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Sample records for kyoto university critical assembly reactor

  1. Accelerator-Driven Subcritical Reactors in Japanese Universities: Experimental Study Using the Kyoto University Critical Assembly

    SciTech Connect

    Shiroya, S.; Unesaki, H.; Misawa, T.

    2001-06-17

    A series of basic experiments for an accelerator-driven sub-critical reactor (ADSR) was officially launched in financial year 2000 at the Kyoto University Critical Assembly (KUCA) as a joint-use program among Japanese universities. These experiments are closely related to the future plan of the Kyoto University Research Reactor Institute. A final goal of this plan is to establish a next-generation neutron source as a substitute for the 5-MW Kyoto University Reactor and based on the ADSR concept to promote joint research among Japanese universities. An attractive point of the ADSR system is that either pulsed or steady neutrons can be provided depending on the accelerator's operation mode.

  2. Biomedical irradiation system for boron neutron capture therapy at the Kyoto University Reactor.

    PubMed

    Kobayashi, T; Kanda, K; Ujeno, Y; Ishida, M R

    1990-01-01

    Physics studies related to radiation source, spectroscopy, beam quality, dosimetry, and biomedical applications using the Kyoto University Reactor Heavy Water Facility are described. Also, described are a Nickel Mirror Neutron Guide Tube and a Super Mirror Neutron Guide Tube that are used both for the measurement of boron concentration in phantom and living tissue and for precise measurements of neutron flux in phantom in the presence of both light and heavy water. Discussed are: (1) spectrum measurements using the time of flight technique, (2) the elimination of gamma rays and fast neutrons from a thermal neutron irradiation field, (3) neutron collimation without producing secondary gamma rays, (4) precise neutron flux measurements, dose estimation, and the measurement of boron concentration in tumor and its periphery using guide tubes, (5) the dose estimation of boron-10 for the first melanoma patient, and (6) special-purpose biological irradiation equipment. Other related subjects are also described. PMID:2176458

  3. Design study of multi-imaging plate system for BNCT irradiation field at Kyoto university reactor.

    PubMed

    Tanaka, Kenichi; Sakurai, Yoshinori; Kajimoto, Tsuyoshi; Tanaka, Hiroki; Takata, Takushi; Endo, Satoru

    2016-09-01

    The converter configuration for a multi-imaging plate system was investigated for the application of quality assurance in the irradiation field profile for boron neutron capture therapy. This was performed by the simulation calculation using the PHITS code in the fields at the Heavy Water Neutron Irradiation Facility of Kyoto University Reactor. The converter constituents investigated were carbon for gamma rays, and polyethylene with and without LiF at varied (6)Li concentration for thermal, epithermal, and fast neutrons. Consequently, potential combinations of the converters were found for two components, gamma rays and thermal neutrons, for the standard thermal neutron mode and three components of gamma rays, epithermal neutrons, and thermal or fast neutrons, for the standard mixed or epithermal neutron modes, respectively. PMID:27423022

  4. Advances in boron neutron capture therapy (BNCT) at kyoto university - From reactor-based BNCT to accelerator-based BNCT

    NASA Astrophysics Data System (ADS)

    Sakurai, Yoshinori; Tanaka, Hiroki; Takata, Takushi; Fujimoto, Nozomi; Suzuki, Minoru; Masunaga, Shinichiro; Kinashi, Yuko; Kondo, Natsuko; Narabayashi, Masaru; Nakagawa, Yosuke; Watanabe, Tsubasa; Ono, Koji; Maruhashi, Akira

    2015-07-01

    At the Kyoto University Research Reactor Institute (KURRI), a clinical study of boron neutron capture therapy (BNCT) using a neutron irradiation facility installed at the research nuclear reactor has been regularly performed since February 1990. As of November 2014, 510 clinical irradiations were carried out using the reactor-based system. The world's first accelerator-based neutron irradiation system for BNCT clinical irradiation was completed at this institute in early 2009, and the clinical trial using this system was started in 2012. A shift of BCNT from special particle therapy to a general one is now in progress. To promote and support this shift, improvements to the irradiation system, as well as its preparation, and improvements in the physical engineering and the medical physics processes, such as dosimetry systems and quality assurance programs, must be considered. The recent advances in BNCT at KURRI are reported here with a focus on physical engineering and medical physics topics.

  5. Experimental study on the thorium-loaded accelerator-driven system at the Kyoto Univ. critical assembly

    SciTech Connect

    Pyeon, C. H.; Yagi, T.; Lim, J. Y.; Misawa, T.

    2012-07-01

    The experimental study on the thorium-loaded accelerator-driven system (ADS) is conducted in the Kyoto Univ. Critical Assembly (KUCA). The experiments are carried out in both the critical and subcritical states for attaining the reaction rates of the thorium capture and fission reactions. In the critical system, the thorium plate irradiation experiment is carried out for the thorium capture and fission reactions. From the results of the measurements, the thorium fission reactions are obtained apparently in the critical system, and the C/E values of reaction rates show the accuracy of relative difference of about 30%. In the ADS experiments with 14 MeV neutrons and 100 MeV protons, the subcritical experiments are carried out in the thorium-loaded cores to obtain the capture reaction rates through the measurements of {sup 115}In(n, {gamma}){sup 116m}In reactions. The results of the experiments reveal the difference between the reaction rate distributions for the change in not only the neutron spectrum but also the external neutron source. The comparison between the measured and calculated reaction rate distributions demonstrates a discrepancy of the accuracy of reaction rate analyses of thorium capture reactions through the thorium-loaded ADS experiments with 14 MeV neutrons. Hereafter, kinetic experiments are planned to be carried out to deduce the delayed neutron decay constants and subcriticality using the pulsed neutron method. (authors)

  6. Controllability of depth dose distribution for neutron capture therapy at the Heavy Water Neutron Irradiation Facility of Kyoto University Research Reactor.

    PubMed

    Sakurai, Yoshinori; Kobayashi, Tooru

    2002-10-01

    The updating construction of the Heavy Water Neutron Irradiation Facility of the Kyoto University Research Reactor has been performed from November 1995 to March 1996 mainly for the improvement in neutron capture therapy. On the performance, the neutron irradiation modes with the variable energy spectra from almost pure thermal to epi-thermal neutrons became available by the control of the heavy-water thickness in the spectrum shifter and by the open-and-close of the cadmium and boral thermal neutron filters. The depth distributions of thermal, epi-thermal and fast neutron fluxes were measured by activation method using gold and indium, and the depth distributions of gamma-ray absorbed dose rate were measured using thermo-luminescent dosimeter of beryllium oxide for the several irradiation modes. From these measured data, the controllability of the depth dose distribution using the spectrum shifter and the thermal neutron filters was confirmed. PMID:12408308

  7. Reactor physics studies in the GCFR Phase III critical assembly

    SciTech Connect

    Morman, J A

    1980-03-01

    The third phase of the gas cooled fast reactor (GCFR) program, ZPR-9 Assembly 30, is based on a multi-zoned core of PuO/sub 2/-UO/sub 2/ with radial and axial blankets of UO/sub 2/. Studies performed in this assembly will be compared to the previous phases of the GCFR program and will help to define parameters in this power-flattened demonstration plant-type core. Measurements in the Phase III program included small sample reactivity worths of various materials, central reaction rates and reaction rate distributions, absorption-to-fission ratios and the central point conversion ratio and the worth of steam entry into a small central zone. The reactivity change associated with the construction of a central pin zone in the core and axial blanket was measured. Reaction rate and steam entry measurements were repeated in the pin environment. Standard analysis methods using ENDF/B-IV data are described and the results are compared to measurements performed during the program.

  8. The medical-irradiation characteristics for neutron capture therapy at the Heavy Water Neutron Irradiation Facility of Kyoto University Research Reactor.

    PubMed

    Sakurai, Yoshinori; Kobayashi, Tooru

    2002-10-01

    At the Heavy Water Neutron Irradiation Facility of the Kyoto University Research Reactor, the mix irradiation of thermal and epi-thermal neutrons, and the solo irradiation of epi-thermal neutrons are available additionally to the thermal neutron irradiation, and then the neutron capture therapy (NCT) at this facility became more flexible, after the update in 1996. The estimation of the depth dose distributions in NCT clinical irradiation, were performed for the standard irradiation modes of thermal, mixed and epi-thermal neutrons, from the both sides of experiment and calculation. On the assumption that the 10B concentration in tumor part was 40 ppm and the ratio of tumor to normal tissue was 3.5, the advantage depth were estimated to 5.4, 6.0, and 8.0, for the respective standard irradiation modes. It was confirmed that the various irradiation conditions can be selected according to the target-volume conditions, such as size, depth, etc. Besides, in the viewpoint of the radiation shielding for patient, it was confirmed that the whole-body exposure is effectively reduced by the new clinical collimators, compared with the old one. PMID:12408307

  9. Analysis of muon radiography of the Toshiba nuclear critical assembly reactor

    SciTech Connect

    Morris, C. L.; Bacon, Jeffery; Borozdin, Konstantin; Fabritius, J. M.; Perry, John; Ramsey, John; Ban, Yuichiro; Izumi, Mikio; Sano, Yuji; Yoshida, Noriyuki; Miyadera, Haruo; Mizokami, Shinya; Otsuka, Yasuyuki; Yamada, Daichi; Sugita, Tsukasa; Yoshioka, Kenichi

    2014-01-13

    A 1.2 × 1.2 m{sup 2} muon tracker was moved from Los Alamos to the Toshiba facility at Kawasaki, Japan, where it was used to take ∼4 weeks of data radiographing the Toshiba Critical Assembly Reactor with cosmic ray muons. In this paper, we describe the analysis procedure, show results of this experiment, and compare the results to Monte Carlo predictions. The results validate the concept of using cosmic rays to image the damaged cores of the Fukushima Daiichi reactors.

  10. Analysis of muon radiography of the Toshiba nuclear critical assembly reactor

    NASA Astrophysics Data System (ADS)

    Morris, C. L.; Bacon, Jeffery; Ban, Yuichiro; Borozdin, Konstantin; Fabritius, J. M.; Izumi, Mikio; Miyadera, Haruo; Mizokami, Shinya; Otsuka, Yasuyuki; Perry, John; Ramsey, John; Sano, Yuji; Sugita, Tsukasa; Yamada, Daichi; Yoshida, Noriyuki; Yoshioka, Kenichi

    2014-01-01

    A 1.2 × 1.2 m2 muon tracker was moved from Los Alamos to the Toshiba facility at Kawasaki, Japan, where it was used to take ˜4 weeks of data radiographing the Toshiba Critical Assembly Reactor with cosmic ray muons. In this paper, we describe the analysis procedure, show results of this experiment, and compare the results to Monte Carlo predictions. The results validate the concept of using cosmic rays to image the damaged cores of the Fukushima Daiichi reactors.

  11. Validation of FSP Reactor Design with Sensitivity Studies of Beryllium-Reflected Critical Assemblies

    SciTech Connect

    John D. Bess; Margaret A. Marshall

    2013-02-01

    The baseline design for space nuclear power is a fission surface power (FSP) system: sodium-potassium (NaK) cooled, fast spectrum reactor with highly-enriched-uranium (HEU)-O2 fuel, stainless steel (SS) cladding, and beryllium reflectors with B4C control drums. Previous studies were performed to evaluate modeling capabilities and quantify uncertainties and biases associated with analysis methods and nuclear data. Comparison of Zero Power Plutonium Reactor (ZPPR)-20 benchmark experiments with the FSP design indicated that further reduction of the total design model uncertainty requires the reduction in uncertainties pertaining to beryllium and uranium cross-section data. Further comparison with three beryllium-reflected HEU-metal benchmark experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) concluded the requirement that experimental validation data have similar cross section sensitivities to those found in the FSP design. A series of critical experiments was performed at ORCEF in the 1960s to support the Medium Power Reactor Experiment (MPRE) space reactor design. The small, compact critical assembly (SCCA) experiments were graphite- or beryllium-reflected assemblies of SS-clad, HEU-O2 fuel on a vertical lift machine. All five configurations were evaluated as benchmarks. Two of the five configurations were beryllium reflected, and further evaluated using the sensitivity and uncertainty analysis capabilities of SCALE 6.1. Validation of the example FSP design model was successful in reducing the primary uncertainty constituent, the Be(n,n) reaction, from 0.28 %dk/k to 0.0004 %dk/k. Further assessment of additional reactor physics measurements performed on the SCCA experiments may serve to further validate FSP design and operation.

  12. Criticality calculations of the Very High Temperature reactor Critical Assembly benchmark with Serpent and SCALE/KENO-VI

    SciTech Connect

    Bostelmann, Friederike; Hammer, Hans R.; Ortensi, Javier; Strydom, Gerhard; Velkov, Kiril; Zwermann, Winfried

    2015-12-30

    Within the framework of the IAEA Coordinated Research Project on HTGR Uncertainty Analysis in Modeling, criticality calculations of the Very High Temperature Critical Assembly experiment were performed as the validation reference to the prismatic MHTGR-350 lattice calculations. Criticality measurements performed at several temperature points at this Japanese graphite-moderated facility were recently included in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, and represent one of the few data sets available for the validation of HTGR lattice physics. Here, this work compares VHTRC criticality simulations utilizing the Monte Carlo codes Serpent and SCALE/KENO-VI. Reasonable agreement was found between Serpent and KENO-VI, but only the use of the latest ENDF cross section library release, namely the ENDF/B-VII.1 library, led to an improved match with the measured data. Furthermore, the fourth beta release of SCALE 6.2/KENO-VI showed significant improvements from the current SCALE 6.1.2 version, compared to the experimental values and Serpent.

  13. Criticality calculations of the Very High Temperature reactor Critical Assembly benchmark with Serpent and SCALE/KENO-VI

    DOE PAGESBeta

    Bostelmann, Friederike; Hammer, Hans R.; Ortensi, Javier; Strydom, Gerhard; Velkov, Kiril; Zwermann, Winfried

    2015-12-30

    Within the framework of the IAEA Coordinated Research Project on HTGR Uncertainty Analysis in Modeling, criticality calculations of the Very High Temperature Critical Assembly experiment were performed as the validation reference to the prismatic MHTGR-350 lattice calculations. Criticality measurements performed at several temperature points at this Japanese graphite-moderated facility were recently included in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, and represent one of the few data sets available for the validation of HTGR lattice physics. Here, this work compares VHTRC criticality simulations utilizing the Monte Carlo codes Serpent and SCALE/KENO-VI. Reasonable agreement was found between Serpent andmore » KENO-VI, but only the use of the latest ENDF cross section library release, namely the ENDF/B-VII.1 library, led to an improved match with the measured data. Furthermore, the fourth beta release of SCALE 6.2/KENO-VI showed significant improvements from the current SCALE 6.1.2 version, compared to the experimental values and Serpent.« less

  14. S-LSR, Cooler Ring Development at Kyoto University

    SciTech Connect

    Shirai, Toshiyuki; Fujimoto, Shinji; Ikegami, Masahiro; Noda, Akira; Souda, Hikaru; Tanabe, Mikio; Tongu, Hiromu; Noda, Koji; Shibuya, Shinji; Takeuchi, Takeshi; Fujimoto, Takeshi; Iwata, Soma; Takubo, Atsushi; Okamoto, Hiromi; Yuri, Yosuke; Grieser, Manfred; Syresin, Evgeny M.

    2006-03-20

    A compact ion cooler ring, S-LSR is under construction in Kyoto University. One of the subjects of S-LSR is a realization of the crystalline beams using the electron beam and the laser cooling. The ring is designed to be satisfied several required conditions for the beam ordering, such as a small betatron phase advance, a small magnetic error and a precise magnet alignment. The design phase advance per a period is less than 127 degree. The calculated closed orbit distortion and the stopband is less than 1 mm and 0.001 without correction, respectively.

  15. Experimental Equipments for Microwave Power Transmission in Kyoto University

    NASA Astrophysics Data System (ADS)

    Matsumoto, H.; Hashimoto, K.; Shinohara, N.; Mitani, T.

    2004-12-01

    RISH, Research Institute of Sustainable Humanosphere, of Kyoto University is one of the most active research laboratories for a microwave power transmission (MPT) and SSPS (Space Solar Power System) in Japan. Since the first MPT rocket experiment in the ionosphere was conducted in early 1980's by Radio Science Center for Space and Atmosphere (RASC), the former institute of the RISH, the RISH has conducted a variety of theoretical and experimental studies on the MPT and SSPS both in laboratory and field. In the paper, we will show the experimental equipments for the MPT and the SSPS research in the RISH. We have two facilities of METLAB (Microwave Energy Transmission LABoratory) and SPSLAB (SPS LABoratory) for the MPT researches. The METLAB is composed of an anechoic radio wave chamber and an experimental system specially designed for the MPT experiment. The SPSLAB is a facility to promote a systematic research for conceptual, technical and coordinative research with our colleagues in all parts of Japan including the industrial partners. These facilities are now open for inter-universities collaborations. The other available experimental facilities are multiple MPT systems with a phased array transmitter using phase controlled magnetrons (PCMs) at both 2.45 GHz and 5.8 GHz, and a semiconductor-based beam control system and a retrodirective target detecting system. These facilities are named SPORTS (Space POwer Radio Transmission System) 2.45 and SPORTS5.8.

  16. A Code to Produce Cell Averaged Cross Sections for Fast Critical Assemblies and Fast Power Reactors.

    Energy Science and Technology Software Center (ESTSC)

    1987-05-14

    Version 00 SLAROM solves the neutron integral transport equations to determine the flux distribution and spectra in a fast reactor lattice and calculates cell averaged effective cross sections. The code uses multigroup data of the type in DLC-111/JFS that use Bondarenko factors for resonance effects.

  17. Methanation assembly using multiple reactors

    DOEpatents

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  18. FEL Beamline for Wide Tunable Range and Beam Sharing System at Kyoto University

    SciTech Connect

    Bakr, Mahmoud; Yoshida, K.; Higashimura, K.; Ueda, S.; Kinjo, R.; Sonobe, T.; Kii, T.; Masuda, K.; Ohgaki, H.; Zen, H.

    2010-02-03

    A mid-infrared free electron laser (MIR-FEL) facility (KU-FEL: Kyoto University Free Electron Laser) has been constructed for developing energy materials in Institute of Advanced Energy (IAE), Kyoto University. The tunable range of KU-FEL was estimated as 5-13.2 {mu}m by numerical calculation to design the MIR-FEL transport line for application purposes. Aiming to increase the number of FEL users with different desires we decided to develop an FEL beam sharing system that is useful for multi-utilization at different end-stations. The transport line and the beam sharing system has been designed and constructed to the user stations. Applications of the MIR-FEL in the renewable energy research at Kyoto University will start as well.

  19. University Reactor Instrumentation Grant

    SciTech Connect

    S. M. Bajorek

    2000-02-01

    A noble gas air monitoring system was purchased through the University Reactor Instrumentation Grant Program. This monitor was installed in the Kansas State TRIGA reactor bay at a location near the top surface of the reactor pool according to recommendation by the supplier. This system is now functional and has been incorporated into the facility license.

  20. ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

    SciTech Connect

    G. Palmiotti

    2011-12-01

    The ENDF/B-VII.1 library is the latest revision to the United States' Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 418 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [1]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected 235U and 239Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as 236U capture. Other deficiencies, such as the overprediction of Pu solution system critical eigenvalues and a decreasing trend in calculated eigenvalue for

  1. ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

    NASA Astrophysics Data System (ADS)

    Kahler, A. C.; MacFarlane, R. E.; Mosteller, R. D.; Kiedrowski, B. C.; Frankle, S. C.; Chadwick, M. B.; McKnight, R. D.; Lell, R. M.; Palmiotti, G.; Hiruta, H.; Herman, M.; Arcilla, R.; Mughabghab, S. F.; Sublet, J. C.; Trkov, A.; Trumbull, T. H.; Dunn, M.

    2011-12-01

    The ENDF/B-VII.1 library is the latest revision to the United States' Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [M. B. Chadwick et al., "ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data," Nuclear Data Sheets, 112, 2887 (2011)]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected 235U and 239Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected

  2. ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

    SciTech Connect

    Kahler, A.C.; Herman, M.; Kahler,A.C.; MacFarlane,R.E.; Mosteller,R.D.; Kiedrowski,B.C.; Frankle,S.C.; Chadwick,M.B.; McKnight,R.D.; Lell,R.M.; Palmiotti,G.; Hiruta,H.; Herman,M.; Arcilla,R.; Mughabghab,S.F.; Sublet,J.C.; Trkov,A.; Trumbull,T.H.; Dunn,M.

    2011-12-01

    The ENDF/B-VII.1 library is the latest revision to the United States Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [M. B. Chadwick et al., 'ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data,' Nuclear Data Sheets, 112, 2887 (2011)]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected {sup 235}U and {sup 239}Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for

  3. Utilisation of British University Research Reactors.

    ERIC Educational Resources Information Center

    Duncton, P. J.; And Others

    British experience relating to the employment of university research reactors and subcritical assemblies in the education of nuclear scientists and technologists, in the training of reactor operators and for fundamental pure and applied research in this field is reviewed. The facilities available in a number of British universities and the uses…

  4. Criticality of spent reactor fuel

    SciTech Connect

    Harris, D.R.

    1987-01-01

    The storage capacity of spent reactor fuel pools can be greatly increased by consolidation. In this process, the fuel rods are removed from reactor fuel assemblies and are stored in close-packed arrays in a canister or skeleton. An earlier study examined criticality consideration for consolidation of Westinghouse fuel, assumed to be fresh, in canisters at the Millstone-2 spent-fuel pool and in the General Electric IF-300 shipping cask. The conclusions were that the fuel rods in the canister are so deficient in water that they are adequately subcritical, both in normal and in off-normal conditions. One potential accident, the water spill event, remained unresolved in the earlier study. A methodology is developed here for spent-fuel criticality and is applied to the water spill event. The methodology utilizes LEOPARD to compute few-group cross sections for the diffusion code PDQ7, which then is used to compute reactivity. These codes give results for fresh fuel that are in good agreement with KENO IV-NITAWL Monte Carlo results, which themselves are in good agreement with continuous energy Monte Carlo calculations. These methodologies are in reasonable agreement with critical measurements for undepleted fuel.

  5. Present status of PACS at Kyoto University Hospital: image workstation for clinical education

    NASA Astrophysics Data System (ADS)

    Minato, Kotaro; Komori, Masaru; Nakano, Yoshihisa; Okajima, Kaoru; Kimura, Ishu; Takahashi, Takashi; Konishi, Junji; Abe, Mituyuki; Gotoh, Yoshihiro; Sato, Kazuhiro

    1990-08-01

    The PAC system: KIDS (Kyoto University Hospital Image Database and Communication System) has been expanded to include several major digital imaging modalities such as X-ray CT, MRI, DSA and CR. The fiber optic high-speed local area network and the workstation with quick image handling are newly designed. The system (new KIDS) is intended to achieve a film-less environment in the department of radiology and to evaluate the feasibility of a hospital-wide PAC system. The present status of the system at the end of 1989 including a image workstation installed in a lecture hall for clinical education is described.

  6. University Reactor Instrumentation Program

    SciTech Connect

    Vernetson, W.G.

    1992-11-01

    Recognizing that the University Reactor Instrumentation Program was developed in response to widespread needs in the academic community for modernization and improvement of research and training reactors at institutions such as the University of Florida, the items proposed to be supported by this grant over its two year period have been selected as those most likely to reduce foreed outages, to meet regulatory concerns that had been expressed in recent years by Nuclear Regulatory Commission inspectors or to correct other facility problems and limitations. Department of Energy Grant Number DE-FG07-90ER129969 was provided to the University of Florida Training Reactor(UFTR) facility through the US Department of Energy's University Reactor Instrumentation Program. The original proposal submitted in February, 1990 requested support for UFTR facility instrumentation and equipment upgrades for seven items in the amount of $107,530 with $13,800 of this amount to be the subject of cost sharing by the University of Florida and $93,730 requested as support from the Department of Energy. A breakdown of the items requested and total cost for the proposed UFTR facility instrumentation and equipment improvements is presented.

  7. The features of neutronic calculations for fast reactors with hybrid cores on the basis of BFS-62-3A critical assembly experiments

    SciTech Connect

    Mitenkova, E. F.; Novikov, N. V.; Blokhin, A. I.

    2012-07-01

    The different (U-Pu) fuel compositions are considered for next generation of sodium fast breeder reactors. The considerable discrepancies in axial and radial neutron spectra for hybrid reactor systems compared to the cores with UO{sub 2} fuel cause increasing uncertainty of generating the group nuclear constants in those reactor systems. The calculation results of BFS-62-3A critical assembly which is considered as full-scale model of BN-600 hybrid core with steel reflector specify quite different spectra in local areas. For those systems the MCNP 5 calculations demonstrate significant sensitivity of effective multiplication factor K{sub eff} and spectral indices to nuclear data libraries. For {sup 235}U, {sup 238}U, {sup 239}Pu the results of calculated radial fission rate distributions against the reconstructed ones are analyzed. Comparative analysis of spectral indices, neutron spectra and radial fission rate distributions are performed using the different versions of ENDF/B, JENDL-3.3, JENDL-4, JEFF-3.1.1 libraries and BROND-3 for Fe, Cr isotopes. For analyzing the fission rate sensitivity to the plutonium presence in the fuel {sup 239}Pu is substituted for {sup 235}U (enrichment 90%) in the FA areas containing the plutonium. For {sup 235}U, {sup 238}U, {sup 239}Pu radial fission rate distributions the explanation of pick values discrepancies is based on the group fission constants analyses and possible underestimation of some features at the experimental data recovery method (Westcott factors, temperature dependence). (authors)

  8. REACTOR NOZZLE ASSEMBLY

    DOEpatents

    Capuder, F.C.; Dearwater, J.R.

    1959-02-10

    An improved nozzle assembly useful in a process for the direct reduction of uranium hexafluoride to uranium tetrafluoride by means of dissociated ammonia in a heated reaction vessel is descrlbed. The nozzle design provides for intimate mixing of the two reactants and at the same time furnishes a layer of dissociated ammonia adjacent to the interior wall of the reaction vessel, thus preventing build-up of the reaction product on the vessel wall.

  9. NRC Targets University Reactors.

    ERIC Educational Resources Information Center

    Marshall, Eliot

    1984-01-01

    The Nuclear Regulatory Commission (NRC) wants universities to convert to low-grade fuel in their research reactions. Researchers claim the conversion, which will bring U.S. reactors in line with a policy the NRC is trying to impress on foreigners, could be financially and scientifically costly. Impact of the policy is considered. (JN)

  10. On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

    SciTech Connect

    Samuel Bays; Ayodeji Alajo

    2010-05-01

    This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

  11. Microdosimetry of epithermal neutron field at the Kyoto University reactor.

    PubMed

    Onizuka, Y; Endo, S; Ishikawa, M; Hoshi, M; Takada, M; Kobayashi, T; Sakurai, Y; Utsumi, H; Uehara, S; Hayabuchi, N; Maeda, N; Takatuji, T; Fujika, K

    2002-01-01

    Microdosimetric spectra were measured in order to gain the microdosimetric parameters of some epithermal neutron fields. Changes in dose mean lineal energy YD as a function of depth of heavy water showed a trend of softening with heavy water of the beam. The neutron absorbed dose was obtained by using the frequency mean lineal energy. Results show good agreement with measurements with the activation method using gold foil. This study demonstrated how microdosimetric parameters change in radiation quality as a function of heavy water depth. PMID:12194334

  12. Simulated nuclear reactor fuel assembly

    DOEpatents

    Berta, V.T.

    1993-04-06

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  13. Simulated nuclear reactor fuel assembly

    DOEpatents

    Berta, Victor T.

    1993-01-01

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  14. Nuclear reactor composite fuel assembly

    DOEpatents

    Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  15. Reactivity control assembly for nuclear reactor

    DOEpatents

    Bollinger, Lawrence R.

    1984-01-01

    Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

  16. Vitiligo vulgaris and autoimmune diseases in Japan: A report from vitiligo clinic in Kyoto University Hospital.

    PubMed

    Tanioka, Miki; Yamamoto, Yosuke; Katoh, Mayumi; Takahashi, Kenzo; Miyachi, Yoshiki

    2009-01-01

    We reviewed the causes of "loss of skin color" in 144 patients, who visited Vitiligo Clinic of Kyoto University Hospital between April 2005 and August 2008. The numbers of patients with generalized and segmental Vitiligo vulgaris were 98 (68.1%) and 26 (18.1%), respectively. Small numbers of the patients suffered from Vogt-Koyanagi-Harada disease, piebaldism, congenital albinism, Hypomelanosis of Ito, post-inflammatory hypopigmentation, white leaf-shaped macules associated with tuberous sclerosis and nevus hypopigmentosus. One forth of the patients with generalized vitiligo had complications, while no complications were found in the patients with segmental vitiligo. Among the complications, autoimmune diseases dominated 43% (10 of 23 cases). Autoimmune thyroid diseases explained for the most of the complicated autoimmune diseases and were associated with 7.4% of the patients with generalized vitiligo. Minor autoimmune complications include myasthenia gravis, Sjogren syndrome and autoimmune nephritis. Reflecting the condition that our clinic is located in a university hospital, vitiligo patients with end-stage non-melanoma cancers of internal organs accounted for 8.4% of the patients of generalized vitiligo. PMID:20046588

  17. Critical Casimir forces for colloidal assembly.

    PubMed

    Nguyen, V D; Dang, M T; Nguyen, T A; Schall, P

    2016-02-01

    Critical Casimir forces attract increasing interest due to their opportunities for reversible particle assembly in soft matter and nano science. These forces provide a thermodynamic analogue of the celebrated quantum mechanical Casimir force that arises from the confinement of vacuum fluctuations of the electromagnetic field. In its thermodynamic analogue, solvent fluctuations, confined between suspended particles, give rise to an attractive or repulsive force between the particles. Due to its unique temperature dependence, this effect allows in situ control of reversible assembly. Both the force magnitude and range vary with the solvent correlation length in a universal manner, adjusting with temperature from fractions of the thermal energy, k B T, and nanometre range to several ten kT and micrometer length scale. Combined with recent breakthroughs in the synthesis of complex particles, critical Casimir forces promise the design and assembly of complex colloidal structures, for fundamental studies of equilibrium and out-of-equilibrium phase behaviour. This review highlights recent developments in this evolving field, with special emphasis on the dynamic interaction control to assemble colloidal structures, in and out of equilibrium. PMID:26750980

  18. Critical Casimir forces for colloidal assembly

    NASA Astrophysics Data System (ADS)

    Nguyen, V. D.; Dang, M. T.; Nguyen, T. A.; Schall, P.

    2016-02-01

    Critical Casimir forces attract increasing interest due to their opportunities for reversible particle assembly in soft matter and nano science. These forces provide a thermodynamic analogue of the celebrated quantum mechanical Casimir force that arises from the confinement of vacuum fluctuations of the electromagnetic field. In its thermodynamic analogue, solvent fluctuations, confined between suspended particles, give rise to an attractive or repulsive force between the particles. Due to its unique temperature dependence, this effect allows in situ control of reversible assembly. Both the force magnitude and range vary with the solvent correlation length in a universal manner, adjusting with temperature from fractions of the thermal energy, k B T, and nanometre range to several ten kT and micrometer length scale. Combined with recent breakthroughs in the synthesis of complex particles, critical Casimir forces promise the design and assembly of complex colloidal structures, for fundamental studies of equilibrium and out-of-equilibrium phase behaviour. This review highlights recent developments in this evolving field, with special emphasis on the dynamic interaction control to assemble colloidal structures, in and out of equilibrium.

  19. The Kyoto Tridimensional Spectrograph II on Subaru and the University of Hawaii 88 in Telescopes

    NASA Astrophysics Data System (ADS)

    Sugai, H.; Hattori, T.; Kawai, A.; Ozaki, S.; Hayashi, T.; Ishigaki, T.; Ishii, M.; Ohtani, H.; Shimono, A.; Okita, Y.; Matsubayashi, K.; Kosugi, G.; Sasaki, M.; Takeyama, N.

    2010-01-01

    In order to investigate physical conditions of ionized gas in galaxies, as well as its kinematics, we have developed the Kyoto tridimensional spectrograph II. It is a multimode optical instrument, including integral field spectrograph (IFS) and Fabry-Perot imager modes. We have designed it compact so that we can mount it on 2 m class telescopes as well as on the 8.2 m Subaru telescope. Special care was taken to obtain high-quality calibrations in the IFS mode. In order to remove the chromatic aberration of micropupil images produced by a lenslet array, we have introduced a corrector lens system behind the lenslet array. The internal calibration system simulates the telescope optics so that the system provides micropupil images identical to those produced by the telescope. The rigidness of the instrument provides the positional stability of micropupil images. We have succeeded in test observations of all the modes on Subaru and the University of Hawaii 88 in (UH88) telescopes and have verified the performance of the instrument. This includes the instrument efficiencies as well as the effective sky background subtraction and the minimization of crosstalk effects in the IFS mode. In the IFS mode a spatial resolution of 0.4'' was obtained in good seeing conditions. Each of 37 × 37 lenslets subtends 0.1'' in Subaru's case. This samples the image size well. A wider field of view is emphasized in the case of UH88.

  20. NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

    DOEpatents

    Stengel, F.G.

    1963-12-24

    A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

  1. Advanced absorber assembly design for breeder reactors

    SciTech Connect

    Pitner, A.L.; Birney, K.R.

    1980-01-01

    An advanced absorber assembly design has been developed for breeder reactor control rod applications that provides for improved in-reactor performance, longer lifetimes, and reduced fabrication costs. The design comprises 19 vented pins arranged in a circular array inside of round duct tubes. The absorber material is boron carbide; cladding and duct components are constructed from the modified Type 316 stainless steel alloy. Analyses indicate that this design will scram 30 to 40% faster than the reference FFTF absorber assembly. The basic design characteristics of this advanced FFTF absorber assembly are applicable to large core breeder reactor design concepts.

  2. University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor

    SciTech Connect

    Eric C. Woolstenhulme; Dana M. Hewit

    2008-09-01

    The Department of Energy’s Idaho National Laboratory, under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at Purdue University Reactor. With this work completed and in anticipation of other impending conversion projects, the INL convened and engaged the project participants in a structured discussion to capture the lessons learned. The lessons learned process has allowed us to capture gaps, opportunities, and good practices, drawing from the project team’s experiences. These lessons will be used to raise the standard of excellence, effectiveness, and efficiency in all future conversion projects.

  3. An international survey of physicians regarding clinical trials: a comparison between Kyoto University Hospital and Seoul National University Hospital

    PubMed Central

    2013-01-01

    Background International clinical trials are now rapidly expanding into Asia. However, the proportion of global trials is higher in South Korea compared to Japan despite implementation of similar governmental support in both countries. The difference in clinical trial environment might influence the respective physicians’ attitudes and experience towards clinical trials. Therefore, we designed a questionnaire to explore how physicians conceive the issues surrounding clinical trials in both countries. Methods A questionnaire survey was conducted at Kyoto University Hospital (KUHP) and Seoul National University Hospital (SNUH) in 2008. The questionnaire consisted of 15 questions and 2 open-ended questions on broad key issues relating to clinical trials. Results The number of responders was 301 at KUHP and 398 at SNUH. Doctors with trial experience were 196 at KUHP and 150 at SNUH. Among them, 12% (24/196) at KUHP and 41% (61/150) at SUNH had global trial experience. Most respondents at both institutions viewed clinical trials favorably and thought that conducting clinical trials contributed to medical advances, which would ultimately lead to new and better treatments. The main reason raised as a hindrance to conducting clinical trials was the lack of personnel support and time. Doctors at both university hospitals thought that more clinical research coordinators were required to conduct clinical trials more efficiently. KUHP doctors were driven mainly by pure academic interest or for their desire to find new treatments, while obtaining credits for board certification and co-authorship on manuscripts also served as motivation factors for doctors at SNUH. Conclusions Our results revealed that there might be two different approaches to increase clinical trial activity. One is a social level approach to establish clinical trial infrastructure providing sufficient clinical research professionals. The other is an individual level approach that would provide incentives to

  4. Scanning tunneling microscope assembly, reactor, and system

    DOEpatents

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  5. Fuel assembly for nuclear reactors

    DOEpatents

    Creagan, Robert J.; Frisch, Erling

    1977-01-01

    A new and improved fuel assembly is formed to minimize the amount of parasitic structural material wherein a plurality of hollow tubular members are juxtaposed to the fuel elements of the assembly. The tubular members may serve as guide tubes for control elements and are secured to a number of longitudinally spaced grid members along the fuel assembly. The grid members include means thereon engaging each of the fuel elements to laterally position the fuel elements in a predetermined array. Openings in the bottom of each hollow member serve as a shock absorber to cushion shock transmitted to the structure when the control elements are rapidly inserted in their corresponding tubular members.

  6. Electromagnetic Near Field Measurements of Two Critical Assemblies

    NASA Astrophysics Data System (ADS)

    Goettee, Jeffrey; Goorley, Tim; Mayo, Douglas; Myers, William; Goda, Joetta; Sage, Frank

    2015-04-01

    Preliminary measurements of the fast metal nuclear reactors at the National Criticality Experiments Research Center (NCERC) and at White Sands Missile Range (WSMR) within the past year characterize the very near field environment of these critical assemblies. Both reactors are fast, highly enriched uranium metal reactors and can be operated in a burst mode above prompt supercritical. Initial measurements of the electric and the magnetic fields within the reactor cell are consistent between the two facilities, and begin to describe the dependance on distance and polarization as might be assumed from initial Monte Carlo modelling of these facilities. The amplitude and time variation of the electric and magnetic fields are consistent with burst time scales. The polarization is consistent with the geometry of the source and with Compton scattering from fission gammas as the dominant ionization mechanism. An overview of the two fast neutron sources and the excursion dynamics, the experimental details, and summary of the modelling calculations will be provided as background.

  7. Dismantlement of the TSF-SNAP Reactor Assembly

    SciTech Connect

    Peretz, Fred J

    2009-01-01

    This paper describes the dismantlement of the Tower Shielding Facility (TSF)?Systems for Nuclear Auxiliary Power (SNAP) reactor, a SNAP-10A reactor used to validate radiation source terms and shield performance models at Oak Ridge National Laboratory (ORNL) from 1967 through 1973. After shutdown, it was placed in storage at the Y-12 National Security Complex (Y-12), eventually falling under the auspices of the Highly Enriched Uranium (HEU) Disposition Program. To facilitate downblending of the HEU present in the fuel elements, the TSF-SNAP was moved to ORNL on June 24, 2006. The reactor assembly was removed from its packaging, inspected, and the sodium-potassium (NaK) coolant was drained. A superheated steam process was used to chemically react the residual NaK inside the reactor assembly. The heat exchanger assembly was removed from the top of the reactor vessel, and the criticality safety sleeve was exchanged for a new safety sleeve that allowed for the removal of the vessel lid. A chain-mounted tubing cutter was used to separate the lid from the vessel, and the 36 fuel elements were removed and packaged in four U.S. Department of Transportation 2R/6M containers. The fuel elements were returned to Y-12 on July 13, 2006. The return of the fuel elements and disposal of all other reactor materials accomplished the formal objectives of the dismantlement project. In addition, a project model was established for the handling of a fully fueled liquid-metal?cooled reactor assembly. Current criticality safety codes have been benchmarked against experiments performed by Atomics International in the 1950s and 1960s. Execution of this project provides valuable experience applicable to future projects addressing space and liquid-metal-cooled reactors.

  8. Reactivity control assembly for nuclear reactor. [LMFBR

    DOEpatents

    Bollinger, L.R.

    1982-03-17

    This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

  9. Lateral restraint assembly for reactor core

    DOEpatents

    Gorholt, Wilhelm; Luci, Raymond K.

    1986-01-01

    A restraint assembly for use in restraining lateral movement of a reactor core relative to a reactor vessel wherein a plurality of restraint assemblies are interposed between the reactor core and the reactor vessel in circumferentially spaced relation about the core. Each lateral restraint assembly includes a face plate urged against the outer periphery of the core by a plurality of compression springs which enable radial preloading of outer reflector blocks about the core and resist low-level lateral motion of the core. A fixed radial key member cooperates with each face plate in a manner enabling vertical movement of the face plate relative to the key member but restraining movement of the face plate transverse to the key member in a plane transverse to the center axis of the core. In this manner, the key members which have their axes transverse to or subtending acute angles with the direction of a high energy force tending to move the core laterally relative to the reactor vessel restrain such lateral movement.

