA mutation in one noninvasive corticotroph microadenoma. PIK3CA amplification was observed in 21.2% (7/33) of the cases. This study demonstrates the presence of somatic mutations and amplifications of the PIK3CA gene in a second series of pituitary adenomas, corroborating the previously described involvement of the PI3K/Akt signaling pathway in the tumorigenic process of this gland.

  9. Oklo: The fossil nuclear reactors. Physics study - Translation of chapters 6, 13 and conclusions

    Energy Technology Data Exchange (ETDEWEB)

    Naudet, R. [CEA, Paris (France)

    1996-09-01

    Three parts of the 1991 book `Oklo: reacteurs nucleaires fossiles. Etude physique` have been translated in this report. The chapters bear the titles `Study of criticality`(45 p.), `Some problems with the overall functioning of the reactor zones`(45 p.) and `Conclusions` (15 p.), respectively.

  10. V.S.O.P. (99/09) Computer Code System for Reactor Physics and Fuel Cycle Simulation; Version 2009

    OpenAIRE

    Rütten, H.-J.; Haas, K. A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

    2010-01-01

    V.S.O.P.(99/ 09) represents the further development of V.S.O.P.(99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of...

  11. Time-variability of alpha from realistic models of Oklo reactors

    CERN Document Server

    Gould, C R; Lamoreaux, S K

    2006-01-01

    We reanalyze Oklo $^{149}$Sm data using realistic models of the natural nuclear reactors. Disagreements among recent Oklo determinations of the time evolution of $\\alpha$, the electromagnetic fine structure constant, are shown to be due to different reactor models, which led to different neutron spectra used in the calculations. We use known Oklo reactor epithermal spectral indices as criteria for selecting realistic reactor models. Two Oklo reactors, RZ2 and RZ10, were modeled with MCNP. The resulting neutron spectra were used to calculate the change in the $^{149}$Sm effective neutron capture cross section as a function of a possible shift in the energy of the 97.3-meV resonance. We independently deduce ancient $^{149}$Sm effective cross sections, and use these values to set limits on the time-variation of $\\alpha$. Our study resolves a contradictory situation with previous Oklo $\\alpha$-results. Our suggested $2 \\sigma$ bound on a possible time variation of $\\alpha$ over two billion years is stringent: $ -...

  12. Modeling of the HiPco process for carbon nanotube production. II. Reactor-scale analysis

    Science.gov (United States)

    Gokcen, Tahir; Dateo, Christopher E.; Meyyappan, M.

    2002-01-01

    The high-pressure carbon monoxide (HiPco) process, developed at Rice University, has been reported to produce single-walled carbon nanotubes from gas-phase reactions of iron carbonyl in carbon monoxide at high pressures (10-100 atm). Computational modeling is used here to develop an understanding of the HiPco process. A detailed kinetic model of the HiPco process that includes of the precursor, decomposition metal cluster formation and growth, and carbon nanotube growth was developed in the previous article (Part I). Decomposition of precursor molecules is necessary to initiate metal cluster formation. The metal clusters serve as catalysts for carbon nanotube growth. The diameter of metal clusters and number of atoms in these clusters are some of the essential information for predicting carbon nanotube formation and growth, which is then modeled by the Boudouard reaction with metal catalysts. Based on the detailed model simulations, a reduced kinetic model was also developed in Part I for use in reactor-scale flowfield calculations. Here this reduced kinetic model is integrated with a two-dimensional axisymmetric reactor flow model to predict reactor performance. Carbon nanotube growth is examined with respect to several process variables (peripheral jet temperature, reactor pressure, and Fe(CO)5 concentration) with the use of the axisymmetric model, and the computed results are compared with existing experimental data. The model yields most of the qualitative trends observed in the experiments and helps to understanding the fundamental processes in HiPco carbon nanotube production.

  13. Preliminary Reactor Physics Assessment of the HTR Module with 14% Enriched UCO Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard Strydom; Hans D. Gougar

    2013-03-01

    The high temperature reactor (HTR) Module (Lohnert, 1990) is a graphite-moderated, helium cooled pebble bed design that has been extensively used as a reference template for the former South African (Matzner, 2004) and current Chinese (Zhang et al., 2009) HTR programs. This design utilizes spherical fuel elements packed into a dynamic pebble bed, consisting of tri-structural isotropic (TRISO) coated uranium oxide (UO2) 500 µm fuel kernels with a U-235 enrichment of 7.8% and a heavy metal loading of 7 g per pebble. This fuel type was previously qualified for use in Germany for pebble bed HTRs, as well as undergoing re-qualification in South Africa for the PBMR project. It is also the fuel type being tested for use in the high temperature reactor (HTR-PM) under construction in China. In the United States, however, a different TRISO fuel form is the subject of a qualification program. The U.S. experience with HTRs has been focused upon the batch-fueled prismatic reactor in which TRISO particles are embedded in cylindrical compacts and stacked inside the graphite blocks which comprise the core. Under this type of operating regime, a smaller TRISO with a different composition and enrichment performs better than the fuel historically used in PBRs. Fuel kernels and compacting techniques more suited to prismatic core duty are currently being developed and qualified under the U.S. Department of Energy's Advanced Gas Reactor (AGR) fuel development program and in support of the Next Generation Nuclear Plant project. Interest in the pebble bed concept remains high, however, and a study was undertaken by the authors to assess the viability of using AGR fuel in a pebble bed reactor. Using the German HTR Module as the reference plant, key neutronic and thermal-hydraulic parameters were compared between the nominal design and one fueled with the fuel that is the focus of the AGR program.

  14. Investigating the Inhibitory Effect of Wortmannin in the Hotspot Mutation at Codon 1047 of PIK3CA Kinase Domain: A Molecular Docking and Molecular Dynamics Approach.

    Science.gov (United States)

    Kumar, D Thirumal; Doss, C George Priya

    2016-01-01

    Oncogenic mutations in phosphatidylinositol-4,5-bisphosphate 3-kinase, catalytic subunit alpha (PIK3CA) are the most frequently reported in association with various forms of cancer. Several studies have reported the significance of hotspot mutations in a catalytic subunit of PIK3CA in association with breast cancer. Mutations are frequently observed in the highly conserved region of the kinase domain (797-1068 amino acids) of PIK3CA are activating or gain-of-function mutations. Mutation in codon 1047 occurs in the C-terminal region of the kinase domain with histidine (H) replaced by arginine (R), lysine (L), and tyrosine (Y). Pathogenicity and protein stability predictors PhD-SNP, Align GVGD, HANSA, iStable, and MUpro classified H1047R as highly deleterious when compared to H1047L and H1047Y. To explore the inhibitory activity of Wortmannin toward PIK3CA, the three-dimensional structure of the mutant protein was determined using homology modeling followed by molecular docking and molecular dynamics analysis. Docking studies were performed for the three mutants and native with Wortmannin to measure the differences in their binding pattern. Comparative docking study revealed that H1047R-Wortmannin complex has a higher number of hydrogen bonds as well as the best binding affinity next to the native protein. Furthermore, 100 ns molecular dynamics simulation was initiated with the docked complexes to understand the various changes induced by the mutation. Though Wortmannin was found to nullify the effect of H1047R over the protein, further studies are required for designing a better compound. As SNPs are major genetic variations observed in disease condition, personalized medicine would provide enhanced drug therapy.