  10. Beryllium reflected cavity reactor for UF6 critical experiments

    NASA Technical Reports Server (NTRS)

    Jarvis, G. A.; Bernard, W.; Helmick, H. H.; White, R.

    1975-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diam by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials are available. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

  11. Criticality Safety Evaluation of the LLNL Inherently Safe Subcritical Assembly (ISSA)

    SciTech Connect

    Percher, Catherine

    2012-06-19

    The LLNL Nuclear Criticality Safety Division has developed a training center to illustrate criticality safety and reactor physics concepts through hands-on experimental training. The experimental assembly, the Inherently Safe Subcritical Assembly (ISSA), uses surplus highly enriched research reactor fuel configured in a water tank. The training activities will be conducted by LLNL following the requirements of an Integration Work Sheet (IWS) and associated Safety Plan. Students will be allowed to handle the fissile material under the supervision of LLNL instructors. This report provides the technical criticality safety basis for instructional operations with the ISSA experimental assembly.

  12. Nuclear reactor shutdown control rod assembly

    DOEpatents

    Bilibin, Konstantin

    1988-01-01

    A temperature responsive, self-actuated nuclear reactor shutdown control rod assembly 10. The upper end 18 of a lower drive line 17 fits within the lower end of an upper drive line 12. The lower end (not shown) of the lower drive line 17 is connected to a neutron absorber. During normal temperature conditions the lower drive line 17 is supported by detent means 22,26. When an overtemperature condition occurs thermal actuation means 34 urges ring 26 upwardly sufficiently to allow balls 22 to move radially outwardly thereby allowing lower drive line 17 to move downwardly toward the core of the nuclear reactor resulting in automatic reduction of the reactor powder.

  13. Universal Assembly for Captive Bolts

    NASA Technical Reports Server (NTRS)

    Marke, M. L.; Hagopian, B.

    1982-01-01

    New method allows for virtually any bolt to be easily converted to "captive" bolt. Method eliminates need for separate design for each application. Cup-shaped washer that is flattened secures tap to bolt. Wire attached to tab holds bolt assembly captive. Flattening washer can also be done during installation of bolt. Wash, tab and spacer are all made of corrosion-resistant steel.

  14. Nuclear criticality research at the University of New Mexico

    SciTech Connect

    Busch, R.D.

    1997-06-01

    Two projects at the University of New Mexico are briefly described. The university`s Chemical and Nuclear Engineering Department has completed the final draft of a primer for MCNP4A, which it plans to publish soon. The primer was written to help an analyst who has little experience with the MCNP code to perform criticality safety analyses. In addition, the department has carried out a series of approach-to-critical experiments on the SHEBA-II, a UO{sub 2}F{sub 2} solution critical assembly at Los Alamos National Laboratory. The results obtained differed slightly from what was predicted by the TWODANT code.

  15. Student research in criticality safety at the University of Arizona

    SciTech Connect

    Hetrick, D.L.

    1997-06-01

    A very brief progress report on four University of Arizona student projects is given. Improvements were made in simulations of power pulses in aqueous solutions, including the TWODANT model. TWODANT calculations were performed to investigate the effect of assembly shape on the expansion coefficient of reactivity for solutions. Preliminary calculations were made of critical heights for the Los Alamos SHEBA assembly. Calculations to support French experiments to measure temperature coefficients of dilute plutonium solutions confirmed feasibility.

  16. Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

    SciTech Connect

    Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.; Bounds, J.A.; Kimpland, R.H.; Damjanovich, R.P.; Jaegers, P.J.

    1997-08-01

    Experiments were performed to measure a variety of parameters for SHEBA: behavior of the facility during transient and steady-state operation; characteristics of the SHEBA fuel; delayed-critical solution height vs solution temperature; initial reactor period and reactivity vs solution height; calibration of power level vs reactor power instrumentation readings; flux profile in SHEBA; radiation levels and neutron spectra outside the assembly for code verification and criticality alarm and dosimetry purposes; and effect on reactivity of voids in the fuel.

  17. Reactor building assembly and method of operation

    SciTech Connect

    Fennern, L.E.; Caraway, H.A.; Hsu, Li C.

    1993-06-01

    A reactor building assembly is described comprising: a reactor pressure vessel containing a reactor core for generating heat in the form of steam; a containment vessel enclosing said pressure vessel; a first enclosure surrounding said containment vessel and spaced laterally therefrom to define a first chamber there between, and having a top and a bottom; a second enclosure surrounding said first enclosure and spaced laterally therefrom to define a second chamber there between, and having a top and a bottom; a building inlet for receiving into said second chamber fresh air from outside said second enclosure; a building outlet for discharging stale air from said first chamber; a transfer duct disposed through said first enclosure selectively joining in flow communication said first and second chambers; said building inlet being disposed at said second enclosure top, said building outlet being disposed at said first enclosure top, and said transfer duct being disposed adjacent said first enclosure bottom for allowing said fresh air to flow downwardly by gravity through said second chamber and through said transfer duct into said first chamber for cooling said first chamber, said stale air flowing upwardly by natural buoyancy for discharger from said first chamber through said building outlet; an exhaust stack disposed above said building outlet and in flow communication therewith for channeling upwardly said stale air from said first chamber for discharge into the surrounding environs; and a passive first driving means for increasing flow of said stale air from said building outlet comprising: an isolation pool containing isolation water; an isolation condenser disposed in said isolation pool, and joined in flow communication with said reactor pressure vessel for receiving primary steam therefrom, said primary steam being cooled in said isolation condenser for heating said isolation water to generate secondary steam.

  18. Test observations of the Kyoto Tridimensional Spectrograph II at the University of Hawaii 88-in and Subaru Telescopes

    NASA Astrophysics Data System (ADS)

    Sugai, Hajime; Hattori, Takashi; Kawai, Atsushi; Ozaki, Shinobu; Kosugi, George; Ohtani, Hiroshi; Hayashi, Tadashi; Ishigaki, Tsuyoshi; Ishii, Motomi; Sasaki, Minoru; Shimono, Atsushi; Okita, Yoshiko; Sudo, Jun; Takeyama, Norihide

    2004-09-01

    In order to investigate the physical conditions of ionized gas in galaxies, as well as its kinematics, we have developed the Kyoto tridimensional spectrograph II (3DII). It is a multi-mode instrument designed for Cassegrain focus, including integral field spectrograph (IFS) and Fabry-Perot imager modes. We have designed it compact so that we can mount it at 2-m class telescopes as well as at 8-m Subaru telescope. We have succeeded in test observations of the 3DII. In the IFS mode the spatial resolution of ~ 0".5 and 0".4 was obtained in 30-minute exposures at University of Hawaii 88-inch (UH88) and Subaru, respectively, in relatively good weather conditions. Each of 37 × 37 microlenses subtends ~ 0".1 in Subaru's case. This samples well the image size. A wider field of view is emphasized in the case of UH88. Because our micropupil spectroscopy is free from a slit effect, we have reached the accuracy of an order of one tenth of a pixel for deriving velocity fields in terms of velocity center while the full width at half maximum of the instrumental profile corresponds to two pixels. At Subaru we have used a container designed in a collaboration with National Astronomical Observatory, Japan: it fits with a robotic instrument exchanger. The containerincludes two heat exchangers to keep its surface cool and void degrading the image quality. We have established effective observational equences by realizing a software interface with Subaru operating system. ome results from target observations are shown.

  19. University Budgeting for Critical Mass and Competition.

    ERIC Educational Resources Information Center

    Jones, L. R.

    A critical mass strategy for university academic budgeting, resource planning, and management is described, and state government budgetary and program policy for universities is analyzed. The use of "rational systems management" in universities is discussed as a means of placing in perspective the critical mass resource decision model. Attention…

  20. Experience With A Small Scale All Digital CT And MRI Clinical Service Unit: Present Status Of Kyoto University Hospital Image Database And Communication System

    NASA Astrophysics Data System (ADS)

    Minato, K.; Komori, M.; Nakano, Y.

    1988-06-01

    Kyoto University Hospital is currently developing a prototype PAC system named KIDS (Kyoto univ. hosp. Image Database and communication System). The present goal of the system is to achieve the totally digital CT and MRI unit in the radiological department. Because KIDS is designed as a first step of a long-range plan towards a hospital wide system, it includes all of the basic functions required in realizing the PAC system, such as communication networks, a long term archiving unit, a laser film printer and image workstations. The system concept, architecture and current status are described in this paper. Our early experience and evaluations with the system in a clinical environment are also mentioned.

  1. Fast critical experiment data for space reactors

    NASA Astrophysics Data System (ADS)

    Collins, P. J.; McFarlane, H. F.; Olsen, D. N.; Atkinson, C. A.; Ross, J. R.

    Data from a number of previous critical experiments exist that are relevant to the design concepts being considered for SP-100 and MMW space reactors. Although substantial improvements in experiment techniques have since made some of the measured quantities somewhat suspect, the basic criticality data are still useful in most cases. However, experiments require recalculation with modern computational methods and nuclear cross section data before they can be applied to today's designs. Recently, about 20 fast benchmark critical experiments were calculated with the latest ENDF/B data and modern transport codes. These calculations were undertaken as a part of the planning process for a new series of benchmark experiments aimed at supporting preliminary designs of SP-100 and MMW space reactors.

  2. University of Virginia Reactor Facility Decommissioning Results

    SciTech Connect

    Ervin, P. F.; Lundberg, L. A.; Benneche, P. E.; Mulder, R. U.; Steva, D. P.

    2003-02-24

    The University of Virginia Reactor Facility started accelerated decommissioning in 2002. The facility consists of two licensed reactors, the CAVALIER and the UVAR. This paper will describe the progress in 2002, remaining efforts and the unique organizational structure of the project team.

  3. University Reactor Matching Grants Program

    SciTech Connect

    John Valentine; Farzad Rahnema; Said Abdel-Khalik

    2003-02-14

    During the 2002 Fiscal year, funds from the DOE matching grant program, along with matching funds from the industrial sponsors, have been used to support research in the area of thermal-hydraulics. Both experimental and numerical research projects have been performed. Experimental research focused on two areas: (1) Identification of the root cause mechanism for axial offset anomaly in pressurized water reactors under prototypical reactor conditions, and (2) Fluid dynamic aspects of thin liquid film protection schemes for inertial fusion reactor chambers. Numerical research focused on two areas: (1) Multi-fluid modeling of both two-phase and two-component flows for steam conditioning and mist cooling applications, and (2) Modeling of bounded Rayleigh-Taylor instability with interfacial mass transfer and fluid injection through a porous wall simulating the ''wetted wall'' protection scheme in inertial fusion reactor chambers. Details of activities in these areas are given.

  4. Safety analysis report for the Hanford Critical Mass Laboratory: Supplement No. 2. Experiments with heterogeneous assemblies

    SciTech Connect

    Gore, B.F.; Davenport, L.C.

    1981-04-01

    Factors affecting the safety of criticality experiments using heterogeneous assemblies are described and assessed. It is concluded that there is no substantial change in safety from experiments already being routinely performed at the Critical Mass Laboratory (CML), and that laboratory and personnel safety are adequately provided by the combination of engineered and administrative safety limits enforced at the CML. This conclusion is based on the analysis of operational controls, potential hazards, and the consequences of accidents. Contingencies considered that could affect nuclear criticality include manual changes in fuel loadings, water flooding, fire, explosion, loss of services, earthquake, windstorm, and flood. Other potential hazards considered include radiation exposure to personnel, and potential releases within the Assembly Room and outside to the environment. It is concluded that the Maximum Credible Nuclear Burst of 3 x 10/sup 18/ fissions (which served as the design basis for the CML) is valid for heterogeneous assemblies as well as homogeneous assemblies. This is based upon examination of the results of reactor destructive tests and the results of the SL-1 reactor destructive accident. The production of blast effects which might jeopardize the CML critical assembly room (of thick reinforced concrete) is not considered credible due to the extreme circumstances required to produce blast effects in reactor destructive tests. Consequently, it is concluded that, for experiments with heterogeneous assemblies, the consequences of the Maximum Credible Burst are unchanged from those previously estimated for experiments with homogeneous systems.

  5. Analysis of subcritical experiments using fresh and spent research reactor fuel assemblies

    NASA Astrophysics Data System (ADS)

    Zino, John Frederick

    1999-11-01

    This research investigated the concepts associated with crediting the burnup of spent nuclear fuel assemblies for the purposes of criticality safety. To accomplish this, a collaborative experimental research program was undertaken between Westinghouse, the University of Missouri Research Reactor (MURR) facility and Oak Ridge National Laboratory (ORNL). The purpose of the program was to characterize the subcritical behavior of a small array of fresh and spent MURR fuel assemblies using the 252Cf Source-driven noise technique. An aluminum test rig was built which was capable of holding up to four, highly enriched (93.15 wt.% 235U) MURR fuel assemblies in a 2 x 2 array. The rig was outfitted with one source and four detector drywells which allowed researchers to perform active neutron noise measurements on the array of fuel assemblies. The 1 atmosphere gas 3He neutron detectors used to perform the measurements were quenched with CF4 gas to allow improved discrimination of the neutron signals in the very high gamma-ray fields associated with spent fuel (˜8000 R/hr). In addition, the detector drywells were outfitted with 1″ lead collars to provide additional gamma-ray shielding from the spent fuel. Reactivity changes were induced in the subcritical lattice by replacing individual fresh assemblies (in a 4-assembly array) with spent assemblies of known, maximum burnup (143 Mw-D). The absolute and relative measured reactivity changes were then compared to those predicted by three-dimensional Monte Carlo calculations. The purpose of these comparisons was to investigate the accuracy of modern transport theory depletion calculations to accurately simulate the reactivity effects of burnup in spent nuclear fuel. A total of seven subcritical measurements were performed at the MURR reactor facility on July 20th and 27th, 1998. These measurements generated several estimates of prompt neutron decay constants (alpha) and ratios of spectral densities through frequency correlations

  6. Nuclear reactor removable radial shielding assembly having a self-bowing feature

    DOEpatents

    Pennell, William E.; Kalinowski, Joseph E.; Waldby, Robert N.; Rylatt, John A.; Swenson, Daniel V.

    1978-01-01

    A removable radial shielding assembly for use in the periphery of the core of a liquid-metal-cooled fast-breeder reactor, for closing interassembly gaps in the reactor core assembly load plane prior to reactor criticality and power operation to prevent positive reactivity insertion. The assembly has a lower nozzle portion for inserting into the core support and a flexible heat-sensitive bimetallic central spine surrounded by blocks of shielding material. At refueling temperature and below the spine is relaxed and in a vertical position so that the tolerances permitted by the interassembly gaps allow removal and replacement of the various reactor core assemblies. During an increase in reactor temperature from refueling to hot standby, the bimetallic spine expands, bowing the assembly toward the core center line, exerting a radially inward gap-closing-force on the above core load plane of the reactor core assembly, closing load plane interassembly gaps throughout the core prior to startup and preventing positive reactivity insertion.

  7. Universities and Globalization: Critical Perspectives.

    ERIC Educational Resources Information Center

    Currie, Jan, Ed.; Newson, Janice, Ed.

    The 14 papers in this collection examine how a globalizing political economy affects the way universities are governed, discussing practices such as managerialism, accountability, and privatization which represent a shift toward business values and a market agenda. Part 1 gives a theoretical overview of the globalization agenda. Part 2 gives three…

  8. University Rankings in Critical Perspective

    ERIC Educational Resources Information Center

    Pusser, Brian; Marginson, Simon

    2013-01-01

    This article addresses global postsecondary ranking systems by using critical-theoretical perspectives on power. This research suggests rankings are at once a useful lens for studying power in higher education and an important instrument for the exercise of power in service of dominant norms in global higher education. (Contains 1 table and 1…

  9. Benchmark Evaluation of the Medium-Power Reactor Experiment Program Critical Configurations

    SciTech Connect

    Margaret A. Marshall; John D. Bess

    2013-02-01

    A series of small, compact critical assembly (SCCA) experiments were performed in 1962-1965 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for the Medium-Power Reactor Experiment (MPRE) program. The MPRE was a stainless-steel clad, highly enriched uranium (HEU)-O2 fuelled, BeO reflected reactor design to provide electrical power to space vehicles. Cooling and heat transfer were to be achieved by boiling potassium in the reactor core and passing vapor directly through a turbine. Graphite- and beryllium-reflected assemblies were constructed at ORCEF to verify the critical mass, power distribution, and other reactor physics measurements needed to validate reactor calculations and reactor physics methods. The experimental series was broken into three parts, with the third portion of the experiments representing the beryllium-reflected measurements. The latter experiments are of interest for validating current reactor design efforts for a fission surface power reactor. The entire series has been evaluated as acceptable benchmark experiments and submitted for publication in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and in the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  10. SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configurations

    SciTech Connect

    Gauld, I.C.

    2000-03-01

    Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the US, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k{sub eff} values within about 1% {Delta}k/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models.

  11. SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configuration

    SciTech Connect

    Gauld, I.C.

    2000-03-16

    Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the United States, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k{sub eff} values within about 1% {Delta}k/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models.

  12. Automatic coolant flow control device for a nuclear reactor assembly

    DOEpatents

    Hutter, E.

    1984-01-27

    A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

  13. Critical Capability Pedagogies and University Education

    ERIC Educational Resources Information Center

    Walker, Melanie

    2010-01-01

    The article argues for an alliance of the capability approach developed by Amartya Sen with ideas from critical pedagogy for undergraduate university education which develops student agency and well being on the one hand, and social change towards greater justice on the other. The purposes of a university education in this article are taken to…

  14. Criticality calculations for the VR-1 reactor with IRT-3M-HEU fuel and IRT-4MLEU fuel.

    SciTech Connect

    Hanan, N. A.; Matos, J. E.

    2007-01-17

    At The request of the Czech Technical University in Prague, ANL has performed independent verification calculations using the MCNP Monte Carlo code for three core configurations of the VR-1 reactor: a current core configuration B1 with HEU (36%) IRT-3M fuel assemblies and planned core configurations C1 and C2 with LEU (19.7%) IRT-4M fuel assemblies. Details of these configurations were provided to ANL by CTU. For core configuration B1, criticality calculations were performed for two sets of control rod positions provided to ANL by CTU. For core configurations C1 and C2, criticality calculations were done for cases with all control rods at the top positions, all control rods at the bottom positions, and two critical states of the reactor for different control rod positions. In addition, sensitivity studies for variation of the {sup 235}U mass in each fuel assembly and variation of the fuel meat and cladding thicknesses in each of the fuel tubes were done for the C1 core configuration. Finally the reactivity worth of the individual control rods was calculated for the B1, C1, and C2 core configurations.

  15. Criticality safety analysis on fissile materials in Fukushima reactor cores

    SciTech Connect

    Liu, Xudong; Lemaitre-Xavier, E.; Ahn, Joonhong; Hirano, Fumio

    2013-07-01

    The present study focuses on the criticality analysis for geological disposal of damaged fuels from Fukushima reactor cores. Starting from the basic understanding of behaviors of plutonium and uranium, a scenario sequence for criticality event is considered. Due to the different mobility of plutonium and uranium in geological formations, the criticality safety is considered in two parts: (1) near-field plutonium system and (2) far-field low enriched uranium (LEU) system. For the near-field plutonium system, a mathematical analysis for pure-solute transport was given, assuming a particular buffer material and waste form configuration. With the transport and decay of plutonium accounted, the critical mass of plutonium was compared with the initial load of a single canister. Our calculation leads us to the conclusion that our system with the initial loading being the average mass of plutonium in an assembly just before the accident is very unlikely to become critical over time. For the far-field LEU system, due to the uncertainties in the geological and geochemical conditions, calculations were made in a parametric space that covers the variation of material compositions and different geometries. Results show that the LEU system could not remain sub-critical within the entire parameter space assumed, although in the iron-rich rock, the neutron multiplicity is significantly reduced.

  16. TRIPOLI-4 criticality calculations for MOX fuelled SNEAK 7A and 7B fast critical assemblies

    SciTech Connect

    Lee, Y. K.

    2012-07-01

    A prototype Generation IV fast neutron reactor is under design and development in France. The MOX fuel will be introduced into this self-generating core in order to demonstrate low net plutonium production. To support the TRIPOLI-4 Monte Carlo transport code in criticality calculations of fast reactors, the effective delayed neutron fraction {beta}eff estimation and the Probability Tables (PT) option to treat the unresolved resonance region of cross-sections are two essentials. In this study, TRIPOLI-4 calculations have been made using current nuclear data libraries JEFF-3.1.1 and ENDF/B-VII.0 to benchmark the reactor physics parameters of the MOX fuelled SNEAK 7A and 7B fast critical assemblies. TRIPOLI-4 calculated K{sub eff} and {beta}eff of the homogeneous R-Z models and the 3D multi-cell models have been validated against the measured ones. The impact of the PT option on K{sub eff} is 340 {+-} 10 pcm for SNEAK 7A core and 410 {+-} 12 pcm for 7B. Four-group spectra and energy spectral indices, f8/f5, f9/f5, and c8/f5 in the two SNEAK cores have also been calculated with the TRIPOLI-4 mesh tally. Calculated spectrum-hardening index f8/f5 is 0.0418 for SNEAK 7A and 0.0315 for 7B. From this study the SNEAK 3D models have been verified for the next revision of IRPhE (International Handbook of Evaluated Reactor Physics Benchmark Experiments). (authors)

  17. Use of Sensitivity and Uncertainty Analysis in the Design of Reactor Physics and Criticality Benchmark Experiments for Advanced Nuclear Fuel

    SciTech Connect

    Rearden, B.T.; Anderson, W.J.; Harms, G.A.

    2005-08-15

    Framatome ANP, Sandia National Laboratories (SNL), Oak Ridge National Laboratory (ORNL), and the University of Florida are cooperating on the U.S. Department of Energy Nuclear Energy Research Initiative (NERI) project 2001-0124 to design, assemble, execute, analyze, and document a series of critical experiments to validate reactor physics and criticality safety codes for the analysis of commercial power reactor fuels consisting of UO{sub 2} with {sup 235}U enrichments {>=}5 wt%. The experiments will be conducted at the SNL Pulsed Reactor Facility.Framatome ANP and SNL produced two series of conceptual experiment designs based on typical parameters, such as fuel-to-moderator ratios, that meet the programmatic requirements of this project within the given restraints on available materials and facilities. ORNL used the Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI) to assess, from a detailed physics-based perspective, the similarity of the experiment designs to the commercial systems they are intended to validate. Based on the results of the TSUNAMI analysis, one series of experiments was found to be preferable to the other and will provide significant new data for the validation of reactor physics and criticality safety codes.

  18. University Reactor Conversion Lessons Learned Workshop for Texas A&M University Nuclear Science Center Reactor

    SciTech Connect

    Eric C. Woolstenhulme; Dana M. Meyer

    2007-04-01

    The objectives of this meeting were to capture the observations, insights, issues, concerns, and ideas of those involved in the Texas A&M University Nuclear Science Center (TAMU NSC) TRIGA Reactor Conversion so that future efforts can be conducted with greater effectiveness, efficiency, and with fewer challenges. This workshop was held in conjunction with a similar workshop for the University of Florida Reactor Conversion. Some of the generic lessons from that workshop are included in this report for completeness.

  19. The World Nuclear University Alumni Assembly

    SciTech Connect

    White-Horton, Jessica L; Lynch, Patrick D; Gilligan, Kimberly V; Garner, James R; Guzzardo, Tyler; Kuhn, Michael J; Rowe, Nathan C

    2014-01-01

    The World Nuclear University Summer Institute was established by the World Nuclear Association in 2005 as a program for future leaders in the nuclear field. Since the Summer Institute s inception in 2005, a total of some 800 fellows from more than 70 countries have participated in the program. In 2012, the World Nuclear University held its first ever alumni event at the IAEA in Vienna, Austria, and at that time, the precedent was set that the reunion would be held biennially. The 2014 alumni assembly was held at Oak Ridge National Laboratory from March 31 April 4, 2014. The event offered three separate areas of opportunities for the participating alumni: professional development, leadership, and peer-to-peer engagement. The professional development consisted of training groups, while the leadership will involve discussions with invited leaders, including members of the Blue Ribbon Commission. The peer-to-peer engagement not only give past fellows a chance to reconnect with their own classmates, but it allowed for further international engagement, between the speakers and alumni, as well as between the classes themselves.

  20. Sharing of the RPI Reactor Critical Facility (RCF). Final summary report, January 1988--September 1995

    SciTech Connect

    Harris, D.R.

    1995-09-01

    Rensselaer Polytechnic Institute (RPI) has participated for a number of years in Sharing of the Reactor Critical Facility (RCF) under the U.S. Department of Energy University Reactor Sharing Program. In September of each year a Sharing invitation is sent to 92 public and private high schools and to 74 colleges and universities within about a 3 hour drive to the RCF (Appendix B). Each year about 10 such educational institutions send groups to share the RCF.

  1. Temperature measuring analysis of the nuclear reactor fuel assembly

    SciTech Connect

    Urban, F. E-mail: zdenko.zavodny@stuba.sk; Kučák, L. E-mail: zdenko.zavodny@stuba.sk; Bereznai, J. E-mail: zdenko.zavodny@stuba.sk; Závodný, Z. E-mail: zdenko.zavodny@stuba.sk; Muškát, P. E-mail: zdenko.zavodny@stuba.sk

    2014-08-06

    Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuel assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.

  2. Detachable connection for a nuclear reactor fuel assembly

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.

    1986-01-01

    A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engageable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

  3. Detachable connection for a nuclear reactor fuel assembly

    DOEpatents

    Christiansen, D.W.; Karnesky, R.A.

    1983-08-29

    A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engagable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

  4. Fast critical assembly safeguards. Summary report, October 1978-September 1979

    SciTech Connect

    Not Available

    1980-02-01

    Nuclear material inventory verification techniques for large split-table type fast critical assemblies are being studied under this program. Emphasis has been given to techniques that minimize fuel handling in order to reduce facility down time and radiation exposure to the inventory team. The techniques studied include autoradiography, reactivity, and spectral index measurements. Autoradiographic techniques capable of verifying the core loading diagram for a uranium or plutonium core have been demonstrated and examples of results are given. Results of reactivity and spectral index measurements for detection of simulated diversion of fuel from a large plutonium core are presented. These measurements indicate that the sensitivity of reactivity measurements to fuel removal is sufficient to detect a removal of 2 kg in a 2600 kg core. Reactivity compensation by the addition of polyethylene was readily detected by spectral index measurements using indium foils, if the foils were located within 10 cm of the moderator. The measurements made in this study were relative to a known reactor configuration and were made using facility equipment. Independent verification of reactivity and spectral index measurements by a safeguards inspector may be difficult. 27 figures, 5 tables.

  5. University Reactor Sharing Program. Final report, September 30, 1992--September 29, 1994

    SciTech Connect

    Wehring, B.W.

    1995-01-01

    Over the past 20 years, the number of nuclear reactors on university campuses in the US declined from more than 70 to less than 40. Contrary to this trend, The University of Texas at Austin constructed a new reactor facility at a cost of $5.8 million. The new reactor facility houses a new TRIGA Mark II reactor which replaces an in-ground TRIGA Mark I reactor located in a 50-year old building. The new reactor facility was constructed to strengthen the instruction and research opportunities in nuclear science and engineering for both undergraduate and graduate students at The University of Texas. On January 17, 1992, The University of Texas at Austin received a license for operation of the new reactor. Initial criticality was achieved on March 12, 1992, and full power operation, on March 25, 1992. The UT-TRIGA research reactor provides hands-on education, multidisciplinary research and unique service activities for academic, medical, industrial, and government groups. Support by the University Reactor Sharing Programs increases the availability of The University of Texas reactor facility for use by other educational institutions which do not have nuclear reactors.

  6. Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask

    SciTech Connect

    Wagner, J. C.

    2008-01-31

    The Interim Staff Guidance on bumup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission's Spent Fuel Project Office, recommends a bumup measurement for each assembly to confirm the reactor record and compliance with the assembly bumup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. This report presents a computational criticality safety analysis of the consequences of misloading fuel assemblies in a highcapacity cask that relies on burnup credit for criticality safety. The purpose of this report is to provide a quantitative understanding of the effects of fuel misloading events on safety margins. A wide variety of fuel-misloading configurations are investigated and results are provided for informational purposes. This report does not address the likelihood of occurrence for any of the misload configurations considered. For representative, qualified bumup-enrichment combinations, with and without fission products included, misloading two assemblies that are underburned by 75% results in an increase in keff of 0.025-0.045, while misloading four assemblies that are underburned by 50% also results in an increase in keff of 0.025-0.045. For the cask and conditions considered, a reduction in bumup of 20% in all assemblies results in an increase in kff of less than 0.035. Misloading a single fresh assembly with 3, 4, or 5 wt% 235U enrichment results in an increase in keffof--0.02, 0.04, or 0.06, respectively. The report concludes with a summary of these and other important findings, as well as a discussion of relevant issues that should be considered when assessing the appropriate role of burnup measurements.

  7. Advanced Plasma Pyrolysis Assembly (PPA) Reactor and Process Development

    NASA Technical Reports Server (NTRS)

    Wheeler, Richard R., Jr.; Hadley, Neal M.; Dahl, Roger W.; Abney, Morgan B.; Greenwood, Zachary; Miller, Lee; Medlen, Amber

    2012-01-01

    Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA's Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development.

  8. Reactor cell assembly for use in spectroscopy and microscopy applications

    SciTech Connect

    Grindstaff, Quirinus; Stowe, Ashley Clinton; Smyrl, Norm; Powell, Louis; McLane, Sam

    2015-08-04

    The present disclosure provides a reactor cell assembly that utilizes a novel design and that is wholly or partially manufactured from Aluminum, such that reactions involving Hydrogen, for example, including solid-gas reactions and thermal decomposition reactions, are not affected by any degree of Hydrogen outgassing. This reactor cell assembly can be utilized in a wide range of optical and laser spectroscopy applications, as well as optical microscopy applications, including high-temperature and high-pressure applications. The result is that the elucidation of the role of Hydrogen in the reactions studied can be achieved. Various window assemblies can be utilized, such that high temperatures and high pressures can be accommodated and the signals obtained can be optimized.

  9. Critical experiments at Sandia National Laboratories : technical meeting on low-power critical facilities and small reactors.

    SciTech Connect

    Harms, Gary A.; Ford, John T.; Barber, Allison Delo

    2010-11-01

    Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-III is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide benchmark

  10. Sharing of Rensselaer Polytechnic Institute Reactor Critical Facility (RCF)

    SciTech Connect

    1995-11-30

    The RPI Reactor Critical Facility (RCF) operated successfully over the period fall 1994 - fall 1995. During this period, the RCF was used for Critical Reactor Laboratory spring 1995 (12 students); Reactor Operations Training fall 1994 (3 students); Reactor Operations Training spring 1995 (3 students); and Reactor Operations Training fall 1995 (3 students). Thirty-two Instrumentation and Measurement students used the RCF for one class for hands-on experiments with nuclear instruments. In addition, a total of nine credits of PhD thesis work were carried out at the RCF. This document constitutes the 1995 Report of the Rensselaer Polytechnic Institute`s Reactor Critical Facility (RCF) to the USNRC, to the USDOE, and to RPI management.

  11. Nuclear-accident dosimetry: measurements at the Los Alamos SHEBA critical assembly

    SciTech Connect

    Vasilik, D.G.; Martin, R.W.; Fuller, D.

    1981-07-01

    Criticality dosimeters were exposed to different degraded neutron and gamma-ray energy spectra from the Los Alamos Solution High Energy Burst Assembly (SHEBA). The liquid critical test assembly was operated in the continuous mode to provide a mixed source of neutron and gamma-ray radiation for the evaluation of Los Alamos criticality detector systems. Different neutron and gamma-ray spectra were generated by operating the reactor (a) shielded by 12 cm of Lucite, (b) unshielded, (c) shielded by 20 cm of concrete, and (d) shielded by 15 cm of steel. This report summarizes the dosimetry measurements conducted for these different configurations. In-air measurements were conducted with shielded and unshielded area and personnel dosimeters. Phantom measurements were made using personnel dosimeters. Combined blood-sodium and hair sulfur activation measurements of absorbed dose were also made. In addition, indium foils placed on phantoms were evaluated for the purpose of screening personnel for radiation exposure.

  12. Replacing ODCs in a Critical Hand Cleaning Manual Electronics Assembly Operation

    NASA Technical Reports Server (NTRS)

    Bonner, J. K.; Walton, Sharon

    1997-01-01

    The manufacture of high reliability electronics assemblies for spacecraft and ground support equipment still often involves manual assembly processes. In addition, rework and repair of critical assemblies aslo often entails manual assembly processes.

  13. Criticality safety evaluation report for FFTF 42% fuel assemblies

    SciTech Connect

    Richard, R.F.

    1997-10-28

    An FFTF tritium/isotope production mission will require a new fuel supply. The reference design core will use a mixed oxide fuel nominally enriched to 40 wt% Pu. This enrichment is significantly higher than that of the standard Driver Fuel Assemblies used in past operations. Consequently, criticality safety for handling and storage of this fuel must be addressed. The purpose of this document is to begin the process by determining the minimum critical number for these new fuel assemblies in water, sodium and air. This analysis is preliminary and further work can be done to refine the results reported here. Analysis was initially done using 45 wt 5 PuO. Additionally, a preliminary assessment is done concerning storage of these fuel assemblies in Interim Decay Storage (IDS), Fuel Storage Facility (FSF), and Core Component Containers/Interim Storage Casks (CCC/ISC).

  14. [Prof. Michiharu Matsuoka, founder of the Department of Orthopaedic Surgery at Kyoto University and his achievements in orthopaedic surgery in the Meiji era of Japan (Part 5, Faculty members and training of doctors from Nagoya)].

    PubMed

    Hirotani, Hayato

    2010-09-01

    During the years when Dr. M. Matsuoka was professor of the Department of Orthopaedic Surgery, Kyoto Medical School, Kyoto Imperial University (June, 1907-January, 1914), seven doctors worked as his faculty members and founded the base of the current development and reputation of the Department. After resignation from their academic positions, they served in orthopaedic practice in several areas in Japan where orthopaedic surgery was not well recognized. In addition, Prof. Matsuoka trained three doctors from the Aichi Prefectural Medical College (School of Medicine, Nagoya University) in the orthopaedic practice, including x-ray technique and they contributed to the development of orthopaedic surgery in the areas of Nagoya city and Tokai. Backgrounds and achievements of these ten doctors are described. PMID:21560319

  15. Consistent Pl Analysis of Aqueous Uranium-235 Critical Assemblies

    NASA Technical Reports Server (NTRS)

    Fieno, Daniel

    1961-01-01

    The lethargy-dependent equations of the consistent Pl approximation to the Boltzmann transport equation for slowing down neutrons have been used as the basis of an IBM 704 computer program. Some of the effects included are (1) linearly anisotropic center of mass elastic scattering, (2) heavy element inelastic scattering based on the evaporation model of the nucleus, and (3) optional variation of the buckling with lethargy. The microscopic cross-section data developed for this program covered 473 lethargy points from lethargy u = 0 (10 Mev) to u = 19.8 (0.025 ev). The value of the fission neutron age in water calculated here is 26.5 square centimeters; this value is to be compared with the recent experimental value given as 27.86 square centimeters. The Fourier transform of the slowing-down kernel for water to indium resonance energy calculated here compared well with the Fourier transform of the kernel for water as measured by Hill, Roberts, and Fitch. This method of calculation has been applied to uranyl fluoride - water solution critical assemblies. Theoretical results established for both unreflected and fully reflected critical assemblies have been compared with available experimental data. The theoretical buckling curve derived as a function of the hydrogen to uranium-235 atom concentration for an energy-independent extrapolation distance was successful in predicting the critical heights of various unreflected cylindrical assemblies. The critical dimensions of fully water-reflected cylindrical assemblies were reasonably well predicted using the theoretical buckling curve and reflector savings for equivalent spherical assemblies.

  16. Reactivity effects of void formations in a solution critical assembly

    SciTech Connect

    Walters, S.G.

    1994-01-01

    SHEBA II (Solution High Energy Burst Assembly) was constructed in order to better understand the neutronics of solutions of fissile materials. In order to estimate the effect on criticality from the formation of bubbles, models were devised in MCNP (Monte Carlo Neutron Photon transport code) and THREEDANT (THREE dimensional, Diffusion-Accelerated, Neutral-Particle Transport). It was found that the formation of voids in all but the outside bottom edge of the assembly cylinder tend to act as a negative insertion of reactivity. Also, an experiment has been designed which will verify the results of the codes.

  17. Estimation of Critical Flow Velocity for Collapse of Gas Test Loop Booster Fuel Assembly

    SciTech Connect

    Guillen; Mark J. Russell

    2006-07-01

    This paper presents calculations performed to determine the critical flow velocity for plate collapse due to static instability for the Gas Test Loop booster fuel assembly. Long, slender plates arranged in a parallel configuration can experience static divergence and collapse at sufficiently high coolant flow rates. Such collapse was exhibited by the Oak Ridge High Flux Reactor in the 1940s and the Engineering Test Reactor at the Idaho National Laboratory in the 1950s. Theoretical formulas outlined by Miller, based upon wide-beam theory and Bernoulli’s equation, were used for the analysis. Calculations based upon Miller’s theory show that the actual coolant flow velocity is only 6% of the predicted critical flow velocity. Since there is a considerable margin between the theoretically predicted plate collapse velocity and the design velocity, the phenomena of plate collapse due to static instability is unlikely.

  18. Criticality Safety Evaluation of a LLNL Training Assembly for Criticality Safety (TACS)

    SciTech Connect

    Heinrichs, D P

    2006-06-26

    Hands-on experimental training in the physical behavior of multiplying systems is one of ten key areas of training required for practitioners to become qualified in the discipline of criticality safety as identified in DOE-STD-1135-99, ''Guidance for Nuclear Criticality Safety Engineer Training and Qualification''. This document is a criticality safety evaluation of the training activities (or operations) associated with HS-3200, ''Laboratory Class for Criticality Safety''. These activities utilize the Training Assembly for Criticality Safety (TACS). The original intent of HS-3200 was to provide LLNL fissile material handlers with a practical hands-on experience as a supplement to the academic training they receive biennially in HS-3100, ''Fundamentals of Criticality Safety'', as required by ANSI/ANS-8.20-1991, ''Nuclear Criticality Safety Training''. HS-3200 is to be enhanced to also address the training needs of nuclear criticality safety professionals under the auspices of the NNSA Nuclear Criticality Safety Program.

  19. Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

    DOE PAGESBeta

    Williams, M. L.; Wiarda, D.; Ilas, G.; Marshall, W. J.; Rearden, B. T.

    2014-06-15

    Recently, we processed a new covariance data library based on ENDF/B-VII.1 for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. Moreover, the cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

  20. Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

    SciTech Connect

    Williams, M.L. Wiarda, D.; Ilas, G.; Marshall, W.J.; Rearden, B.T.

    2015-01-15

    A new covariance data library based on ENDF/B-VII.1 was recently processed for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. The cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

  1. Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

    NASA Astrophysics Data System (ADS)

    Williams, M. L.; Wiarda, D.; Ilas, G.; Marshall, W. J.; Rearden, B. T.

    2015-01-01

    A new covariance data library based on ENDF/B-VII.1 was recently processed for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. The cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

  2. Self-assembled nanoparticle-stabilized photocatalytic reactors.

    PubMed

    Burdyny, Thomas; Riordon, Jason; Dinh, Cao-Thang; Sargent, Edward H; Sinton, David

    2016-01-28

    The efficiency of nanostructured photocatalysts continues to improve at an impressive pace and is closing in on those needed for commercial applications; however, present-day reactor strategies used to deploy these nanostructures fail to achieve the sufficient areas (>1 m(2)) needed for solar application. Here we report the Self-assembled Nanoparticle-stabilized Photocatalytic Reactor (SNPR), a fully-scalable reactor strategy comprised only of nanoparticles adsorbed at the fluid-fluid interfaces of oil-in-water emulsions, water-in-oil emulsions, and CO2-in-water foams. We show that SNPRs naturally disperse over open water and need no physical substrate, requiring only photocatalysts and fluid. In environmental applications the SNPR provides more than double the reaction rate of a comparable single-phase reactor. In continuous mode, the SNPR achieves 100% photocatalyst retention and processes 96% of the stream over 20 hours; in contrast, the performance of a comparable aqueous suspension declines to zero over this interval, losing all photocatalyst to the outlet stream. We further characterize the photoactivity of individual photocatalytic droplets, with reactants in both the continuous and dispersed phases. These results demonstrate SNPRs as a robust and flexible reactor strategy and a route-to-scale for nanomaterials. PMID:26700375

  3. Passive gamma analysis of the boiling-water-reactor assemblies

    NASA Astrophysics Data System (ADS)

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

    2016-09-01

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative-Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  4. Passive gamma analysis of the boiling-water-reactor assemblies

    DOE PAGESBeta

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

    2016-06-04

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden’s Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in themore » past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.« less

  5. Neutronics for critical fission reactors and subcritical fission in hybrids

    SciTech Connect

    Salvatores, Massimo

    2012-06-19

    The requirements of future innovative nuclear fuel cycles will focus on safety, sustainability and radioactive waste minimization. Critical fast neutron reactors and sub-critical, external source driven systems (accelerator driven and fusion-fission hybrids) have a potential role to meet these requirements in view of their physics characteristics. This paper provides a short introduction to these features.

  6. Neutronics for critical fission reactors and subcritical fission in hybrids

    NASA Astrophysics Data System (ADS)

    Salvatores, Massimo

    2012-06-01

    The requirements of future innovative nuclear fuel cycles will focus on safety, sustainability and radioactive waste minimization. Critical fast neutron reactors and sub-critical, external source driven systems (accelerator driven and fusion-fission hybrids) have a potential role to meet these requirements in view of their physics characteristics. This paper provides a short introduction to these features.

  7. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods

    SciTech Connect

    Margaret A. Marshall

    2013-03-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.”(Reference 1) The experiment studied in this evaluation was the first of the series and had the fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Once the critical configurations had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U, , and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configurations are described in Sections 1.3, 1.4 and 1.7, respectively. Information for this

  8. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods

    SciTech Connect

    Margaret A. Marshall

    2012-09-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.”(Reference 1) The experiment studied in this evaluation was the first of the series and had the fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Once the critical configurations had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U, , and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configurations are described in Sections 1.3, 1.4 and 1.7, respectively. Information for this

  9. Critical partnerships: Los Alamos, universities, and industry

    SciTech Connect

    Berger, C.L.

    1997-04-01

    Los Alamos National Laboratory, situated 35 miles northwest of Santa Fe, NM, is one of the Department of Energy`s three Defense Programs laboratories. It encompasses 43 square miles, employees approximately 10,000 people, and has a budget of approximately $1.1B in FY97. Los Alamos has a strong post-cold war mission, that of reducing the nuclear danger. But even with that key role in maintaining the nation`s security, Los Alamos views partnerships with universities and industry as critical to its future well being. Why is that? As the federal budget for R&D comes under continued scrutiny and certain reduction, we believe that the triad of science and technology contributors to the national system of R&D must rely on and leverage each others capabilities. For us this means that we will rely on these partners to help us in 5 key ways: We expect that partnerships will help us maintain and enhance our core competencies. In doing so, we will be able to attract the best scientists and engineers. To keep on the cutting edge of research and development, we have found that partnerships maintain the excellence of staff through new and exciting challenges. Additionally, we find that from our university and corporate partners we often learn and incorporate {open_quotes}best practices{close_quotes} in organizational management and operations. Finally, we believe that a strong national system of R&D will ensure and enhance our ability to generate revenues.

  10. Howard University Assembles Fund-Raising Juggernaut

    ERIC Educational Resources Information Center

    Masterson, Kathryn

    2008-01-01

    As a dental student 35 years ago, Leo E. Rouse and his Howard University classmates learned to fill cavities and cap teeth by crowding around one faculty member and angling for a clear view of the day's demonstration. Today students at Howard's College of Dentistry, where Dr. Rouse is now the dean, get an unobstructed view of dental procedures…

  11. Flow excursion experiments with a production reactor assembly mockup

    SciTech Connect

    Rush, G.C.; Blake, J.E. ); Nash, C.A. )

    1990-01-01

    A series of power ramp and loss-of-coolant accidents were simulated with an electrically heated mockup of a Savannah River Site production reactor assembly. The one-to-one scale mockup had full multichannel annular geometry in its heated section in addition to prototypical inlet and outlet endfitting hardware. Power levels causing void generation and flow instability in the water coolant flowing through the mockup were found under different transient and quasi-steady state test conditions. A reasonably sharp boundary between initial operating powers leading to or not leading to flow instability were found: that being 0.2 MW or less on power levels of 4 to 6.3 MW. Void generation occurred before, but close to, the point of flow instability. The data were taken in support of the Savannah River reactor limits program and will be used in continuing code benchmarking efforts. 6 refs., 12 figs., 2 tabs.

  12. Individual source positioning mechanism for a nuclear reactor fuel assembly

    SciTech Connect

    Wilson, J.F.; Gjertsen, R.K.; Cerni, S.

    1987-07-07

    A nuclear reactor is described including a fuel assembly, at lest one elongated neutron source rod and an upper core plate. The fuel assembly has top and bottom nozzles with a guide thimbles extending between and interconnecting the nozzles. The upper core plate is positioned adjacent to and above the top nozzle of the fuel assembly and having flow openings to allow passage of coolant from the fuel assembly. At least some of the openings is aligned over respective ones of the guide thimbles with seating means defined about the openings on a lower side of the core plate, a separate mechanism for positioning each individual neutron source rod in a respective guide thimble aligned with one of the openings defined through the upper core plate, comprising: (a) locating means registering against the core plate seating means; and (b) resilient holddown means extending partially into the guide thimble and coupling the source rod with the locating means in a manner which restrains the source rod in a lateral direction and positions the rod in a stationary axial relationship within the guide thimble.

  13. Test Suite for Nuclear Data I: Deterministic Calculations for Critical Assemblies and Replacement Coefficients

    SciTech Connect

    Pruet, J; Brown, D A; Descalle, M

    2006-05-22

    The authors describe tools developed by the Computational Nuclear Physics group for testing the quality of internally developed nuclear data and the fidelity of translations from ENDF formatted data to ENDL formatted data used by Livermore. These tests include S{sub n} calculations for the effective k value characterizing critical assemblies and for replacement coefficients of different materials embedded in the Godiva and Jezebel critical assemblies. For those assemblies and replacement materials for which reliable experimental information is available, these calculations provide an integral check on the quality of data. Because members of the ENDF and reactor communities use calculations for these same assemblies in their validation process, a comparison between their results with ENDF formatted data and their results with data translated into the ENDL format provides a strong check on the accuracy of translations. As a first application of the test suite they present a study comparing ENDL 99 and ENDF/B-V. They also consider the quality of the ENDF/B-V translation previously done by the Computational Nuclear Physics group. No significant errors are found.

  14. Benchmarking of Graphite Reflected Critical Assemblies of UO2

    SciTech Connect

    Margaret A. Marshall; John D. Bess

    2011-11-01

    A series of experiments were carried out in 1963 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for use in space reactor research programs. A core containing 93.2% enriched UO2 fuel rods was used in these experiments. The first part of the experimental series consisted of 253 tightly-packed fuel rods (1.27 cm triangular pitch) with graphite reflectors [1], the second part used 253 graphite-reflected fuel rods organized in a 1.506 cm triangular pitch [2], and the final part of the experimental series consisted of 253 beryllium-reflected fuel rods with a 1.506 cm triangular pitch. [3] Fission rate distribution and cadmium ratio measurements were taken for all three parts of the experimental series. Reactivity coefficient measurements were taken for various materials placed in the beryllium reflected core. The first part of this experimental series has been evaluated for inclusion in the International Reactor Physics Experiment Evaluation Project (IRPhEP) [4] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbooks, [5] and is discussed below. These experiments are of interest as benchmarks because they support the validation of compact reactor designs with similar characteristics to the design parameters for a space nuclear fission surface power systems. [6

  15. Self-assembled nanoparticle-stabilized photocatalytic reactors

    NASA Astrophysics Data System (ADS)

    Burdyny, Thomas; Riordon, Jason; Dinh, Cao-Thang; Sargent, Edward H.; Sinton, David

    2016-01-01

    The efficiency of nanostructured photocatalysts continues to improve at an impressive pace and is closing in on those needed for commercial applications; however, present-day reactor strategies used to deploy these nanostructures fail to achieve the sufficient areas (>1 m2) needed for solar application. Here we report the Self-assembled Nanoparticle-stabilized Photocatalytic Reactor (SNPR), a fully-scalable reactor strategy comprised only of nanoparticles adsorbed at the fluid-fluid interfaces of oil-in-water emulsions, water-in-oil emulsions, and CO2-in-water foams. We show that SNPRs naturally disperse over open water and need no physical substrate, requiring only photocatalysts and fluid. In environmental applications the SNPR provides more than double the reaction rate of a comparable single-phase reactor. In continuous mode, the SNPR achieves 100% photocatalyst retention and processes 96% of the stream over 20 hours; in contrast, the performance of a comparable aqueous suspension declines to zero over this interval, losing all photocatalyst to the outlet stream. We further characterize the photoactivity of individual photocatalytic droplets, with reactants in both the continuous and dispersed phases. These results demonstrate SNPRs as a robust and flexible reactor strategy and a route-to-scale for nanomaterials.The efficiency of nanostructured photocatalysts continues to improve at an impressive pace and is closing in on those needed for commercial applications; however, present-day reactor strategies used to deploy these nanostructures fail to achieve the sufficient areas (>1 m2) needed for solar application. Here we report the Self-assembled Nanoparticle-stabilized Photocatalytic Reactor (SNPR), a fully-scalable reactor strategy comprised only of nanoparticles adsorbed at the fluid-fluid interfaces of oil-in-water emulsions, water-in-oil emulsions, and CO2-in-water foams. We show that SNPRs naturally disperse over open water and need no physical substrate

  16. The Promotion of Peace Education through Guides in Peace Museums. A Case Study of the Kyoto Museum for World Peace, Ritsumeikan University

    ERIC Educational Resources Information Center

    Tanigawa, Yoshiko

    2015-01-01

    This paper focuses on how peace education at a peace museum is promoted by a volunteer guide service for visitors. Peace museums are places where many materials related to war and peace history are on display. To support the learning experience of museum visitors, many peace museums in Japan provide a volunteer guide service. The Kyoto Museum for…

  17. Experimental physics characteristics of a heavy-metal-reflected fast-spectrum critical assembly

    NASA Technical Reports Server (NTRS)

    Heneveld, W. H.; Paschall, R. K.; Springer, T. H.; Swanson, V. A.; Thiele, A. W.; Tuttle, R. J.

    1972-01-01

    A zero-power critical assembly was designed, constructed, and operated for the purpose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power electric conversion system. The range of the previous experimental investigations has been expanded to include the reactivity effects of:(1) surrounding the reactor with 15.24 cm (6 in.) of polyethylene, (2) reducing the heights of a portion of the upper and lower axial reflectors by factors of 2 and 4, (3) adding 45 kg of W to the core uniformly in two steps, (4) adding 9.54 kg of Ta to the core uniformly, and (5) inserting 2.3 kg of polyethylene into the core proper and determining the effect of a Ta addition on the polyethylene worth.

  18. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 3-Surry Unit 1 Cycle 2

    SciTech Connect

    Bowman, S.M.

    1995-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using selected critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations in this report is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and to provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of two reactor critical configurations for Surry Unit 1 Cycle 2. This unit and cycle were chosen for a previous analysis using a different methodology because detailed isotopics from multidimensional reactor calculations were available from the Virginia Power Company. These data permitted a direct comparison of criticality calculations using the utility-calculated isotopics with those using the isotopics generated by the SCALE-4

  19. Intrinsic universality and the computational power of self-assembly.

    PubMed

    Woods, Damien

    2015-07-28

    Molecular self-assembly, the formation of large structures by small pieces of matter sticking together according to simple local interactions, is a ubiquitous phenomenon. A challenging engineering goal is to design a few molecules so that large numbers of them can self-assemble into desired complicated target objects. Indeed, we would like to understand the ultimate capabilities and limitations of this bottom-up fabrication process. We look to theoretical models of algorithmic self-assembly, where small square tiles stick together according to simple local rules in order to carry out a crystal growth process. In this survey, we focus on the use of simulation between such models to classify and separate their computational and expressive powers. Roughly speaking, one model simulates another if they grow the same structures, via the same dynamical growth processes. Our journey begins with the result that there is a single intrinsically universal tile set that, with appropriate initialization and spatial scaling, simulates any instance of Winfree's abstract Tile Assembly Model. This universal tile set exhibits something stronger than Turing universality: it captures the geometry and dynamics of any simulated system in a very direct way. From there we find that there is no such tile set in the more restrictive non-cooperative model, proving it weaker than the full Tile Assembly Model. In the two-handed model, where large structures can bind together in one step, we encounter an infinite set of infinite hierarchies of strictly increasing simulation power. Towards the end of our trip, we find one tile to rule them all: a single rotatable flipable polygonal tile that simulates any tile assembly system. We find another tile that aperiodically tiles the plane (but with small gaps). These and other recent results show that simulation is giving rise to a kind of computational complexity theory for self-assembly. It seems this could be the beginning of a much longer journey

  20. Experimental physics characteristics of a heavy-metal-reflected fast-spectrum critical assembly

    NASA Technical Reports Server (NTRS)

    Heneveld, W. H.; Paschall, R. K.; Springer, T. H.; Swanson, V. A.; Thiele, A. W.; Tuttle, R. J.

    1971-01-01

    A zero-power critical assembly was designed, constructed, and operated for the purpose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7 cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power electric conversion system. The experimental program consisted basically of measuring the differential neutron spectra and the changes in critical mass that accompanied the stepwise addition of (Li-7)3N, Hf, Ta, and W to a basic core fueled with U metal in a pin-type Ta honeycomb structure. In addition, experimental results were obtained on power distributions, control characteristics, neutron lifetime, and reactivity worths of numerous absorber, structural, and scattering materials.

  1. The Secular University and Its Critics

    ERIC Educational Resources Information Center

    Jobani, Yuval

    2016-01-01

    Universities in the USA have become bastions of secularity in a distinctly religious society. As such, they are subjected to a variety of robust and rigorous religious critiques. In this paper I do not seek to engage in the debate between the supporters of the secular university and its opponents. Furthermore, I do not claim to summarize the…

  2. A Photo-neutron Source for a Sub-Critical Nuclear Reactor Program

    SciTech Connect

    Reda, M.A.; Harmon, J.F.; Sadineni, S.B.

    2003-08-26

    Experiments to benchmark photo-neutron production calculations for an Accelerator Driven Sub-Critical System (ADS) are described. A photo-nuclear based neutron source with output > 1013 n/sec has been proposed as a driver for a program using the sub-critical assembly at Idaho State University. The program is intended to study ADS control issues arising from coupling an accelerator neutron source with a sub-critical assembly. The experiments were performed using the 20 MeV electron linear accelerator at the Idaho Accelerator Center (IAC). Results of calculations, that were made using ACCEPT, PINP, MCNP, and MCNPX codes to optimize photo-nuclear based neutron conversion targets, are compared to experimental data for a single energy measurement.

  3. Unique educational opportunities at the Missouri University research reactor

    SciTech Connect

    Ketring, A.R.; Ross, F.K.; Spate, V.

    1997-12-01

    Since the Missouri University Research Reactor (MURR) went critical in 1966, it has been a center where students from many departments conduct their graduate research. In the past three decades, hundreds of graduate students from the MU departments of chemistry, physics, anthropology, nuclear engineering, etc., have received masters and doctoral degrees based on research using neutrons produced at MURR. More recently, the educational opportunities at MURR have been expanded to include undergraduate students and local high school students. Since 1989 MURR has participated in the National Science Foundation-funded Research Experience for Undergraduates (REU) program. As part of this program, undergraduate students from universities and colleges throughout the United States come to MURR and get hands-on research experience during the summer. Another program, started in 1994 by the Nuclear Analysis Program at MURR, allows students from a local high school to conduct a neutron activation analysis (NAA) experiment. We also conduct tours of the center, where we describe the research and educational programs at MURR to groups of elementary school children, high school science teachers, state legislators, professional organizations, and many other groups.

  4. Validation of HELIOS Neutron Cross-Section Library for RBMK Reactors Against the Data From the Critical Facility Experiments

    SciTech Connect

    Jasiulevicius, Audrius; Sehgal, Bal Raj

    2002-07-01

    The RBMK reactors are channel type, water-cooled and graphite moderated reactors. The first RBMK type electricity production reactor was put on-line in 1973. Currently there are 13 operating reactors of this type. Two of the RBMK-1500 reactors are at the Ignalina NPP in Lithuania. Experimental Critical Facility for RBMK reactors, located at Kurchatov Institute, Moscow was designed to carry out critical reactivity experiments on assemblies, which imitate parts of the RBMK reactor core. The facility is composed of Control and Protection Rods (CPR's), fuel assemblies with different enrichment in U-235 and other elements, typical for RBMK reactor core loadings, e.g. additional absorber assemblies, CPR imitators, etc. A simulation of a set of the experiments, performed at the Experimental Critical Facility, was carried out at the Royal Institute of Technology (RIT), Nuclear Power Safety Division, using CORETRAN 3-D neutron dynamics code. The neutron cross sections for assemblies were calculated using HELIOS code. The aim of this work was to evaluate capabilities of the HELIOS code to provide correct cross section data for the RBMK reactor. The calculation results were compared to the similar CORETRAN calculations, when employing WIMS-D4 code generated cross section data. For some of the experiments, where calculation results with CASMO-4 code generated cross sections are available, the comparison is also performed against CASMO-4 results. Eleven different experiments were simulated. Experiments differ in size of the facility core (number of assemblies loaded): from simple core loadings, composed only of a few fuel assemblies, to complicated configurations, which represent a part of the RBMK reactor core. Diverse types of measurements were carried out during these experiments: reactivity, neutron flux distributions (both axial and radial), rod reactivity worth and the voiding effects. Results of the reactivity measurements and relative neutron flux distributions were

  5. The Quest for World Class Universities in China: Critical Reflections

    ERIC Educational Resources Information Center

    Ngok, Kinglun; Guo, Weiqing

    2008-01-01

    Building world-class universities has become a national policy priority in China since then-President Jiang Zemin announced in May 1998 that China must have several world-class universities of international advanced level. This article aims to offer critical reflections on the policy in relation to building world-class universities in China. It…

  6. Adapting the Critical Thinking Assessment Test for Palestinian Universities

    ERIC Educational Resources Information Center

    Basha, Sami; Drane, Denise; Light, Gregory

    2016-01-01

    Critical thinking is a key learning outcome for Palestinian students. However, there are no validated critical thinking tests in Arabic. Suitability of the US developed Critical Thinking Assessment Test (CAT) for use in Palestine was assessed. The test was piloted with university students in English (n = 30) and 4 questions were piloted in Arabic…

  7. In Comparative Analysis for Fuel Burnup of Fuel Assembly Designs for the 300 kW Small Medical Reactor

    NASA Astrophysics Data System (ADS)

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-01

    A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996 [1]. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel burnup calculations were carried out.

  8. In Comparative Analysis for Fuel Burnup of Fuel Assembly Designs for the 300 kW Small Medical Reactor

    SciTech Connect

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-31

    A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel burnup calculations were carried out.

  9. [Dr. Michiharu Matsuoka, founder of the Department of Orthopaedic Surgery, Kyoto University, and his achievements (Part 6: Studying abroad of Dr. Matsuoka and opening to public, reputation and achievement of the department)].

    PubMed

    Hirotani, Hayato

    2011-03-01

    Dr. Michiharu Matsuoka studied orthopaedic surgery in Germany, Austria and other countries during the period from August, 1902 to May, 1906. He visited many university pathological institutes and surgical and orthopaedic clinics to study pathology and to learn the practice of orthopaedic surgery. After that, he started his practice at the newly established Department of Orthopaedic Surgery in the Medical School of Kyoto Imperial University in June, 1906. The department was opened in 1907 and in 1911 it was opened to all citizens and practical doctors in Kyoto City and exhibited many orthopaedic specimens and instruments. In particular, the x-ray apparatus of the Department was so well equipped that a German radiologist who visited the Department admired it in his article that was published in the journal of radiology in 1911. The Department was not surpassed by others for the number of patients with the dislocation of the hip and tuberculous spondylitis as well as the advanced quality and variety of roentgenological and pathological researches on these diseases. PMID:21797054

  10. Possible criticality of marine reactors dumped in the Kara Sea

    SciTech Connect

    Warden, J.M.; Mount, M.; Lynn, N.M.

    1997-05-01

    The largest inventory of radioactive materials dumped in the Kara Sea by the former Soviet Union comes from the spent nuclear fuel (SNF) of seven marine reactors. Using corrosion models derived for the International Arctic Seas Assessment Project (IASAP), the possibility of some of the SNF achieving criticality through structural and material changes has been investigated. Although remote, the possibility cannot at this stage be ruled out.

  11. Performance of boiling water reactor fuel lead test assemblies to 35 MWd/kg U

    SciTech Connect

    Rowland, T.C.; Ikemoto, R.N.; Gehl, S.

    1986-01-01

    This joint Electric Power Research Institute/General Electric (EPRI/GE) fuel performance program involved thorough preirradiation characterization of fuel used in lead test assemblies (LTAs), detailed surveillance of their operation, and interim site examinations of the assemblies during reactor outages. The program originally included four GE-5 LTAs operating in the Peach Bottom-2 (PB-2) reactor. The program was later modified to include the pressurized fuel rod test assembly in the Peach Bottom-3 (PB-3) reactor. The program modification also included extending the operation of the PB-2 and PB-3 LTA fuel beyond normal discharge exposures. Results are summarized in the paper.

  12. Weightless experiments to probe universality of fluid critical behavior.

    PubMed

    Lecoutre, C; Guillaument, R; Marre, S; Garrabos, Y; Beysens, D; Hahn, I

    2015-06-01

    Near the critical point of fluids, critical opalescence results in light attenuation, or turbidity increase, that can be used to probe the universality of critical behavior. Turbidity measurements in SF6 under weightlessness conditions on board the International Space Station are performed to appraise such behavior in terms of both temperature and density distances from the critical point. Data are obtained in a temperature range, far (1 K) from and extremely close (a few μK) to the phase transition, unattainable from previous experiments on Earth. Data are analyzed with renormalization-group matching classical-to-critical crossover models of the universal equation of state. It results that the data in the unexplored region, which is a minute deviant from the critical density value, still show adverse effects for testing the true asymptotic nature of the critical point phenomena. PMID:26172640

  13. Weightless experiments to probe universality of fluid critical behavior

    NASA Astrophysics Data System (ADS)

    Lecoutre, C.; Guillaument, R.; Marre, S.; Garrabos, Y.; Beysens, D.; Hahn, I.

    2015-06-01

    Near the critical point of fluids, critical opalescence results in light attenuation, or turbidity increase, that can be used to probe the universality of critical behavior. Turbidity measurements in SF6 under weightlessness conditions on board the International Space Station are performed to appraise such behavior in terms of both temperature and density distances from the critical point. Data are obtained in a temperature range, far (1 K) from and extremely close (a few μ K ) to the phase transition, unattainable from previous experiments on Earth. Data are analyzed with renormalization-group matching classical-to-critical crossover models of the universal equation of state. It results that the data in the unexplored region, which is a minute deviant from the critical density value, still show adverse effects for testing the true asymptotic nature of the critical point phenomena.

  14. Critical Casimir interactions and colloidal self-assembly in near-critical solvents.

    PubMed

    Tasios, Nikos; Edison, John R; van Roij, René; Evans, Robert; Dijkstra, Marjolein

    2016-08-28

    A binary solvent mixture close to critical demixing experiences fluctuations whose correlation length, ξ, diverges as the critical point is approached. The solvent-mediated (SM) interaction that arises between a pair of colloids immersed in such a near-critical solvent can be long-ranged and this so-called critical Casimir interaction is well-studied. How a (dense) suspension of colloids will self-assemble under these conditions is poorly understood. Using a two-dimensional lattice model for the solvent and hard disks to represent the colloids, we perform extensive Monte Carlo simulations to investigate the phase behaviour of this model colloidal suspension as a function of colloid size and wettability under conditions where the solvent reservoir is supercritical. Unlike most other approaches, where the solvent is modelled as an implicit background, our model employs an explicit solvent and treats the suspension as a ternary mixture. This enables us to capture important features, including the pronounced fractionation of the solvent in the coexisting colloidal phases, of this complex system. We also present results for the partial structure factors; these shed light on the critical behaviour in the ternary mixture. The degree to which an effective two-body pair potential description can describe the phase behaviour and structure of the colloidal suspension is discussed briefly. PMID:27586941

  15. Analysis of the pool critical assembly pressure vessel benchmark using pentran

    SciTech Connect

    Edgar, C. A.; Sjoden, G. E.

    2012-07-01

    The internationally circulated Pool Critical Assembly (PCA) Pressure Vessel Benchmark was analyzed using the PENTRAN Parallel Sn code system for the geometry, material, and source specifications as described in the PCA Benchmark documentation. This research focused on utilizing the BUGLE-96 cross section library and accompanying reaction rates, while examining both adaptive differencing on a coarse mesh basis as well as Directional Theta Weighted Sn differencing in order to compare the calculated PENTRAN results to measured data. The results show good comparison with the measured data as well as to the calculated results provided from TORT for the BUGLE-96 cross sections and reaction rates, which suggests PENTRAN is a viable and reliable code system for calculation of light water reactor neutron shielding and dosimetry calculations. (authors)

  16. Critical Path Modeling: University Planning in an Urban Context.

    ERIC Educational Resources Information Center

    Vasse, William W.; And Others

    1983-01-01

    The experiences of the University of Michigan in developing its Flint campus since the mid-1970s and the use of the Critical Path Method are described. The usefulness of the method during several phases of development is highlighted. (MSE)

  17. Comparison of KENO-VI and MCNP5 Criticality Analyses for a Lunar Regolith Clustered-Reactor System

    SciTech Connect

    Bess, John Darrell

    2008-01-21

    The Lunar Regolith Clustered-Reactor System design has been presented as an alternative method for providing surface power to a lunar facility using a fast-fission, heatpipe-cooled nuclear reactor. The reactor system is divided into subcritical units that can be safely launched into orbit without risk of inadvertent criticality in the event of a launch accident. The reactor subunits are emplaced into the lunar surface to form a clustered-reactor system, utilizing the regolith as both radiation shielding and neutron-reflector material. Coordinated placement of multiple subunits can provision a critical reactor system proportional to localized lunar surface power demand. Reactor units assembled using proven and tested materials in radiation environments such as UO{sub 2} fuel, stainless-steel cladding and support, and compatible liquid-metal heatpipes promote safety and reliability, with ease of manufacture and testing. Reactor power levels of approximately 100 kWth per subunit significantly reduces the negative effects of elevated temperature and radiation environments associated with single nuclear power reactors operated at higher power levels. The analysis of subunit criticality in various accident scenarios differs by up to 4% ({approx}$6 in reactivity) between results generated using conventional criticality analysis codes, MCNP5 and KENO-VI. A demonstrated trend exists between results of the two criticality codes as accident conditions approach a multiplication factor of one. Code comparison of a tri-cluster system on the lunar surface provides comparable results with calculated system reactivity within 0.5%. Iron concentration is confirmed as the dominant element in the lunar regolith influencing system reactivity.

  18. Comparison of KENO-VI and MCNP5 Criticality Analyses for a Lunar Regolith Clustered-Reactor System

    NASA Astrophysics Data System (ADS)

    Bess, John Darrell

    2008-01-01

    The Lunar Regolith Clustered-Reactor System design has been presented as an alternative method for providing surface power to a lunar facility using a fast-fission, heatpipe-cooled nuclear reactor. The reactor system is divided into subcritical units that can be safely launched into orbit without risk of inadvertent criticality in the event of a launch accident. The reactor subunits are emplaced into the lunar surface to form a clustered-reactor system, utilizing the regolith as both radiation shielding and neutron-reflector material. Coordinated placement of multiple subunits can provision a critical reactor system proportional to localized lunar surface power demand. Reactor units assembled using proven and tested materials in radiation environments such as UO2 fuel, stainless-steel cladding and support, and compatible liquid-metal heatpipes promote safety and reliability, with ease of manufacture and testing. Reactor power levels of approximately 100 kWth per subunit significantly reduces the negative effects of elevated temperature and radiation environments associated with single nuclear power reactors operated at higher power levels. The analysis of subunit criticality in various accident scenarios differs by up to 4% (~$6 in reactivity) between results generated using conventional criticality analysis codes, MCNP5 and KENO-VI. A demonstrated trend exists between results of the two criticality codes as accident conditions approach a multiplication factor of one. Code comparison of a tri-cluster system on the lunar surface provides comparable results with calculated system reactivity within 0.5%. Iron concentration is confirmed as the dominant element in the lunar regolith influencing system reactivity.

  19. Electromagnetic Near Field Measurements of Two Critical Assemblies

    SciTech Connect

    Goettee, Jeffrey David

    2015-11-03

    The reactors employed, Godiva IV and WSMR Fast Burst Reactor, are described first. Then the point reactor kinetics model, electromagnetic potential, and the measurement of kinetics quantities are successively discussed. In summary, reactor power produces measurable electric energy. The electric signal mimics power curve for prompt burst operations - features in logarithmic derivatives match. The electric signature should be dependent on the power and not the derivative; therefore, steady-state modes should be measurable.

  20. University Reactor Instrumentation Program. Final report, 1990--1992

    SciTech Connect

    Vernetson, W.G.

    1992-11-01

    Recognizing that the University Reactor Instrumentation Program was developed in response to widespread needs in the academic community for modernization and improvement of research and training reactors at institutions such as the University of Florida, the items proposed to be supported by this grant over its two year period have been selected as those most likely to reduce foreed outages, to meet regulatory concerns that had been expressed in recent years by Nuclear Regulatory Commission inspectors or to correct other facility problems and limitations. Department of Energy Grant Number DE-FG07-90ER129969 was provided to the University of Florida Training Reactor(UFTR) facility through the US Department of Energy`s University Reactor Instrumentation Program. The original proposal submitted in February, 1990 requested support for UFTR facility instrumentation and equipment upgrades for seven items in the amount of $107,530 with $13,800 of this amount to be the subject of cost sharing by the University of Florida and $93,730 requested as support from the Department of Energy. A breakdown of the items requested and total cost for the proposed UFTR facility instrumentation and equipment improvements is presented.

  1. Continuity and Change: Kyoto Chefs Engage with Science.

    PubMed

    de St Maurice, Greg

    2015-01-01

    Kyoto's chefs have reacted proactively to changes brought about by the most recent phase of globalization, hoping to ensure the continued existence and resonance of Kyoto cuisine by using science to adapt it to contemporary circumstances. These chefs are breaking new ground in their pursuit of a scientific understanding of how Kyoto cuisine works. They meet once a month in a kitchen laboratory at Kyoto University to present and analyze culinary experiments in keeping with a predetermined theme. They use their acquired knowledge to more precisely hone their culinary skills and to explain Kyoto cuisine to a global audience. Chefs visit local elementary schools, appear on national television, and welcome chefs from abroad into their kitchens so that people across the world will better understand what authentic Kyoto cuisine consists of. Although these chefs' efforts are groundbreaking, there is also remarkable continuity to their approach. Not only has Kyoto cuisine always been in a steady state of transformation, but the chefs in the Laboratory are engaging with science and a global audience specifically so that they can ascertain Kyoto cuisine's continued existence and importance. Though their means of understanding and articulating what Kyoto cuisine is differs from that of their predecessors, concepts like shun (seasonality) and hin (refinement) still guide chefs today. Ultimately, then, based on interviews and participant observation conducted in and outside of the Japanese Cuisine Laboratory in 2012 and 2013, I argue that by engaging with contemporary food science, Kyoto's chefs achieve a strategic balance of protecting their culinary heritage while adapting it to contemporary circumstances. PMID:26598840

  2. PROCEEDINGS OF RIKEN BNL RESEARCH CENTER WORKSHOP ON RHIC SPIN PHYSICS III AND IV, POLARIZED PARTONS AT HIGH Q2 REGION, AUGUST 3, 2000 AT BNL, OCTOBER 14, 2000 AT KYOTO UNIVERSITY.

    SciTech Connect

    BUNCE, G.; VIGDOR, S.

    2001-03-15

    International workshop on II Polarized Partons at High Q2 region 11 was held at the Yukawa Institute for Theoretical Physics, Kyoto University, Kyoto, Japan on October 13-14, 2000, as a satellite of the international conference ''SPIN 2000'' (Osaka, Japan, October 16-21,2000). This workshop was supported by RIKEN (The Institute of Physical and Chemical Research) and by Yukawa Institute. The scientific program was focused on the upcoming polarized collider RHIC. The workshop was also an annual meeting of RHIC Spin Collaboration (RSC). The number of participants was 55, including 28 foreign visitors and 8 foreign-resident Japanese participants, reflecting the international nature of the RHIC spin program. At the workshop there were 25 oral presentations in four sessions, (1) RHIC Spin Commissioning, (2) Polarized Partons, Present and Future, (3) New Ideas on Polarization Phenomena, (4) Strategy for the Coming Spin Running. In (1) the successful polarized proton commissioning and the readiness of the accelerator for the physics program impressed us. In (2) and (3) active discussions were made on the new structure function to be firstly measured at RHIC, and several new theoretical ideas were presented. In session (4) we have established a plan for the beam time requirement toward the first collision of polarized protons. These proceedings include the transparencies presented at the workshop. The discussion on ''Strategy for the Coming Spin Running'' was summarized by the chairman of the session, S. Vigdor and G. Bunce.

  3. An Investigation into the Critical Thinking Skills of University Students

    ERIC Educational Resources Information Center

    Asude, Bilgin; Jale, Eldelekioglu

    2007-01-01

    The aim of the present study was to investigate into the critical thinking skills of late adolescent Turkish university students. The subjects of the study were the 39 students from the Department of Counseling Psychology and Guidance, Faculty of Education, Uludag University. Two separate discussion groups, each including five students, were…

  4. The Call for an African University: A Critical Reflection

    ERIC Educational Resources Information Center

    van Wyk, Berte; Higgs, Philip

    2007-01-01

    In this paper, we draw on philosophy (particularly African philosophy) to analyse the call for an African university. The call for an African university may be viewed as a call that insists that all critical and transformative educators in Africa embrace an indigenous African worldview and root their nation's educational paradigms in an indigenous…

  5. Canadian University, Inc., and the Role of Canadian Criticism

    ERIC Educational Resources Information Center

    Milz, Sabine

    2005-01-01

    In this article, the author seeks to address the present function of Canadian criticism by undertaking a meditation on the contemporary Canadian university and stating his own position as a critic of Canadian literature in this institutional framework. The author asks: What are the connections between neoliberalism and cultural nationalism in…

  6. An integrated approach for the verification of fresh mixed oxide fuel (MOX) assemblies at light water reactor MOX recycle reactors

    SciTech Connect

    Menlove, Howard O; Lee, Sang - Yoon

    2009-01-01

    This paper presents an integrated approach for the verification of mixed oxide (MOX) fuel assemblies prior to their being loaded into the reactor. There is a coupling of the verification approach that starts at the fuel fabrication plant and stops with the transfer of the assemblies into the thermal reactor. The key measurement points are at the output of the fuel fabrication plant, the receipt at the reactor site, and the storage in the water pool as fresh fuel. The IAEA currently has the capability to measure the MOX fuel assemblies at the output of the fuel fabrication plants using a passive neutron coincidence counting systems of the passive neutron collar (PNCL) type. Also. at the MOX reactor pool, the underwater coincidence counter (UWCC) has been developed to measure the MOX assemblies in the water. The UWCC measurement requires that the fuel assembly be lifted about two meters up in the storage rack to avoid interference from the fuel that is stored in the rack. This paper presents a new method to verify the MOX fuel assemblies that are in the storage rack without the necessity of moving the fuel. The detector system is called the Underwater MOX Verification System (UMVS). The integration and relationship of the three measurements systems is described.

  7. Spectral measurements in critical assemblies: MCNP specifications and calculated results

    SciTech Connect

    Stephanie C. Frankle; Judith F. Briesmeister

    1999-12-01

    Recently, a suite of 86 criticality benchmarks for the Monte Carlo N-Particle (MCNP) transport code was developed, and the results of testing the ENDF/B-V and ENDF/B-VI data (through Release 2) were published. In addition to the standard k{sub eff} measurements, other experimental measurements were performed on a number of these benchmark assemblies. In particular, the Cross Section Evaluation Working Group (CSEWG) specifications contain experimental data for neutron leakage and central-flux measurements, central-fission ratio measurements, and activation ratio measurements. Additionally, there exists another set of fission reaction-rate measurements performed at the National Institute of Standards and Technology (NIST) utilizing a {sup 252}Cf source. This report will describe the leakage and central-flux measurements and show a comparison of experimental data to MCNP simulations performed using the ENDF/B-V and B-VI (Release 2) data libraries. Central-fission and activation reaction-rate measurements will be described, and the comparison of experimental data to MCNP simulations using available data libraries for each reaction of interest will be presented. Finally, the NIST fission reaction-rate measurements will be described. A comparison of MCNP results published previously with the current MCNP simulations will be presented for the NIST measurements, and a comparison of the current MCNP simulations to the experimental measurements will be presented.

  8. 75 FR 56597 - University of Wisconsin; University of Wisconsin Nuclear Reactor Environmental Assessment and...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-16

    ... Engineering Physics. The Mechanical Engineering Building is near the southwestern border of the University of.... The UWNR is located in the Mechanical Engineering Building on the main campus of the University of... conventional construction within the Mechanical Engineering Building. Throughout most of the Reactor...

  9. University Reactor Conversion Lessons Learned Workshop for the University of Florida

    SciTech Connect

    Eric C. Woolstenhulme; Dana M. Meyer

    2007-04-01

    The Department of Energy’s (DOE) Idaho National Laboratory (INL), under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at the University of Florida. This project was successfully completed through an integrated and collaborative effort involving the INL, Argonne National Laboratory (ANL), DOE (Headquarters and Field Office), the Nuclear Regulatory Commission, the Universities, and contractors involved in analyses, fuel design and fabrication, and SNF shipping and disposition. With the work completed with these two universities, and in anticipation of other impending conversion projects, INL convened and engaged the project participants in a structured discussion to capture lessons learned. The objectives of this meeting were to capture the observations, insights, issues, concerns, and ideas of those involved in the reactor conversions so that future efforts can be conducted with greater effectiveness, efficiency, and with fewer challenges.

  10. Disposition of fuel elements from the Aberdeen and Sandia pulse reactor (SPR-II) assemblies

    SciTech Connect

    Mckerley, Bill; Bustamante, Jacqueline M; Costa, David A; Drypolcher, Anthony F; Hickey, Joseph

    2010-01-01

    We describe the disposition of fuel from the Aberdeen (APR) and the Sandia Pulse Reactors (SPR-II) which were used to provide intense neutron bursts for radiation effects testing. The enriched Uranium - 10% Molybdenum fuel from these reactors was shipped to the Los Alamos National Laboratory (LANL) for size reduction prior to shipment to the Savannah River Site (SRS) for final disposition in the H Canyon facility. The Shipper/Receiver Agreements (SRA), intra-DOE interfaces, criticality safety evaluations, safety and quality requirements and key materials management issues required for the successful completion of this project will be presented. This work is in support of the DOE Consolidation and Disposition program. Sandia National Laboratories (SNL) has operated pulse nuclear reactor research facilities for the Department of Energy since 1961. The Sandia Pulse Reactor (SPR-II) was a bare metal Godiva-type reactor. The reactor facilities have been used for research and development of nuclear and non-nuclear weapon systems, advanced nuclear reactors, reactor safety, simulation sources and energy related programs. The SPR-II was a fast burst reactor, designed and constructed by SNL that became operational in 1967. The SPR-ll core was a solid-metal fuel enriched to 93% {sup 235}U. The uranium was alloyed with 10 weight percent molybdenum to ensure the phase stabilization of the fuel. The core consisted of six fuel plates divided into two assemblies of three plates each. Figure 1 shows a cutaway diagram of the SPR-II Reactor with its decoupling shroud. NNSA charged Sandia with removing its category 1 and 2 special nuclear material by the end of 2008. The main impetus for this activity was based on NNSA Administrator Tom D'Agostino's six focus areas to reenergize NNSA's nuclear material consolidation and disposition efforts. For example, the removal of SPR-II from SNL to DAF was part of this undertaking. This project was in support of NNSA's efforts to consolidate the

  11. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 1-Summary

    SciTech Connect

    DeHart, M.D.