  15. Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

    Science.gov (United States)

    Turinsky, Paul J.; Kothe, Douglas B.

    2016-05-01

    The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics "core simulator" based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL's evolving M

  16. Instream Physical Habitat Modelling Types

    DEFF Research Database (Denmark)

    Conallin, John; Boegh, Eva; Krogsgaard, Jørgen

    2010-01-01

    management tools, but require large amounts of data and the model structure is complex. It is concluded that the use of habitat suitability indices (HSIs) and fuzzy rules in hydraulic-habitat modelling are the most ready model types to satisfy WFD demands. These models are well documented, transferable, user-friendly...... and disadvantages as management tools for member states in relation to the requirements of the WFD, but due to their different model structures they are distinct in their data needs, transferability, user-friendliness and presentable outputs. Water resource managers need information on what approaches will best...

  17. Description and results of a two-dimensional lattice physics code benchmark for the Canadian Pressure Tube Supercritical Water-cooled Reactor (PT-SCWR)

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, D.W.; Langton, S.E.; Ball, M.R.; Novog, D.R.; Buijs, A., E-mail: hummeld@mcmaster.ca [McMaster Univ., Hamilton, Ontario (Canada)

    2013-07-01

    Discrepancies have been observed among a number of recent reactor physics studies in support of the PT-SCWR pre-conceptual design, including differences in lattice-level predictions of infinite neutron multiplication factor, coolant void reactivity, and radial power profile. As a first step to resolving these discrepancies, a lattice-level benchmark problem was designed based on the 78-element plutonium-thorium PT-SCWR fuel design under a set of prescribed local conditions. This benchmark problem was modeled with a suite of both deterministic and Monte Carlo neutron transport codes. The results of these models are presented here as the basis of a code-to-code comparison. (author)

  18. Development and application of modeling tools for sodium fast reactor inspection

    Science.gov (United States)

    Le Bourdais, Florian; Marchand, Benoît; Baronian, Vahan

    2014-02-01

    To support the development of in-service inspection methods for the Advanced Sodium Test Reactor for Industrial Demonstration (ASTRID) project led by the French Atomic Energy Commission (CEA), several tools that allow situations specific to Sodium cooled Fast Reactors (SFR) to be modeled have been implemented in the CIVA software and exploited. This paper details specific applications and results obtained. For instance, a new specular reflection model allows the calculation of complex echoes from scattering structures inside the reactor vessel. EMAT transducer simulation models have been implemented to develop new transducers for sodium visualization and imaging. Guided wave analysis tools have been developed to permit defect detection in the vessel shell. Application examples and comparisons with experimental data are presented.

  19. Application of spectral tuning on the dynamic model of the reactor VVER 1000 support cylinder

    Directory of Open Access Journals (Sweden)

    Musil A.

    2007-10-01

    Full Text Available The paper deals with the optimization of parameters of the dynamic model of the reactor VVER 1000 support cylinder. Within the model of the whole reactor, support cylinder appears to be a significant subsystem for its modal properties having dominant influence on the behaviour of the reactor as a whole. Relative sensitivities of eigenfrequencies to a change of the discrete parameters of the model were determined. Obtained values were applied in the following spectral tuning process of the (selected discrete parameters. Since the past calculations have shown that spectral tuning by the changes of mass parameters is not effective, the presented paper demonstrates what results are achieved when the set of the tuning parameters is extended by the geometric parameters. Tuning itself is then formulated as an optimization problem with inequalities.

  20. Modeling of organic pollutant destruction in a stirred-tank reactor by ozonation

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    Destruction of organic contaminants in water by ozonation is a gas-liquid process which involves ozone mass transfer and fast irreversible chemical reactions. Ozonation reactor design and process optimizing require the modeling of the gas-liquid interactions within the reactor. In this paper a theoretical model combining the fluid dynamic and reaction kineticparameters is proposed for predicting the destruction rates of organic pollutants in a semi-batch stirred-tank reactor by ozonation. A simple expression for the enhancement factor as ourprevious work (Cheng, 2000) has been applied to evaluate the chemical mass transfer coefficient in ozone absorption.2,4-dichlorophenol (2,4-DCP) and 2,6-DCP or their mixture are chosen as the model compounds for simulating, and the predicted DCP oundation item: The National Natural Science Foundation of China (No. 20006006) ncentrations are compared with some measured data.

  1. Computational fluid dynamics modelling of biomass fast pyrolysis in fluidised bed reactors, focusing different kinetic schemes.

    Science.gov (United States)

    Ranganathan, Panneerselvam; Gu, Sai

    2016-08-01

    The present work concerns with CFD modelling of biomass fast pyrolysis in a fluidised bed reactor. Initially, a study was conducted to understand the hydrodynamics of the fluidised bed reactor by investigating the particle density and size, and gas velocity effect. With the basic understanding of hydrodynamics, the study was further extended to investigate the different kinetic schemes for biomass fast pyrolysis process. The Eulerian-Eulerian approach was used to model the complex multiphase flows in the reactor. The yield of the products from the simulation was compared with the experimental data. A good comparison was obtained between the literature results and CFD simulation. It is also found that CFD prediction with the advanced kinetic scheme is better when compared to other schemes. With the confidence obtained from the CFD models, a parametric study was carried out to study the effect of biomass particle type and size and temperature on the yield of the products.

  2. Kinetic model for torrefaction of wood chips in a pilot-scale continuous reactor

    DEFF Research Database (Denmark)

    Shang, Lei; Ahrenfeldt, Jesper; Holm, Jens Kai

    2014-01-01

    accordance with the model data. In an additional step a continuous, pilot scale reactor was built to produce torrefied wood chips in large quantities. The "two-step reaction in series" model was applied to predict the mass yield of the torrefaction reaction. Parameters used for the calculation were......Torrefaction is a mild thermal treatment (200-300 °C) in an inert atmosphere, known to increase the energy density of biomass by evaporation of water and a proportion of the volatiles. In this work a "two-step reaction in series" model was used to describe the thermal degradation kinetics of pine...... the temperature along the reactor and the biomass feeding rate in combination with the kinetic parameters obtained from the tests in the TGA. Together with results from a laboratory scale, batch torrefaction reactor that was used to determine the higher heating value (HHV) and mass loss (y) of the same material...