    1995-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original ''fresh'' composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized- water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Isotopic densities for spent fuel assemblies in the core were calculated using the SAS2H analytical sequence in SCALE-4. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code sequence was used to extract the necessary isotopic densities from SAS2H results and to provide the data in the format required for SCALE-4 criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) for the critical configuration. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for analysis of each critical configuration. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power's Surry Unit 1 reactor for the

  12. Consolidated fuel reprocessing program: Criticality experiments with fast test reactor fuel pins in an organic moderator

    SciTech Connect

    Bierman, S.R.

    1986-12-01

    The results obtained in a series of criticality experiments performed as part of a joint program on criticality data development between the United States Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan are presented in this report along with a complete description of the experiments. The experiments involved lattices of Fast Test Reactor (FTR) fuel pins in an organic moderator mixture similar to that used in the solvent extraction stage of fuel reprocessing. The experiments are designed to provide data for direct comparison with previously performed experimental measurements with water moderated lattices of FTR fuel pins. The same lattice arrangements and FTR fuel pin types are used in these organic moderated experimental assemblies as were used in the water moderated experiments. The organic moderator is a mixture of 38 wt % tributylphosphate in a normal paraffin hydrocarbon mixture of C{sub 11}H{sub 24} to C{sub 15}H{sub 32} molecules. Critical sizes of 1054.8, 599.2, 301.8, 199.5 and 165.3 fuel pins were obtained respectively for organic moderated lattices having 0.761 cm, 0.968 cm, 1.242 cm, 1.537 cm and 1.935 cm square lattice pitches as compared to 1046.9, 571.9, 293.9, 199.7 and 165.1 fuel pins for the same lattices water moderated.

  13. Computer modeling of the dynamic processes in the Maryland University Training Reactor - (MUTR)

    SciTech Connect

    White, Bernard H. IV; Ebert, David

    1988-07-01

    The simulator described in this paper models the behaviour of the Maryland University Training Reactor (MUTR). The reactor is a 250 kW, TRIGA reactor. The computer model is based on a system of five primary equations and eight auxiliary equations. The primary equations consist of the prompt jump approximation, a heat balance equation for the fuel and the moderator, and iodine and xenon buildup equations. For the comparison with the computer program, data from the reactor was acquired by using a personal computer (pc) which contained a Strawberry Tree data acquisition Card, connected to the reactor. The systems monitored by the pc were: two neutron detectors, fuel temperature, water temperature, three control rod positions and the period meter. The time differenced equations were programmed in the basic language. It has been shown by this paper, that the MUTR power rise from low power critical to high power, can be modelled by a relatively simple computer program. The program yields accurate agreement considering the simplicity of the program. The steady state error between the reactor and computer power is 4.4%. The difference in steady state temperatures, 112 deg. C and 117 deg. C, of the reactor and computer program, respectively, also yields a 4.5% error. Further fine tuning of the coefficients will yield higher accuracies.

  14. Inverted rank distributions: Macroscopic statistics, universality classes, and critical exponents

    NASA Astrophysics Data System (ADS)

    Eliazar, Iddo; Cohen, Morrel H.

    2014-01-01

    An inverted rank distribution is an infinite sequence of positive sizes ordered in a monotone increasing fashion. Interlacing together Lorenzian and oligarchic asymptotic analyses, we establish a macroscopic classification of inverted rank distributions into five “socioeconomic” universality classes: communism, socialism, criticality, feudalism, and absolute monarchy. We further establish that: (i) communism and socialism are analogous to a “disordered phase”, feudalism and absolute monarchy are analogous to an “ordered phase”, and criticality is the “phase transition” between order and disorder; (ii) the universality classes are characterized by two critical exponents, one governing the ordered phase, and the other governing the disordered phase; (iii) communism, criticality, and absolute monarchy are characterized by sharp exponent values, and are inherently deterministic; (iv) socialism is characterized by a continuous exponent range, is inherently stochastic, and is universally governed by continuous power-law statistics; (v) feudalism is characterized by a continuous exponent range, is inherently stochastic, and is universally governed by discrete exponential statistics. The results presented in this paper yield a universal macroscopic socioeconophysical perspective of inverted rank distributions.

  15. Experiments on small-size fast critical fuel assemblies at the AKSAMIT facility and their use for development of computational models

    NASA Astrophysics Data System (ADS)

    Glushkov, E. S.; Glushkov, A. E.; Gomin, E. A.; Daneliya, S. B.; Zimin, A. A.; Kalugin, M. A.; Kapitonova, A. V.; Kompaniets, G. V.; Moroz, N. P.; Nosov, V. I.; Petrushenko, R. P.; Smirnov, O. N.

    2013-12-01

    Small-size fast critical assemblies with highly enriched fuel at the AKSAMIT facility are described in detail. Computational models of the critical assemblies at room temperature are given. The calculation results for the critical parameters are compared with the experimental data. A good agreement between the calculations and the experimental data is shown. The physical models developed for the critical assemblies, as well as the experimental results, can be applied to verify various codes intended for calculation of the neutronic characteristics of small-size fast nuclear reactors. For these experiments, the results computed using the codes of the MCU family show a high quality of the neutron data and of the physical models used.

  16. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    SciTech Connect

    Pond, R.B.; Matos, J.E.

    1996-12-31

    This document has been prepared to assist research reactor operators possessing spent fuel containing enriched uranium of United States origin to prepare part of the documentation necessary to ship this fuel to the United States. Data are included on the nuclear mass inventory, photon dose rate, and thermal decay heat of spent research reactor fuel assemblies. Isotopic masses of U, Np, Pu and Am that are present in spent research reactor fuel are estimated for MTR, TRIGA and DIDO-type fuel assembly types. The isotopic masses of each fuel assembly type are given as functions of U-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly. Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg[sup 235]U, and for fission product decay times of up to 20 years. Thermal decay heat loads are estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time.

  17. Radioisotope research, production, and processing at the University of Missouri Research Reactor

    SciTech Connect

    Ehrhardt, G.J.; Ketring, A.R.; Ja, Wei; Ma, D.; Zinn, K.; Lanigan, J.

    1995-12-31

    The University of Missouri Research Reactor (MURR) is a 10 MW, light-water-cooled and moderated research reactor which first achieved criticality in 1996 and is currently the highest powered university-owned research reactor in the U.S. For many years a major supplier of reactor-produced isotopes for research and commercial purposes, in the last 15 years MURR has concentrated on development of reactor-produced beta-particle emitters for experimental use in nuclear medicine therapy of cancer and rheumatoid arthritis. MURR has played a major role in the development of bone cancer pain palliation with the agents {sup 153}Sm EDTMP and {sup 186}Re/{sup 188}Re HEDP, as well as in the use of {sup 186}Re, {sup 177}Lu, {sup 166}Ho, and {sup 105}Rh for radioimmunotherapy and receptor-agent-guided radiotherapy. MURR is also responsible for the development of therapeutic, {sup 90}Y-labeled glass microspheres for the treatment of liver tumors, a product ({sup 90}Y Therasphere{trademark}) which is currently an approved drug in Canada. MURR has also pioneered the development of {sup 188}W/{sup 188}Re and {sup 99}Mo/{sup 99m}Tc gel generators, which make the use of low specific activity {sup 188}W and {sup 99}Mo practical for such isotope generators.

  18. Kyoto Tridimensional Spectrograph II

    NASA Astrophysics Data System (ADS)

    Sugai, Hajime; Ohtani, Hiroshi; Ishigaki, Tsuyoshi; Hayashi, Tadashi; Ozaki, Shinobu; Hattori, Takashi; Ishii, M.; Sasaki, Minoru; Takeyama, Norihide

    1998-07-01

    We are building the second version of the Kyoto Tridimensional Spectrograph (Ohtani et al., this symposium). This will be mounted on the MAGNUM, a 2-m telescope under construction at Haleakala, and also on the SUBARU. The spectrograph has four observational modes: Fabry-Perot imager, integral field spectrograph (IFS) with a microlens array, long-slit spectrograph, and filter-imaging modes. The new spectrograph is significantly better than the first version in several ways. The IFS has as many as 37 X 37 microlenses, each of which subtends 0' .39 at the MAGNUM. The optics is designed to be used in wide wavelength ranges from 360 nm to 900 nm. The transmission at any wavelength between 370 and 900 nm is designed to exceed 50% for the collimator plus camera system, and to reach almost 40% even at 360 nm. In order to achieve high efficiency at short wavelengths, we use an anti- reflection coated backside-illuminated 2K X 2K CCD. We are also planning a further improvement by using multi-layer anti- reflection coatings for lenses, in collaboration with National Astronomical Observatory, Japan. In order to assure good image quality under a severe weight limit of 150 kg for this instrument, we have carried out mechanical design by calculating the flexure of the instrument for all telescope attitudes with finite element analysis, and succeeded in limiting the maximum flexure to 30 micrometer. This does not degrade image quality. The movements on the CCD of the light from the center of the focal plane have also been simulated, depending on the telescope attitudes. This is important to obtain not only a good image, but also a correct flat field and wavelength calibration in the IFS mode. The movements are expected to be confined almost within one pixel for an attitude, which is considered to be small enough.

  19. Universal Entanglement Entropy in 2D Conformal Quantum Critical Points

    SciTech Connect

    Hsu, Benjamin; Mulligan, Michael; Fradkin, Eduardo; Kim, Eun-Ah

    2008-12-05

    We study the scaling behavior of the entanglement entropy of two dimensional conformal quantum critical systems, i.e. systems with scale invariant wave functions. They include two-dimensional generalized quantum dimer models on bipartite lattices and quantum loop models, as well as the quantum Lifshitz model and related gauge theories. We show that, under quite general conditions, the entanglement entropy of a large and simply connected sub-system of an infinite system with a smooth boundary has a universal finite contribution, as well as scale-invariant terms for special geometries. The universal finite contribution to the entanglement entropy is computable in terms of the properties of the conformal structure of the wave function of these quantum critical systems. The calculation of the universal term reduces to a problem in boundary conformal field theory.

  20. Fast Reactor Spent Fuel Processing: Experience and Criticality Safety

    SciTech Connect

    Chad Pope

    2007-05-01

    This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day

  1. Universal quantum correlation close to quantum critical phenomena.

    PubMed

    Qin, Meng; Ren, Zhong-Zhou; Zhang, Xin

    2016-01-01

    We study the ground state quantum correlation of Ising model in a transverse field (ITF) by implementing the quantum renormalization group (QRG) theory. It is shown that various quantum correlation measures and the Clauser-Horne-Shimony-Holt inequality will highlight the critical point related with quantum phase transitions, and demonstrate nonanalytic phenomena and scaling behavior when the size of the systems becomes large. Our results also indicate a universal behavior of the critical exponent of ITF under QRG theory that the critical exponent of different measures is identical, even when the quantities vary from entanglement measures to quantum correlation measures. This means that the two kinds of quantum correlation criterion including the entanglement-separability paradigm and the information-theoretic paradigm have some connections between them. These remarkable behaviors may have important implications on condensed matter physics because the critical exponent directly associates with the correlation length exponent. PMID:27189504

  2. A universal mechanism of extreme events and critical phenomena

    PubMed Central

    Wu, J. H.; Jia, Q.

    2016-01-01

    The occurrence of extreme events and critical phenomena is of importance because they can have inquisitive scientific impact and profound socio-economic consequences. Here we show a universal mechanism describing extreme events along with critical phenomena and derive a general expression of the probability distribution without concerning the physical details of individual events or critical properties. The general probability distribution unifies most important distributions in the field and demonstrates improved performance. The shape and symmetry of the general distribution is determined by the parameters of the fluctuations. Our work sheds judicious insights into the dynamical processes of complex systems with practical significance and provides a general approach of studying extreme and critical episodes in a combined and multidisciplinary scheme. PMID:26880219

  3. Universal quantum correlation close to quantum critical phenomena

    PubMed Central

    Qin, Meng; Ren, Zhong-Zhou; Zhang, Xin

    2016-01-01

    We study the ground state quantum correlation of Ising model in a transverse field (ITF) by implementing the quantum renormalization group (QRG) theory. It is shown that various quantum correlation measures and the Clauser-Horne-Shimony-Holt inequality will highlight the critical point related with quantum phase transitions, and demonstrate nonanalytic phenomena and scaling behavior when the size of the systems becomes large. Our results also indicate a universal behavior of the critical exponent of ITF under QRG theory that the critical exponent of different measures is identical, even when the quantities vary from entanglement measures to quantum correlation measures. This means that the two kinds of quantum correlation criterion including the entanglement-separability paradigm and the information-theoretic paradigm have some connections between them. These remarkable behaviors may have important implications on condensed matter physics because the critical exponent directly associates with the correlation length exponent. PMID:27189504

  4. Universal critical dynamics in high resolution neuronal avalanche data.

    PubMed

    Friedman, Nir; Ito, Shinya; Brinkman, Braden A W; Shimono, Masanori; DeVille, R E Lee; Dahmen, Karin A; Beggs, John M; Butler, Thomas C

    2012-05-18

    The tasks of neural computation are remarkably diverse. To function optimally, neuronal networks have been hypothesized to operate near a nonequilibrium critical point. However, experimental evidence for critical dynamics has been inconclusive. Here, we show that the dynamics of cultured cortical networks are critical. We analyze neuronal network data collected at the individual neuron level using the framework of nonequilibrium phase transitions. Among the most striking predictions confirmed is that the mean temporal profiles of avalanches of widely varying durations are quantitatively described by a single universal scaling function. We also show that the data have three additional features predicted by critical phenomena: approximate power law distributions of avalanche sizes and durations, samples in subcritical and supercritical phases, and scaling laws between anomalous exponents. PMID:23003192

  5. Universal quantum correlation close to quantum critical phenomena

    NASA Astrophysics Data System (ADS)

    Qin, Meng; Ren, Zhong-Zhou; Zhang, Xin

    2016-05-01

    We study the ground state quantum correlation of Ising model in a transverse field (ITF) by implementing the quantum renormalization group (QRG) theory. It is shown that various quantum correlation measures and the Clauser-Horne-Shimony-Holt inequality will highlight the critical point related with quantum phase transitions, and demonstrate nonanalytic phenomena and scaling behavior when the size of the systems becomes large. Our results also indicate a universal behavior of the critical exponent of ITF under QRG theory that the critical exponent of different measures is identical, even when the quantities vary from entanglement measures to quantum correlation measures. This means that the two kinds of quantum correlation criterion including the entanglement-separability paradigm and the information-theoretic paradigm have some connections between them. These remarkable behaviors may have important implications on condensed matter physics because the critical exponent directly associates with the correlation length exponent.

  6. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect

    Gregg L. Sharp; R. T. McCracken

    2003-06-01

    The regulatory requirement to develop an upgraded safety basis for a DOE nuclear facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830).1 Subpart B of 10 CFR 830, “Safety Basis Requirements,” requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements.1 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, “Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants”2 as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  7. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect

    Sharp, G.L.; McCracken, R.T.

    2003-05-13

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  8. 3D laser inspection of fuel assembly grid spacers for nuclear reactors based on diffractive optical elements

    NASA Astrophysics Data System (ADS)

    Finogenov, L. V.; Lemeshko, Yu A.; Zav'yalov, P. S.; Chugui, Yu V.

    2007-06-01

    Ensuring the safety and high operation reliability of nuclear reactors takes 100% inspection of geometrical parameters of fuel assemblies, which include the grid spacers performed as a cellular structure with fuel elements. The required grid spacer geometry of assembly in the transverse and longitudinal cross sections is extremely important for maintaining the necessary heat regime. A universal method for 3D grid spacer inspection using a diffractive optical element (DOE), which generates as the structural illumination a multiple-ring pattern on the inner surface of a grid spacer cell, is investigated. Using some DOEs one can inspect the nomenclature of all produced grids. A special objective has been developed for forming the inner surface cell image. The problems of diffractive elements synthesis, projecting optics calculation, adjusting methods as well as calibration of the experimental measuring system are considered. The algorithms for image processing for different constructive elements of grids (cell, channel hole, outer grid spacer rim) and the experimental results are presented.

  9. In-Situ Safeguards Verification of Low Burn-up Pressurized Water Reactor Spent Fuel Assemblies

    SciTech Connect

    Ham, Y S; Sitaraman, S; Park, I; Kim, J; Ahn, G

    2008-04-16

    A novel in-situ gross defect verification method for light water reactor spent fuel assemblies was developed and investigated by a Monte Carlo study. This particular method is particularly effective for old pressurized water reactor spent fuel assemblies that have natural uranium in their upper fuel zones. Currently there is no method or instrument that does verification of this type of spent fuel assemblies without moving the spent fuel assemblies from their storage positions. The proposed method uses a tiny neutron detector and a detector guiding system to collect neutron signals inside PWR spent fuel assemblies through guide tubes present in PWR assemblies. The data obtained in such a manner are used for gross defect verification of spent fuel assemblies. The method uses 'calibration curves' which show the expected neutron counts inside one of the guide tubes of spent fuel assemblies as a function of fuel burn-up. By examining the measured data in the 'calibration curves', the consistency of the operator's declaration is verified.

  10. Nondestructive verification with minimal movement of irradiated light-water-reactor fuel assemblies

    SciTech Connect

    Phillips, J.R.; Bosler, G.E.; Halbig, J.K.; Klosterbuer, S.F.; Menlove, H.O.

    1982-10-01

    Nondestructive verification of irradiated light-water reactor fuel assemblies can be performed rapidly and precisely by measuring their gross gamma-ray and neutron signatures. A portable system measured fuel assemblies with exposures ranging from 18.4 to 40.6 GWd/tU and with cooling times ranging from 1575 to 2638 days. Differences in the measured results for side or corner measurements are discussed. 25 figures, 20 tables.

  11. Fuel assembly for the production of tritium in light water reactors

    DOEpatents

    Cawley, W.E.; Trapp, T.J.

    1983-06-10

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  12. Fuel assembly for the production of tritium in light water reactors

    DOEpatents

    Cawley, William E.; Trapp, Turner J.

    1985-01-01

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  13. CONVECTION REACTOR

    DOEpatents

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  14. Criticality safety research at the University of Tennessee-Knoxville

    SciTech Connect

    Dodds, H.L.

    1997-06-01

    A list of seven research projects in nuclear criticality safety being conducted at the University of Tennessee is given. One of the projects is very briefly described. The study of space-dependent kinetics analysis of a hypothetical criticality accident involving an array of bottles containing UO{sub 2}F{sub 2} is being conducted for the US DOE, Oak Ridge National Laboratory, K-25 plant. Preliminary results for power versus time are presented, which indicate that space-time effects are significant after approximately 70 seconds. 2 refs., 1 fig.

  15. Application of international safeguards to fast critical assembly facilities. FY 1980 summary report

    SciTech Connect

    Not Available

    1980-12-01

    Nuclear materials inventory-verification techniques for large split-table fast critical assemblies are being studied in this program. Emphasis has been given to techniques that minimize fuel handling in order to reduce facility downtime and radiation exposure to the inventory team. The techniques studied include drawer seals, autoradiography, and spectral index measurements. Two-drawer sealing techniques have been studied, and the relative strengths and weaknesses are pointed out. The rod-type locking mechanism would not disrupt the reactor cooling air flow or interfere with autoradiography but is more expensive to implement. Passive autoradiography was used in a ZPPR inventory to verify to a 93% confidence level that less than 8-kg Pu was missing. The inventory was completed in four days by a five-member team with radiation exposures well within acceptable limits. Two autoradiographic film packages were developed to distinguish HEU from a DU matrix. The 30-mil pack requires an exposure between 4 and 16 hours and fits into most of the drawers. The 40-mil pack requires only a two-hour exposure but fits into less than half the drawers.

  16. Numerical simulation of gas dynamics and heat exchange tasks in fuel assemblies of the nuclear reactors

    SciTech Connect

    Zhuchenko, S. V.

    2014-11-12

    This report presents a PC-based program for solution gas dynamics and heat exchange mathematical tasks in fuel assemblies of the fast-neutron nuclear reactors. A fuel assembly consisting of bulk heat-generating elements, which are integrated together by the system of supply and pressure manifolds, is examined. Spherical heat-generating microelements, which contain nuclear fuel, are pulled into the heat-generating elements. Gaseous coolant proceed from supply manifolds to heat-generating elements, where it withdraws the nuclear reaction heat and assembles in pressure manifolds.

  17. Progress of Integral Experiments in Benchmark Fission Assemblies for a Blanket of Hybrid Reactor

    NASA Astrophysics Data System (ADS)

    Liu, R.; Zhu, T. H.; Yan, X. S.; Lu, X. X.; Jiang, L.; Wang, M.; Han, Z. J.; Wen, Z. W.; Lin, J. F.; Yang, Y. W.

    2014-04-01

    This article describes recent progress in integral neutronics experiments in benchmark fission assemblies for the blanket design in a hybrid reactor. The spherical assemblies consist of three layers of depleted uranium shells and several layers of polyethylene shells, separately. In the assemblies with centralizing the D-T neutron source, the plutonium production rates, uranium fission rates and leakage neutron spectra are measured. The measured results are compared to the calculated ones with the MCNP-4B code and ENDF/B-VI library data, available.

  18. Universal critical wrapping probabilities in the canonical ensemble

    NASA Astrophysics Data System (ADS)

    Hu, Hao; Deng, Youjin

    2015-09-01

    Universal dimensionless quantities, such as Binder ratios and wrapping probabilities, play an important role in the study of critical phenomena. We study the finite-size scaling behavior of the wrapping probability for the Potts model in the random-cluster representation, under the constraint that the total number of occupied bonds is fixed, so that the canonical ensemble applies. We derive that, in the limit L → ∞, the critical values of the wrapping probability are different from those of the unconstrained model, i.e. the model in the grand-canonical ensemble, but still universal, for systems with 2yt - d > 0 where yt = 1 / ν is the thermal renormalization exponent and d is the spatial dimension. Similar modifications apply to other dimensionless quantities, such as Binder ratios. For systems with 2yt - d ≤ 0, these quantities share same critical universal values in the two ensembles. It is also derived that new finite-size corrections are induced. These findings apply more generally to systems in the canonical ensemble, e.g. the dilute Potts model with a fixed total number of vacancies. Finally, we formulate an efficient cluster-type algorithm for the canonical ensemble, and confirm these predictions by extensive simulations.

  19. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods (1.506-cm Pitch)

    SciTech Connect

    Margaret A. Marshall

    2013-03-01

    A series of critical experiments were completed from 1962–1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967.a The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were performed to determine critical reflector arrangements, relative fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector” (see Reference 1). The experiment studied in this evaluation was the second of the series and had the fuel rods in a 1.506-cm-triangular pitch. One critical configuration was found (see Reference 3). Once the critical configuration had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U,bc and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configuration are described in Sections 1.3, 1.4, and 1.7, respectively.

  20. Criticality experiments and analysis of molybdenum reflected cylindrical uranyl fluoride water solution reactors

    NASA Technical Reports Server (NTRS)

    Fieno, D.; Fox, T.; Mueller, R.

    1972-01-01

    Clean criticality data were obtained from molybdenum-reflected cylindrical uranyl-fluoride-water solution reactors. Using ENDF/B molybdenum cross sections, a nine energy group two-dimensional transport calculation of a reflected reactor configuration predicted criticality to within 7 cents of the experimental value. For these reactors, it was necessary to compute the reflector resonance integral by a detailed transport calculation at the core-reflector interface volume in the energy region of the two dominant resonances of natural molybdenum.

  1. Critical current versus strain research at the University of Twente

    NASA Astrophysics Data System (ADS)

    van Eck, H. J. N.; van der Laan, D. C.; Dhallé, M.; ten Haken, B.; ten Kate, H. H. J.

    2003-09-01

    At the University of Twente a U-shaped spring has been used to investigate the mechanical properties of a large variety of superconducting tapes and wires. Several mechanisms are responsible for the degradation of critical current as a function of applied strain. A change in its intrinsic parameters causes a reversible critical current dependence in Nb3Sn. The critical current reaches a maximum at a wire-dependent tensile strain level, and decreases when this tensile strain is either released or further increased. In Bi-based tapes the critical current is virtually insensitive to tensile strain up to a sample-dependent irreversible strain limit. When this limit is exceeded, the critical current decreases steeply and irreversibly. This behaviour is attributed to microstructural damage to the filaments. This cracking of the filaments is verified by a magneto-optical strain experiment. Recent experiments suggest that in MgB2 the degradation of critical current is caused by a change in intrinsic properties and damage to the microstructure. Magneto-optical imaging can be used to investigate the influence of applied strain on the microstructure of MgB2, as is done successfully with Bi-based tapes. In all these conductors the thermal precompression of the filaments plays an important role. In Nb3Sn it determines the position of the maximum and in Bi-based and MgB2 conductors it is closely related to the irreversible strain limit.

  2. A fuel for sub-critical fast reactor

    NASA Astrophysics Data System (ADS)

    Moiseenko, V. E.; Chernitskiy, S. V.; Ågren, O.; Noack, K.

    2012-06-01

    Along with the problem of the nuclear waste transmutation, the problem of minimization of waste production is of current interest. It is not possible to eliminate production of waste at a nuclear power plant, but, as is shown in this report, it is in principle possible to arrange a fuel composition with no net production of transuranic elements. The idea is to find the transuranic elements composition to which the depleted uranium is continuously supplied during frequent reprocessing, and amount of each other transuranic fuel component remains unchanged in time. For each transuranic component, the balance is achieved by equating burnup and production rates. The production is due to neutron capture by the neighboring lighter isotope and subsequent beta-decay. The burnup includes fission, neutron capture and decays. For the calculations a simplified burnup model which accounts for 9 isotopes of uranium, neptunium, plutonium and americium is used. The calculated fuel composition consists mainly of uranium with minority of plutonium isotopes. Such a fuel, after usage in a sub-critical fast reactor, should be reprocessed. The fission product content increases during burnup, representing a net production of waste, while the transuranic elements and 238U should be recycled into a new fuel. For such a fuel cycle, the net consumption is only for 238U, and the net waste production is just fission products.

  3. A fuel for sub-critical fast reactor

    SciTech Connect

    Moiseenko, V. E.; Chernitskiy, S. V.; Agren, O.; Noack, K.

    2012-06-19

    Along with the problem of the nuclear waste transmutation, the problem of minimization of waste production is of current interest. It is not possible to eliminate production of waste at a nuclear power plant, but, as is shown in this report, it is in principle possible to arrange a fuel composition with no net production of transuranic elements. The idea is to find the transuranic elements composition to which the depleted uranium is continuously supplied during frequent reprocessing, and amount of each other transuranic fuel component remains unchanged in time. For each transuranic component, the balance is achieved by equating burnup and production rates. The production is due to neutron capture by the neighboring lighter isotope and subsequent beta-decay. The burnup includes fission, neutron capture and decays. For the calculations a simplified burnup model which accounts for 9 isotopes of uranium, neptunium, plutonium and americium is used. The calculated fuel composition consists mainly of uranium with minority of plutonium isotopes. Such a fuel, after usage in a sub-critical fast reactor, should be reprocessed. The fission product content increases during burnup, representing a net production of waste, while the transuranic elements and {sup 238}U should be recycled into a new fuel. For such a fuel cycle, the net consumption is only for 238U, and the net waste production is just fission products.

  4. Critical Issues on Materials for Gen-IV Reactors

    SciTech Connect

    Caro, M; Marian, J; Martinez, E; Erhart, P

    2009-02-27

    Within the LDRD on 'Critical Issues on Materials for Gen-IV Reactors' basic thermodynamics of the Fe-Cr alloy and accurate atomistic modeling were used to help develop the capability to predict hardening, swelling and embrittlement using the paradigm of Multiscale Materials Modeling. Approaches at atomistic and mesoscale levels were linked to build-up the first steps in an integrated modeling platform that seeks to relate in a near-term effort dislocation dynamics to polycrystal plasticity. The requirements originated in the reactor systems under consideration today for future sources of nuclear energy. These requirements are beyond the present day performance of nuclear materials and calls for the development of new, high temperature, radiation resistant materials. Fe-Cr alloys with 9-12% Cr content are the base matrix of advanced ferritic/martensitic (FM) steels envisaged as fuel cladding and structural components of Gen-IV reactors. Predictive tools are needed to calculate structural and mechanical properties of these steels. This project represents a contribution in that direction. The synergy between the continuous progress of parallel computing and the spectacular advances in the theoretical framework that describes materials have lead to a significant advance in our comprehension of materials properties and their mechanical behavior. We took this progress to our advantage and within this LDRD were able to provide a detailed physical understanding of iron-chromium alloys microstructural behavior. By combining ab-initio simulations, many-body interatomic potential development, and mesoscale dislocation dynamics we were able to describe their microstructure evolution. For the first time in the case of Fe-Cr alloys, atomistic and mesoscale were merged and the first steps taken towards incorporating ordering and precipitation effects into dislocation dynamics (DD) simulations. Molecular dynamics (MD) studies of the transport of self-interstitial, vacancy and

  5. Critical factors for assembling a high volume of DNA barcodes

    PubMed Central

    Hajibabaei, Mehrdad; deWaard, Jeremy R; Ivanova, Natalia V; Ratnasingham, Sujeevan; Dooh, Robert T; Kirk, Stephanie L; Mackie, Paula M; Hebert, Paul D.N

    2005-01-01

    Large-scale DNA barcoding projects are now moving toward activation while the creation of a comprehensive barcode library for eukaryotes will ultimately require the acquisition of some 100 million barcodes. To satisfy this need, analytical facilities must adopt protocols that can support the rapid, cost-effective assembly of barcodes. In this paper we discuss the prospects for establishing high volume DNA barcoding facilities by evaluating key steps in the analytical chain from specimens to barcodes. Alliances with members of the taxonomic community represent the most effective strategy for provisioning the analytical chain with specimens. The optimal protocols for DNA extraction and subsequent PCR amplification of the barcode region depend strongly on their condition, but production targets of 100K barcode records per year are now feasible for facilities working with compliant specimens. The analysis of museum collections is currently challenging, but PCR cocktails that combine polymerases with repair enzyme(s) promise future success. Barcode analysis is already a cost-effective option for species identification in some situations and this will increasingly be the case as reference libraries are assembled and analytical protocols are simplified. PMID:16214753

  6. Replacement of split-pin assemblies in nuclear reactors

    SciTech Connect

    Nee, J.D.; Green, R.A.

    1989-12-12

    This patent describes a pin-insertion/torque tool for the replacement of old split-pin assemblies. Each of the new split-pin assemblies including a new split-pin having times and a new nut for securing the new split pin in the guide tube, a new nut being inserted in the guide tube in position to receive a split pin. The the pin-insertion/torque tool including a blade means for engaging a new split pin with the blade with the tines of the new split pins straddling the blade, means, connected to the blade, for advancing the split-pin into the guide tube into threading engagement with the new nut positioned to receive a new split pin and means, to be connected to the nut for securing the new nut onto the new split pin while the split pin is engaged by the blade.

  7. Silicon carbide composite for light water reactor fuel assembly applications

    NASA Astrophysics Data System (ADS)

    Yueh, Ken; Terrani, Kurt A.

    2014-05-01

    The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  8. 75 FR 27372 - University of New Mexico; University of New Mexico AGN-201M Reactor; Environmental Assessment and...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-05-14

    ... COMMISSION University of New Mexico; University of New Mexico AGN-201M Reactor; Environmental Assessment and... considering issuance of a renewed Facility Operating License No. R-102, to the University of New Mexico (the licensee), which would authorize continued operation of the University of New Mexico AGN-201M...

  9. PBF Reactor Building (PER620). Detail of fuel test assembly in ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    PBF Reactor Building (PER-620). Detail of fuel test assembly in preparation for test. When complete, it will fit into in-pile tube. The maximum outside diameter of which must be about 8.25 inches. Date: 1982. INEEL negative no. 82-4908 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  10. Critical dynamics of randomly assembled and diluted threshold networks

    NASA Astrophysics Data System (ADS)

    Kürten, Karl E.; Clark, John W.

    2008-04-01

    The dynamical behavior of a class of randomly assembled networks of binary threshold units subject to random deletion of connections is studied based on the annealed approximation suitable in the thermodynamic limit. The dynamical phase diagram is constructed for several forms of the probability density distribution of nonvanishing connection strengths. The family of power-law distribution functions ρ0(x)=(1-α)/(2|x|α) is found to play a special role in expanding the domain of stable, ordered dynamics at the expense of the disordered, “chaotic” phase. Relationships with other recent studies of the dynamics of complex networks allowing for variable in-degree of the units are explored. The relevance of the pruning of network connections to neural modeling and developmental neurobiology is discussed.

  11. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    SciTech Connect

    Nash, C.A. ); Blake, J.E.; Rush, G.C. )

    1990-01-01

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m{sup 2}) (1.1E+6 BTU/(ft{sup 2}hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient.

  12. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    SciTech Connect

    Nash, C.A.; Blake, J.E.; Rush, G.C.

    1990-12-31

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m{sup 2}) (1.1E+6 BTU/(ft{sup 2}hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient.

  13. Universal thermodynamics at the liquid-vapor critical point.

    PubMed

    Sanchez, Isaac C; Boening, Kevin L

    2014-11-26

    For 68 fluids that include hydrogen bonding and quantum fluids, the fugacity coefficient that defines the residual chemical potential adopts a near universal value of 2/3 at the critical point. More precisely, the reciprocal of the fugacity coefficient equals 1.52 ± 0.02 and includes fluids as diverse as helium (1.50), dodecafluoropentane (1.50), and water (1.53). For 65 classical fluids, a dimensionless thermal pressure coefficient and internal pressure attain critical values of 1.88 ± 0.11 and 1.61 ± 0.11, respectively. From equations of state, values of these new critical constants have been calculated and agree favorably with experimental values. Specifically, for the critical fugacity coefficient, the following results were obtained for its reciprocal: van der Waals (1.44), lattice gas (1.43), scaled particle theory (1.46), and the Redlich-Kwong eq (1.50). The semiempirical Redlich-Kwong equation is also the most accurate for the thermal pressure coefficient (1.86) and internal pressure (1.53). Physical interpretations of these results are discussed as well as their implications for other critical phenomena. PMID:25369319

  14. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 5 - North Anna Unit 1 Cycle 5

    SciTech Connect

    Bowman, S.M.

    1993-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor (AFR) criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial pressurized-water reactors (PWR). The analysis methodology selected for all calculations reported herein was the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and to provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. This unit and cycle were chosen for a previous analysis using a different methodology because detailed isotopics from multidimensional reactor calculations were available from the Virginia Power Company. These data permitted comparison of criticality calculations directly using the utility-calculated isotopics to those using the isotopics generated by the SCALE-4 SAS2H

  15. A new flooding correlation development and its critical heat flux predictions under low air-water flow conditions in Savannah River Site assembly channels

    SciTech Connect

    Lee, S.Y. )

    1993-10-01

    The upper limit to countercurrent flow, namely, flooding, is important to analyze the reactor coolability during an emergency cooling system (ECS) phase as a result of a large-break loss-of-coolant accident (LOCA) such as a double-ended guillotine break in the Savannah River Site (SRS) reactor system. During normal operation, the reactor coolant system utilizes downward flow through concentric heated tubes with ribs, which subdivided each annular channel into four subchannels. In this paper, a new flooding correlation has been developed based on the analytical models and literature data for adiabatic, steady-state, one-dimensional, air-water flow to predict flooding phenomenon in the SRS reactor assembly channel, which may have a counter-current air-water flow pattern during the ECS phase. In addition, the correlation was benchmarked against the experimental data conducted under the Oak Ridge National Laboratory multislit channel, which is close to the SRS assembly geometry. Furthermore, the correlation has also been used as a constitutive relationship in a new two-component two-phase thermal-hydraulics code FLOWTRAN-TF, which has been developed for a detailed analysis of SRS reactor assembly behavior during LOCA scenarios. Finally, the flooding correlation was applied to the predictions of critical heat flux, and the results were compared with the data taken by the SRS heat transfer laboratory under a single annular channel with ribs and a multiannular prototypic test rig.

  16. Planar ceramic membrane assembly and oxidation reactor system

    DOEpatents

    Carolan, Michael Francis; Dyer, legal representative, Kathryn Beverly; Wilson, Merrill Anderson; Ohrn, Ted R.; Kneidel, Kurt E.; Peterson, David; Chen, Christopher M.; Rackers, Keith Gerard; Dyer, Paul Nigel

    2009-04-07

    Planar ceramic membrane assembly comprising a dense layer of mixed-conducting multi-component metal oxide material, wherein the dense layer has a first side and a second side, a porous layer of mixed-conducting multi-component metal oxide material in contact with the first side of the dense layer, and a ceramic channeled support layer in contact with the second side of the dense layer. The planar ceramic membrane assembly can be used in a ceramic wafer assembly comprising a planar ceramic channeled support layer having a first side and a second side; a first dense layer of mixed-conducting multi-component metal oxide material having an inner side and an outer side, wherein the inner side is in contact with the first side of the ceramic channeled support layer; a first outer support layer comprising porous mixed-conducting multi-component metal oxide material and having an inner side and an outer side, wherein the inner side is in contact with the outer side of the first dense layer; a second dense layer of mixed-conducting multi-component metal oxide material having an inner side and an outer side, wherein the inner side is in contact with the second side of the ceramic channeled layer; and a second outer support layer comprising porous mixed-conducting multi-component metal oxide material and having an inner side and an outer side, wherein the inner side is in contact with the outer side of the second dense layer.

  17. Planar ceramic membrane assembly and oxidation reactor system

    DOEpatents

    Carolan, Michael Francis; Dyer, legal representative, Kathryn Beverly; Wilson, Merrill Anderson; Ohm, Ted R.; Kneidel, Kurt E.; Peterson, David; Chen, Christopher M.; Rackers, Keith Gerard; Dyer, deceased, Paul Nigel

    2007-10-09

    Planar ceramic membrane assembly comprising a dense layer of mixed-conducting multi-component metal oxide material, wherein the dense layer has a first side and a second side, a porous layer of mixed-conducting multi-component metal oxide material in contact with the first side of the dense layer, and a ceramic channeled support layer in contact with the second side of the dense layer. The planar ceramic membrane assembly can be used in a ceramic wafer assembly comprising a planar ceramic channeled support layer having a first side and a second side; a first dense layer of mixed-conducting multi-component metal oxide material having an inner side and an outer side, wherein the inner side is in contact with the first side of the ceramic channeled support layer; a first outer support layer comprising porous mixed-conducting multi-component metal oxide material and having an inner side and an outer side, wherein the inner side is in contact with the outer side of the first dense layer; a second dense layer of mixed-conducting multi-component metal oxide material having an inner side and an outer side, wherein the inner side is in contact with the second side of the ceramic channeled layer; and a second outer support layer comprising porous mixed-conducting multi-component metal oxide material and having an inner side and an outer side, wherein the inner side is in contact with the outer side of the second dense layer.

  18. Summary Report of Commercial reactor Criticality Data for Three Mile Island Unit 1

    SciTech Connect

    Larry B. Wimmer

    2001-08-29

    The objective of the ''Summary Report of Commercial Reactor Criticality Data for Three Mile Island Unit I'' is to present the CRC data for the TMI-1 reactor. Results from the CRC evaluations will support the development and validation of the neutronics models used for criticality analyses involving commercial spent nuclear fuel. These models and their validation are discussed in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000).

  19. Method of preparing gas tags for identification of single and multiple failures of nuclear reactor fuel assemblies

    DOEpatents

    McCormick, Norman J.

    1976-01-01

    For use in the identification of failed fuel assemblies in a nuclear reactor, the ratios of the tag gas isotopic concentrations are located on curved surfaces to enable the ratios corresponding to failure of a single fuel assembly to be distinguished from those formed from any combination of two or more failed assemblies.

  20. Agreement between University of Hawaii Professional Assembly and the Board of Regents of the University of Hawaii, 1987-1989.

    ERIC Educational Resources Information Center

    National Education Association, Washington, DC.