  3. Model Development and Simulation of Nitrification in SHARON Reactor in Moderate Temperature by Simulink

    Directory of Open Access Journals (Sweden)

    Dr. Adnan Abbas Al-Samawi

    2015-11-01

    Full Text Available In order to reduce the nitrogen compounds in WWTP effluent according to legislations, nitrogen of reject water is removed in separate unit by applying innovative cost effective process named SHARON (Single reactor High activity Ammonium Removal Over Nitrite process which is feasible to apply in moderate weather and more cost effective process due to elimination the heat exchanger required to keep the reject water of high temperature. In addition to the save in oxygen requirement to oxide ammonium by preventing nitrite oxidation and the saving in external COD addition for denitrification. Also, there is no need for large reactor volume because HRT equal to SRT. Significant mathematical model of nitrification process in SHARON reactor was developed based on substances and organisms mass balance as well as organisms kinetics. A relatively favorable consistency was obtained between the experimental and the predicted results of model. A high correlation of (R2=0.946 between model predictions and experimental data sets.

  4. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg, Principal Investigator; Kevin A. Steuhm, Project Manager

    2012-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core

  5. The Capsicum annuum class IV chitinase ChitIV interacts with receptor-like cytoplasmic protein kinase PIK1 to accelerate PIK1-triggered cell death and defence responses.

    Science.gov (United States)

    Kim, Dae Sung; Kim, Nak Hyun; Hwang, Byung Kook

    2015-04-01

    The pepper receptor-like cytoplasmic protein kinase, CaPIK1, which mediates signalling of plant cell death and defence responses was previously identified. Here, the identification of a class IV chitinase, CaChitIV, from pepper plants (Capsicum annuum), which interacts with CaPIK1 and promotes CaPIK1-triggered cell death and defence responses, is reported. CaChitIV contains a signal peptide, chitin-binding domain, and glycol hydrolase domain. CaChitIV expression was up-regulated by Xanthomonas campestris pv. vesicatoria (Xcv) infection. Notably, avirulent Xcv infection rapidly induced CaChitIV expression in pepper leaves. Bimolecular fluorescence complementation and co-immunoprecipitation revealed that CaPIK1 interacts with CaChitIV in planta, and that the CaPIK1-CaChitIV complex is localized mainly in the cytoplasm and plasma membrane. CaChitIV is also localized in the endoplasmic reticulum. Transient co-expression of CaChitIV with CaPIK1 enhanced CaPIK1-triggered cell death response and reactive oxygen species (ROS) and nitric oxide (NO) bursts. Co-silencing of both CaChitIV and CaPIK1 in pepper plants conferred enhanced susceptibility to Xcv infection, which was accompanied by a reduced induction of cell death response, ROS and NO bursts, and defence response genes. Ectopic expression of CaPIK1 in Arabidopsis enhanced basal resistance to Hyaloperonospora arabidopsidis infection. Together, the results suggest that CaChitIV positively regulates CaPIK1-triggered cell death and defence responses through its interaction with CaPIK1.

  6. Modelling of a large scale reactor for plasma deposition of silicon

    NARCIS (Netherlands)

    Nienhuis, G. J.; W. Goedheer,

    1999-01-01

    A 2D fluid model for RF discharges in a mixture of silane and hydrogen is applied to a cylindrically symmetric reactor with an electrode radius large compared to the electrode separation. In the model the electron kinetics are included by solving the two-term Boltzmann equation to obtain the electro

  7. Recommendations concerning models and parameters best suited to breeder reactor environmental radiological assessments

    Energy Technology Data Exchange (ETDEWEB)

    Miller, C.W.; Baes, C.F. III; Dunning, D.E. Jr.

    1980-05-01

    Recommendations are presented concerning the models and parameters best suited for assessing the impact of radionuclide releases to the environment by breeder reactor facilities. These recommendations are based on the model and parameter evaluations performed during this project to date. Seven different areas are covered in separate sections.

  8. Bridging Physics and Biology Teaching through Modeling

    CERN Document Server

    Hoskinson, Anne-Marie; Zwickl, Benjamin M; Hinko, Kathleen; Caballero, Marcos D

    2013-01-01

    As the frontiers of biology become increasingly interdisciplinary, the physics education community has engaged in ongoing efforts to make physics classes more relevant to life sciences majors. These efforts are complicated by the many apparent differences between these fields, including the types of systems that each studies, the behavior of those systems, the kinds of measurements that each makes, and the role of mathematics in each field. Nonetheless, physics and biology are both fundamental sciences that rely on observations and measurements to construct models of the natural world. In the present theoretical article, we propose that efforts to bridge the teaching of these two disciplines must emphasize shared scientific practices, particularly scientific modeling. We define modeling using language common to both disciplines and highlight how an understanding of the modeling process can help reconcile apparent differences between physics and biology. We elaborate how models can be used for explanatory, pre...

  9. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

    Science.gov (United States)

    Dautray, Robert

    2011-06-01

    The author firstly gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the 1950s. "Neutronics", thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, taps, waste cock, safety circuits, heat exchange units, etc.) have now attained maturity, sufficient to implement sodium cooling circuits. Notwithstanding, the use of metallic sodium still raises certain severe questions in terms of safe handling (i.e. inflammability) and other important security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchange devices) are undergoing in-depth research so as to last longer. The fuel cycle, notably the refabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. Among the priorities for a fully operational system (power station - the fuel cycle - operation-maintenance - the spent fuel pool and its cooling system-emergency cooling system-emergency electric power-transportation movements-equipment handling - final disposal of radioactive matter, independent safety barriers), the author includes materials (fabrication of targets, an irradiation and inspection instrument), the chemistry of all sorting processes, equipment "refabrication" or rehabilitation

  10. The Physical Internet and Business Model Innovation

    Directory of Open Access Journals (Sweden)

    Diane Poulin

    2012-06-01

    Full Text Available Building on the analogy of data packets within the Digital Internet, the Physical Internet is a concept that dramatically transforms how physical objects are designed, manufactured, and distributed. This approach is open, efficient, and sustainable beyond traditional proprietary logistical solutions, which are often plagued by inefficiencies. The Physical Internet redefines supply chain configurations, business models, and value-creation patterns. Firms are bound to be less dependent on operational scale and scope trade-offs because they will be in a position to offer novel hybrid products and services that would otherwise destroy value. Finally, logistical chains become flexible and reconfigurable in real time, thus becoming better in tune with firm strategic choices. This article focuses on the potential impact of the Physical Internet on business model innovation, both from the perspectives of Physical-Internet enabled and enabling business models.

  11. Analysis of transients in advanced heavy water reactor using lumped parameter models

    Energy Technology Data Exchange (ETDEWEB)

    Manmohan Pandey; Venkata Ramana Eaga; Sankar Sastry, P. [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati (India); Gupta, S.K.; Lele, H.G.; Chatterjee, B. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)

    2005-07-01

    Full text of publication follows: Analysis of transients occurring in nuclear power plants, arising from the complex interplay between core neutronics and thermal-hydraulics, is important for their operation and safety. Numerical simulations of such transients can be carried out extensively at very low computational cost by using lumped parameter mathematical models. The Advanced Heavy Water Reactor (AHWR), being developed in India, is a vertical pressure tube type reactor cooled by boiling light water under natural circulation, using thorium as fuel and heavy water as moderator. In the present work, nonlinear and linear lumped parameter dynamic models for AHWR have been developed and validated with a distributed parameter model. The nonlinear lumped model is based on point reactor kinetics equations and one-dimensional homogeneous equilibrium model of two-phase flow. The distributed model is built with RELAP5/MOD3.2 code. Various types of transients have been simulated numerically, using the lumped model as well as RELAP5. The results have been compared and parameters tuned to make the lumped model match the distributed model (RELAP5) in terms of steady state as well as dynamic behaviour. The linear model has been derived by linearizing the nonlinear model for small perturbations about the steady state. Numerical simulations of transients using the linear model have been compared with results obtained from the nonlinear model. Thus, the range of validity of the linear model has been determined. Stability characteristics of AHWR have been investigated using the lumped parameter models. (authors)

  12. Physics of the Hilbert Book Model

    Energy Technology Data Exchange (ETDEWEB)

    Leunen, Hans van

    2014-07-01

    The Hilbert Book Model is the name of a personal project of the author. The model is deduced from a foundation that is based on quantum logic and that is subsequently extended with trustworthy mathematical methods. What is known from conventional physics is used as a guideline, but the model is not based on the methodology of contemporary physics. In this way the model can reach deeper into the basement of physics. The ambition of the model is rather modest. It limits its scope to the lowest levels of the physical hierarchy. Thus fields and elementary particles are treated in fair detail, but composites are treated marginally and only some aspects of cosmology are touched. Still the model dives into the origins of gravitation and inertia and explains the diversity of the elementary particles. It explains what photons are and introduces a lower level of physical objects and a new kind of ultra-high frequency waves that carry information about their emitters. It explains entanglement and the Pauli principle. Above all the HBM introduces a new way of looking at space and time. Where contemporary physics applies the spacetime model, the HBM treats space and progression as a paginated model.

  13. Instream Physical Habitat Modelling Types

    DEFF Research Database (Denmark)

    Conallin, John; Boegh, Eva; Krogsgaard, Jørgen

    2010-01-01

    -friendly and have flexible data needs. They can easily be implemented in new regions using expert information or different types of local data. Furthermore, they are easily presentable to stakeholders and have the potential to be applied over large spatial scales. Integral care must be taken in the use...... and disadvantages as management tools for member states in relation to the requirements of the WFD, but due to their different model structures they are distinct in their data needs, transferability, user-friendliness and presentable outputs. Water resource managers need information on what approaches will best...... management tools, but require large amounts of data and the model structure is complex. It is concluded that the use of habitat suitability indices (HSIs) and fuzzy rules in hydraulic-habitat modelling are the most ready model types to satisfy WFD demands. These models are well documented, transferable, user...

  14. A new model for anaerobic processes of up-flow anaerobic sludge blanket reactors based on cellular automata

    DEFF Research Database (Denmark)

    Skiadas, Ioannis V.; Ahring, Birgitte Kiær

    2002-01-01

    characteristics and lead to different reactor behaviour. A dynamic mathematical model has been developed for the anaerobic digestion of a glucose based synthetic wastewater in UASB reactors. Cellular automata (CA) theory has been applied to simulate the granule development process. The model takes...... into consideration that granule diameter and granule microbial composition are functions of the reactor operational parameters and is capable of predicting the UASB performance and the layer structure of the granules....

  15. Model Reduction Using Proper Orthogonal Decomposition and Predictive Control of Distributed Reactor System

    Directory of Open Access Journals (Sweden)

    Alejandro Marquez

    2013-01-01

    Full Text Available This paper studies the application of proper orthogonal decomposition (POD to reduce the order of distributed reactor models with axial and radial diffusion and the implementation of model predictive control (MPC based on discrete-time linear time invariant (LTI reduced-order models. In this paper, the control objective is to keep the operation of the reactor at a desired operating condition in spite of the disturbances in the feed flow. This operating condition is determined by means of an optimization algorithm that provides the optimal temperature and concentration profiles for the system. Around these optimal profiles, the nonlinear partial differential equations (PDEs, that model the reactor are linearized, and afterwards the linear PDEs are discretized in space giving as a result a high-order linear model. POD and Galerkin projection are used to derive the low-order linear model that captures the dominant dynamics of the PDEs, which are subsequently used for controller design. An MPC formulation is constructed on the basis of the low-order linear model. The proposed approach is tested through simulation, and it is shown that the results are good with regard to keep the operation of the reactor.

  16. Mutational analyses of the BRAF, KRAS, and PIK3CA genes in oral squamous cell carcinoma

    Science.gov (United States)

    Bruckman, Karl C.; Schönleben, Frank; Qiu, Wanglong; Woo, Victoria L.; Su, Gloria H.

    2010-01-01

    OBJECTIVES The development of oral squamous cell carcinoma (OSCC) is a complex, multistep process. To date, numerous oncogenes and tumor-suppressor genes have been implicated in oral carcinogenesis. Of particular interest in this regard are genes involved in cell cycling and apoptosis, such BRAF, KRAS, and PIK3CA genes. STUDY DESIGN Mutations of BRAF, KRAS, and PIK3CA were evaluated by direct genomic sequencing of exons 1 of KRAS, 11 and 15 of BRAF, and 9 and 20 of PIK3CA in OSCC specimens. RESULTS Both BRAF and KRAS mutations were detected with a mutation frequency of 2% (1/42). PIK3CA mutations were detected at 3% (1/35). CONCLUSIONS This is the first report implicating BRAF mutation in OSCC. Our study supports that mutations in the BRAF, KRAS, and PIK3CA genes make at least a minor contribution to OSCC tumorigenesis, and pathway-specific therapies targeting these two pathways should be considered for OSCC in a subset of patients with these mutations. PMID:20813562

  17. Are Physical Education Majors Models for Fitness?

    Science.gov (United States)

    Kamla, James; Snyder, Ben; Tanner, Lori; Wash, Pamela

    2012-01-01

    The National Association of Sport and Physical Education (NASPE) (2002) has taken a firm stance on the importance of adequate fitness levels of physical education teachers stating that they have the responsibility to model an active lifestyle and to promote fitness behaviors. Since the NASPE declaration, national initiatives like Let's Move…

  18. V. S. O. P. ('94) Computer Code System for Reactor Physics and Fuel Cycle Simulation

    OpenAIRE

    Teuchert, E.; Haas, K. A.; Rütten, H. J.; Brockmann, Hans; Gerwin, Helmut; Ohlig, U.; Scherer, Winfried

    1994-01-01

    V. S. O. P. ('Very Superior Old Programs) is a system of codes lurked together for the simulationof reactor life histories and temporary in-depth research. In comprises neutron cross sectionlibraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculationwith depletion and shut-down features, in-core and out-of--pile fuel management, fuel cyclecost analysis, and thermal hydraulics (at present restricted to 's). Various techniques havebeen employed to accelerat...