    The collective bargaining agreement between University of Hawaii Professional Assembly and the Board of Regents of the University of Hawaii for the period 1987-1989 is presented. Items covered in the agreement include: unit recognition, outside employment, exemption from tuition, duty period, paid and unpaid leaves of absence, sabbatical leaves,…

  1. Temperature actuated shutdown assembly for a nuclear reactor

    DOEpatents

    Sowa, Edmund S.

    1976-01-01

    Three identical bimetallic disks, each shaped as a spherical cap with its convex side composed of a layer of metal such as molybdenum and its concave side composed of a metal of a relatively higher coefficient of thermal expansion such as stainless steel, are retained within flanges attached to three sides of an inner hexagonal tube containing a neutron absorber to be inserted into a nuclear reactor core. Each disk holds a metal ball against its normally convex side so that the ball projects partially through a hole in the tube located concentrically with the center of each disk; at a predetermined temperature an imbalance of thermally induced stresses in at least one of the disks will cause its convex side to become concave and its concave side to become convex, thus pulling the ball from the hole in which it is located. The absorber has a conical bottom supported by the three balls and is small enough in relation to the internal dimensions of the tube to allow it to slip toward the removed ball or balls, thus clearing the unremoved balls or ball so that it will fall into the reactor core.

  2. EVALUATION OF THE INITIAL CRITICAL CONFIGURATION OF THE HTR-10 PEBBLE-BED REACTOR

    SciTech Connect

    William K. Terry

    2005-11-01

    This report describes the evaluation of data from the initial criticality measurement of the HTR-10 pebble-bed reactor at the Institute of Nuclear Energy Technology in China to determine whether the data are of sufficient quality to use as benchmarks for reactor physics computer codes intended for pebble-bed reactor analysis. The evaluation applied the INL pebble-bed reactor physics code PEBBED to perform an uncertainty analysis on the core critical height. The overall uncertainty in k-effective was slightly over 0.5%, which is considered adequate for an experimental benchmark.

  3. Testing universality in critical exponents: The case of rainfall

    NASA Astrophysics Data System (ADS)

    Deluca, Anna; Puig, Pedro; Corral, Álvaro

    2016-04-01

    One of the key clues to consider rainfall as a self-organized critical phenomenon is the existence of power-law distributions for rain-event sizes. We have studied the problem of universality in the exponents of these distributions by means of a suitable statistic whose distribution is inferred by several variations of a permutational test. In contrast to more common approaches, our procedure does not suffer from the difficulties of multiple testing and does not require the precise knowledge of the uncertainties associated to the power-law exponents. When applied to seven sites monitored by the Atmospheric Radiation Measurement Program the tests lead to the rejection of the universality hypothesis, despite the fact that the exponents are rather close to each other. We discuss the reasons of the rejection.

  4. Universal quantum criticality in Hubbard models with massless Dirac dispersion

    NASA Astrophysics Data System (ADS)

    Otsuka, Yuichi; Yunoki, Seiji; Sorella, Sandro

    We investigate the metal-insulator transition of two-dimensional interacting electrons with massless Dirac-like dispersion, describe by the Hubbard models on two geometrically different lattices: honeycomb and π-flux square lattices. By performing large-scale quantum Monte Carlo simulations followed by careful finite-size scaling analyses, we find that the transition from semi-metallic to antiferromagnetic insulating phases is continuous and evaluate the critical exponents with a high degree of accuracy for the corresponding universality class, which is described in the continuous limit by the Gross-Neveu model. We furthermore discuss the fate of the quasiparticle weight and the Fermi velocity across this transition.

  5. Three-dimensional discrete ordinates reactor assembly calculations on GPUs

    SciTech Connect

    Evans, Thomas M; Joubert, Wayne; Hamilton, Steven P; Johnson, Seth R; Turner, John A; Davidson, Gregory G; Pandya, Tara M

    2015-01-01

    In this paper we describe and demonstrate a discrete ordinates sweep algorithm on GPUs. This sweep algorithm is nested within a multilevel comunication-based decomposition based on energy. We demonstrated the effectiveness of this algorithm on detailed three-dimensional critical experiments and PWR lattice problems. For these problems we show improvement factors of 4 6 over conventional communication-based, CPU-only sweeps. These sweep kernel speedups resulted in a factor of 2 total time-to-solution improvement.

  6. Dose rate estimates from irradiated light-water-reactor fuel assemblies in air

    SciTech Connect

    Lloyd, W.R.; Sheaffer, M.K.; Sutcliffe, W.G.

    1994-01-31

    It is generally considered that irradiated spent fuel is so radioactive (self-protecting) that it can only be moved and processed with specialized equipment and facilities. However, a small, possibly subnational, group acting in secret with no concern for the environment (other than the reduction of signatures) and willing to incur substantial but not lethal radiation doses, could obtain plutonium by stealing and processing irradiated spent fuel that has cooled for several years. In this paper, we estimate the dose rate at various distances and directions from typical pressurized-water reactor (PWR) and boiling-water reactor (BWR) spent-fuel assemblies as a function of cooling time. Our results show that the dose rate is reduced rapidly for the first ten years after exposure in the reactor, and that it is reduced by a factor of {approx}10 (from the one year dose rate) after 15 years. Even for fuel that has cooled for 15 years, a lethal dose (LD50) of 450 rem would be received at 1 m from the center of the fuel assembly after several minutes. However, moving from 1 to 5 m reduces the dose rate by over a factor of 10, and moving from 1 to 10 m reduces the dose rate by about a factor of 50. The dose rates 1 m from the top or bottom of the assembly are considerably less (about 10 and 22%, respectively) than 1 m from the center of the assembly, which is the direction of the maximum dose rate.

  7. Critical Configuration and Physics Mesaurements for Graphite Reflected Assemblies of U(93.15)O2 Fuel Rods (1.27-CM Pitch)

    SciTech Connect

    Margaret A. Marshall

    2011-09-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory's Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950's efforts were made to study 'power plants for the production of electrical power in space vehicles'. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in FY 1964, 1965, and 1966. A summary of the program's effort was compiled in 1967. The delayed critical experiments served as a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated 253 stainless steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. 'The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.' The experiment studied within this evaluation was the first of the series and had the 253 fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Information for this evaluation was compiled from Reference 1 and 2, reports on subsequent experiments in the series, and the experimental logbook as well as from communication with the experimenter, John T. Mihalczo.

  8. Measurements of control rod efficiency in RBMK critical assembly upon dropping of the rods

    SciTech Connect

    Zhitarev, V. E. Kachanov, V. M.; Sergevnin, A. Yu.; Lebedev, G. V.

    2014-12-15

    The efficiency of control rods in the RBMK critical assembly was measured in the case where one manual-control rod (MCR) is dropped from a steady critical state, and several other MCRs were additionally dropped after 44 s. The measured number of neutrons in the assembly during and after dropping of the rods was used to calculate the efficiency values of the rods by solution of the system of point kinetics equations. A series of methods of the initial data treatment for determination of the desired values of reactivity without the calculated corrections were used.

  9. Critical length scales and strain localization govern the mechanical performance of multi-layer graphene assemblies

    NASA Astrophysics Data System (ADS)

    Xia, Wenjie; Ruiz, Luis; Pugno, Nicola M.; Keten, Sinan

    2016-03-01

    Multi-layer graphene assemblies (MLGs) or fibers with a staggered architecture exhibit high toughness and failure strain that surpass those of the constituent single sheets. However, how the architectural parameters such as the sheet overlap length affect these mechanical properties remains unknown due in part to the limitations of mechanical continuum models. By exploring the mechanics of MLG assemblies under tensile deformation using our established coarse-grained molecular modeling framework, we have identified three different critical interlayer overlap lengths controlling the strength, plastic stress, and toughness of MLGs, respectively. The shortest critical length scale Lsc governs the strength of the assembly as predicted by the shear-lag model. The intermediate critical length Lpc is associated with a dynamic frictional process that governs the strain localization propensity of the assembly, and hence the failure strain. The largest critical length scale LTc corresponds to the overlap length necessary to achieve 90% of the maximum theoretical toughness of the material. Our analyses provide the general guidelines for tuning the constitutive properties and toughness of multilayer 2D nanomaterials using elasticity, interlayer adhesion energy and geometry as molecular design parameters.Multi-layer graphene assemblies (MLGs) or fibers with a staggered architecture exhibit high toughness and failure strain that surpass those of the constituent single sheets. However, how the architectural parameters such as the sheet overlap length affect these mechanical properties remains unknown due in part to the limitations of mechanical continuum models. By exploring the mechanics of MLG assemblies under tensile deformation using our established coarse-grained molecular modeling framework, we have identified three different critical interlayer overlap lengths controlling the strength, plastic stress, and toughness of MLGs, respectively. The shortest critical length scale

  10. Radiopharmaceuticals developed at the University of Missouri research reactor

    SciTech Connect

    Ketring, A.R.; Ehrhardt, G.J.; Day, D.E.

    1997-12-01

    The University of Missouri Research Reactor (MURR) has put a great deal of effort in the last two decades into development of radiotherapeutic beta emitters as nuclear medicine radiotherapeutics for malignancies. This paper describes the development of two of these drugs, {sup 153}Sm ethylenediaminetetra-methylene phosphonic acid (EDTMP) (Quadramet{trademark}) and {sup 90}Y glass microspheres (TheraSphere{trademark}). Samarium-153 EDTMP is a palliative used to treat the pain of metastatic bone cancer without the side effects of narcotic pain killers. Yttrium-90 glass microspheres are delivered via hepatic artery catheter to embolize the capillaries of liver tumors and deliver a large radiation dose for symptom palliation and life prolonging purposes.

  11. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    SciTech Connect

    Pond, R.B.; Matos, J.E.

    1996-05-01

    As part of the Department of Energy`s spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, are not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report.

  12. 77 FR 27487 - License Amendment Request From The State University of New York, University of Buffalo Reactor...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-10

    ... accordance with the NRC E-Filing rule (72 FR 49139, August 28, 2007). The E-Filing process requires... COMMISSION License Amendment Request From The State University of New York, University of Buffalo Reactor....resource@nrc.gov . The University of Buffalo Decommissioning Plan and License Amendment Request...

  13. A Theoretical Investigation of Oxidation Efficiency of a Volatile Removal Assembly Reactor Under Microgravity Conditions

    NASA Technical Reports Server (NTRS)

    Guo, Boyun

    2005-01-01

    Volatile Removal Assembly (VRA) is a subsystem of the Closed Environment Life Support System (CELSS) installed in the International Space Station. It is used for removing contaminants (volatile organics) in the wastewater produced by the space station crews. The major contaminants are formic acid, ethanol, and propylene glycol. The VRA contains a slim packbed reactor (3.5 cm diameter and four 28 cm long tubes in series) to perform catalyst oxidation of wastewater at elevated pressure and temperature under microgravity conditions. In the reactor, the contaminants are burned with oxygen gas (O2) to form water and carbon dioxide (CO2) that dissolves in the water stream. Optimal design of the reactor requires a thorough understanding about how the reactor performs under microgravity conditions. The objective of this study was to develop a mathematical model to interpret experimental data obtained from normal and microgravity conditions, and to predict the performance of VRA reactor under microgravity conditions. Catalyst oxidation kinetics and the total oxygen-water contact area control the efficiency of catalyst oxidation for mass transfer, which depends on oxygen gas holdup and distribution in the reactor. The process involves bubbly flow in porous media with chemical reactions in microgravity environment. This presents a unique problem in fluid dynamics that has not been studied. Guo et al. (2004) developed a mathematical model that predicts oxygen holdup in the VRA reactor. No mathematical model has been found in the literature that can be used to predict the efficiency of catalyst oxidation under microgravity conditions.

  14. Waterproofed photomultiplier tube assemblies for the Daya Bay reactor neutrino experiment

    NASA Astrophysics Data System (ADS)

    Chow, Ken; Cummings, John; Edwards, Emily; Edwards, William; Ely, Ry; Hoff, Matthew; Lebanowski, Logan; Li, Bo; Li, Piyi; Lin, Shih-Kai; Liu, Dawei; Liu, Jinchang; Luk, Kam-Biu; Miao, Jiayuan; Napolitano, Jim; Ochoa-Ricoux, Juan Pedro; Peng, Jen-Chieh; Qi, Ming; Steiner, Herbert; Stoler, Paul; Stuart, Mary; Wang, Lingyu; Yang, Changgen; Zhong, Weili

    2015-09-01

    In the Daya Bay Reactor Neutrino Experiment 960 20-cm-diameter waterproof photomultiplier tubes are used to instrument three water pools as Cherenkov detectors for detecting cosmic-ray muons. Of these 960 photomultiplier tubes, 341 are recycled from the MACRO experiment. A systematic program was undertaken to refurbish them as waterproof assemblies. In the context of passing the water leakage check, a success rate better than 97% was achieved. Details of the design, fabrication, testing, operation, and performance of these waterproofed photomultiplier-tube assemblies are presented.

  15. Research reactors

    SciTech Connect

    Tonneson, L.C.; Fox, G.J.

    1996-04-01

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world`s research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted.

  16. CSER 01-011 Criticality Safety Evaluation for Light Water Reactor Fuel in NAC-1 Casks

    SciTech Connect

    ERICKSON, D.G.

    2002-06-26

    Document presents analysis performed to demonstrate criticality safety of packaging spent PWR fuel assemblies currently located at the 324 Building into a NAC-1 cask. Interim storage of the cask is also documented.

  17. Simulation of reactor pulses in fast burst and externally driven nuclear assemblies

    NASA Astrophysics Data System (ADS)

    Green, Taylor Caldwell, IV

    The following research contributes original concepts to the fields of deterministic neutron transport modeling and reactor power excursion simulation. A deterministic neutron transport code was created to assess the value of new methods of determining neutron current, fluence, and flux values through the use of view factor and average path length calculations. The neutron transport code is also capable of modeling the highly anisotropic neutron transport of deuterium-tritium fusion external source neutrons using diffusion theory with the aid of a modified first collision source term. The neutron transport code was benchmarked with MCNP, an industry standard stochastic neutron transport code. Deterministic neutron transport methods allow users to model large quantities of neutrons without simulating their interactions individually. Subsequently, deterministic methods allow users to more easily couple neutron transport simulations with other physics simulations. Heat transfer and thermoelastic mechanics physics simulation modules were each developed and benchmarked using COMSOL, a commercial heat transfer and mechanics simulation software. The physics simulation modules were then coupled and used to simulate reactor pulses in fast burst and externally driven nuclear assemblies. The coupled system of equations represents a new method of simulating reactor pulses that allows users to more fully characterize pulsed assemblies. Unlike older methods of reactor pulse simulation, the method presented in this research does not require data from the operational reactor in order to simulate its behavior. The ability to simulate the coupled neutron transport and thermo-mechanical feedback present in pulsed reactors prior their construction would significantly enhance the quality of pulsed reactor pre-construction safety analysis. Additionally, a graphical user interface is created to allow users to run simulations and visualize the results using the coupled physics simulation

  18. Nuclear reactor fuel assembly duct-tube-to-inlet-nozzle attachment system

    DOEpatents

    Christiansen, David W.; Smith, Bob G.

    1982-01-01

    A reusable system for removably attaching the lower end 21 of a nuclear reactor fuel assembly duct tube to an upper end 11 of a nuclear reactor fuel assembly inlet nozzle. The duct tube's lower end 21 has sides terminating in locking tabs 22 which end in inwardly-extending flanges 23. The flanges 23 engage recesses 13 in the top section 12 of the inlet nozzle's upper end 11. A retaining collar 30 slides over the inlet nozzle's upper end 11 to restrain the flanges 23 in the recesses 13. A locking nut 40 has an inside threaded portion 41 which engages an outside threaded portion 15 of the inlet nozzle's upper end 11 to secure the retaining collar 30 against protrusions 24 on the duct tube's sides.

  19. Assemblies and methods for mitigating effects of reactor pressure vessel expansion

    DOEpatents

    Challberg, Roy C.; Gou, Perng-Fei; Chu, Cherk Lam; Oliver, Robert P.

    1999-01-01

    Support assemblies for allowing RPV radial expansion while simultaneously limiting horizontal, vertical, and azimuthal movement of the RPV within a nuclear reactor are described. In one embodiment, the support assembly includes a support block and a guide block. The support block includes a first portion and a second portion, and the first portion is rigidly coupled to the RPV adjacent the first portion. The guide block is rigidly coupled to a reactor pressure vessel support structure and includes a channel sized to receive the second portion of the support block. The second portion of the support block is positioned in the guide block channel to movably couple the guide block to the support block.

  20. Assemblies and methods for mitigating effects of reactor pressure vessel expansion

    DOEpatents

    Challberg, R.C.; Gou, P.F.; Chu, C.L.; Oliver, R.P.

    1999-07-27

    Support assemblies for allowing RPV radial expansion while simultaneously limiting horizontal, vertical, and azimuthal movement of the RPV within a nuclear reactor are described. In one embodiment, the support assembly includes a support block and a guide block. The support block includes a first portion and a second portion, and the first portion is rigidly coupled to the RPV adjacent the first portion. The guide block is rigidly coupled to a reactor pressure vessel support structure and includes a channel sized to receive the second portion of the support block. The second portion of the support block is positioned in the guide block channel to movably couple the guide block to the support block. 6 figs.

  1. The Ohio State University Reactor Sharing Program [August 15, 2000 - May 31, 2001

    SciTech Connect

    Myser, Richard D.

    2002-01-04

    During the period from August 15, 2000 through may 31, 2001 the Ohio State University (OSU) Research Reactor participated in the Reactor Sharing Program by providing services to nine colleges and universities and four secondary school organizations. A total of about 17 faculty and 170 students utilized their facilities. The staff of the OSU Research Reactor is generally involved in four types of experiments at the Nuclear Reactor Laboratory. Included are introductions to nuclear research, neutron activation analysis, material irradiation, and classes that measure various reactor parameters.

  2. Critical length scales and strain localization govern the mechanical performance of multi-layer graphene assemblies.

    PubMed

    Xia, Wenjie; Ruiz, Luis; Pugno, Nicola M; Keten, Sinan

    2016-03-17

    Multi-layer graphene assemblies (MLGs) or fibers with a staggered architecture exhibit high toughness and failure strain that surpass those of the constituent single sheets. However, how the architectural parameters such as the sheet overlap length affect these mechanical properties remains unknown due in part to the limitations of mechanical continuum models. By exploring the mechanics of MLG assemblies under tensile deformation using our established coarse-grained molecular modeling framework, we have identified three different critical interlayer overlap lengths controlling the strength, plastic stress, and toughness of MLGs, respectively. The shortest critical length scale L governs the strength of the assembly as predicted by the shear-lag model. The intermediate critical length L is associated with a dynamic frictional process that governs the strain localization propensity of the assembly, and hence the failure strain. The largest critical length scale L corresponds to the overlap length necessary to achieve 90% of the maximum theoretical toughness of the material. Our analyses provide the general guidelines for tuning the constitutive properties and toughness of multilayer 2D nanomaterials using elasticity, interlayer adhesion energy and geometry as molecular design parameters. PMID:26935048

  3. Two distinct domains of protein 4.1 critical for assembly of functional nuclei in vitro.

    PubMed

    Krauss, Sharon Wald; Heald, Rebecca; Lee, Gloria; Nunomura, Wataru; Gimm, J Aura; Mohandas, Narla; Chasis, Joel Anne

    2002-11-15

    Protein 4.1R, a multifunctional structural protein, acts as an adaptor in mature red cell membrane skeletons linking spectrin-actin complexes to plasma membrane-associated proteins. In nucleated cells protein 4.1 is not associated exclusively with plasma membrane but is also detected at several important subcellular locations crucial for cell division. To identify 4.1 domains having critical functions in nuclear assembly, 4.1 domain peptides were added to Xenopus egg extract nuclear reconstitution reactions. Morphologically disorganized, replication deficient nuclei assembled when spectrin-actin-binding domain or NuMA-binding C-terminal domain peptides were present. However, control variant spectrin-actin-binding domain peptides incapable of binding actin or mutant C-terminal domain peptides with reduced NuMA binding had no deleterious effects on nuclear reconstitution. To test whether 4.1 is required for proper nuclear assembly, 4.1 isoforms were depleted with spectrin-actin binding or C-terminal domain-specific antibodies. Nuclei assembled in the depleted extracts were deranged. However, nuclear assembly could be rescued by the addition of recombinant 4.1R. Our data establish that protein 4.1 is essential for nuclear assembly and identify two distinct 4.1 domains, initially characterized in cytoskeletal interactions, that have crucial and versatile functions in nuclear assembly. PMID:12171917

  4. Partial Defect Verification of the Pressurized Water Reactor Spent Fuel Assemblies

    SciTech Connect

    Ham, Y S; Sitaraman, S

    2010-02-05

    The International Atomic Energy Agency (IAEA) has the responsibility to carry out independent inspections of all nuclear material and facilities subject to safeguards agreements in order to verify compliance with non-proliferation commitments. New technologies have been continuously explored by the IAEA and Member States to improve the verification measures to account for declared inventory of nuclear material and detect clandestine diversion and production of nuclear materials. Even with these efforts, a technical safeguards challenge has remained for decades for the case of developing a method in identifying possible diversion of nuclear fuel pins from the Light Water Reactor (LWR) spent fuel assemblies. We had embarked on this challenging task and successfully developed a novel methodology in detecting partial removal of fuel from pressurized water reactor spent fuel assemblies. The methodology uses multiple tiny neutron and gamma detectors in the form of a cluster and a high precision driving system to obtain underwater radiation measurements inside a Pressurized Water Reactor (PWR) spent fuel assembly without any movement of the fuel. The data obtained in such a manner can provide spatial distribution of neutron and gamma flux within a spent fuel assembly. The combined information of gamma and neutron signature is used to produce base signatures and they are principally dependent on the geometry of the detector locations, and exhibit little sensitivity to initial enrichment, burn-up or cooling time. A small variation in the fuel bundle such as a few missing pins changes the shape of the signature to enable detection. This resulted in a breakthrough method which can be used to detect pin diversion without relying on the nuclear power plant operator's declared operation data. Presented are the results of various Monte Carlo simulation studies and experiments from actual commercial PWR spent fuel assemblies.

  5. A CFD M&S PROCESS FOR FAST REACTOR FUEL ASSEMBLIES

    SciTech Connect

    Kurt D. Hamman; Ray A. Berry

    2008-09-01

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-e and SST (Menter) k-? were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  6. Structural analysis of fuel assembly clads for the Upgraded Transient Reactor Test Facility (TREAT Upgrade)

    SciTech Connect

    Ewing, T.F.; Wu, T.S.

    1986-01-01

    The Upgraded Transient Reactor Test Facility (TREAT Upgrade) is designed to test full-length, pre-irradiated fuel pins of the type used in large LMFBRs under accident conditions, such as severe transient overpower and loss-of-coolant accidents. In TREAT Upgrade, the central core region is to contain new fuel assemblies of higher fissile loadings to maximize the energy deposition to the test fuel. These fuel assemblies must withstand normal peak clad temperatures of 850/sup 0/C for hundreds of test transients. Due to high temperatures and gradients predicted in the clad, creep and plastic strain effects are significant, and the clad structural behavior cannot be analyzed by conventional linear techniques. Instead, the detailed elastic-plastic-creep behavior must be followed along the time-dependent load history. This paper presents details of the structural evaluations of the conceptual TREAT Upgrade fuel assembly clads.

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

    SciTech Connect

    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.

  8. Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

    SciTech Connect

    Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh; Ba Vien Luong; Kien Cuong Nguyen

    2008-07-15

    The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configuration with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)

  9. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 4-Three Mile Island Unit 1 Cycle 5

    SciTech Connect

    DeHart, M.D.

    1995-01-01

    The requirements of ANSI/ANS-8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original ''fresh'' composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using relevant and well-documented critical configurations from commercial pressurized water reactors. The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Isotopic densities for spent fuel assemblies in the core were calculated using the SCALE-4 SAS2H analytical sequence. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code family was used to extract the necessary isotopic densities from SAS2H results and to provide the data in the format required for SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) for the critical configuration. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all calculations. This volume of the report documents a reactor critical calculation for GPU Nuclear Corporation's Three Mile Island Unit 1 (TMI-1) during hot, zero-power startup testing for the beginning of cycle 5. This unit and cycle were selected because of their relevance in spent fuel benchmark applications: (1) cycle 5 startup occurred after an especially long downtime of 6.6 years; and (2) the core consisted primarily (75%) of burned fuel, with

  10. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  11. Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio

    SciTech Connect

    Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik; Suzuki, Mitsutoshi

    2014-09-30

    Criticality condition of the reactors is one of the important factors for evaluating reactor operation and nuclear fuel breeding ratio is another factor to show nuclear fuel sustainability. This study analyzes the effect of burnup steps and cycle operation step for evaluating the criticality condition of the reactor as well as the performance of nuclear fuel breeding or breeding ratio (BR). Burnup step is performed based on a day step analysis which is varied from 10 days up to 800 days and for cycle operation from 1 cycle up to 8 cycles reactor operations. In addition, calculation efficiency based on the variation of computer processors to run the analysis in term of time (time efficiency in the calculation) have been also investigated. Optimization method for reactor design analysis which is used a large fast breeder reactor type as a reference case was performed by adopting an established reactor design code of JOINT-FR. The results show a criticality condition becomes higher for smaller burnup step (day) and for breeding ratio becomes less for smaller burnup step (day). Some nuclides contribute to make better criticality when smaller burnup step due to individul nuclide half-live. Calculation time for different burnup step shows a correlation with the time consuming requirement for more details step calculation, although the consuming time is not directly equivalent with the how many time the burnup time step is divided.

  12. Beta- and gamma-dose measurements of the Godiva IV critical assembly.

    PubMed

    Hankins, D E

    1984-03-01

    To aid in the re-evaluation of an exposure that occurred in 1963, information was required on the response of film badges to the beta- and gamma-ray doses from a critical assembly. Of particular interest was the beta spectra from the assembly. The techniques used and the results obtained in this study are of interest to health physicists at facilities where exposures to betas occur. The dose rates from the Los Alamos National Laboratory Godiva IV Critical Assembly were measured at numerous distances from the assembly four and 12 days following a burst. Information was obtained on the beta-particle spectra using absorption curve studies. The beta/gamma dose-rate ratio as a function of distance from the assembly was determined. Shielding provided by various metals, gloves and clothing was measured. The beta- and gamma-ray doses measured were compared with a film packet used in the past at the Nevada Test Site with two types of current TLD personnel badges. Measurements made with a commercial thin-window ion chamber instrument are compared with the dose rates obtained using other dosimeters. PMID:6698784

  13. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    SciTech Connect

    Not Available

    1993-11-01

    This document contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE non-reactor nuclear facilities. Adherence to these guidelines will provide consistency and uniformity in criticality safety evaluations (CSEs) across the complex and will document compliance with the requirements of DOE Order 5480.24.

  14. Requalification of SPERT (Special Power Excursion Reactor Test) pins for use in university reactors

    SciTech Connect

    Snelgrove, J.L.; Domagala, R.F.; Dates, L.R.

    1986-12-01

    A series of nondestructive and destructive examinations have been performed on a representative sample of stainless steel-clad UO/sub 2/ fuel pins procured in the early-to-mid 1960s for the SPERT program. These examinations were undertaken in order to requalify the SPERT pins for use in converting university research reactors from the use of highly enriched uranium to the use of low-enriched uranium. The requalification program included visual and dimensional inspections of fuel pins and fuel pellets, radiographic inspections of welds, fill gas analyses, and chemical and spectrographic analyses of fuel and cladding materials. In general all attributes tested were within or very close to specified values, although some weld defects not covered by the original specifications were found. 1 ref., 4 figs., 11 tabs.

  15. DOE Lab-to-Lab MPC&A workshop for cooperative tasks with Russian institutes: Focus on critical assemblies and item facilities

    SciTech Connect

    Bieber, A.M. Jr.; Fishbone, L.G.; Kato, W.Y.; Lazareth, O.W.; Suda, S.C.; Garcia, D.; Haga, R.

    1995-12-01

    Seventeen Russian scientists and engineers representing five different institutes participated in a Workshop on material control and accounting as part of the US-Russian Lab-to-Lab Cooperative Program in Nuclear Materials Protection, Control, and Accounting (MPC&A). In addition to presentations and discussions, the Workshop included an exercise at Brookhaven National Laboratory (BNL) and demonstrations at the Zero Power Physics Reactor (critical-assembly facility) of Argonne National Laboratory-West (ANL-W). The Workshop particularly emphasized procedures for physical inventory-taking at critical assemblies and item facilities, with associated supporting techniques and methods. By learning these topics and applying the methods and experience at their own institutes, the Russian scientists and engineers will be able to determine and verify nuclear material inventories based on sound procedures, including measurements. This will constitute a significant enhancement to MPC&A at the Russian institutes.

  16. Forced-to-natural convection transition tests in parallel simulated liquid metal reactor fuel assemblies

    SciTech Connect

    Levin, A.E. ); Montgomery, B.H. )

    1990-01-01

    The Thermal-Hydraulic Out of Reactor Safety (THORS) Program at Oak Ridge National Laboratory (ORNL) had as its objective the testing of simulated, electrically heated liquid metal reactor (LMR) fuel assemblies in an engineering-scale, sodium loop. Between 1971 and 1985, the THORS Program operated 11 simulated fuel bundles in conditions covering a wide range of normal and off-normal conditions. The last test series in the Program, THORS-SHRS Assembly 1, employed two parallel, 19-pin, full-length, simulated fuel assemblies of a design consistent with the large LMR (Large Scale Prototype Breeder -- LSPB) under development at that time. These bundles were installed in the THORS Facility, allowing single- and parallel-bundle testing in thermal-hydraulic conditions up to and including sodium boiling and dryout. As the name SHRS (Shutdown Heat Removal System) implies, a major objective of the program was testing under conditions expected during low-power reactor operation, including low-flow forced convection, natural convection, and forced-to-natural convection transition at various powers. The THORS-SHRS Assembly 1 experimental program was divided up into four phases. Phase 1 included preliminary and shakedown tests, including the collection of baseline steady-state thermal-hydraulic data. Phase 2 comprised natural convection testing. Forced convection testing was conducted in Phase 3. The final phase of testing included forced-to-natural convection transition tests. Phases 1, 2, and 3 have been discussed in previous papers. The fourth phase is described in this paper. 3 refs., 2 figs.

  17. Modular assembly for supporting, straining, and directing flow to a core in a nuclear reactor

    DOEpatents

    Pennell, William E.

    1977-01-01

    A reactor core support arrangement for supporting, straining, and providing fluid flow to the core and periphery of a nuclear reactor during normal operation. A plurality of removable inlet modular units are contained within permanent liners in the lower supporting plate of the reactor vessel lower internals. During normal operation (1) each inlet modular unit directs main coolant flow to a plurality of core assemblies, the latter being removably supported in receptacles in the upper portion of the modular unit and (2) each inlet modular unit may direct bypass flow to a low pressure annular region of the reactor vessel. Each inlet modular unit may include special fluid seals interposed between mating surfaces of the inlet modular units and the core assemblies and between the inlet modular units and the liners, to minimize leakage and achieve an hydraulic balance. Utilizing the hydraulic balance, the modular units are held in the liners and the assemblies are held in the modular unit receptacles by their own respective weight. Included as part of the permanent liners below the horizontal support plate are generally hexagonal axial debris barriers. The axial debris barriers collectively form a bottom boundary of a secondary high pressure plenum, the upper boundary of which is the bottom surface of the horizontal support plate. Peripheral liners include radial debris barriers which collectively form a barrier against debris entry radially. During normal operation primary coolant inlet openings in the liner, below the axial debris barriers, pass a large amount of coolant into the inlet modular units, and secondary coolant inlet openings in the portion of the liners within the secondary plenum pass a small amount of coolant into the inlet modular units. The secondary coolant inlet openings also provide alternative coolant inlet flow paths in the unlikely event of blockage of the primary inlet openings. The primary inlet openings have characteristics which limit the

  18. Independent Confirmatory Survey Report for the University of Arizona Nuclear Reactor Laboratory, Tucson, Arizona

    SciTech Connect

    Nick A. Altic

    2011-11-11

    The University of Arizona (University) research reactor is a TRIGA swimming pool type reactor designed by General Atomics and constructed at the University in 1958. The reactor first went into operation in December of 1958 under U.S. Nuclear Regulatory Commission (NRC) license R-52 until final shut down on May 18, 2010. Initial site characterization activities were conducted in February 2009 during ongoing reactor operations to assess the radiological status of the Nuclear Reactor Laboratory (NRL) excluding the reactor tank, associated components, and operating systems. Additional post-shutdown characterization activities were performed to complete characterization activities as well as verify assumptions made in the Decommissioning Plan (DP) that were based on a separate activation analysis (ESI 2009 and WMG 2009). Final status survey (FSS) activities began shortly after the issuance of the FSS plan in May 2011. The contractor completed measurement and sampling activities during the week of August 29, 2011.

  19. Application of S{sub N} and Monte Carlo codes to the SHEBA critical assemblies

    SciTech Connect

    O`Dell, R.D.

    1993-07-01

    The Solution High-Energy Burst Assembly (SHEBA) at Los Alamos is a low-enriched (4.95 wt. %) aqueous uranyl fluoride solution critical assembly. There are two SHEBA configurations, both consisting of right circular cylinders with a central control rod. The first configuration, hereafter called the old SHEBA, had a fuel solution diameter of 54.6 cm and a measured critical solution height of 36.5 cm. An improved modification, hereafter called the new SHEBA, has a fuel solution diameter of 48.9 cm but since it is not yet operational, the critical solution height has not yet been measured. In this presentation the application of the discrete-ordinates (S{sub N}) code TWODANT using Hansen-Roach cross sections and the MCNP Monte Carlo code using continuous-energy cross sections for calculating the critical solution heights for both the old and new SHEBA assemblies is described. The code`s predictions are compared and it is shown that a single calculation with a standard computer code may yield misleading results, especially when using a Monte Carlo code.

  20. 76 FR 69296 - University of Utah, University of Utah TRIGA Nuclear Reactor, Notice of Issuance of Renewed...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-11-08

    ... published in the Federal Register on July 21, 2011 (76 FR 43733-43737). The NRC received no request for a..., 2011 (76 FR 60091-60094), and concluded that renewal of the facility operating license will not have a... COMMISSION University of Utah, University of Utah TRIGA Nuclear Reactor, Notice of Issuance of...

  1. Estimation of critical flow velocity for collapse of booster fuel assembly

    SciTech Connect

    Donna Guillen; Mark J. Russell

    2005-09-01

    A Gas Test Loop (GTL) system is currently being designed to provide a high intensity fast-flux irradiation environment for testing fuels and materials for advanced concept nuclear reactors. To assess the performance of candidate reactor fuels, these fuels must be irradiated under actual fast reactor flux conditions and operating environments, preferably in an existing irradiation facility. The GTL system is being designed for operation in the northwest test lobe of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The Technical and Functional Requirements (T&FRs) for the GTL stipulate a minimum neutron flux intensity (10{sup 15} n/cm{sup 2} {center_dot} s) and fast to thermal neutron ratio (>15) for the test environment. The incorporation of booster fuel within the test lobe is necessary to achieve these neutron flux requirements. The current design of the booster fuel assembly for the GTL calls for 3 concentric rings of 4 ft long uranium silicide fuel plates clad with 6061 aluminum.

  2. Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

    SciTech Connect

    S. Bragg-Sitton; J. Bess; J. Werner; G. Harms; S. Bailey

    2011-09-01

    Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al., 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).

  3. EXPERIMENT OPERATIONS PLAN FOR A LOSS-OF-COOLANT ACCIDENT SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    SciTech Connect

    Russcher, G. E.; Cannon, L. W.; Goodman, R. L.; Hesson, G. M.; King, L. L.; McDuffie, P. N.; Marshall, R. K.; Nealley, C.; Pilger, J. P.; Mohr, C. L.

    1981-04-01

    Pressurized water reactor loss-of-coolant accident phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship between the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. This document contains both experiment proposal and assembly proposal information. The intent of this document is to supply information required by the Chalk River Nuclear Laboratories (CRNL), and to identify the planned procedures and data that will be used both to establish readiness to proceed from one test phase to the next and to operate the experiment. Operating control settings and limits are provided for both experimenter systems and CRNL systems. A hazards review summarizes safety issues that have been addressed during the development of the experiment plan.

  4. The Problematic Potential of Universities to Advance Critical Urban Politics

    ERIC Educational Resources Information Center

    Pendras, Mark; Dierwechter, Yonn

    2012-01-01

    Recent research has explored the connections between universities and the cities/places in which they are located. Increasingly, emphasis is placed on the economic role of the university and on universities as urban stabilizers that can mobilize investment and advance development goals. This article explores a different charge for the university:…

  5. Neutronics assessment of stringer fuel assembly designs for the liquid-salt-cooled very high temperature reactor (LS-VHTR).

    SciTech Connect

    Szakaly, F. J.; Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2007-01-01

    Neutronic studies of 18-pin and 36-pin stringer fuel assemblies have been performed to ascertain that core design requirements for the Liquid-Salt Cooled Very High Temperature Reactor (LS-VHTR) can be met. Parametric studies were performed to determine core characteristics required to achieve a target core cycle length of 18 months and fuel discharge burnup greater than 100 GWd/t under the constraint that the uranium enrichment be less than 20% in order to support non-proliferation goals. The studies were done using the WIMS9 lattice code and the linear reactivity model to estimate the core reactivity balance, fuel composition, and discharge burnup. The results show that the design goals can be met using a 1-batch fuel management scheme, uranium enrichment of 15% and a fuel packing fraction of 30% or greater for the 36-pin stringer fuel assembly design.Evaluations of a liquid-salt- (molten-salt-) cooled version of the prismatic-block type VHTR, the LS-VHTR, are ongoing at U.S. national laboratories, universities, and industry. These evaluations have included core and passive safety studies and balance of plant conceptual designs.

  6. Faculty Perceptions of Critical Thinking at a Health Sciences University

    ERIC Educational Resources Information Center

    Rowles, Joie; Morgan, Christine; Burns, Shari; Merchant, Christine

    2013-01-01

    The fostering of critical thinking skills has become an expectation of faculty, especially those teaching in the health sciences. The manner in which critical thinking is defined by faculty impacts how they will address the challenge to promote critical thinking among their students. This study reports the perceptions of critical thinking held by…

  7. Analysis of Critical Reactor Response for TOPAZ-II Water Immersion Scenarios

    NASA Astrophysics Data System (ADS)

    Ponomarev-Stepnoi, Nikolai N.; Glushkov, Yevgeny S.; Yermoshin, Mikhail Y.; Skorlygin, Vladimir V.

    1994-07-01

    The unmodified TOPAZ-II water immersion event leading to surrounding the reactor with water and filling with water all internal core cavities (including TFE NaK channels) may hypothetically result in criticality. This paper presents results of preliminary studies of such an accident. Possible scenarios have been analyzed as well as reactivity effects involving the water presence in internal core cavities. A preliminary coupled model has been developed to describe accident transients in the reactor and TFE. The model is based on assumptions that result in overestimating possible consequences. The numerical simulations results point at the TOPAZ-II reactor capability to quench effectively possible power bursts and predict stable periodic oscillations as a final system state, wherein steaming and then refilling up some internal core cavities occurs. That may be considered to be demonstration of the TOPAZ-II reactor self-control capability if its criticality involves water immersion event.

  8. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    SciTech Connect

    Spelt, P.F.

    1992-01-01

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs.

  9. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    SciTech Connect

    Spelt, P.F.

    1992-08-01

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs.

  10. An improved resonance self-shielding method for heterogeneous fast reactor assembly and core calculations

    SciTech Connect

    Lee, C.; Yang, W. S.

    2013-07-01

    An improved resonance self-shielding method has been developed to accurately estimate the effective multigroup cross sections for heterogeneous fast reactor assembly and core calculations. In the method, the heterogeneity effect is considered by the use of isotopic escape cross sections while the resonance interference effect is accounted for through the narrow resonance approximation or slowing-down calculations for specific compositions. The isotopic escape cross sections are calculated by solving fixed-source transport equations with the method of characteristics for the whole problem domain. This method requires no pre-calculated resonance integral tables or parameters that are typically necessary in the subgroup method. Preliminary results for multi pin-cell fast reactor problems show that the escape cross sections estimated from the explicit-geometry fixed source calculations produce more accurate eigenvalue and self-shielded effective cross sections than those from conventional one-dimensional geometry models. (authors)

  11. CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

    SciTech Connect

    Pytel, K.; Mieleszczenko, W.; Lechniak, J.; Moldysz, A.; Andrzejewski, K.; Kulikowska, T.; Marcinkowska, A.; Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division; Institute of Atomic Energy

    2010-03-01

    The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

  12. Steady state temperature profiles in two simulated liquid metal reactor fuel assemblies with identical design specifications

    SciTech Connect

    Levin, A.E.; Carbajo, J.J.; Lloyd, D.B.; Montgomery, B.H.; Rose, S.D.; Wantland, J.L.