  19. V. S. O. P. - Computer Code System for Reactor Physics and Fuel Cycle Simulation

    OpenAIRE

    Teuchert, E.; Hansen, U.; Haas, K. A.

    1980-01-01

    V .S .O .P . (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprisesneutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based onneutron flux synthesis with depletion and shut-down features, incore and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employe...

  20. Proposal and analysis of the benchmark problem suite for reactor physics study of LWR next generation fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-10-01

    In order to investigate the calculation accuracy of the nuclear characteristics of LWR next generation fuels, the Research Committee on Reactor Physics organized by JAERI has established the Working Party on Reactor Physics for LWR Next Generation Fuels. The next generation fuels mean the ones aiming for further extended burn-up such as 70 GWd/t over the current design. The Working Party has proposed six benchmark problems, which consists of pin-cell, PWR fuel assembly and BWR fuel assembly geometries loaded with uranium and MOX fuels, respectively. The specifications of the benchmark problem neglect some of the current limitations such as 5 wt% {sup 235}U to achieve the above-mentioned target. Eleven organizations in the Working Party have carried out the analyses of the benchmark problems. As a result, status of accuracy with the current data and method and some problems to be solved in the future were clarified. In this report, details of the benchmark problems, result by each organization, and their comparisons are presented. (author)

  1. Mathematical Modeling and Simulation of the Dehydrogenation of Ethyl Benzene to Form Styrene Using Steady-State Fixed Bed Reactor

    Directory of Open Access Journals (Sweden)

    Zaidon M. Shakoor

    2013-05-01

    Full Text Available In this research, two models are developed to simulate the steady state fixed bed reactor used for styrene production by ethylbenzene dehydrogenation. The first is one-dimensional model, considered axial gradient only while the second is two-dimensional model considered axial and radial gradients for same variables.The developed mathematical models consisted of nonlinear simultaneous equations in multiple dependent variables. A complete description of the reactor bed involves partial, ordinary differential and algebraic equations (PDEs, ODEs and AEs describing the temperatures, concentrations and pressure drop across the reactor was given. The model equations are solved by finite differences method. The reactor models were coded with Mat lab 6.5 program and various numerical techniques were used to obtain the desired solution.The simulation data for both models were validated with industrial reactor results with a very good concordance.

  2. Model-Based Analysis and Efficient Operation of a Glucose Isomerization Reactor Plant

    DEFF Research Database (Denmark)

    2015-01-01

    efficiency. The objective of this study is the application of the developed framework on an industrial case study of a glucose isomerization (GI) reactor plant that is part of a corn refinery, with the objective to improve the productivity of the process. Therefore, a multi-scale reactor model......The application of computer-aided model based methods within an integrated systematic framework is illustrated with the objective to assist the multi-purpose pharmaceutical/biochemical industry to systematically solve the complex problems that are experienced when aiming at improving the process...... is developedfor use as a building block for the GI reactor plant simulation. An optimal operation strategy is proposed on the basis of the simulation results...

  3. Modeling of Bubble Column Slurry Reactor for Dimethyl Ether Synthesis from Syngas

    Institute of Scientific and Technical Information of China (English)

    张海涛; 应卫勇; 房鼎业

    2005-01-01

    A mathematical model for a bubble column slurry reactor is presented for dimethyl ether synthesis from syngas. Methanol synthesis from carbon monoxide and carbon dioxide by hydrogenation and the methanol dehydration are considered as independent reactions, in which methanol, dimethyl ether and carbon dioxide are the key components. In this model, the gas phase is considered to be in plug flow and the liquid phase to be in partly back mixing with axial distribution of solid catalyst. The simulation results show that the axial dispersion of solid catalysts, the operational height of the slurry phase in the bubble column slurry reactor, and the reaction results are influenced by the reaction temperature and pressure, which are the basic data for the scale-up of reactor.

  4. Reactor models for a series of continuous stirred tank reactors with a gas-liquid-solid leaching system: Part II. Gas-transfer control

    Science.gov (United States)

    Papangelakis, V. G.; Demopoulos, G. P.

    1992-12-01

    This article is the second in a three-article series devoted to the development of comprehensive three-phase steady-state reactor models. In this article in particular, model equations are developed for the case of a leaching reactor operating under pure gas-transfer control. That is, the transfer of a gaseous reactant at the g-1 interface is considered to be the controlling step of the process rather than the particle dissolution reaction itself. For the derivation of the appropriate model equations, the gas-transfer capacity of the reactor is coupled with the particle dissolution kinetics. Two model versions are developed. In model version 1, the dissolved gas is assumed to be distributed equally among all particles. On the basis of this assumption, a gastransfer control-shrinking core model (GTC-SCM) equation is formulated which, along with the segregated flow model, helps to calculate the conversion of the solid phase. The size distribution of the particles at the exit of the reactor is computed via a mass-particle size density (PSD) function derived with the use of the population balance model (PBM). In model version 2, the dissolved gas is assumed to be distributed among particles in proportion to their surface area. Using the PBM, equations are developed suitable for the calculation of the total specific surface area of the reacting solids and their conversion. Single as well as multiple parallel leaching reactions are considered in developing the two model versions.

  5. Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong-beom [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Park, No-Cheol, E-mail: pnch@yonsei.ac.kr [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Lee, Sang-Jeong; Park, Young-Pil [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Choi, Youngin [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 34142 (Korea, Republic of)

    2017-03-15

    Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.

  6. Modeling Physics with Easy Java Simulations

    Science.gov (United States)

    Christian, Wolfgang; Esquembre, Francisco

    2007-01-01

    Modeling has been shown to correct weaknesses of traditional instruction by engaging students in the design of physical models to describe, explain, and predict phenomena. Although the modeling method can be used without computers, the use of computers allows students to study problems that are difficult and time consuming, to visualize their…

  7. Anaerobic tapered fluidized bed reactor for starch wastewater treatment and modeling using multilayer perceptron neural network

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Anaerobic treatability of synthetic sago wastewater was investigated in a laboratory anaerobic tapered fluidized bed reactor (ATFBR) with a mesoporous granular activated carbon (GAC) as a support material. The experimental protocol was defined to examine the effect of the maximum organic loading rate (OLR), hydraulic retention time (HRT), the efficiency of the reactor and to report on its steady-state performance. The reactor was subjected to a steady-state operation over a range of OLR up to 85.44 kg COD/(m3·d). The COD removal efficiency was found to be 92% in the reactor while the biogas produced in the digester reached 25.38 m3/(m3·d) of the reactor. With the increase of OLR from 83.7 kg COD/(m3·d), the COD removal efficiency decreases. Also an artificial neural network (ANN) model using multilayer perceptron (MLP) has been developed for a system of two input variable and five output dependent variables. For the training of the input-output data, the experimental values obtained have been used. The output parameters predicted have been found to be much closer to the corresponding experimental ones and the model was validated for 30% of the untrained data. The mean square error (MSE) was found to be only 0.0146.

  8. A dynamic model of a passively cooled small modular reactor for controller design purposes

    Energy Technology Data Exchange (ETDEWEB)

    Arda, Samet E., E-mail: s.e.arda@asu.edu; Holbert, Keith E., E-mail: holbert@asu.edu

    2015-08-15

    Highlights: • A mathematical dynamic model is developed for a passively cooled small modular reactor. • Reactor response associated single-phase natural circulation is analyzed. • A moving boundary model for a helical-coil steam generator is analyzed. • Dynamic responses of the overall model to representative perturbations are evaluated. • This compact model can be utilized for control system design. - Abstract: An analytical dynamic model for a passively cooled small modular reactor (SMR) is developed using a state-variable lumped parameter approach. Reactor power is represented by the generation time formulation of the point kinetics equations with a single combined neutron precursor group. The heat transfer process in the core is described via an overall heat transfer coefficient by defining two coolant lumps paired to a single fuel lump. In addition, a thermal–hydraulics model for single-phase natural circulation is incorporated. For the helical-coil steam generator, a moving-boundary model including subcooled, two-phase, and superheated regions is utilized. Finally, the hot leg riser and downcomer regions are expressed by first-order lags. The performance of the overall system described by ordinary differential equations (ODEs) is evaluated by the Simulink dynamic environment and directly using a MATLAB ODE solver recommended for stiff systems. Simulation results based on NuScale SMR design data show that the initial steady-state values for 100% power are within range of the design data and the model can predict the system dynamics due to typical perturbations, e.g., control rod movement and change in feedwater mass flow rate and temperature. The model developed in this work can be utilized as a foundation for designing and testing a suitable control algorithm for reactor thermal power.

  9. A simplified model of aerosol removal by natural processes in reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Powers, D.A.; Washington, K.E.; Sprung, J.L. [Sandia National Labs., Albuquerque, NM (United States); Burson, S.B. [Nuclear Regulatory Commission, Washington, DC (United States)

    1996-07-01

    Simplified formulae are developed for estimating the aerosol decontamination that can be achieved by natural processes in the containments of pressurized water reactors and in the drywells of boiling water reactors under severe accident conditions. These simplified formulae were derived by correlation of results of Monte Carlo uncertainty analyses of detailed models of aerosol behavior under accident conditions. Monte Carlo uncertainty analyses of decontamination by natural aerosol processes are reported for 1,000, 2,000, 3,000, and 4,000 MW(th) pressurized water reactors and for 1,500, 2,500, and 3,500 MW(th) boiling water reactors. Uncertainty distributions for the decontamination factors and decontamination coefficients as functions of time were developed in the Monte Carlo analyses by considering uncertainties in aerosol processes, material properties, reactor geometry and severe accident progression. Phenomenological uncertainties examined in this work included uncertainties in aerosol coagulation by gravitational collision, Brownian diffusion, turbulent diffusion and turbulent inertia. Uncertainties in aerosol deposition by gravitational settling, thermophoresis, diffusiophoresis, and turbulent diffusion were examined. Electrostatic charging of aerosol particles in severe accidents is discussed. Such charging could affect both the coagulation and deposition of aerosol particles. Electrostatic effects are not considered in most available models of aerosol behavior during severe accidents and cause uncertainties in predicted natural decontamination processes that could not be taken in to account in this work. Median (50%), 90 and 10% values of the uncertainty distributions for effective decontamination coefficients were correlated with time and reactor thermal power. These correlations constitute a simplified model that can be used to estimate the decontamination by natural aerosol processes at 3 levels of conservatism. Applications of the model are described.

  10. Artificial Neural Networks Based Modeling and Control of Continuous Stirred Tank Reactor

    Directory of Open Access Journals (Sweden)

    R. S.M.N. Malar

    2009-01-01

    Full Text Available Continuous Stirred Tank Reactor (CSTR is one of the common reactors in chemical plant. Problem statement: Developing a model incorporating the nonlinear dynamics of the system warrants lot of computation. An efficient control of the product concentration can be achieved only through accurate model. Approach: In this study, attempts were made to alleviate the above mentioned problem using “Artificial Intelligence” (AI techniques. One of the AI techniques namely Artificial Neural Networks (ANN was used to model the CSTR incorporating its non-linear characteristics. Two nonlinear models based control strategies namely internal model control and direct inverse control were designed using the neural networks and applied to the control of isothermal CSTR. Results: The simulation results for the above control schemes with set point tracking were presented. Conclusion: Results indicated that neural networks can learn accurate models and give good non-linear control when model equations are not known.

  11. Physics of the Quark Model

    Science.gov (United States)

    Young, Robert D.

    1973-01-01

    Discusses the charge independence, wavefunctions, magnetic moments, and high-energy scattering of hadrons on the basis of group theory and nonrelativistic quark model with mass spectrum calculated by first-order perturbation theory. The presentation is explainable to advanced undergraduate students. (CC)

  12. Physics of the Quark Model

    Science.gov (United States)

    Young, Robert D.

    1973-01-01

    Discusses the charge independence, wavefunctions, magnetic moments, and high-energy scattering of hadrons on the basis of group theory and nonrelativistic quark model with mass spectrum calculated by first-order perturbation theory. The presentation is explainable to advanced undergraduate students. (CC)

  13. DynMo: Dynamic Simulation Model for Space Reactor Power Systems

    Science.gov (United States)

    El-Genk, Mohamed; Tournier, Jean-Michel

    2005-02-01

    A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK® platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed.