    1985-01-01

    Temperature data from steady state tests in two parallel, simulated liquid metal reactor fuel assemblies with identical design specifications have been compared to determine the extent to which they agree. In general, good agreement was found in data at low flows and in bundle-center data at higher flows. Discrepancies in the data wre noted near the bundle edges at higher flows. An analysis of bundle thermal boundary conditions showed that the possible eccentric placement of one bundle within the housing could account for these discrepancies.

  13. Analysis of a boron-carbide-drum-controlled critical reactor experiment

    NASA Technical Reports Server (NTRS)

    Mayo, W. T.

    1972-01-01

    In order to validate methods and cross sections used in the neutronic design of compact fast-spectrum reactors for generating electric power in space, an analysis of a boron-carbide-drum-controlled critical reactor was made. For this reactor the transport analysis gave generally satisfactory results. The calculated multiplication factor for the most detailed calculation was only 0.7-percent Delta k too high. Calculated reactivity worth of the control drums was $11.61 compared to measurements of $11.58 by the inverse kinetics methods and $11.98 by the inverse counting method. Calculated radial and axial power distributions were in good agreement with experiment.

  14. Conversion and standardization of US university reactor fuels using LEU, status 1989

    SciTech Connect

    Brown, K.R.; Matos, J.E.; Argonne National Lab., IL )

    1989-01-01

    In 1986, the US Department of Energy initiated a program to change the fuel used in most of the US university research reactors using HEU (93%) to LEU({lt}20{percent}) in order to minimize the risk of theft or diversion of this weapons-useable material. An important consideration in the LEU conversion planning process has been the desire to standardize the fuels that are used and to enhance the performance and utilization of the reactors. This paper describes the current status of this conversion process and the plans and schedules to complete an orderly transition from HEU to LEU fuel in most of these reactors. To date, three university reactors have been converted to LEU fuel, completed safety documentation for three reactors is being evaluated by the USNRC, and work on the safety documentation for six reactors is in progress. 13 refs., 9 tabs.

  15. Characterization of the CALIBAN Critical Assembly Neutron Spectra using Several Adjustment Methods Based on Activation Foils Measurement

    NASA Astrophysics Data System (ADS)

    Casoli, Pierre; Grégoire, Gilles; Rousseau, Guillaume; Jacquet, Xavier; Authier, Nicolas

    2016-02-01

    CALIBAN is a metallic critical assembly managed by the Criticality, Neutron Science and Measurement Department located on the French CEA Center of Valduc. The reactor is extensively used for benchmark experiments dedicated to the evaluation of nuclear data, for electronic hardening or to study the effect of the neutrons on various materials. Therefore CALIBAN irradiation characteristics and especially its central cavity neutron spectrum have to be very accurately evaluated. In order to strengthen our knowledge of this spectrum, several adjustment methods based on activation foils measurements are being studied for a few years in the laboratory. Firstly two codes included in the UMG package have been tested and compared: MAXED and GRAVEL. More recently, the CALIBAN cavity spectrum has been studied using CALMAR, a new adjustment tool currently under development at the CEA Center of Cadarache. The article will discuss and compare the results and the quality of spectrum rebuilding obtained with the UMG codes and with the CALMAR software, from a set of activation measurements carried out in the CALIBAN irradiation cavity.

  16. The first critical experiment with a LEU Russian fuel IRT-4M at the training reactor VR-1

    SciTech Connect

    Frybort, Jan

    2008-07-15

    A critical experiment is a standard part of training of students at the Training Reactor VR-1 operated within the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague. In autumn 2005 the HEU fuel IRT-3M with enrichment 36 % {sup 235}U was replaced by the LEU fuel IRT-4M with enrichment 19.7 % {sup 235}U. The fuel replacement at the VR-1 Reactor is a part of an international program RERTR. This Paper presents basic information about preparation for the fuel replacement and approaching of the first critical state with the new zone configuration C1 which replaced B1 core with the old IRT-3M fuel. The whole process was carried out according to the Czech law and the relevant international recommendations. The experience with the VR-1 operation confirms the assumption that the C1 core configuration will be suitable from the point of view of the reactivity balance for the long term safe operation of the Training Reactor VR-1. (author)

  17. Final report. U.S. Department of Energy University Reactor Sharing Program

    SciTech Connect

    Bernard, John A

    2003-01-21

    Activities supported at the MIT Nuclear Reactor Laboratory under the U.S. DOE University Reactor Sharing Program are reported for Grant DE FG02-95NE38121 (September 16, 1995 through May 31, 2002). These activities fell under four subcategories: support for research at thesis and post-doctoral levels, support for college-level laboratory exercises, support for reactor tours/lectures on nuclear energy, and support for science fair participants.

  18. As Universities Close Their Reactors, Energy Dept. Considers a Policy Shift.

    ERIC Educational Resources Information Center

    Southwick, Ron

    2001-01-01

    Discusses how, as many universities shut down their nuclear reactors used for research and training, the Energy Department considers moving its support to regional facilities, a change that might lead to more shutdowns. (EV)

  19. Mixed enrichment core design for the NC State University PULSTAR Reactor

    SciTech Connect

    Mayo, C.W.; Verghese, K.; Huo, Y.G.

    1997-12-01

    The North Carolina State University PULSTAR Reactor license was renewed for an additional 20 years of operation on April 30, 1997. The relicensing period added additional years to the facility operating time through the end of the second license period, increasing the excess reactivity needs as projected in 1988. In 1995, the Nuclear Reactor Program developed a strategic plan that addressed the future maintenance, development, and utilization of the facility. Goals resulting from this plan included increased academic utilization of the facility in accordance with its role as a university research facility, and increased industrial service use in accordance with the mission of a land grant university. The strategic plan was accepted, and it is the intent of the College of Engineering to operate the PULSTAR Reactor as a going concern through at least the end of the current license period. In order to reach the next relicensing review without prejudice due to low excess reactivity, it is desired to maintain sufficient excess reactivity so that, if relicensed again, the facility could continue to operate without affecting users until new fuel assistance was provided. During the NC State University license renewal, the operation of the PULSTAR Reactor at the State University of New York at Buffalo (SUNY Buffalo) was terminated. At that time, the SUNY Buffalo facility had about 240 unused PULSTAR Reactor fuel pins with 6% enrichment. The objective of the work reported here was to develop a mixed enrichment core design for the NC State University PULSTAR reactor which would: (1) demonstrate that 6% enriched SUNY buffalo fuel could be used in the NC State University PULSTAR Reactor within the existing technical specification safety limits for core physics parameters; (2) show that use of this fuel could permit operating the NC State University PULSTAR Reactor to 2017 with increased utilization; and (3) assure that the decision whether or not to relicense the facility would

  20. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  1. A Re-Analysis of Historical Los Alamos Critical Assembly Reaction Rate Measurements

    NASA Astrophysics Data System (ADS)

    Kahler, A. C.; MacInnes, M.; Chadwick, M. B.

    2016-02-01

    Starting in the 1950s and continuing into the early 1970s, a number of foil irradiations and fission chamber measurements were made in a variety of Fast critical assemblies at Los Alamos National Laboratory. These include (i) Godiva, a bare HEU spherical assembly; (ii) Flattop-25, a spherical assembly consisting of an HEU core and a natural uranium reflector; (iii) Jezebel, a bare 239Pu assembly; and (iv) Flattop-Pu, a spherical assembly consisting of a 239Pu core and a natural uranium reflector. In most instances the irradiations occur at or near the center of the assembly, but in selected instances data were obtained for a radial traverse extending into the Flattop reflector region. Measurements were made for a number of threshold reactions, including 45Sc(n,2n)44mSc, 51V(n,α)48Sc, 75As(n,2n)74As, 89Y(n,2n)88Y, 90Zr(n,2n)89Zr, 103Rh(n,2n)102gRh, 107Ag(n,2n)106mAg, 169Tm(n,2n)168Tm, 175Lu(n,2n)174Lu, 191Ir(n,2n)190Ir, 197Au(n,2n)196Au, 203Tl(n,2n)202Tl, 204Pb(n,2n)203Pb and 238U(n,2n)237U. Fission ratio data for 238U(n,f)/235U(n,f) and 239Pu(n,f)/235U(n,f) were also obtained. We report C/E values from MCNP6 calculations using ENDF/B-VII.1 and IRDFF-v1.03 cross section data.

  2. Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Tests 1 and 2

    SciTech Connect

    Russcher, G. E.; Wilson, C. L.; Marshall, R, K.; King, L. L.; Parchen, L. J.; Pilger, J. P.; Hesson, G. M.; Mohr, C. L.

    1981-09-01

    A loss of Coolant Accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects of LOCA conditions on pressurized water reactor test fuel bundles. This experiment operation plan for the second and third experiments of the program will provide peak fuel cladding temperatures of up to 1172K (1650{degree}F) and 1061K (1450{degree}) respectively. for a long enough time to cause test fuel cladding deformation and rupture in both. Reflood coolant delay times and the reflooding rates for the experiments were selected from thermal-hydraulic data measured in the National Research Universal (NRU) reactor facilities and test train assembly during the first experiment.

  3. Critical assembly: A technical history of Los Alamos during the Oppenheimer years, 1943--1945

    SciTech Connect

    Hoddeson, L.; Henriksen, P.W.; Meade, R.A.; Westfall, C.

    1993-11-01

    This volume treats the technical research that led to the first atomic bombs. The authors explore how the ``critical assembly`` of scientists, engineers, and military Personnel at Los Alamos collaborated during World War II, blending their traditions to create a new approach to large-scale research. The research was characterized by strong mission orientation, multidisciplinary teamwork, expansion of the scientists` traditional methodology with engineering techniques, and a trail-and-error methodology responding to wartime deadlines. The book opens with an introduction laying out major themes. After a synopsis of the prehistory of the bomb project, from the discovery of nuclear fission to the start of the Manhattan Engineer District, and an overview of the early materials program, the book examines the establishment of the Los Alamos Laboratory, the implosion and gun assembly programs, nuclear physics research, chemistry and metallurgy, explosives, uranium and plutonium development, confirmation of spontaneous fission in pile-produced plutonium, the thermonuclear bomb, critical assemblies, the Trinity test, and delivery of the combat weapons.

  4. Graphite and Beryllium Reflector Critical Assemblies of UO2 (Benchmark Experiments 2 and 3)

    SciTech Connect

    Margaret A. Marshall; John D. Bess

    2012-11-01

    INTRODUCTION A series of experiments was carried out in 1962-65 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for use in space reactor research programs. A core containing 93.2 wt% enriched UO2 fuel rods was used in these experiments. The first part of the experimental series consisted of 252 tightly-packed fuel rods (1.27-cm triangular pitch) with graphite reflectors [1], the second part used 252 graphite-reflected fuel rods organized in a 1.506-cm triangular-pitch array [2], and the final part of the experimental series consisted of 253 beryllium-reflected fuel rods in a 1.506-cm-triangular-pitch configuration and in a 7-tube-cluster configuration [3]. Fission rate distribution and cadmium ratio measurements were taken for all three parts of the experimental series. Reactivity coefficient measurements were taken for various materials placed in the beryllium reflected core. All three experiments in the series have been evaluated for inclusion in the International Reactor Physics Experiment Evaluation Project (IRPhEP) [4] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbooks, [5]. The evaluation of the first experiment in the series was discussed at the 2011 ANS Winter meeting [6]. The evaluations of the second and third experiments are discussed below. These experiments are of interest as benchmarks because they support the validation of compact reactor designs with similar characteristics to the design parameters for a space nuclear fission surface power systems [7].

  5. Two distinct domains of protein 4.1 critical for assembly offunctional nuclei in Vitro

    SciTech Connect

    Krauss, Sharon Wald; Heald, Rebecca; Lee, Gloria; Nunomura, Wataru; Gimm,J. Aura; Mohandas, Narla; Chasis, Joel AnneJ. Aura; Mohandas, Narla; Chasis, Joel Anne

    2002-11-15

    Protein 4.1R, a multifunctional structural protein, acts asan adaptor in mature red cell membrane skeletons linking spectrin-actincomplexes to plasma membrane-associated proteins. In nucleated cellsprotein 4.1 is not associated exclusively with plasma membrane but isalso detected at several important subcellular locations crucial for celldivision. To identify 4.1 domains having critical functions in nuclearassembly, 4.1 domain peptides were added to Xenopus egg extract nuclearreconstitution reactions. Morphologically disorganized, replicationdeficient nuclei assembled when spectrin-actin binding domain orNuMA-binding C-terminal domain peptides were present. However, controlvariant spectrin-actin binding domain peptides incapable of bindingactin, or mutant C-terminal domain peptides with reduced NuMA binding,had no deleterious effects on nuclear reconstitution. To test if 4.1 isrequired for proper nuclear assembly, 4.1 isoforms were depleted withspectrin-actin binding or C-terminal domain-specific antibodies. Nucleiassembled in depleted extracts ha d deranged phenotypes. However, nuclearassembly could be rescued by addition of recombinant 4.1R. Our dataestablishes that protein 4.1 is essential for nuclear assembly andidentifies two distinct 4.1 domains, initially characterized incytoskeletal interactions, that have crucial and versatile functions innuclear assembly.

  6. CRITICAL CONFIGURATION FOR BERYLLIUM REFLECTED ASSEMBLIES OF U(93.15)O2 FUEL RODS (1.506-CM PITCH AND 7-TUBE CLUSTERS)

    SciTech Connect

    Margaret A. Marshall

    2012-05-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950’s efforts were made to study “power plants for the production of electrical power in space vehicles”. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of 253 unmoderated stainless steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.” The first two experiments in the series were evaluated in HEU-COMP-FAST-001 (SCCA-FUND-EXP-001) and HEU-COMP-FAST-002 (SCCA-FUND-EXP-002). The first experiment had the 253 fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank (References 1 and 2). The second experiment in the series, performed in early 1963, had the 253 fuel tubes at a 1.506-cm triangular lattice in a 25.96 cm OD core tank and graphite reflectors on all sides. The third set of experiments in the series, performed in mid-1963, which is studied in this evaluation, used beryllium reflectors. The beryllium reflected system was the preferred reactor configuration for this application because of the small thickness of the reflector. The two core configurations had the 253 fuel tubes

  7. Final report for U.S. Department of Energy Grant DE-FG02-95NE38118-5 University Reactor Sharing Program [Purdue University

    SciTech Connect

    Bean, R.S.

    2001-06-01

    Under the Reactor Sharing Program, a total of 350 high school students participated in the neutron activation experiment and 484 high school and university students and members of the general public participated in reactor tours.

  8. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    SciTech Connect

    1998-09-01

    This Department of Energy (DOE) is approved for use by all components of DOE. It contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE Non-Reactor Nuclear Facilities. Adherence with these guidelines will provide consistency and uniformity in Criticality Safety Evaluations (CSEs) across the complex and will document compliance with DOE Order 5480.24 requirements as they pertain to CSEs.

  9. A critical assembly designed to measure neutronic benchmarks in support of the Space Nuclear Thermal Propulsion program

    NASA Astrophysics Data System (ADS)

    Parma, E. J.; Ball, R. M.; Hoovler, G. S.; Selcow, E. C.; Cerbone, R. J.

    1992-10-01

    A reactor designed to perform criticality experiments in support of the Space Nuclear Thermal Propulsion program is currently in operation at the Sandia National Laboratories' reactor facility. The reactor is a small, water-moderated system that uses highly enriched uranium particle fuel in a 19-element configuration. Its purpose is to obtain neutronic measurements under a variety of experimental conditions that are subsequently used to benchmark reactor-design computer codes. Brookhaven National Laboratory, Babcock & Wilcox, and Sandia National Laboratories participated in determining the reactor's performance requirements, design, follow on experimentation, and in obtaining the licensing approvals. Brookhaven National Laboratory is primarily responsible for the analytical support, Babcock & Wilcox the hardware design, and Sandia National Laboratories the operational safety. All of the team members participate in determining the experimentation requirements, performance, and data reduction. Initial criticality was achieved in October 1989. An over-all description of the reactor is presented along with key design features and safety-related aspects.

  10. Characteristics of Neutron Fields for Radiation Protection and Other Applications at the Kinki University Reactor

    NASA Astrophysics Data System (ADS)

    Ogawa, Yoshihiro; Fujiwara, Tatsuya; Morishima, Hiroshige; Urabe, Itsumasa; Sagawa, Hiroyuki

    2003-06-01

    In order to get useful information about neutron energy spectrum and neutron dose, a versatile and accurate reactor model of the Kinki University Reactor (UTR-KINKI) was developed under the three-dimensional continuous-energy MCNP Monte Carlo code. The agreement between MCNP predictions and the experimentally determined values was very good. This paper describes characteristics of neutron fields at the Kinki University Reactor calculated with the present MCNP model of the UTR-KINKI. From the results obtained it was clear that these neutron fields are applicable to development and performance evaluation of personnel dosimeters and experimental studies on biological effects of low levels of radiation.

  11. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    SciTech Connect

    Chodak, P. III

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO{sub 2} assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the {sup 239}Pu and {ge}90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  12. Development of reactivity feedback effect measurement techniques under sub-critical condition in fast reactors

    SciTech Connect

    Kitano, A.; Nishi, H.; Suzuki, T.; Okajima, S.; Kanemoto, S.

    2012-07-01

    The first-of-a-kind reactor has been licensed by a safety examination of the plant design based on the measured data in precedent mock-up experiments. The validity of the safety design can be confirmed without a mock-up experiment, if the reactor feed-back characteristics can be measured before operation, with the constructed reactor itself. The 'Synthesis Method', a systematic and sophisticated method of sub-criticality measurement, is proposed in this work to ensure the safety margin before operation. The 'Synthesis Method' is based on the modified source multiplication method (MSM) combined with the noise analysis method to measure the reference sub-criticality level for MSM. A numerical simulation for the control-rod reactivity worth and the isothermal feed-back reactivity was conducted for typical fast reactors of 100 MWe-size, 300 MWe-size, 750 MWe-size, and 1500 MWe-size to investigate the applicability of Synthesis Method. The number of neutron detectors and their positions necessary for the measurement were investigated for both methods of MSM and the noise analysis by a series of parametric survey calculations. As a result, it was suggested that a neutron detector located above the core center and three or more neutron detectors located above the radial blanket region enable the measurement of sub-criticality within 10% uncertainty from -$0.5 to -$2 and within 15% uncertainty for the deeper sub-criticality. (authors)

  13. Multiple lead seal assembly for a liquid-metal-cooled fast-breeder nuclear reactor

    DOEpatents

    Hutter, Ernest; Pardini, John A.

    1977-03-15

    A reusable multiple lead seal assembly provides leak-free passage of stainless-steel-clad instrument leads through the cover on the primary tank of a liquid-metal-cooled fast-breeder nuclear reactor. The seal isolates radioactive argon cover gas and sodium vapor within the primary tank from the exterior atmosphere and permits reuse of the assembly and the stainless-steel-clad instrument leads. Leads are placed in flutes in a seal body, and a seal shell is then placed around the seal body. Circumferential channels in the body and inner surface of the shell are contiguous and together form a conduit which intersects each of the flutes, placing them in communication with a port through the wall of the seal shell. Liquid silicone rubber sealant is injected into the flutes through the port and conduit; the sealant fills the space in the flutes not occupied by the leads themselves and dries to a rubbery hardness. A nut, threaded onto a portion of the seal body not covered by the seal shell, jacks the body out of the shell and shears the sealant without damage to the body, shell, or leads. The leads may then be removed from the body. The sheared sealant is cleaned from the body, leads, and shell and the assembly may then be reused with the same or different leads.

  14. Modular head assembly and method of retrofitting existing nuclear reactor facilities

    SciTech Connect

    Malandra, L.J.; Ledue, R.J.; Hankinson, M.F.; Kowalski, E.F.

    1987-07-07

    A method is described of retrofitting existing nuclear reactor facilities so as to form a modular closure head assembly for a nuclear reactor pressure vessel, where the existing nuclear reactor facilities comprise control rod drive mechanism cooling systems which include vertically extending elbow air ducts inter-connecting vertically spaced upper and lower air manifolds. The elbow air ducts extend radially beyond the peripheral envelope of the closure head, comprising the steps of: removing the upper air manifold; removing the vertically extending elbow air ducts; capping the air ports of the lower air manifold which ports were previously fluidically connecting the lower air manifold to the vertically extending elbow air ducts; disposing vertically upwardly extending air exhaust ducts above the lower air manifold in such an manner that the air exhaust ducts are disposed within the peripheral envelope of the closure head; fluidically connecting exhaust fans to the upper regions of the air exhaust ducts; fluidically connecting the lower regions of the air exhaust ducts the lower air manifold; permanently securing lift rods to the closure head at positions disposed radially outwardly of the lower air manifold; attaching a seismic support platform to the lift rods; proving fluidic passage of the vertically extending air exhaust ducts through the seismic support platform; attaching a missile shield plate to the lift rods; and proving fluidic passage of the vertically extending air exhaust ducts through the missile shield plate.

  15. Conversion and standardization of university reactor fuels using low-enrichment uranium: Plans and schedules

    SciTech Connect

    Young, H.H.; Brown, K.R.; Matos, J.E.

    1986-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. To minimize this risk, the US Nuclear Regulatory Commission issued its final rule on ''Limiting the Use of Highly Enriched Uranium in Domestically Licensed Research and Test Reactors,'' in February 1986. This paper describes the plans and schedules developed by the US Department of Energy to coordinate an orderly transition from HEU to LEU fuel in most of these reactors. An important element in the planning process has been the desire to standardize the LEU fuels used in US university reactors and to enhance the performance and utilization of a number of these reactors. The program is estimated to cost about $10 million and to last about five years.

  16. Remote Sensing and the Kyoto Protocol: A Workshop Summary

    NASA Technical Reports Server (NTRS)

    Rosenqvist, Ake; Imhoff, Marc; Milne, Anthony; Dobson, Craig

    2000-01-01

    The Kyoto Protocol to the United Nations Framework Convention on Climate Change contains quantified, legally binding commitments to limit or reduce greenhouse gas emissions to 1990 levels and allows carbon emissions to be balanced by carbon sinks represented by vegetation. The issue of using vegetation cover as an emission offset raises a debate about the adequacy of current remote sensing systems and data archives to both assess carbon stocks/sinks at 1990 levels, and monitor the current and future global status of those stocks. These concerns and the potential ratification of the Protocol among participating countries is stimulating policy debates and underscoring a need for the exchange of information between the international legal community and the remote sensing community. On October 20-22 1999, two working groups of the International Society for Photogrammetry and Remote Sensing (ISPRS) joined with the University of Michigan (Michigan, USA) to convene discussions on how remote sensing technology could contribute to the information requirements raised by implementation of, and compliance with, the Kyoto Protocol. The meeting originated as a joint effort between the Global Monitoring Working Group and the Radar Applications Working Group in Commission VII of the ISPRS, co-sponsored by the University of Michigan. Tile meeting was attended by representatives from national government agencies and international organizations and academic institutions. Some of the key themes addressed were: (1) legal aspects of transnational remote sensing in the context of the Kyoto Protocol; (2) a review of the current and future and remote sensing technologies that could be applied to the Kyoto Protocol; (3) identification of areas where additional research is needed in order to advance and align remote sensing technology with the requirements and expectations of the Protocol; and 94) the bureaucratic and research management approaches needed to align the remote sensing

  17. Universal Pragmatics: A Critical Approach to Image Ethics.

    ERIC Educational Resources Information Center

    Craig, Robert L.

    Visual criticism is a major component of the new visual communication. Visual communication has changed through the advent of new technology which allows images to be combined and manipulated with relative ease. Visual criticism analyzes the forms and practices of image production and examines the roles of images in society creating a new dialogue…

  18. Creative Writing and Critical Response in the University Literature Class

    ERIC Educational Resources Information Center

    Wilson, Peter

    2011-01-01

    Concerns about the relation between critical and creative writing are reviewed in the context of encouraging students to engage in both kinds of writing as a response to literature in undergraduate degree courses. In particular the paper seeks to illustrate and promote good practice in the integration of creative and critical written responses to…

  19. Sulfur activation at the Little Boy-Comet Critical Assembly: A replica of the Hiroshima bomb

    NASA Astrophysics Data System (ADS)

    Kerr, G. D.; Emergy, J. F.; Pace, J. V., III

    1985-04-01

    Studies have been completed on the activation of sulfur by fast neutrons from the Little Boy-Comet Critical Assembly which replicates the general features of the Hiroshima bomb. The complex effects of the bomb's design and construction of leakage of sulfur-activation neutrons were investigated both experimentally and theoretically. Our sulfur activation studies were performed as part of a larger program to provide benchmark data for testing of methods used in recent source-term calculations for the Hiroshima bomb. Source neutrons capable of activating sulfur play an important role in determining neutron doses in Hiroshima at a kilometer or more from the point of explosion.

  20. Supplementary neutron-flux calculations for the ORNL Pool Critical Assembly Pressure Vessel Facility

    SciTech Connect

    Maudlin, P.J.; Maerker, R.E.

    1982-01-01

    A three-dimensional Monte Carlo calculation using the MORSE code was performed to validate a procedure previously adopted in the ORNL discrete ordinate analysis of measurements made in the ORNL Pool Critical Assembly Pressure Vessel Facility. The results of these flux calculations agree, within statistical undertainties of about 5%, with those obtained from a discrete ordinate analysis employing the same procedure. This study therefore concludes that the procedure for combining several one- and two-dimensional discrete ordinate calculations into a three-dimensional flux is sufficiently accurate that it does not account for the existing discrepancies observed between calculations and measurements in this facility.

  1. A Critical Appraisal of RAFT-Mediated Polymerization-Induced Self-Assembly

    PubMed Central

    2016-01-01

    Recently, polymerization-induced self-assembly (PISA) has become widely recognized as a robust and efficient route to produce block copolymer nanoparticles of controlled size, morphology, and surface chemistry. Several reviews of this field have been published since 2012, but a substantial number of new papers have been published in the last three years. In this Perspective, we provide a critical appraisal of the various advantages offered by this approach, while also pointing out some of its current drawbacks. Promising future research directions as well as remaining technical challenges and unresolved problems are briefly highlighted. PMID:27019522

  2. [Dr. Michiharu Matsuoka, founder of the Department of Orthopaedic Surgery, Kyoto University, and his achievements. (Part 7: The academic carrier of Dr. Michiharu Matsuoka--from elementary school to the graduate school, Imperial University of Tokyo)].

    PubMed

    Hirotani, Hayato

    2011-12-01

    The background of the higher education of Dr. Michiharu Matsuoka shown on the official resume was disclosed by Dr. Kazuo Naito in 1986, but the courses of the elementary and secondary schools were not described in it. In regard to his lower educational courses, the author referred to the laws and regulations issued by the Ministry of Education of the Japan Government and the Yamaguchi Prefectural Office. Those were often revised with times. The author presumed the elementary school (Murozumi Primary School [the first established primary school at the birthplace; Murozumi, Hikari-City, Yamaguchi Prefecture]) and middle schools (Prefectural Yamaguchi Middle School and Yamaguchi High School) to which he had been admitted. These presumptions were made to explain his whole educational course without unreasonableness. After finishing the first school year of the Yamaguchi High School, he was transferred to the Preparatory Course of the Yamaguchi Higher School (Yamaguchi Kotô Chugakkô, Yoka), because of the amendment of the educational system. Then he was transferred to the Preparatory Course of the Daisan Higher School (Daisan Kotô Chugakkô, Yoka), and to the Preparatory Course of Daiichi Higher School (Daiichi Kotô Chugakkô, Yoka). After his graduation from the Regular Course of the Daiichi Higher School (Daiichi Kotô Chugakkô, Honka), he was admitted to the Medical College of the Imperial University from which he graduated in 1897. In addition, he was a medical student of the Graduate School of the Imperial University of Tokyo just before he left Japan for studying abroad. The whole academic carrier of Dr. Matsuoka is not only clearly clarified, but it is also indicated that he was one of the successful examples of the educational system proposed by Yamaguchi Prefecture in Meiji era which articulated the local primary and middle schools with the Imperial University of Tokyo. PMID:22586890

  3. PREFACE: Beyond Kyoto - the necessary road

    NASA Astrophysics Data System (ADS)

    Margrethe Basse, Ellen

    2009-03-01

    The Beyond Kyoto conference in Aarhus March 2009 was organised in collaboration with other knowledge institutions, businesses and authorities. It brought together leading scientists, policy-makers, authorities, intergovernmental organisations, NGO's, business stakeholders and business organisations. The conference was a joint interdisciplinary project involving many academic areas and disciplines. These conference proceedings are organised in central and recurring themes that cut across many debates on climate change, the climatic challenges as well as the solutions. In the front there is a short presentation of the conference concept. Part I of the proceedings focuses on issues related to the society - covering climate policy, law, market based instruments, financial structure, behaviour and consumption, public participation, media communication and response from indigenous peoples etc. Part II of the proceedings concerns the scientific knowledge base on climate related issues - covering climate change processes per se, the potential impacts of projected climate change on biodiversity and adaptation possibilities, the interplay between climate, agriculture and biodiversity, emissions, agricultural systems, increasing pressure on the functioning of agriculture and natural areas, vulnerability to extreme weather events and risks in respect to sea-level rise etc. The conference proceedings committee consists of four professors from Aarhus University: Jens-Christian Svenning, Jørgen E Olesen, Mads Forchhammer and Ellen Margrethe Basse. Aarhus University's Climate Secretariat has had the overall responsibility for coordinating the many presentations, as well as the practical side of arranging the conference and supporting the publication of papers. As Head of the Climate Secretariat and Chair of Aarhus University's Climate Panel, I would like to thank everyone for their contribution. This applies both to the scientific and the practical efforts. Special thanks to

  4. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-01

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  5. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    SciTech Connect

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  6. Kyoto tridimensional spectrograph II: progress

    NASA Astrophysics Data System (ADS)

    Sugai, Hajime; Ohtani, Hiroshi; Ozaki, Shinobu; Hattori, Takashi; Ishii, Motomi; Ishigaki, Tsuyoshi; Hayashi, Tadashi; Sasaki, Minoru; Takeyama, Norihide

    2000-08-01

    We are building the Kyoto tridimensional spectrograph II and are planning to mount it on Subaru telescope. The spectrograph has four observational modes: Fabry-Perot imager, integral field spectrograph (IFS) with a microlens array, long-slit spectrograph, and filter-imaging modes. The optics is designed to be used in wide wavelength range from 360 nm to 900 nm. The design well matches with high spatial resolution of Subaru: 0 inch .06 pixel-1 in Fabry- Perot mode, for which we actually will use binning before adaptive optics at optical wavelengths becomes available, and 0 inch .1 lens-1 in microlens array mode. These well sample image sizes obtained by Subaru, which are about 0 inch .4 in relatively good conditions. We have evaluated a point spread function of our cylindrical microlens array and found that it consists of a diffraction pattern and more extended component which probably comes from border regions between microlenses. With a suitable mask at the micro pupil position, the crosstalk between spectra will be limited down to a few percent. With a suitable mask at the micro pupil position, the crosstalk between spectra will be limited down to a few percent. We have succeeded in synchronizing frequency switching of Fabry-Perot etalons with the movement of charge on the CCD. This technique enables to average out all temporal variations between each passband.

  7. Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

    PubMed

    Mohammadi, A; Hassanzadeh, M; Gharib, M

    2016-02-01

    In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified. PMID:26720262

  8. The Internationalisation Agenda: A Critical Examination of Internationalisation Strategies in Public Universities in Ghana

    ERIC Educational Resources Information Center

    Gyamera, Gifty Oforiwaa

    2015-01-01

    Recently, various strategies have been adopted and adapted by universities in Ghana to re/position themselves in the international arena. Utilising postcolonial and neoliberal theories, this paper critically examines the internationalisation strategies of three public universities in Ghana. Although all the universities have adopted strategies to…

  9. Innovation for Transformation in Nigeria University Education: Implications for the Production of Critical and Creative Thinkers

    ERIC Educational Resources Information Center

    Onu, V. C.; Eskay, M. K.; Obiyo, N. O.; Igbo, J. N.; Ezeanwu, A. B.

    2012-01-01

    This descriptive survey research studied innovation for transformation in Nigeria university education: implications for the production of critical and creative thinkers. Thus, students' perception of knowledge generation and dissemination by university lecturers were elicited. From a population of registered students in a Nigerian university, 200…

  10. Analysis of Np-237 ENDF for the theortical interpretation of critical assembly experiments.

    SciTech Connect

    Mihaila, B.; Chadwick, M. B.; MacFarlane, R. E.; Kawano, T.

    2004-01-01

    We report on the present status of our effort toward an improved Np-237 evaluated nuclear data file (ENDF). The aim here is to bridge the gap between calculated and observed k-eff values, as measured at the Np-U critical assembly at LANL, TA-18. As such, we perform a critical analysis of the existing body of experimental data and recommended evaluations. We are targeting in principal the fission nu-bar and cross section in Np-237, as well as the inelastic scattering which is particularly important since Np-237 is a threshold fissioner. This analysis will be employed in a future sensitivity study of the calculated k-eff with respect to variations of the afore mentioned nuclear data.

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

    SciTech Connect

    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 ICSBEP and the IRPh

  12. CRITICAL CONFIGURATION AND PHYSICS MEASUREMENTS FOR GRAPHITE REFLECTED ASSEMBLIES OF U(93.15)O2 FUEL RODS (1.506-CM PITCH)

    SciTech Connect

    Margaret A. Marshall

    2012-03-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950’s efforts were made to study “power plants for the production of electrical power in space vehicles”. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of 253 unmoderated stainless steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.” The first experiment in the series was evaluated in HEU-COMP-FAST-001. It had the 253 fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank (References 1 and 2). The second experiment in the series, performed in early 1963, which is studied in this evaluation, had the 253 fuel tubes at a 1.506-cm triangular lattice in a 25.96 cm OD core tank and graphite reflectors on all sides. The experiment has been determined to represent an acceptable benchmark experiment. Information for this evaluation was compiled from published reports on all three parts of the experimental series (Reference 1-5) and the experimental logbook as

  13. Agreement between the University of Hawaii Professional Assembly and the Board of Regents of the University of Hawaii, 1983-1985.

    ERIC Educational Resources Information Center

    American Association of Univ. Professors, Washington, DC.

    The collective bargaining agreement between the University of Hawaii Board of Regents and the University of Hawaii Professional Assembly covering the period July 1, 1983-June 30, 1985 is presented. The American Association of University Professors affiliated union has 2,810 members. Items covered in the agreement include: unit recognition,…

  14. Monte Carlo testing of unresolved resonance treatment for fast and intermediate critical assemblies

    SciTech Connect

    Weinman, J.P.

    1998-10-01

    The purpose of this study is to investigate the eigenvalue sensitivity to changes in unresolved resonance treatment by comparing RACER Monte Carlo calculations for several fast and intermediate spectrum critical experiments. Calculations performed using smooth, dilute-average, tabulated cross sections were compared with calculations using the probability table method to produce stochastically generated resonance cross sections in the unresolved resonance region. The use of the probability table method is superior to the dilute-average cross section method for representing the unresolved resonance region because the table method properly accounts for resonance self shielding; thereby, reducing the effectiveness of the cross sections in the region. The unresolved resonance region is typically found in the intermediate and fast energy range. Eleven benchmark critical assemblies that span a range of {sup 235}U enrichments (93.8 to 10.2%) and four highly enriched {sup 239}Pu and {sup 233}U assemblies were analyzed. These benchmarks were chosen to accentuate the reactivity importance of the unresolved resonance range.

  15. Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

    SciTech Connect

    Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.

    1997-10-01

    The Solution High-Energy Burst Assembly (SHEBA) was originally constructed during 1980 and was designed to be a clean free-field geometry, right-circular, cylindrically symmetric critical assembly employing U(5%)O{sub 2}F{sub 2} solution as fuel. A second version of SHEBA, employing the same fuel but equipped with a fuel pump and shielding pit, was commissioned in 1993. This report includes data and operating experience for the 1993 SHEBA only. Solution-fueled benchmark work focused on the development of experimental measurements of the characterization of SHEBA; a summary of the results are given. A description of the system and the experimental results are given in some detail in the report. Experiments were designed to: (1) study the behavior of nuclear excursions in a low-enrichment solution, (2) evaluate accidental criticality alarm detectors for fuel-processing facilities, (3) provide radiation spectra and dose measurements to benchmark radiation transport calculations on a low-enrichment solution system similar to centrifuge enrichment plants, and (4) provide radiation fields to calibrate personnel dosimetry. 15 refs., 37 figs., 10 tabs.

  16. A Multi-Phased Sampling Effort to Characterize a University TRIGA Research Reactor

    SciTech Connect

    Taylor, K.E.; Holm, R.L.

    2006-07-01

    A radiological characterization project was conducted at the University of Illinois (University) TRIGA research nuclear reactor in July 2005 as part of the long-term facility decommissioning project. The characterization effort included multiple survey and sampling techniques designed to assess both contamination of the reactor building and equipment and activation of reactor components and the reactor bio-shield. Radiation measurements included alpha and beta surface contamination measurements, gamma dose rate measurements, and gross gamma radiation measurements. Modeling was conducted based on the field measurements to predict concentrations of activation products in reactor components that were not directly sampled. The sampling effort included collecting removable contamination swipes, concrete samples from the reactor room floor and bio-shield, soil samples from below and around the perimeter of the reactor building, graphite samples from graphite moderator, and metal samples from reactor components. Concrete samples were obtained using an innovative technology that allowed for quick sample collection and analysis. Concrete, soil, graphite, and metal samples were analyzed on-site using liquid scintillation counters and gamma spectroscopy. Additional samples were sent off-site for analysis. (authors)

  17. Critical Role of the Fusion Protein Cytoplasmic Tail Sequence in Parainfluenza Virus Assembly

    PubMed Central

    Stone, Raychel; Takimoto, Toru

    2013-01-01

    Interactions between viral glycoproteins, matrix protein and nucleocapsid sustain assembly of parainfluenza viruses at the plasma membrane. Although the protein interactions required for virion formation are considered to be highly specific, virions lacking envelope glycoprotein(s) can be produced, thus the molecular interactions driving viral assembly and production are still unclear. Sendai virus (SeV) and human parainfluenza virus type 1 (hPIV1) are highly similar in structure, however, the cytoplasmic tail sequences of the envelope glycoproteins (HN and F) are relatively less conserved. To unveil the specific role of the envelope glycoproteins in viral assembly, we created chimeric SeVs whose HN (rSeVhHN) or HN and F (rSeVh(HN+F)) were replaced with those of hPIV1. rSeVhHN grew as efficiently as wt SeV or hPIV1, suggesting that the sequence difference in HN does not have a significant impact on SeV replication and virion production. In sharp contrast, the growth of rSeVh(HN+F) was significantly impaired compared to rSeVhHN. rSeVh(HN+Fstail) which expresses a chimeric hPIV1 F with the SeV cytoplasmic tail sequence grew similar to wt SeV or rSeVhHN. Further analysis indicated that the F cytoplasmic tail plays a critical role in cell surface expression/accumulation of HN and F, as well as NP and M association at the plasma membrane. Trafficking of nucelocapsids in infected cells was not significantly affected by the origin of F, suggesting that F cytoplasmic tail is not involved in intracellular movement. These results demonstrate the role of the F cytoplasmic tail in accumulation of structural components at the plasma membrane assembly sites. PMID:23593451

  18. Nuclear reactors built, being built, or planned, 1991

    SciTech Connect

    Simpson, B.