  14. Contribution to the modelling of gas-solid reactions and reactors; Contribution a la modelisation des reactions et des reacteurs gaz-solide

    Energy Technology Data Exchange (ETDEWEB)

    Patisson, F

    2005-09-15

    Gas-solid reactions control a great number of major industrial processes involving matter transformation. This dissertation aims at showing that mathematical modelling is a useful tool for both understanding phenomena and optimising processes. First, the physical processes associated with a gas-solid reaction are presented in detail for a single particle, together with the corresponding available kinetic grain models. A second part is devoted to the modelling of multiparticle reactors. Different approaches, notably for coupling grain models and reactor models, are illustrated through various case studies: coal pyrolysis in a rotary kiln, production of uranium tetrafluoride in a moving bed furnace, on-grate incineration of municipal solid wastes, thermogravimetric apparatus, nuclear fuel making, steel-making electric arc furnace. (author)

  15. THE MODELING OF COUNTER-ROTATING TWIN-SCREW EXTRUDERS AS REACTORS FOR SINGLE-COMPONENT REACTIONS

    NARCIS (Netherlands)

    GANZEVELD, KJ; CAPEL, JE; VANDERWAL, DJ; JANSSEN, LPBM

    1994-01-01

    Numerical models are useful to study the behaviour of the extruder as a polymerization reactor. With a correct numerical model a theoretical analysis of the influence of several reaction and extruder parameters can be made, the limitations of the use of the extruder reactor can be determined and the

  16. THE MODELING OF COUNTER-ROTATING TWIN-SCREW EXTRUDERS AS REACTORS FOR SINGLE-COMPONENT REACTIONS

    NARCIS (Netherlands)

    GANZEVELD, KJ; CAPEL, JE; VANDERWAL, DJ; JANSSEN, LPBM

    Numerical models are useful to study the behaviour of the extruder as a polymerization reactor. With a correct numerical model a theoretical analysis of the influence of several reaction and extruder parameters can be made, the limitations of the use of the extruder reactor can be determined and the

  17. Ontology modeling in physical asset integrity management

    CERN Document Server

    Yacout, Soumaya

    2015-01-01

    This book presents cutting-edge applications of, and up-to-date research on, ontology engineering techniques in the physical asset integrity domain. Though a survey of state-of-the-art theory and methods on ontology engineering, the authors emphasize essential topics including data integration modeling, knowledge representation, and semantic interpretation. The book also reflects novel topics dealing with the advanced problems of physical asset integrity applications such as heterogeneity, data inconsistency, and interoperability existing in design and utilization. With a distinctive focus on applications relevant in heavy industry, Ontology Modeling in Physical Asset Integrity Management is ideal for practicing industrial and mechanical engineers working in the field, as well as researchers and graduate concerned with ontology engineering in physical systems life cycles. This book also: Introduces practicing engineers, research scientists, and graduate students to ontology engineering as a modeling techniqu...

  18. Modeling of reaction kinetics in bubbling fluidized bed biomass gasification reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thapa, R.K.; Halvorsen, B.M. [Telemark University College, Kjolnes ring 56, P.O. Box 203, 3901 Porsgrunn (Norway); Pfeifer, C. [University of Natural Resources and Life Sciences, Vienna (Austria)

    2013-07-01

    Bubbling fluidized beds are widely used as biomass gasification reactors as at the biomass gasification plant in Gussing, Austria. The reactor in the plant is a dual circulating bubbling fluidized bed gasification reactor. The plant produces 2MW electricity and 4.5MW heat from the gasification of biomass. Wood chips as biomass and olivine particles as hot bed materials are fluidized with high temperature steam in the reactor. As a result, biomass undergoes endothermic chemical reaction to produce a mixture of combustible gases in addition to some carbon-dioxide (CO2). The combustible gases are mainly hydrogen (H2), carbon monoxide (CO) and methane (CH4). The gas is used to produce electricity and heat via utilization in a gas engine. Alternatively, the gas is further processed for gaseous or liquid fuels, but still on the process of development level. Composition and quality of the gas determine the efficiency of the reactor. A computational model has been developed for the study of reaction kinetics in the gasification rector. The simulation is performed using commercial software Barracuda virtual reactor, VR15. Eulerian-Lagrangian approach in coupling of gas-solid flow has been implemented. Fluid phase is treated with an Eulerian formulation. Discrete phase is treated with a Lagrangian formulation. Particle-particle and particle-wall interactions and inter-phase heat and mass transfer have been taken into account. Series of simulations have been performed to study model prediction of the gas composition. The composition is compared with data from the gasifier at the CHP plant in Güssing, Austria. The model prediction of the composition of gases has good agreements with the result of the operating plant.

  19. Modeling of reaction kinetics in bubbling fluidized bed biomass gasification reactor

    Directory of Open Access Journals (Sweden)

    R.K. Thapa, C. Pfeifer, B. M. Halvorsen

    2014-01-01

    Full Text Available Bubbling fluidized beds are widely used as biomass gasification reactors as at the biomass gasification plant in Güssing, Austria. The reactor in the plant is a dual circulating bubbling fluidized bed gasification reactor. The plant produces 2MW electricity and 4.5MW heat from the gasification of biomass. Wood chips as biomass and olivine particles as hot bed materials are fluidized with high temperature steam in the reactor. As a result, biomass undergoes endothermic chemical reaction to produce a mixture of combustible gases in addition to some carbon-dioxide (CO2. The combustible gases are mainly hydrogen (H2, carbon monoxide (CO and methane (CH4. The gas is used to produce electricity and heat via utilization in a gas engine. Alternatively, the gas is further processed for gaseous or liquid fuels, but still on the process of development level. Composition and quality of the gas determine the efficiency of the reactor. A computational model has been developed for the study of reaction kinetics in the gasification rector. The simulation is performed using commercial software Barracuda virtual reactor, VR15. Eulerian-Lagrangian approach in coupling of gas-solid flow has been implemented. Fluid phase is treated with an Eulerian formulation. Discrete phase is treated with a Lagrangian formulation. Particle-particle and particle-wall interactions and inter-phase heat and mass transfer have been taken into account. Series of simulations have been performed to study model prediction of the gas composition. The composition is compared with data from the gasifier at the CHP plant in Güssing, Austria. The model prediction of the composition of gases has good agreements with the result of the operating plant.

  20. Cold Model Study and Commercial Test on Novel Vapor-Liquid Distributor of Hydroprocessing Reactor

    Institute of Scientific and Technical Information of China (English)

    Wang Shaobing; Zhang Zhanzhu; Wu Defei; Guo Qingming

    2007-01-01

    A novel vapor-liquid distributor was developed on the basis of sufficient study on the existing distributors applied in hydroprocessing reactors.The cold model test data showed that the fluid distribution performance of the novel vapor-liquid distributor was evidently better than the traditional one.Commercial tests of the new distributor were carried out in the 300 kt/a gas oil hydrotreating reactor at SINOPEC Changling Branch Company,showing that the new vapor-liquid distributor could improve the fluid distribution,promote the hydrotreating efficiency and lead to better performance than the traditional one.

  1. Assessment of uncertainty in full core reactor physics calculations using statistical methods

    Energy Technology Data Exchange (ETDEWEB)

    McEwan, C., E-mail: mcewac2@mcmaster.ca [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    The best estimate method of safety analysis involves choosing a realistic set of input parameters for a proposed safety case and evaluating the uncertainty in the results. Determining the uncertainty in code outputs remains a challenge and is the subject of a benchmarking exercise proposed by the Organization for Economic Cooperation and Development. The work proposed in this paper will contribute to this benchmark by assessing the uncertainty in a depletion calculation of the final nuclide concentrations for an experiment performed in the Fukushima-2 reactor. This will be done using lattice transport code DRAGON and a tool known as DINOSAUR. (author)

  2. Thermal-hydraulic modeling needs for passive reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, J.M. [Nuclear Regulatory Commission, Washington, DC (United States)

    1997-07-01

    The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.

  3. Comparison between triangular and hexagonal modeling of a hexagonal-structured reactor core using box method

    Energy Technology Data Exchange (ETDEWEB)

    Malmir, Hessam, E-mail: malmir@energy.sharif.edu [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); Moghaddam, Nader Maleki [Department of Nuclear Engineering and Physics, Amir Kabir University of Technology (Tehran Polytechnique), Hafez Street, Tehran (Iran, Islamic Republic of); Zahedinejad, Ehsan [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of)

    2011-02-15

    A hexagonal-structured reactor core (e.g. VVER-type) is mostly modeled by structured triangular and hexagonal mesh zones. Although both the triangular and hexagonal models give good approximations over the neutronic calculation of the core, there are some differences between them that seem necessary to be clarified. For this purpose, the neutronic calculations of a hexagonal-structured reactor core have to be performed using the structured triangular and hexagonal meshes based on box method of discretisation and then the results of two models should be benchmarked in different cases. In this paper, the box method of discretisation is derived for triangular and hexagonal meshes. Then, two 2-D 2-group static simulators for triangular and hexagonal geometries (called TRIDIF-2 and HEXDIF-2, respectively) are developed using the box method. The results are benchmarked against the well-known CITATION computer code in case of a VVER-1000 reactor core. Furthermore, the relative powers calculated by the TRIDIF-2 and HEXDIF-2 along with the ones obtained by the CITATION code are compared with the verified results which have been presented in the Final Safety Analysis Report (FSAR) of the aforementioned reactor. Different benchmark cases revealed the reliability of the box method in contrast with the CITATION code. Furthermore, it is shown that the triangular modeling of the core is more acceptable compared with the hexagonal one.

  4. Modified kinetic-hydraulic UASB reactor model for treatment of wastewater containing biodegradable organic substrates.

    Science.gov (United States)

    El-Seddik, Mostafa M; Galal, Mona M; Radwan, A G; Abdel-Halim, Hisham S

    2016-01-01

    This paper addresses a modified kinetic-hydraulic model for up-flow anaerobic sludge blanket (UASB) reactor aimed to treat wastewater of biodegradable organic substrates as acetic acid based on Van der Meer model incorporated with biological granules inclusion. This dynamic model illustrates the biomass kinetic reaction rate for both direct and indirect growth of microorganisms coupled with the amount of biogas produced by methanogenic bacteria in bed and blanket zones of reactor. Moreover, the pH value required for substrate degradation at the peak specific growth rate of bacteria is discussed for Andrews' kinetics. The sensitivity analyses of biomass concentration with respect to fraction of volume of reactor occupied by granules and up-flow velocity are also demonstrated. Furthermore, the modified mass balance equations of reactor are applied during steady state using Newton Raphson technique to obtain a suitable degree of freedom for the modified model matching with the measured results of UASB Sanhour wastewater treatment plant in Fayoum, Egypt.

  5. CHANGE IN DEFORMATION PROPERTIES MODELING OF CONCRETE IN PROTECTIVE STRUCTURES OF NUCLEAR REACTOR BY IONIZING RADIATION

    Directory of Open Access Journals (Sweden)

    E. K. Agakhanov

    2016-01-01

    Full Text Available The necessity of studying the effect impact of elementary particles impact on the strength and deformation materials properties used in protective constructions nuclear reactors and reactor technology has been stipulated. A nuclear reactor pressure vessel from prestressed concrete, combining the functions of biological protection is to be considered. The neutron flux problem distribution in the pressure vessel of a nuclear reactor has been solved. The solution is made in axisymmetric with the finite element method using a flat triangular finite element. Computing has been conducted in Matlab package. The comparison with the results has been obtained using the finite difference method, as well as the graphs of changes under the influence of radiation exposure and the elastic modulus of concrete radiation deformations have been constructed. The proposed method allows to simulate changes in the deformation properties of concrete under the influence of neutron irradiation. Results of the study can be used in the calculation of stress-strain state of structures, taking into account indirect heterogeneity caused by the physical fields influence.

  6. Modelling anaerobic digestion in a biogas reactor: ADM1 model development with lactate as an intermediate (Part I).

    Science.gov (United States)

    Satpathy, Preseela; Biernacki, Piotr; Uhlenhut, Frank; Cypionka, Heribert; Steinigeweg, Sven

    2016-12-05

    The Anaerobic Digestion Model No. 1 (ADM1) was extended to include lactate, a crucial metabolic product during sugar fermentation. This study tests the validity of the modified ADM1 model in improving the predictions of a standard biogas reactor. This reactor was prepared in the laboratory with simple organic substrates with an intention to represent an 'average biogas plant'. Kinetic parameters were determined from a lactic acid enriched steady-state reactor. The parameters were adjusted further in order to acquire satisfying simulation results systematically with the batch experiments and then against the standard biogas reactor. Arresting methanogenesis revealed that lactate degradation occurred majorly via acetate followed by propionate, and a non-negligible proportion of butyrate too was found, which were further updated in the model. The modified ADM1 provided a successful correlation with the experimental results for the batch and continuous experiments. We justified that inclusion of lactate in the model resulted in optimized simulation for both biogas and methane content in the standard biogas reactor.

  7. Improvement and modelling of hexenal transfer in liquid-gas reactor.

    Science.gov (United States)

    Ben Akacha, Najla; Guizani, Om Elkhir; Gargouri, Mohamed

    2007-12-01

    The aim of the present work is to improve the extraction of an aromatic compound, 2E-hexenal, from a continuous liquid-gas reactor. Having an improved process to recover hexenal could be of interest to obtain this chemical if produced by any biotechnological process. The experimental program proposed on the basis of a full 23 factorial design demonstrated that the conditions optima for the extraction are 40 degrees C as a reactor temperature