    1992-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  19. The Potential of Critical Race Theory in Decolonizing University Curricula

    ERIC Educational Resources Information Center

    McLaughlin, Juliana; Whatman, Susan

    2011-01-01

    This paper critiques our experiences as non-Indigenous Australian educators of working with numerous embedding Indigenous perspectives curricular projects at an Australian university. Reporting on these project outcomes alone, while useful in identifying limitations, does not illustrate ways in which future embedding and decolonizing projects can…

  20. Young Children Demystifying and Remaking the University through Critical Play

    ERIC Educational Resources Information Center

    Campano, Gerald; Ngo, Lan; Low, David E.; Bartow Jacobs, Katrina

    2016-01-01

    This article, part of a four-year research partnership with a multilingual faith community and its school, explores what happened when we invited young children in an aftercare program to inquire into the university from their perspectives. Through a sociocultural literacy framework and realist theories of identity and experience, we examine the…

  1. Creativity and Critical Thinking in the Globalised University

    ERIC Educational Resources Information Center

    Clegg, Phil

    2008-01-01

    This paper outlines the dynamic life of the university in the era of neo-liberal globalisation, and within this context, discusses the nature of "creativity" as a life force or power, similar to the Ancient Greek idea of "Eros". This power is contrasted with functionalist and bureaucratic notions of creativity, and a disjuncture is identified…

  2. Resisting the English Neoliberalising University: What Critical Pedagogy Can Offer

    ERIC Educational Resources Information Center

    Canaan, Joyce E.

    2013-01-01

    This paper seeks to contribute to current efforts of left academics to shift the English public university away from its present state of what I below call "deep neoliberalisation." I utilise the concept of neoliberalisation rather than the more common concept of neoliberalism to frame what was an initially gradual and, under the current…

  3. Critical Factors in the Use of Evaluation in Italian Universities

    ERIC Educational Resources Information Center

    Rebora, Gianfranco; Turri, Matteo

    2011-01-01

    The use made of evaluation output is crucial for understanding the position and effectiveness of evaluation systems. This article examines the development of evaluation in the Italian university system from the 1990s onwards where serious problems have been and continue to be addressed in the use of evaluation output to improve academic activities…

  4. Building Technology Transfer Capacity in Turkish Universities: A Critical Analysis

    ERIC Educational Resources Information Center

    Ranga, Marina; Temel, Serdal; Ar, Ilker Murat; Yesilay, Rustem Baris; Sukan, Fazilet Vardar

    2016-01-01

    University technology transfer has been receiving significant government funding since 2012. Results of this major investment are now expected by the Turkish government and society, not only in terms of better teaching and research performance, but also of new jobs, new products and services, enhanced regional development and contribution to…

  5. Questing for Internationalization of Universities in Asia: Critical Reflections

    ERIC Educational Resources Information Center

    Mok, Ka Ho

    2007-01-01

    Globalization and the evolution of the knowledge-based economy have caused dramatic changes to the character and functions of higher education in most countries around the world. One major trend related to reforming and restructuring universities in Asia that has emerged is the adoption of strategies along the lines of the Anglo-Saxon paradigm in…

  6. Nuclear reactor fuel assembly duct-tube-to-handling-socket attachment system

    DOEpatents

    Christiansen, David W.; Smith, Bob G.

    1982-01-01

    A reusable system for removably attaching the upper end 10of a nuclear reactor duct tube to the lower end 30 of a nuclear reactor fuel assembly handling socket. A transition ring 20, fixed to the duct tube's upper end 10, has an interior-threaded section 22 with a first locking hole segment 24. An adaptor ring 40, fixed to the handling socket's lower end 30 has an outside-threaded section 42 with a second locking hole segment 44. The inside 22 and outside 42 threaded sections match and can be joined so that the first 24 and second 44 locking hole segments can be aligned to form a locking hole. A locking ring 50, with a locking pin 52, slides over the adaptor ring 40 so that the locking pin 52 fits in the locking hole. A swage lock 60 or a cantilever finger lock 70 is formed from the locking cup collar 26 to fit in a matching groove 54 or 56 in the locking ring 50 to prevent the locking ring's locking pin 52 from backing out of the locking hole.

  7. Decommissioning Small Research and Training Reactors; Experience on Three Recent University Projects - 12455

    SciTech Connect

    Gilmore, Thomas; DeWitt, Corey; Miller, Dustin; Colborn, Kurt

    2012-07-01

    Decommissioning small reactors within the confines of an active University environment presents unique challenges. These range from the radiological protection of the nearby University population and grounds, to the logistical challenges of working in limited space without benefit of the established controlled, protected, and vital areas common to commercial facilities. These challenges, and others, are discussed in brief project histories of three recent (calendar year 2011) decommissioning activities at three University training and research reactors. These facilities include three separate Universities in three states. The work at each of the facilities addresses multiple phases of the decommissioning process, from initial characterization and pre-decommissioning waste removal, to core component removal and safe storage, through to complete structural dismantlement and site release. The results of the efforts at each University are presented, along with the challenges that were either anticipated or discovered during the decommissioning efforts, and results and lessons learned from each of the projects. (authors)

  8. Modeling & analysis of criticality-induced severe accidents during refueling for the Advanced Neutron Source Reactor

    SciTech Connect

    Georgevich, V.; Kim, S.H.; Taleyarkhan, R.P.; Jackson, S.

    1992-10-01

    This paper describes work done at the Oak Ridge National Laboratory (ORNL) for evaluating the potential and resulting consequences of a hypothetical criticality accident during refueling of the 330-MW Advanced Neutron Source (ANS) research reactor. The development of an analytical capability is described. Modeling and problem formulation were conducted using concepts of reactor neutronic theory for determining power level escalation, coupled with ORIGEN and MELCOR code simulations for radionuclide buildup and containment transport Gaussian plume transport modeling was done for determining off-site radiological consequences. Nuances associated with modeling this blast-type scenario are described. Analysis results for ANS containment response under a variety of postulated scenarios and containment failure modes are presented. It is demonstrated that individuals at the reactor site boundary will not receive doses beyond regulatory limits for any of the containment configurations studied.

  9. A human reliability analysis of the University of New Mexico's AGN- 201M nuclear research reactor

    SciTech Connect

    Brumburgh, G.P. ); Heger, A.S. . Dept. of Chemical and Nuclear Engineering)

    1992-10-15

    During 1990--1991, a probabilistic risk assessment was conducted on the University of New Mexico's AGN-201M nuclear research reactor to address the risk and consequence of a maximum hypothetical release accident. The assessment indicated a potential for consequential human error to precipitate Chis scenario. Subsequently, a human reliability analysis was performed to evaluate the significance of human interaction on the reactor's safety systems. This paper presents the results of that investigation.

  10. CSER 01-011 Criticality Safety Evaluation for Light Water Reactor Fuel in NAC-1 Casks

    SciTech Connect

    ERICKSON, D.G.

    2001-11-01

    This analysis references a previous analysis (Larson, 1999) for a qualitative acceptability argument, and an appropriate ANS standard (ANS, 1998), and a reference (Clark 1966) for safe cylinder diameters as a function of {sup 235}U enrichment in UO{sub 2} fuel rods (pins) in water, for a quantitative acceptability argument. The previous analysis established the criticality safety of PWR assemblies without broken pins, pin segments or powdered fuel. This addendum extends the previous range of applicability in accordance with the controlling procedure (FH 2001b). The operation, when conducted according to the established limits stated in this document, complies with the incredibility principle. The evaluations demonstrated criticality safety for PWR assemblies with broken pins and pin segments of UO{sub 2} fuel.

  11. Conversion and standardization of university reactor fuels using low-enrichment uranium - options and costs

    SciTech Connect

    Harris, D.R.; Matos, J.E.; Young, H.H.

    1985-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. The US Nuclear Regulatory Commission has issued a policy statement expressing its concern and has published a proposed rule on limiting the use of HEU in NRC-licensed non-power reactors. The fuel options, functional impacts, licensing, and scheduling of conversion and standardization of these reactor fuels to use of low-enrichment uranium (LEU) have been assessed. The university reactors span a wide range in form and function, from medium-power intense neutron sources where HEU fuel may be required, to low-power training and research facilities where HEU fuel is unnecessary. Conversion provides an opportunity to standardize university reactor fuels and improve reactor utilization in some cases. The entire program is estimated to cost about $10 million and to last about five years. Planning for conversion and standardization is facilitated by the US Department of Energy. 20 refs., 1 tab.

  12. FAST TRACK COMMUNICATION: Criticality-induced universality in ratchets

    NASA Astrophysics Data System (ADS)

    Chacón, Ricardo

    2010-08-01

    Conclusive mathematical arguments are presented supporting the ratchet conjecture (Chacón 2007 J. Phys. A: Math. Theor. 40 F413), i.e. the existence of a universal force waveform which optimally enhances directed transport by symmetry breaking. Specifically, such a particular waveform is shown to be unique for both temporal and spatial biharmonic forces, and general (non-perturbative) laws providing the dependence of the strength of directed transport on the force parameters are deduced for these forces. The theory explains previous results for a great diversity of systems subjected to such biharmonic forces and provides a universal quantitative criterion to optimize any application of the ratchet effect induced by symmetry breaking of temporal and spatial biharmonic forces.

  13. Calculated k-effectives for light water reactor typical, U + Pu nitrate solution critical experiments

    SciTech Connect

    Primm, R.T. III; Mincey, J.F.

    1982-01-01

    The Department of Energy's Consolidated Fuel Reprocessing Program has as a goal the design of nuclear fuel reprocessing equipment. In order to validate computer codes used for criticality analyses in the design of such equipment, k-effectives have been calculated for several U + Pu nitrate solution critical experiments. As of January 1981, descriptions of 45 unpoisoned, U + Pu solution experiments were available in the open literature. Twelve of these experiments were performed with solutions which have physical characteristics typical of dissolved, light water reactor fuel. This paper contains a discussion of these twelve experiments, a review of the calculational procedure used to determine k-effectives, and the results of the calculations.

  14. Modular Pebble Bed Reactor Project, University Research Consortium Annual Report

    SciTech Connect

    Petti, David Andrew

    2000-07-01

    This project is developing a fundamental conceptual design for a gas-cooled, modular, pebble bed reactor. Key technology areas associated with this design are being investigated which intend to address issues concerning fuel performance, safety, core neutronics and proliferation resistance, economics and waste disposal. Research has been initiated in the following areas: · Improved fuel particle performance · Reactor physics · Economics · Proliferation resistance · Power conversion system modeling · Safety analysis · Regulatory and licensing strategy Recent accomplishments include: · Developed four conceptual models for fuel particle failures that are currently being evaluated by a series of ABAQUS analyses. Analytical fits to the results are being performed over a range of important parameters using statistical/factorial tools. The fits will be used in a Monte Carlo fuel performance code, which is under development. · A fracture mechanics approach has been used to develop a failure probability model for the fuel particle, which has resulted in significant improvement over earlier models. · Investigation of fuel particle physio-chemical behavior has been initiated which includes the development of a fission gas release model, particle temperature distributions, internal particle pressure, migration of fission products, and chemical attack of fuel particle layers. · A balance of plant, steady-state thermal hydraulics model has been developed to represent all major components of a MPBR. Component models are being refined to accurately reflect transient performance. · A comparison between air and helium for use in the energy-conversion cycle of the MPBR has been completed and formed the basis of a master’s degree thesis. · Safety issues associated with air ingress are being evaluated. · Post shutdown, reactor heat removal characteristics are being evaluated by the Heating-7 code. · PEBBED, a fast deterministic neutronic code package suitable for

  15. BFS, a Legacy to the International Reactor Physics, Criticality Safety, and Nuclear Data Communities

    SciTech Connect

    J. Blair Briggs; Anatoly Tsibulya; Yevgeniy Rozhikhin

    2012-03-01

    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. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. Data provided by these two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades The Russian Federation has been a major contributor to both projects with the Institute of Physics and Power Engineering (IPPE) as the major contributor from the Russian Federation. Included in the benchmark specifications from the BFS facilities are 34 critical configurations from BFS-49, 61, 62, 73, 79, 81, 97, 99, and 101; spectral characteristics measurements from BFS-31, 42, 57, 59, 61, 62, 73, 97, 99, and 101; reactivity effects measurements from BFS-62-3A; reactivity coefficients and kinetics measurements from BFS-73; and reaction rate measurements from BFS-42, 61, 62, 73, 97, 99, and 101.

  16. Scale-4 analysis of pressurized water reactor critical configurations: Volume 5, North Anna Unit 1 Cycle 5

    SciTech Connect

    Bowman, S.M.; Suto, T. |

    1996-10-01

    ANSI/ANS 8.1 requires that calculational methods for away-from- reactor (AFR) criticality safety analyses be validated against experiment. This report summarizes part of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial PWRs. Codes and data in the SCALE-4 code system were used. This volume documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. The KENO V.a criticality calculations for the North Anna 1 Cycle 5 beginning-of-cycle model yielded a value for k{sub eff} of 1. 0040{+-}0.0005.

  17. Study of neutron noise from reflected, metal assemblies with criticality safety applications in mind

    SciTech Connect

    Barnett, C.S.

    1985-08-20

    The author studied the statistics of detected neutrons that leaked from four subcritical reflected, enriched-uranium assemblies, to explore the feasibility of developing a criticality warning system based on neutron noise analysis. The calculated multiplication factors of the assemblies are 0.59, 0.74, 0.82, and 0.92. The author studied three possible discriminators, i.e., three signatures that might be used to discriminate among assemblies of various multiplications. They are: (1) variance-to-mean ratio of the counts in a time bin (V/M); (2) covariance-to-mean ratio of the counts in a common time bin from two different detectors (C/M); and (3) covariance-to-mean ratio of the counts from a single detector in two adjacent time bins of equal length, which the author calls the serial-covariance-to-mean ratio (SC/M). The performances of the three discriminators were not greatly different, but a hierarchy did emerge: SC/M greater than or equal to V/M greater than or equal to C/M. An example of some results: in the neighborhood of k = 0.6 the ..delta..k required for satisfactory discrimination varies from about 3% to 7% as detector solid angle varies from 19% to 5%. In the neighborhood of k = 0.8 the corresponding ..delta..ks are 1% and 2%. The noise analysis techniques studied performed well enough in deeply subcritical situations to deserve testing in an applications environment. They have a good chance of detecting changes in reactivity that are potentially dangerous. One can expect sharpest results when doing comparisons, i.e., when comparing two records, one taken in the past under circumstances known to be normal and one taken now to search for change.

  18. Investigation of Coded Source Neutron Imaging at the North Carolina State University PULSTAR Reactor

    SciTech Connect

    Xiao, Ziyu; Mishra, Kaushal; Hawari, Ayman; Bingham, Philip R; Bilheux, Hassina Z; Tobin Jr, Kenneth William

    2010-10-01

    A neutron imaging facility is located on beam-tube #5 of the 1-MWth PULSTAR reactor at the North Carolina State University. An investigation has been initiated to explore the application of coded imaging techniques at the facility. Coded imaging uses a mosaic of pinholes to encode an aperture, thus generating an encoded image of the object at the detector. To reconstruct the image recorded by the detector, corresponding decoding patterns are used. The optimized design of coded masks is critical for the performance of this technique and will depend on the characteristics of the imaging beam. In this work, Monte Carlo (MCNP) simulations were utilized to explore the needed modifications to the PULSTAR thermal neutron beam to support coded imaging techniques. In addition, an assessment of coded mask design has been performed. The simulations indicated that a 12 inch single crystal sapphire filter is suited for such an application at the PULSTAR beam in terms of maximizing flux with good neutron-to-gamma ratio. Computational simulations demonstrate the feasibility of correlation reconstruction methods on neutron transmission imaging. A gadolinium aperture with thickness of 500 m was used to construct the mask using a 38 34 URA pattern. A test experiment using such a URA design has been conducted and the point spread function of the system has been measured.

  19. Sulfur activation at the Little Boy-Comet Critical Assembly: a replica of the Hiroshima bomb

    SciTech Connect

    Kerr, G.D.; Emery, J.F.; Pace, J.V. III

    1985-04-01

    Studies have been completed on the activation of sulfur by fast neutrons from the Little Boy-Comet Critical Assembly which replicates the general features of the Hiroshima bomb. The complex effects of the bomb's design and construction on leakage of sulfur-activation neutrons were investigated both experimentally and theoretically. Our sulfur activation studies were performed as part of a larger program to provide benchmark data for testing of methods used in recent source-term calculations for the Hiroshima bomb. Source neutrons capable of activating sulfur play an important role in determining neutron doses in Hiroshima at a kilometer or more from the point of explosion. 37 refs., 5 figs., 6 tabs.

  20. Functional Assembly of Accessory Optic System Circuitry Critical for Compensatory Eye Movements

    PubMed Central

    Sun, Lu O.; Brady, Colleen M.; Cahill, Hugh; Al-Khindi, Timour; Sakuta, Hiraki; Dhande, Onkar S.; Noda, Masaharu; Huberman, Andrew D.; Nathans, Jeremy; Kolodkin, Alex L.

    2015-01-01

    SUMMARY Accurate motion detection requires neural circuitry that compensates for global visual field motion. Select subtypes of retinal ganglion cells perceive image motion and connect to the accessory optic system (AOS) in the brain, which generates compensatory eye movements that stabilize images during slow visual field motion. Here, we show that the murine transmembrane semaphorin 6A (Sema6A) is expressed in a subset of On direction-selective ganglion cells (On DSGCs) and is required for retinorecipient axonal targeting to the medial terminal nucleus (MTN) of the AOS. Plexin A2 and A4, two Sema6A binding partners, are expressed in MTN cells, attract Sema6A+ On DSGC axons, and mediate MTN targeting of Sema6A+ RGC projections. Furthermore, Sema6A/Plexin-A2/A4 signaling is required for the functional output of the AOS. These data reveal molecular mechanisms underlying the assembly of AOS circuits critical for moving image perception. PMID:25959730

  1. Critical seeding density improves the properties and translatability of self-assembling anatomically shaped knee menisci.

    PubMed

    Hadidi, Pasha; Yeh, Timothy C; Hu, Jerry C; Athanasiou, Kyriacos A

    2015-01-01

    A recent development in the field of tissue engineering is the rise of all-biologic, scaffold-free engineered tissues. Since these biomaterials rely primarily upon cells, investigation of initial seeding densities constitutes a particularly relevant aim for tissue engineers. In this study, a scaffold-free method was used to create fibrocartilage in the shape of the rabbit knee meniscus. The objectives of this study were to: (i) determine the minimum seeding density, normalized by an area of 44 mm(2), necessary for the self-assembling process of fibrocartilage to occur; (ii) examine relevant biomechanical properties of engineered fibrocartilage, such as tensile and compressive stiffness and strength, and their relationship to seeding density; and (iii) identify a reduced, or optimal, number of cells needed to produce this biomaterial. It was found that a decreased initial seeding density, normalized by the area of the construct, produced superior mechanical and biochemical properties. Collagen per wet weight, glycosaminoglycans per wet weight, tensile properties and compressive properties were all significantly greater in the 5 million cells per construct group as compared to the historical 20 million cells per construct group. Scanning electron microscopy demonstrated that a lower seeding density results in a denser tissue. Additionally, the translational potential of the self-assembling process for tissue engineering was improved though this investigation, as fewer cells may be used in the future. The results of this study underscore the potential for critical seeding densities to be investigated when researching scaffold-free engineered tissues. PMID:25234157

  2. Modeling the effect in of criticality from changes in key parameters for small High Temperature Nuclear Reactor (U-BatteryTM) using MCNP4C

    NASA Astrophysics Data System (ADS)

    Pauzi, A. M.

    2013-06-01

    The neutron transport code, Monte Carlo N-Particle (MCNP) which was wellkown as the gold standard in predicting nuclear reaction was used to model the small nuclear reactor core called "U-batteryTM", which was develop by the University of Manchester and Delft Institute of Technology. The paper introduces on the concept of modeling the small reactor core, a high temperature reactor (HTR) type with small coated TRISO fuel particle in graphite matrix using the MCNPv4C software. The criticality of the core were calculated using the software and analysed by changing key parameters such coolant type, fuel type and enrichment levels, cladding materials, and control rod type. The criticality results from the simulation were validated using the SCALE 5.1 software by [1] M Ding and J L Kloosterman, 2010. The data produced from these analyses would be used as part of the process of proposing initial core layout and a provisional list of materials for newly design reactor core. In the future, the criticality study would be continued with different core configurations and geometries.

  3. University-School-Community Partnership as Vehicle for Leadership, Service, and Change: A Critical Brokerage Perspective

    ERIC Educational Resources Information Center

    Hopson, Rodney; Miller, Peter; Lovelace, Temple S.

    2016-01-01

    Using a critical brokerage perspective to advance theoretical insights in the development of a community university partnership and understanding of the organizational embeddedness of a community empowerment agency in Pittsburgh, PA, USA, this article suggests that partnerships between American universities and communities are perfect vehicles for…

  4. Universities and Regional Development: A Critical Assessment of Tensions and Contradictions. International Studies in Higher Education

    ERIC Educational Resources Information Center

    Pinheiro, Romulo, Ed.; Benneworth, Paul, Ed.; Jones, Glen A., Ed.

    2012-01-01

    Universities are under increasing pressure to help promote socio-economic growth in their local communities. However until now, no systematic, critical attention has been paid to the factors and mechanisms that currently make this process so daunting. In Universities and Regional Development, scholars from Europe, the Americas, Africa, and Asia…

  5. Adapting, Not Adopting: Barriers Affecting Teaching for Critical Thinking at Two Rwandan Universities

    ERIC Educational Resources Information Center

    Schendel, Rebecca

    2016-01-01

    A recent study of student learning at three of Rwanda's most prestigious public universities has suggested that Rwandan students are not improving in their critical thinking ability during their time at university. This article reports on a series of faculty-level case studies, which were conducted at two of the participating institutions in order…

  6. Critical Discourse Analysis of Moderated Discussion Board of Virtual University of Pakistan

    ERIC Educational Resources Information Center

    Perveen, Ayesha

    2015-01-01

    The paper critically evaluated the discursive practices on the Moderated Discussion Board (MDB) of Virtual University of Pakistan (VUP). The paramount objective of the study was to conduct a critical discourse analysis (CDA) of the MDB on the Learning Management System (LMS) of VUP. For this purpose, the academic power relations of the students…

  7. Developing Critical Thinking in E-Learning Environment: Kuwait University as a Case Study

    ERIC Educational Resources Information Center

    Al-Fadhli, Salah; Khalfan, Abdulwahed

    2009-01-01

    This article investigated the impact of using e-learning models' with the principles of constructivism to enhance the critical thinking skills of students in higher education institutions. The study examines the effectiveness of e-learning model in enhancing critical thinking of students at university level. This effectiveness is measured by a…

  8. Critical Thinking in the University Curriculum--The Impact on Engineering Education

    ERIC Educational Resources Information Center

    Ahern, A.; O'Connor, T.; McRuairc, G.; McNamara, M.; O'Donnell, D.

    2012-01-01

    Critical thinking is a graduate attribute that many courses, including engineering courses, claim to produce in students. As a graduate attribute it is seen by academics as a particularly desirable outcome of student learning and is said by researchers to be a defining characteristic of university education. However, how critical thinking is…

  9. White Students at the Historically Black University: Toward Developing a Critical Consciousness

    ERIC Educational Resources Information Center

    Henry, Wilma J.; Closson, Rosemary B.

    2010-01-01

    The purpose of this article is to examine the potential of historically Black colleges and universities (HBCUs) to facilitate the development of a critical consciousness among their White students. It discusses philosophical views regarding the process of unveiling "Whiteness," including White critical studies and White identity development…

  10. Teaching and Learning Critical Reading with Transnational Texts at a Mexican University: An Emergentist Case Study

    ERIC Educational Resources Information Center

    Perales Escudero, Moises Damian

    2011-01-01

    This dissertation project examines the implementation of a critical reading intervention in a Mexican university, and the emergence of target critical reading processes in Mexican college-level EFL readers. It uses a Complexity Theory-inspired, qualitative methodology. Orienting the selection and design of materials is a deep view of culture that…

  11. Helping ESL and EFL University Students Read Critically: A 2000's Challenge.

    ERIC Educational Resources Information Center

    Crismore, Avon

    This paper discusses how most students are not yet competent critical readers of academic or electronic texts. This is especially true of English as a Second Language (ESL) and English as a Foreign Language (EFL) university students. After describing the extent of the lack of critical reading skills, an extensive literature review on the subject…

  12. University Reactor Instrumentation grant program. Final report, September 7, 1990--August 31, 1995

    SciTech Connect

    Talnagi, J.W.

    1998-06-17

    The Ohio State University Nuclear Reactor Laboratory (OSU NRL) participated in the Department of Energy (DOE) grant program commonly denoted as the University Reactor Instrumentation (URI) program from the period September 1990 through August 1995, after which funding was terminated on a programmatic basis by DOE. This program provided funding support for acquisition of capital equipment targeted for facility upgrades and improvements, including modernizing reactor systems and instrumentation, improvements in research and instructional capabilities, and infrastructure enhancements. The staff of the OSU NRL submitted five grant applications during this period, all of which were funded either partially or in their entirety. This report will provide an overview of the activities carried out under these grants and assess their impact on the OSU NRL facilities.

  13. The development of a universal diagnostic probe system for Tokamak fusion test reactor

    NASA Technical Reports Server (NTRS)

    Mastronardi, R.; Cabral, R.; Manos, D.

    1982-01-01

    The Tokamak Fusion Test Reactor (TFTR), the largest such facility in the U.S., is discussed with respect to instrumentation in general and mechanisms in particular. The design philosophy and detailed implementation of a universal probe mechanism for TFTR is discussed.

  14. Two-scale-factor universality near the critical point of fluids

    NASA Technical Reports Server (NTRS)

    Sengers, J. V.; Moldover, M. R.

    1978-01-01

    Thermodynamic data from interferometric density profile studies and light-scattering experiments near the critical isochore of Xe, CO2 and SF6 provide a basis for examining the hypothesized two-scale-factor universality for the correlation function of fluids near the gas-liquid critical point. For the investigation, three-scale-factor universality is assumed, with Ising-like critical exponent values obtained through the renormalization group technique. The two thermodynamic scale factors are found from the density profiles, while the scale factor for the correlation length is obtained from the light-scattering data.

  15. The Kyoto Protocol: A business perspective

    SciTech Connect

    Malin, C.B.

    1998-01-19

    Governments have made a tentative start in responding to climate change. In marathon negotiating sessions that extended into an extra day Dec. 1--11 in Kyoto, Japan, representatives from more than 160 governments hammered out the Kyoto Protocol to the United Nations Framework Convention on Climate Change (FCCC). The protocol calls for developed countries to reduce emissions of greenhouse gases (GHGs) on averaged by 5.2% below 1990 levels by the years 2008--2012. Developing countries have no new obligations. The paper discusses the agreement, ratification, future questions, business role, and the challenge.

  16. 78 FR 5840 - Notice of License Termination for University of Illinois Advanced TRIGA Reactor, License No. R-115

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-01-28

    ... COMMISSION Notice of License Termination for University of Illinois Advanced TRIGA Reactor, License No. R-115... No. R-115, for the University of Illinois Advanced TRIGA Reactor (ATR). The NRC has terminated the..., Facility Operating License No. R-115 is terminated. The above referenced documents may be examined,...

  17. Criticality Safety Evaluation for the Advanced Test Reactor U-Mo Demonstration Elements

    SciTech Connect

    Leland M. Montierth

    2010-12-01

    The Reduced Enrichment Research Test Reactors (RERTR) fuel development program is developing a high uranium density fuel based on a (LEU) uranium-molybdenum alloy. Testing of prototypic RERTR fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. Two RERTR-Full Size Demonstration fuel elements based on the ATR-Reduced YA elements (all but one plate fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). The two fuel elements will be irradiated in alternating cycles such that only one element is loaded in the reactor at a time. Existing criticality analyses have analyzed Standard (HEU) ATR elements (all plates fueled) from which controls have been derived. This criticality safety evaluation (CSE) documents analysis that determines the reactivity of the Demonstration fuel elements relative to HEU ATR elements and shows that the Demonstration elements are bound by the Standard HEU ATR elements and existing HEU ATR element controls are applicable to the Demonstration elements.

  18. N-Reactor (U-metal) Fuel Characteristics for Disposal Criticality Analysis

    SciTech Connect

    Taylor, Larry Lorin

    2000-05-01

    DOE-owned spent nuclear fuels encompass many fuel types. In an effort to facilitate criticality analysis for these various fuel types, they were categorized into nine characteristic fuel groups with emphasis on fuel matrix composition. Out of each fuel group, a representative fuel type was chosen for analysis as a bounding case within that fuel group. Generally, burnup data, fissile enrichments, and total fuel and fissile mass govern the selection of the representative or candidate fuel within that group. Additionally, the criticality analysis will also require data to support design of the canister internals, thermal, and radiation shielding. The purpose of this report is to consolidate and provide in a concise format, material and information/data needed to perform supporting analyses to qualify N-Reactor fuels for acceptance into the designated repository. The N Reactor fuels incorporate zirconium cladding and uranium metal with unique fabrication details in terms of physical size, and method of construction. The fuel construction and post-irradiation handling have created attendant issues relative to cladding failure in the underwater storage environment. These fuels were comprised of low-enriched metal (0.947 to 1.25 wt% 235U) that were originally intended to generate weapons-grade plutonium for national defense. Modifications in subsequent fuel design and changes in the mode of reactor operation in later years were focused more toward power production.

  19. CO2 Reduction Assembly Prototype Using Microlith-Based Sabatier Reactor for Ground Demonstration

    NASA Technical Reports Server (NTRS)

    Junaedi, Christian; Hawley, Kyle; Walsh, Dennis; Roychoudhury, Subir; Abney, Morgan B.; Perry, Jay L.

    2014-01-01

    The utilization of CO2 to produce life support consumables, such as O2 and H2O, via the Sabatier reaction is an important aspect of NASA's cabin Atmosphere Revitalization System (ARS) and In-Situ Resource Utilization (ISRU) architectures for both low-earth orbit and long-term manned space missions. Carbon dioxide can be reacted with H2, obtained from the electrolysis of water, via Sabatier reaction to produce methane and H2O. Methane can be stored and utilized as propellant while H2O can be either stored or electrolyzed to produce oxygen and regain the hydrogen atoms. Depending on the application, O2 can be used to replenish the atmosphere in human-crewed missions or as an oxidant for robotic and return missions. Precision Combustion, Inc. (PCI), with support from NASA, has previously developed an efficient and compact Sabatier reactor based on its Microlith® catalytic technology and demonstrated the capability to achieve high CO2 conversion and CH4 selectivity (i.e., =90% of the thermodynamic equilibrium values) at high space velocities and low operating temperatures. This was made possible through the use of high-heat-transfer and high-surface-area Microlith catalytic substrates. Using this Sabatier reactor, PCI designed, developed, and demonstrated a stand-alone CO2 Reduction Assembly (CRA) test system for ground demonstration and performance validation. The Sabatier reactor was integrated with the necessary balance-of-plant components and controls system, allowing an automated, single "push-button" start-up and shutdown. Additionally, the versatility of the test system prototype was demonstrated by operating it under H2-rich (H2/CO2 of >4), stoichiometric (ratio of 4), and CO2-rich conditions (ratio of <4) without affecting its performance and meeting the equilibrium-predicted water recovery rates. In this paper, the development of the CRA test system for ground demonstration will be discussed. Additionally, the performance results from testing the system at

  20. Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

    SciTech Connect

    Lata

    1996-09-26

    This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

  1. SpRoUTS (Space Robot Universal Truss System): Reversible Robotic Assembly of Deployable Truss Structures of Reconfigurable Length

    NASA Technical Reports Server (NTRS)

    Jenett, Benjamin; Cellucci, Daniel; Cheung, Kenneth

    2015-01-01

    Automatic deployment of structures has been a focus of much academic and industrial work on infrastructure applications and robotics in general. This paper presents a robotic truss assembler designed for space applications - the Space Robot Universal Truss System (SpRoUTS) - that reversibly assembles a truss from a feedstock of hinged andflat-packed components, by folding the sides of each component up and locking onto the assembled structure. We describe the design and implementation of the robot and show that the assembled truss compares favorably with prior truss deployment systems.

  2. Innovative technologies on fuel assemblies cleaning for sodium fast reactors: First considerations on cleaning process

    SciTech Connect

    Simon, N.; Lorcet, H.; Beauchamp, F.; Guigues, E.; Lovera, P.; Fleche, J. L.; Lacroix, M.; Carra, O.; Prele, G.

    2012-07-01

    Within the framework of Sodium Fast Reactor development, innovative fuel assembly cleaning operations are investigated to meet the GEN IV goals of safety and of process development. One of the challenges is to mitigate the Sodium Water Reaction currently used in these processes. The potential applications of aqueous solutions of mineral salts (including the possibility of using redox chemical reactions) to mitigate the Sodium Water Reaction are considered in a first part and a new experimental bench, dedicated to this study, is described. Anhydrous alternative options based on Na/CO{sub 2} interaction are also presented. Then, in a second part, a functional study conducted on the cleaning pit is proposed. Based on experimental feedback, some calculations are carried out to estimate the sodium inventory on the fuel elements, and physical methods like hot inert gas sweeping to reduce this inventory are also presented. Finally, the implementation of these innovative solutions in cleaning pits is studied in regard to the expected performances. (authors)

  3. Recent palladium membrane reactor development at the tritium systems test assembly

    SciTech Connect

    Willms, R.S.; Birdsell, S.A.; Wilhelm, R.C.

    1995-07-01

    The palladium membrane reactor (PMR) is proving to be a simple and effective means for recovering hydrogen isotopes from fusion fuel impurities such as methane and water. This device directly combines two techniques which have long been utilized for hydrogen processing, namely catalytic shift reactions and palladium/silver permeators. A proof-of-principle (PMR) has been constructed and tested at the Tritium Systems Test Assembly of Los Alamos National Laboratory. The first tests with this device showed that is was effective for the proposed purpose. Initial work concluded that a nickel catalyst was an appropriate choice for use in a PMR. More detailed testing of the PMR with such a catalyst was performed and reported in other works. It was shown that a nickel catalyst-packed PMR did, indeed, recover hydrogen from water and methane with efficiencies approaching 100% in a single processing pass. These experiments were conducted over an extended period of time and no failure or need for regeneration was encountered. These positive results have prompted further PMR development. Topics addressed include alternate PMR geometries and initial testing of the PMR with tritium. These are the subjects of this paper.

  4. Bottom head to shell junction assembly for a boiling water nuclear reactor

    DOEpatents

    Fife, Alex Blair; Ballas, Gary J.

    1998-01-01

    A bottom head to shell junction assembly which, in one embodiment, includes an annular forging having an integrally formed pump deck and shroud support is described. In the one embodiment, the annular forging also includes a top, cylindrical shaped end configured to be welded to one end of the pressure vessel cylindrical shell and a bottom, conical shaped end configured to be welded to the disk shaped bottom head. Reactor internal pump nozzles also are integrally formed in the annular forging. The nozzles do not include any internal or external projections. Stubs are formed in each nozzle opening to facilitate welding a pump housing to the forging. Also, an upper portion of each nozzle opening is configured to receive a portion of a diffuser coupled to a pump shaft which extends through the nozzle opening. Diffuser openings are formed in the integral pump deck to provide additional support for the pump impellers. The diffuser opening is sized so that a pump impeller can extend at least partially therethrough. The pump impeller is connected to the pump shaft which extends through the nozzle opening.

  5. Bottom head to shell junction assembly for a boiling water nuclear reactor

    DOEpatents

    Fife, A.B.; Ballas, G.J.

    1998-02-24

    A bottom head to shell junction assembly which, in one embodiment, includes an annular forging having an integrally formed pump deck and shroud support is described. In the one embodiment, the annular forging also includes a top, cylindrical shaped end configured to be welded to one end of the pressure vessel cylindrical shell and a bottom, conical shaped end configured to be welded to the disk shaped bottom head. Reactor internal pump nozzles also are integrally formed in the annular forging. The nozzles do not include any internal or external projections. Stubs are formed in each nozzle opening to facilitate welding a pump housing to the forging. Also, an upper portion of each nozzle opening is configured to receive a portion of a diffuser coupled to a pump shaft which extends through the nozzle opening. Diffuser openings are formed in the integral pump deck to provide additional support for the pump impellers. The diffuser opening is sized so that a pump impeller can extend at least partially therethrough. The pump impeller is connected to the pump shaft which extends through the nozzle opening. 5 figs.

  6. Concept for UF6-fueled self-critical DNPL reactor system

    NASA Technical Reports Server (NTRS)

    Rodgers, R. J.

    1979-01-01

    An analytical study of a self-critical nuclear pumped laser system concept was performed. The primary emphasis was on a reactor concept which would use gaseous uranium hexafluoride (UF6) as the fissioning material. A reference configuration was selected which has a 3.2 cu m lasing volume as the reactor core. The core is composed of a series of hexagonal graphite tubes which are surrounded by a reflector-moderator composed either of heavy water or beryllium. Laser transitions requiring average fission power densities less than approximately 1 kW/cu cm for excitation are most attractive. Operation at wavelengths greater than approximately 400 nm may be required because of limitations imposed by the opacity of gaseous UF6. Further research directed toward identification of UF6 compatible lasing transitions is required.

  7. Intact and Degraded Component Criticality Calculations of N Reactors Spent Nuclear Fuel

    SciTech Connect

    L. Angers

    2001-01-31

    The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k{sub eff}) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR.

  8. BRAIN ACONITASE ACTIVITY IN SPONTANEOUSLY HYPERTENSIVE (SHR) AND WISTAR-KYOTO (WKY) RATS.

    EPA Science Inventory

    Animal models of susceptibility are critical for human health risk assessment. Previous studies indicate that spontaneously hypertensive (SHR) rats are more sensitive than Wistar-Kyoto (WKY) rats to the cholinesterase (ChE) inhibitors such as carbaryl and chlorpyrifos. This diffe...

  9. A U. S. Perspective on Fast Reactor Fuel Fabrication Technology and Experience Part I: Metal Fuels and Assembly Design

    SciTech Connect

    Douglas E. Burkes; Randall S. Fielding; Douglas L. Porter; Douglas C. Crawford; Mitchell K. Meyer

    2009-06-01

    This paper is Part I of a review focusing on the United States experience with metallic fast reactor fuel fabrication and assembly design for the Experimental Breeder Reactor-II and the Fast Flux Test Facility, and it also refers to the impact of development in other nations. Experience with metal fuel fabrication in the United States is extensive, including over 60 years of research conducted by the government, national laboratories, industry, and academia. This experience has culminated into a foundation of research and resulted in significant improvements to the technologies employed to fabricate metallic fast reactor fuel. This part of the review documents the current state of fuel fabrication technologies for metallic fuels, some of the challenges faced by previous researchers, and how these were overcome. Knowledge gained from reviewing previous investigations will aid both researchers and policy makers in forming future decisions relating to nuclear fuel fabrication technologies.

  10. A US perspective on fast reactor fuel fabrication technology and experience part I: metal fuels and assembly design

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Fielding, Randall S.; Porter, Douglas L.; Crawford, Douglas C.; Meyer, Mitchell K.

    2009-06-01

    This paper is part I of a review focusing on the United States experience with metallic fast reactor fuel fabrication and assembly design for the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF). Experience with metal fuel fabrication in the United States is extensive, including over 60 years of research conducted by the government, national laboratories, industry, and academia. This experience has culminated in a considerable amount of research that resulted in significant improvements to the technologies employed to fabricate metallic fast reactor fuel. This part of the review documents the current state of fuel fabrication technologies for metallic fuels, some of the challenges faced by previous researchers, and how these were overcome. Knowledge gained from reviewing previous investigations will aid both researchers and policy makers in forming future decisions relating to nuclear fuel fabrication technologies.

  11. Summary of experimental data for critical arrays of water moderated Fast Test Reactor fuel

    SciTech Connect

    Durst, B.M.; Bierman, S.R.; Clayton, E.D.; Mincey, J.F.; Primm, R.T. III

    1981-05-01

    A research program, funded by the Consolidated Fuel Reprocessing Program (CFRP) of Oak Ridge National Laboratory (ORNL), was initiated at Battelle Pacific Northwest Laboratory (PNL) to acquire experimental data on heterogeneous water moderated arrays of Fast Test Reactor (FTR) fuel pins. The objective of this program is to provide critical experiment data for validating calculational techniques used in criticality assessments of reprocessing equipment containing FTR-type fuels. Consequently, the experiments were designed to permit accurate definition in Monte Carlo computer codes currently used in these assessments. Square and triangular pitched lattices of fuel have been constructed under a variety of conditions covering the range from undermoderated to overmoderated arrays. Experiments were conducted composed of arrays which were water reflected, partially concrete reflected, and arrays with interspersed solid neutron absorbers. The absorbers utilized were Boral, and cadmium plates and gadolinium cylindrical rods. Data from non-CFRP sponsored subcritical experiments (previously performed at Hanford) also are included.

  12. Criticality benchmark guide for light-water-reactor fuel in transportation and storage packages

    SciTech Connect

    Lichtenwalter, J.J.; Bowman, S.M.; DeHart, M.D.; Hopper, C.M.

    1997-03-01

    This report is designed as a guide for performing criticality benchmark calculations for light-water-reactor (LWR) fuel applications. The guide provides documentation of 180 criticality experiments with geometries, materials, and neutron interaction characteristics representative of transportation packages containing LWR fuel or uranium oxide pellets or powder. These experiments should benefit the U.S. Nuclear Regulatory Commission (NRC) staff and licensees in validation of computational methods used in LWR fuel storage and transportation concerns. The experiments are classified by key parameters such as enrichment, water/fuel volume, hydrogen-to-fissile ratio (H/X), and lattice pitch. Groups of experiments with common features such as separator plates, shielding walls, and soluble boron are also identified. In addition, a sample validation using these experiments and a statistical analysis of the results are provided. Recommendations for selecting suitable experiments and determination of calculational bias and uncertainty are presented as part of this benchmark guide.

  13. Optical critical dimension metrology for directed self-assembly assisted contact hole shrink

    NASA Astrophysics Data System (ADS)

    Dixit, Dhairya; Green, Avery; Hosler, Erik R.; Kamineni, Vimal; Preil, Moshe E.; Keller, Nick; Race, Joseph; Chun, Jun Sung; O'Sullivan, Michael; Khare, Prasanna; Montgomery, Warren; Diebold, Alain C.

    2016-01-01

    Directed self-assembly (DSA) is a potential patterning solution for future generations of integrated circuits. Its main advantages are high pattern resolution (˜10 nm), high throughput, no requirement of high-resolution mask, and compatibility with standard fab-equipment and processes. The application of Mueller matrix (MM) spectroscopic ellipsometry-based scatterometry to optically characterize DSA patterned contact hole structures fabricated with phase-separated polystyrene-b-polymethylmethacrylate (PS-b-PMMA) is described. A regression-based approach is used to calculate the guide critical dimension (CD), DSA CD, height of the PS column, thicknesses of underlying layers, and contact edge roughness of the post PMMA etch DSA contact hole sample. Scanning electron microscopy and imaging analysis is conducted as a comparative metric for scatterometry. In addition, optical model-based simulations are used to investigate MM elements' sensitivity to various DSA-based contact hole structures, predict sensitivity to dimensional changes, and its limits to characterize DSA-induced defects, such as hole placement inaccuracy, missing vias, and profile inaccuracy of the PMMA cylinder.

  14. University Multilingualism: A Critical Narrative from the University of the Western Cape, South Africa

    ERIC Educational Resources Information Center

    Antia, Bassey E.

    2015-01-01

    This article offers a narrative of the University of the Western Cape, South Africa, from the prism of the duality of language as a co-modality (with people, protest, policy and practices) for constituting the institution in whole or in part and as a reflection of its co-modalities. For its framing, the narrative eclectically draws on language…

  15. METHOD AND MEANS FOR SUPPORTING REACTOR FUEL CONTAINERS IN AN ASSEMBLY

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.; Coombs, C.A.

    1962-12-11

    This patent relates to means for supporting fuelcontaining tubes in an assembly which include grid means at either end of the fuel element assembly antl improved grid means intermediate of the ends to provide support against lateral displacement. (AEC)

  16. Critical Attractor and Universality in a Renormalization Scheme for Three Frequency Hamiltonian Systems

    NASA Astrophysics Data System (ADS)

    Chandre, C.; Jauslin, H. R.

    1998-12-01

    We study an approximate renormalization-group transformation to analyze the breakup of invariant tori for 3 degrees of freedom Hamiltonian systems. The scheme is implemented for the spiral mean torus. We find numerically that the critical surface is the stable manifold of a critical nonperiodic attractor. We compute scaling exponents associated with this fixed set, and find that they can be expected to be universal.

  17. Reactor Meltdown: Critical Zone Processes In Siliciclastics Unlikely To Be Directly Transferable To Carbonates

    NASA Astrophysics Data System (ADS)

    Gulley, J. D.; Cohen, M. J.; Kramer, M. G.; Martin, J. B.; Graham, W. D.

    2013-12-01

    Carbonate terrains cover 20% of Earth's ice-free land and are modified through interactions between rocks, water and biota that couple ecosystems processes to weathering reactions within the critical zone. Weathering in carbonate systems differs from the Critical Zone Reactor model developed for siliciclastic systems because reactions in siliciclastic critical zones largely consist of incongruent weathering (e.g., feldspar to secondary clay minerals) that typically occur in the soil zone within a few meters of the land surface. These incongruent reactions create regolith, which is removed by physical transport mechanisms that drive landscape denudation. In contrast, carbonate critical zones are mostly composed of homogeneous and soluble minerals, which dissolve congruently with the weathering products exported in solution, limiting regolith in the soil mantle to small amounts of insoluble residues. These reactions can extend to depths greater than 2 km below the surface. As water at the land surface drains preferentially through vertical joints and horizontal bedding planes of the carbonate critical zones, it is 'charged' with biologically-derived carbon dioxide, which decreases pH, dissolves carbonate rock, and enlarges subsurface flowpaths through feedbacks between flow and dissolution. Caves are extreme end products of this process and are key morphological features of carbonate critical zones. Caves link surface processes to the deep subsurface and serve as efficient delivery agents for oxygen, carbon and nutrients to zones within the critical zone that are deficient in all three, interrupting vertical and horizontal chemical gradients that would exist if caves were not present. We present select data from air and water-filled caves in the upper Floridan aquifer, Florida, USA, that demonstrate how caves, acting as very large preferential flow paths, alter processes in carbonate relative to siliciclastic critical zones. While caves represent an extreme end

  18. University of Florida--US Department of Energy 1994-1995 reactor sharing program

    SciTech Connect

    Vernetson, W.G.

    1996-06-01

    The grant support of $24,250 (1994-95?) was well used by the University of Florida as host institution to support various educational institutions in the use of UFTR Reactor. All users and uses were screened to assure the usage was for educational institutions eligible for participation in the Reactor Sharing Program; where research activities were involved, care was taken to assure the research was not funded by grants for contract funding from outside sources. Over 12 years, the program has been a key catalyst for renewing utilization of UFTR both by external users around the State of Florida and the Southeast and by various faculty members within the University of Florida. Tables provide basic information about the 1994-95 program and utilization of UFTR.

  19. University of Virginia {open_quotes}virtual{close_quotes} reactor facility tours

    SciTech Connect

    Krause, D.R.; Mulder, R.U.

    1995-12-31

    An electronic information and tour book has been constructed for the University of Virginia reactor (UVAR) facility. Utilizing the global Internet, the document resides on the University of Virginia World Wide Web (WWW or W) server within the UVAR Homepage at http://www.virginia. edu/{approximately}reactor/. It is quickly accessible wherever an Internet connection exists. The UVAR Homepage files are accessed with the hypertext transfer protocol (http) prefix. The files are written in hypertext markup language (HTML), a very simple method of preparing ASCII text for W3 presentation. The HTML allows use of various hierarchies of headers, indentation, fonts, and the linking of words and/or pictures to other addresses-uniform resource locators. The linking of texts, pictures, sounds, and server addresses is known as hypermedia.

  20. Modeling of Coolant Flow in the Fuel Assembly of the Reactor of a Floating Nuclear Power Plant Using the Logos CFD Program

    NASA Astrophysics Data System (ADS)

    Dmitriev, S. M.; Dobrov, A. A.; Legchanov, M. A.; Khrobostov, A. E.

    2015-09-01

    Results of computer modeling of coolant flow in the fuel assembly of the reactor of a floating nuclear power plant using the LOGOS CFD programs have been given. The possibility of using the obtained results to improve models built into the engineering programs of thermohydraulic calculation of nuclear-reactor cores has been considered.

  1. Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test

    SciTech Connect

    Godfroy, Thomas J.; Bragg-Sitton, Shannon M.; Kapernick, Richard J.

    2004-02-04

    One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

  2. Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test

    NASA Astrophysics Data System (ADS)

    Godfroy, Thomas J.; Kapernick, Richard J.; Bragg-Sitton, Shannon M.

    2004-02-01

    One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

  3. Startup experience at the University of Texas TRIGA Mark II Reactor

    SciTech Connect

    Bauer, Thomas L.; Wehring, Bernard W.

    1992-07-01

    After eight years of singular effort, the UT-TRIGA Mark II research reactor was licensed and is fully operational. This reactor is the focus of a new reactor laboratory facility which is located at the Balcones Research Center, a north Austin campus of The University of Texas at Austin. The UT-TRIGA reactor is licensed for 1.1 MW steady power operation and 3 dollar pulsing. A startup program was implemented upon receipt of the facility license on January 17, 1992. Several facility features are unique to this startup. Among these were the use of fuel with various burnup and a digital control system. The reactor laboratory staff with assistance from a General Atomics instrumentation engineer performed all phases of the startup program. Core loading began in February 1992 with final testing completed in May 1992. Several unusual problems were encountered during this time. Experiment authorizations have been written to resume Neutron Activation Analysis programs and isotope production. Several neutron beam tube experiments are in the design and test phase. (author)

  4. Increasing use of yellow colors in Kyoto

    NASA Astrophysics Data System (ADS)

    Akita, Munehira; Nara, Iwao

    2002-06-01

    Colors used for commercial signboards, displayed outdoors as well as indoors through windows, such as a store sign, an advertising sign, a sky sign, a poster, a placard, and a billboard were extensively surveyed in Kyoto City, Japan, in 1998. The survey showed that various kinds of yellow painted signs have increased rapidly and invaded a center area and suburbs of the city. Vivid yellow, what we called it the Y98 virus, is specially considered a color unpleasantly matched to the city image of Kyoto which was the capital of Japan for nearly 1000 years (794 to 1868) and is endowed with cultural and historic heritage. Discussions trying to find out what we could do to prevent the rapid spread of a big commercial display painted with vivid yellows what we called 'the Y98 virus' over the city will be summarized in a main text.

  5. NUCLEAR REACTOR

    DOEpatents

    Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

    1962-10-23

    A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

  6. Critical Resource Effects on America's Universities: What's behind the Growing Entrepreneurial Orientation?

    ERIC Educational Resources Information Center

    Powers, Joshua B.

    The purpose of this study was to investigate the effects of critical resource flows on technology transfer activity. The investigation focused on the impact on a university's licensing orientation of four sources of research and development (R&D) revenues: federal, state, industry, and institutional. By licensing orientation is meant the number of…

  7. Building Social Inclusion through Critical Arts-Based Pedagogies in University Classroom Communities

    ERIC Educational Resources Information Center

    Chappell, Sharon Verner; Chappell, Drew

    2016-01-01

    In humanities and education university classrooms, the authors facilitated counter-narrative arts-based inquiry projects in order to build critical thought and social inclusion. The first author examines public performance installations created by graduate students in elementary and bilingual education on needs-based and dignity-based rights of…

  8. International Students' Critical Thinking-Related Problem Areas: UK University Teachers' Perspectives

    ERIC Educational Resources Information Center

    Shaheen, Nisbah

    2016-01-01

    This qualitative study aims to understand the areas of international students' critical thinking-related initial difficulties, in order to facilitate their academic experiences in UK universities. Using a sample of 14 British teachers, the findings reveal that students from culturally and linguistically diverse traditions are very different in…

  9. Perceived Learning, Critical Elements and Lasting Impacts on University-Based Wilderness Educational Expeditions

    ERIC Educational Resources Information Center

    Asfeldt, Morten; Hvenegaard, Glen

    2014-01-01

    This study examined participants' perceptions of learning, critical elements, and lasting impacts of their wilderness expeditions. Fifty-seven students, who completed a for-credit wilderness canoe expedition between 1993 and 2007 at the Augustana Campus, University of Alberta, participated in the investigation. Perceived learning most…

  10. Mastering Leadership Concepts through Utilizing Critical Thinking Strategies within Educational Administration Courses at Kuwait University

    ERIC Educational Resources Information Center

    Alqahtani, Abdulmuhsen Ayedh; Al-Enezi, Mutlaq M.

    2012-01-01

    The current study aims at exploring the students' perceptions of mastering leadership concepts and critical thinking strategies implemented by faculty members in the college of education at Kuwait University, and the impact of the later on former. The data was collected using a questionnaire on a sample consisting of 411 students representing…

  11. Critical Success Factors in Crafting Strategic Architecture for E-Learning at HP University

    ERIC Educational Resources Information Center

    Sharma, Kunal; Pandit, Pallvi; Pandit, Parul

    2011-01-01

    Purpose: The purpose of this paper is to outline the critical success factors for crafting a strategic architecture for e-learning at HP University. Design/methodology/approach: A descriptive survey type of research design was used. An empirical study was conducted on students enrolled with the International Centre for Distance and Open Learning…

  12. Exposing Ideology within University Policies: A Critical Discourse Analysis of Faculty Hiring, Promotion and Remuneration Practices

    ERIC Educational Resources Information Center

    Uzuner-Smith, Sedef; Englander, Karen

    2015-01-01

    Using critical discourse analysis (CDA), this paper exposes the neoliberal ideology of the knowledge-based economy embedded within university policies, specifically those that regulate faculty hiring, promotion, and remuneration in two national contexts: Turkey and Mexico. The paper follows four stages of CDA: (1) focus upon a social wrong in its…

  13. Self-Assembled Superparamagnetic Iron Oxide Nanoclusters for Universal Cell Labeling and MRI

    NASA Astrophysics Data System (ADS)

    Chen, Shuzhen; Zhang, Jun; Jiang, Shengwei; Lin, Gan; Luo, Bing; Yao, Huan; Lin, Yuchun; He, Chengyong; Liu, Gang; Lin, Zhongning

    2016-05-01

    Superparamagnetic iron oxide (SPIO) nanoparticles have been widely used in a variety of biomedical applications, especially as contrast agents for magnetic resonance imaging (MRI) and cell labeling. In this study, SPIO nanoparticles were stabilized with amphiphilic low molecular weight polyethylenimine (PEI) in an aqueous phase to form monodispersed nanocomposites with a controlled clustering structure. The iron-based nanoclusters with a size of 115.3 ± 40.23 nm showed excellent performance on cellular uptake and cell labeling in different types of cells, moreover, which could be tracked by MRI with high sensitivity. The SPIO nanoclusters presented negligible cytotoxicity in various types of cells as detected using MTS, LDH, and flow cytometry assays. Significantly, we found that ferritin protein played an essential role in protecting stress from SPIO nanoclusters. Taken together, the self-assembly of SPIO nanoclusters with good magnetic properties provides a safe and efficient method for universal cell labeling with noninvasive MRI monitoring capability.

  14. The assembly of galaxy halo within cosmic web: primordial anisotropy and its universality

    NASA Astrophysics Data System (ADS)

    Kang, Xi

    2015-08-01

    Observations in the past few years have found two interesting facts. One is that satellite galaxies are not randomly distributed around the central galaxies. The other is that galaxies spin are well correlated with the large scales structures, with dependence on galaxy type. Existed theoretical studies have difficult to explain the two observations. In fact they can be understood in the context of galaxy halo assembly with its relation to the cosmic web. Using N-body simulation, we investigate the assembly history of dark matter halo. It is found that satellite galaxy (or subhaloes) are accreted mostly along the cosmic filament, and the current anisotropy of satellites are already set at high-redshift universe at accretion. However, we find that for low-mass haloes ( or spiral galaxy) at z=0, the accretion of their massive progenitors are preferentially be perpendicular to the filament, not like the progenitors of elliptical galaxies. This finding well explains the puzzle that the observed spins of spiral galaxies are aligned with cosmic filament, but that of elliptical galaxies (or their short axes) are perpendicular to filament.

  15. The NASA Constellation University Institutes Project: Thrust Chamber Assembly Virtual Institute

    NASA Technical Reports Server (NTRS)

    Tucker, P. Kevin; Rybak, Jeffry A.; Hulka, James R.; Jones, Gregg W.; Nesman, Tomas; West, Jeffrey S.

    2006-01-01

    This paper documents key aspects of the Constellation University Institutes Project (CUIP) Thrust Chamber Assembly (TCA) Virtual Institute (VI). Specifically, the paper details the TCA VI organizational and functional aspects relative to providing support for Constellation Systems. The TCA VI vision is put forth and discussed in detail. The vision provides the objective and approach for improving thrust chamber assembly design methodologies by replacing the current empirical tools with verified and validated CFD codes. The vision also sets out ignition, performance, thermal environments and combustion stability as focus areas where application of these improved tools is required. Flow physics and a study of the Space Shuttle Main Engine development program are used to conclude that the injector is the key to robust TCA design. Requirements are set out in terms of fidelity, robustness and demonstrated accuracy of the design tool. Lack of demonstrated accuracy is noted as the most significant obstacle to realizing the potential of CFD to be widely used as an injector design tool. A hierarchical decomposition process is outlined to facilitate the validation process. A simulation readiness level tool used to gauge progress toward the goal is described. Finally, there is a description of the current efforts in each focus area. The background of each focus area is discussed. The state of the art in each focus area is noted along with the TCA VI research focus in the area. Brief highlights of work in the area are also included.

  16. Power Distribution Analysis for the ORNL High Flux Isotope Reactor Critical Experiment 3

    SciTech Connect

    Chandler, David; Primm, Trent; Maldonado, G Ivan

    2010-01-01

    The mission of the Reduced Enrichment for Research and Test Reactors Program is to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors, as well as radioisotope production processes, to low-enriched uranium (LEU) fuel and targets. Oak Ridge National Laboratory (ORNL) is currently reviewing the design bases and key operating criteria including fuel operating parameters, enrichment-related safety analyses, fuel performance, and fuel fabrication in regard to converting the fuel of the High Flux Isotope Reactor (HFIR) from HEU to LEU. The purpose of this study is to validate Monte Carlo methods currently in use for conversion analyses. The methods have been validated for the prediction offlux values in the reactor target, reflector, and beam tubes, but this study focuses on the prediction of the power density profile in the core. Power distributions were calculated in the fuel elements of the HFIR, a research reactor at ORNL, via MCNP and were compared to experimentally obtained data. This study was performed to validate Monte Carlo methods for power density calculations and to observe biases. A current three-dimensional MCNP model was modified to replicate the 1965 HFIR Critical Experiment 3 (HFIRCE-3). In this experiment, the power profile was determined by counting the gamma activity at selected locations in the core. 'Foils' (chunks of fuel meat and clad) were punched out of the fuel elements in HFIRCE-3 following irradiation, and experimental relative power densities were obtained by measuring the activity of these foils and comparing each foil's activity to the activity of a normalizing foil. This analysis consisted of calculating corresponding activities by inserting volume tallies into the modified MCNP model to represent the punchings. The average fission density was calculated for each foil location and then normalized to the reference foil

  17. Monte Carlo Modeling of Fast Sub-critical Assembly with MOX Fuel for Research of Accelerator-Driven Systems

    NASA Astrophysics Data System (ADS)

    Polanski, A.; Barashenkov, V.; Puzynin, I.; Rakhno, I.; Sissakian, A.

    It is considered a sub-critical assembly driven with existing 660 MeV JINR proton accelerator. The assembly consists of a central cylindrical lead target surrounded with a mixed-oxide (MOX) fuel (PuO2 + UO2) and with reflector made of beryllium. Dependence of the energetic gain on the proton energy, the neutron multiplication coefficient, and the neutron energetic spectra have been calculated. It is shown that for subcritical assembly with a mixed-oxide (MOX) BN-600 fuel (28%PuO 2 + 72%UO2) with effective density of fuel material equal to 9 g/cm 3 , the multiplication coefficient keff is equal to 0.945, the energetic gain is equal to 27, and the neutron flux density is 1012 cm˜2 s˜x for the protons with energy of 660 MeV and accelerator beam current of 1 uA.

  18. Burst wait time simulation of CALIBAN reactor at delayed super-critical state

    SciTech Connect

    Humbert, P.; Authier, N.; Richard, B.; Grivot, P.; Casoli, P.

    2012-07-01

    In the past, the super prompt critical wait time probability distribution was measured on CALIBAN fast burst reactor [4]. Afterwards, these experiments were simulated with a very good agreement by solving the non-extinction probability equation [5]. Recently, the burst wait time probability distribution has been measured at CEA-Valduc on CALIBAN at different delayed super-critical states [6]. However, in the delayed super-critical case the non-extinction probability does not give access to the wait time distribution. In this case it is necessary to compute the time dependent evolution of the full neutron count number probability distribution. In this paper we present the point model deterministic method used to calculate the probability distribution of the wait time before a prescribed count level taking into account prompt neutrons and delayed neutron precursors. This method is based on the solution of the time dependent adjoint Kolmogorov master equations for the number of detections using the generating function methodology [8,9,10] and inverse discrete Fourier transforms. The obtained results are then compared to the measurements and Monte-Carlo calculations based on the algorithm presented in [7]. (authors)

  19. A human reliability analysis of the University of New Mexico`s AGN- 201M nuclear research reactor. Revision 1

    SciTech Connect

    Brumburgh, G.P.; Heger, A.S.

    1992-10-15

    During 1990--1991, a probabilistic risk assessment was conducted on the University of New Mexico`s AGN-201M nuclear research reactor to address the risk and consequence of a maximum hypothetical release accident. The assessment indicated a potential for consequential human error to precipitate Chis scenario. Subsequently, a human reliability analysis was performed to evaluate the significance of human interaction on the reactor`s safety systems. This paper presents the results of that investigation.

  20. Promoting University Students' Critical Thinking Skills through Peer Feedback Activity in an Online Discussion Forum

    ERIC Educational Resources Information Center

    Ekahitanond, Visara

    2013-01-01

    This study investigated the impact of the critical inquiry model through peer feedback strategies in an online environment on university students' critical thinking skills and examined their attitudes towards learning through the critical inquiry model and peer feedback strategies. Pre-and post-tests were employed to measure critical thinking…

  1. Evidence for Critical Energy for Ion Confinement in Magnetic Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Maglich, Bogdan; Hester, Tim; Scott, Dan; Calsec Collaboration

    2015-03-01

    It is shown here that fusion test reactors could not ignite for half-a-century because trials were conducted at thermonuclear ion energies 10-30 KeV, an order of magnitude lower than critical energy, Ec ~ 200 KeV. At subcritical energies, plasma is destroyed by neutralization of ions via overlooked atomic (non-nuclear) charge transfer collisions with giant cross-section, 109 barns, 100 times greater than that for ionization collisions that counters neutralization. Neutral injection sets limit on ion magnetic confinement time <10-6 s vs. >1 s required for ignition. In contrast, at energies above Ec, ionization prevails; near ~ 1 MeV, stable confinement of 20 s was routinely observed with charged injection. - To render ITER viable, ion energy must be increased to >/ = 1 MeV; neutral radioactive DT fuel replaced with charged, nonradioactive deuterium, giving rise to compact aneutronicreactor with direct conversion into RF power.

  2. Analyzing the thermionic reactor critical experiments. [thermal spectrum of uranium 235 core

    NASA Technical Reports Server (NTRS)

    Niederauer, G. F.

    1973-01-01

    The Thermionic Reactor Critical Experiments (TRCE) consisted of fast spectrum highly enriched U-235 cores reflected by different thicknesses of beryllium or beryllium oxide with a transition zone of stainless steel between the core and reflector. The mixed fast-thermal spectrum at the core reflector interface region poses a difficult neutron transport calculation. Calculations of TRCE using ENDF/B fast spectrum data and GATHER library thermal spectrum data agreed within about 1 percent for the multiplication factor and within 6 to 8 percent for the power peaks. Use of GAM library fast spectrum data yielded larger deviations. The results were obtained from DOT R Theta calculations with leakage cross sections, by region and by group, extracted from DOT RZ calculations. Delineation of the power peaks required extraordinarily fine mesh size at the core reflector interface.

  3. Neutronics, steady-state, and transient analyses for the Poland MARIA reactor for irradiation testing of LEU lead test fuel assemblies from CERCA : ANL independent verification results.

    SciTech Connect

    Garner, P. L.; Hanan, N. A.

    2011-06-07

    The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decide to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process.

  4. Neutronic assessment of stringer fuel assembly design for liquid-salt-cooledvery high temperature reactor (LS-VHTR).

    SciTech Connect

    Szakaly, F. J.; Kim, T. K.; Taiwo, T. A.

    2006-09-15

    Neutronic studies of 18-pin and 36-pin stringer fuel assemblies have been performed to ascertain that core design requirements for the Liquid-Salt Cooled Very High Temperature Reactor (LS-VHTR) can be met. Parametric studies were performed to determine core characteristics required to achieve a target core cycle length of 18 months and fuel discharge burnup greater than 100 GWd/t under the constraint that the uranium enrichment be less than 20% in order to support non-proliferation goals. The studies were done using the WIMS9 lattice code and the linear reactivity model to estimate the core reactivity balance, fuel composition, and discharge burnup. The results show that the design goals can be met using a 1-batch fuel management scheme, uranium enrichment of 15% and a fuel packing fraction of 30% or greater for the 36-pin stringer fuel assembly design.

  5. The problems of mass transfer and formation of deposits of corrosion products on fuel assemblies of a VVER-1200 reactor

    NASA Astrophysics Data System (ADS)

    Rodionov, Yu. A.; Kritskii, V. G.; Berezina, I. G.; Gavrilov, A. V.

    2014-03-01

    On the basis of examination of materials published both in Russia and abroad, as well as their own investigations, the authors explain the reasons for the occurrence of such effects as AOA (Axial Offset Anomalies) and an increase in the coolant pressure difference in the core of nuclear reactors of the VVER type. To detect the occurrence of the AOA effect, the authors suggest using the specific activity of 58Co in the coolant. In the VVER-1200 design the thermohydraulic regime for fuel assemblies in the first year of their service life involves slight boiling of the coolant in the upper part of the core, which may induce the occurrence of the AOA effect, intensification of corrosion of fuel claddings, and abnormal increase in deposition of corrosion products. Radiolysis of the water coolant in the boiling section (boiling in pores of deposits) may intensify not only general corrosion but also a localized (nodular) one. As a result of intensification of the corrosion processes and growth of deposits, deterioration of the radiation situation in the rooms of the primary circuit of a VVER-1200 reactor as compared to that at nuclear power plants equipped with reactors of the VVER-1000 type is possible. Recommendations for preventing the AOA effect at nuclear power plants with VVER-1200 reactors on the matter of the direction of further investigations are made.

  6. Progress made on the University of Missouri research reactor HEU to LEU fuel conversion feasibility study

    SciTech Connect

    McKibben, J. Charles; Kutikkad, Kiratadas; Foyto, Leslie P.

    2008-07-15

    The University of Missouri Research Reactor (MURR), the highest-powered University-owned research reactor in the U.S. designed to operate at a maximum steady-state power level of 10 MW{sub th}, is one of five U.S. high performance research reactors that use HEU fuel that is actively th collaborating with the U.S. Department of Energy to find a suitable LEU fuel replacement. A conversion feasibility study, using U-10Mo monolithic LEU fuel, is currently being performed at MURR. At first, broad scoping studies where conducted using the transport code MCNP, where the core water-to-metal ratio was varied by altering the thickness or width of the plate cladding, fuel meat, and coolant channel gaps, and varying the number of fuel plates. From these studies, an optimal LEU core design was chosen based on the following calculated parameters: power peaking factors, excess reactivity, and the fast and thermal fluxes available to the experimental facilities. Fuel burnup calculations are now being performed using the 3-D diffusion theory code REBUS. Also included in this paper are some preliminary safety analyses, including parametric studies using the reactivity transient code PARET-ANL and hydraulic calculations using the light- and heavy-water thermal-hydraulic transient code RELAP5/MOD3.3. (author)

  7. Universality of One-Dimensional Fermi Systems, I. Response Functions and Critical Exponents

    NASA Astrophysics Data System (ADS)

    Benfatto, G.; Falco, P.; Mastropietro, V.

    2014-08-01

    The critical behavior of one-dimensional interacting Fermi systems is expected to display universality features, called Luttinger liquid behavior. Critical exponents and certain thermodynamic quantities are expected to be related among each other by model-independent formulas. We establish such relations, the proof of which has represented a challenging mathematical problem, for a general model of spinning fermions on a one dimensional lattice; interactions are short ranged and satisfy a positivity condition which makes the model critical at zero temperature. Proofs are reported in two papers: in the present one, we demonstrate that the zero temperature response functions in the thermodynamic limit are Borel summable and have anomalous power-law decay with multiplicative logarithmic corrections. Critical exponents are expressed in terms of convergent expansions and depend on all the model details. All results are valid for the special case of the Hubbard model.

  8. The dynamin middle domain is critical for tetramerization and higher-order self-assembly

    PubMed Central

    Ramachandran, Rajesh; Surka, Mark; Chappie, Joshua S; Fowler, Douglas M; Foss, Ted R; Song, Byeong Doo; Schmid, Sandra L

    2007-01-01

    The large multidomain GTPase dynamin self-assembles around the necks of deeply invaginated coated pits at the plasma membrane and catalyzes vesicle scission by mechanisms that are not yet completely understood. Although a structural role for the ‘middle' domain in dynamin function has been suggested, it has not been experimentally established. Furthermore, it is not clear whether this putative function pertains to dynamin structure in the unassembled state or to its higher-order self-assembly or both. Here, we demonstrate that two mutations in this domain, R361S and R399A, disrupt the tetrameric structure of dynamin in the unassembled state and impair its ability to stably bind to and nucleate higher-order self-assembly on membranes. Consequently, these mutations also impair dynamin's assembly-dependent stimulated GTPase activity. PMID:17170701

  9. The dynamin middle domain is critical for tetramerization and higher-order self-assembly.

    PubMed

    Ramachandran, Rajesh; Surka, Mark; Chappie, Joshua S; Fowler, Douglas M; Foss, Ted R; Song, Byeong Doo; Schmid, Sandra L

    2007-01-24

    The large multidomain GTPase dynamin self-assembles around the necks of deeply invaginated coated pits at the plasma membrane and catalyzes vesicle scission by mechanisms that are not yet completely understood. Although a structural role for the 'middle' domain in dynamin function has been suggested, it has not been experimentally established. Furthermore, it is not clear whether this putative function pertains to dynamin structure in the unassembled state or to its higher-order self-assembly or both. Here, we demonstrate that two mutations in this domain, R361S and R399A, disrupt the tetrameric structure of dynamin in the unassembled state and impair its ability to stably bind to and nucleate higher-order self-assembly on membranes. Consequently, these mutations also impair dynamin's assembly-dependent stimulated GTPase activity. PMID:17170701

  10. Spatial kinetics in fast reactors

    NASA Astrophysics Data System (ADS)

    Seleznev, E. F.; Belov, A. A.; Panova, I. S.; Matvienko, I. P.; Zhukov, A. M.

    2013-12-01

    The analysis of the solution to the spatial nonstationary equation of neutron transport is presented by the example of a fast reactor. Experiments in spatial kinetics conducted recently at the complex of critical assemblies (fast physical stand) and computations of their data using the TIMER code (for solving the nonstationary equation in multidimensional diffusion approximation for direct and inverse problems of reactor kinetics) have shown that kinetics of fast reactors substantially differs from kinetics of thermal reactors. The difference is connected with influence of the delayed neutron spectrum on rates of the process in a fast reactor.

  11. Participation in the United States Department of Energy University Reactor Instrumentation Program. Final report, September 1990--August 1993

    SciTech Connect

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1993-10-01

    The University of Virginia Reactor Facility is an integral part of the Department of Mechanical, Aerospace and Nuclear Engineering and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and one of the most utilized university research reactor in the mid-Atlantic states. A major objective of this facility is to support educational programs in the region. The University of Virginia has received support under the U.S. Department of Energy (DOE) University Reactor Instrumentation Program every year since 1990. The monies from this program have been used to purchase new equipment to replace outdated or inadequate safety-related instrumentation used in conjunction with reactor operations. This report documents the equipment purchased and the status of the installation and use of this equipment from September 1990 through August 1993. This report constitutes the final report for this project period.

  12. Correlations of Nucleate Boiling Heat Transfer and Critical Heat Flux for External Reactor Vessel Cooling

    SciTech Connect

    J. Yang; F. B. Cheung; J. L. Rempe; K. Y. Suh; S. B. Kim

    2005-07-01

    Four types of steady-state boiling experiments were conducted to investigate the efficacy of two distinctly different heat transfer enhancement methods for external reactor vessel cooling under severe accident conditions. One method involved the use of a thin vessel coating and the other involved the use of an enhanced insulation structure. By comparing the results obtained in the four types of experiments, the separate and integral effect of vessel coating and insulation structure were determined. Correlation equations were obtained for the nucleate boiling heat transfer and the critical heat flux. It was found that both enhancement methods were quite effective. Depending on the angular location, the local critical heat flux could be enhanced by 1.4 to 2.5 times using vessel coating alone whereas it could be enhanced by 1.8 to 3.0 times using an enhanced insulation structure alone. When both vessel coating and insulation structure were used simultaneously, the integral effect on the enhancement was found much less than the product of the two separate effects, indicating possible competing mechanisms (i.e., interference) between the two enhancement methods.

  13. Validation of KENO-VI: A comparison with hexagonal lattice light-water-reactor critical experiments

    SciTech Connect

    Lichtenwalter, J.J.

    1998-06-01

    The KENO-VI Monte Carlo code, released with Version 4.3 of the SCALE Code System, provides the capability to model more complex geometries than previously allowed by KENO-V.a. One significant improvement is the simplistic specification of hexprism unit cells and hexagonal arrays, an arduous task to complete in KENO-V.a. This report documents the validation of KENO-VI against 30 critical experiments consisting of low enriched uranium, light water reactor (LWR) fuel rods in hexagonal lattices with no poisons. The reference, enrichment, pitch, cladding, and core identification of the experiments are given. The results indicate that KENO-VI accurately calculates these critical experiments, with a bias of {minus}0.51% for the 238 group cross section library and {minus}0.24% for the 44 group cross section library. If these biases are properly taken into account, the KENO-VI code can be used with confidence for the design and safety analysis of storage and transportation systems of similar LWR type fuels.

  14. Universal critical behavior of the two-dimensional Ising spin glass

    NASA Astrophysics Data System (ADS)

    Fernandez, L. A.; Marinari, E.; Martin-Mayor, V.; Parisi, G.; Ruiz-Lorenzo, J. J.

    2016-07-01

    We use finite size scaling to study Ising spin glasses in two spatial dimensions. The issue of universality is addressed by comparing discrete and continuous probability distributions for the quenched random couplings. The sophisticated temperature dependency of the scaling fields is identified as the major obstacle that has impeded a complete analysis. Once temperature is relinquished in favor of the correlation length as the basic variable, we obtain a reliable estimation of the anomalous dimension and of the thermal critical exponent. Universality among binary and Gaussian couplings is confirmed to a high numerical accuracy.

  15. Universal free-energy distribution in the critical point of a random Ising ferromagnet.

    PubMed

    Dotsenko, Victor; Holovatch, Yurij

    2014-11-01

    We discuss the non-self-averaging phenomena in the critical point of weakly disordered Ising ferromagnet. In terms of the renormalized replica Ginzburg-Landau Hamiltonian in dimensions D<4, we derive an explicit expression for the probability distribution function (PDF) of the critical free-energy fluctuations. In particular, using known fixed-point values for the renormalized coupling parameters, we obtain the universal curve for such PDF in the dimension D=3. It is demonstrated that this function is strongly asymmetric: its left tail is much slower than the right one. PMID:25493758

  16. Universality and criticality of a second-order granular solid-liquid-like phase transition

    NASA Astrophysics Data System (ADS)

    Castillo, Gustavo; Mujica, Nicolás; Soto, Rodrigo

    2015-01-01

    We experimentally study the critical properties of the nonequilibrium solid-liquid-like transition that takes place in vibrated granular matter. The critical dynamics is characterized by the coupling of the density field with the bond-orientational order parameter Q4, which measures the degree of local crystallization. Two setups are compared, which present the transition at different critical accelerations as a result of modifying the energy dissipation parameters. In both setups five independent critical exponents are measured, associated to different properties of Q4: the correlation length, relaxation time, vanishing wavenumber limit (static susceptibility), the hydrodynamic regime of the pair correlation function, and the amplitude of the order parameter. The respective critical exponents agree in both setups and are given by ν⊥=1 ,ν∥=2 ,γ =1 ,η ≈0.6 -0.67 , and β =1 /2 , whereas the dynamical critical exponent is z =ν∥/ν⊥=2 . The agreement on five exponents is an exigent test for the universality of the transition. Thus, while dissipation is strictly necessary to form the crystal, the path the system undergoes toward the phase separation is part of a well-defined universality class. In fact, the local order shows critical properties while density does not. Being the later conserved, the appropriate model that couples both is model C in the Hohenberg and Halperin classification. The measured exponents are in accord with the nonequilibrium extension to model C if we assume that α , the exponent associated in equilibrium to the specific heat divergence but with no counterpart in this nonequilibrium experiment, vanishes.

  17. Amplitude ratios for critical systems in the c=-2 universality class

    NASA Astrophysics Data System (ADS)

    Izmailian, N. Sh.; Hu, Chin-Kun

    2013-01-01

    We study the finite-size corrections of the critical dense polymer (CDP) and the dimer models on ∞×N rectangular lattice. We find that the finite-size corrections in the CDP and dimer models depend in a crucial way on the parity of N, and a change of the parity of N is equivalent to the change of boundary conditions. We present a set of universal amplitude ratios for amplitudes in finite-size correction terms of critical systems in the universality class with central charge c=-2. The results are in perfect agreement with a perturbated conformal field theory under the assumption that all analytical corrections coming from the operators which belongs to the tower of the identity. Our results inspire many interesting problems for further studies.

  18. Final Assembly and Initial Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect

    S. B. Grover

    2007-05-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing.1,2 The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The final design phase for the first experiment was completed in 2005, and the fabrication and assembly of the first experiment test train (designated AGR-1) as well as the support systems and fission product monitoring system that will monitor and control the experiment

  19. Self-organized criticality in cortical assemblies occurs in concurrent scale-free and small-world networks

    PubMed Central

    Massobrio, Paolo; Pasquale, Valentina; Martinoia, Sergio

    2015-01-01

    The spontaneous activity of cortical networks is characterized by the emergence of different dynamic states. Although several attempts were accomplished to understand the origin of these dynamics, the underlying factors continue to be elusive. In this work, we specifically investigated the interplay between network topology and spontaneous dynamics within the framework of self-organized criticality (SOC). The obtained results support the hypothesis that the emergence of critical states occurs in specific complex network topologies. By combining multi-electrode recordings of spontaneous activity of in vitro cortical assemblies with theoretical models, we demonstrate that different ‘connectivity rules’ drive the network towards different dynamic states. In particular, scale-free architectures with different degree of small-worldness account better for the variability observed in experimental data, giving rise to different dynamic states. Moreover, in relationship with the balance between excitation and inhibition and percentage of inhibitory hubs, the simulated cortical networks fall in a critical regime. PMID:26030608

  20. Results of the MCNP analysis of 20/20 LEU fuel for the Oregon State University TRIGA Reactor

    SciTech Connect

    Dodd, B.; Klein, A.C.; Lewis, B.R.; Merritt, P.A.

    1995-12-31

    The Monte Carlo Neutron/Photon (MCNP) code has been used to perform the neutronics analysis required to support revision of the Oregon State University TRIGA Reactor (OSTR) Safety Analysis Report (SAR). The SAR revision is a necessary part of the preparation of the application for authorization to convert the OSTR core from High Enriched Uranium (HEU) FLIP fuel to a Low Enriched Uranium (LEU) fuel. Before MCNP was applied to LEU-fueled cores, it was first validated by comparing MCNP calculations on FLIP cores to historical, measured values for these cores. The LEU fuel considered was the 20 wt%, 20% enriched (20/20) TRIGA fuel approved by the Nuclear Regulatory Commission (NRC) in NUREG 1282. The results show that the 20/20 fuel is much more reactive than FLIP fuel. A just-critical OSTR FLIP core contains 65 elements, while a just-critical 20/20 core only needs 51 elements. Similarly, the current operational FLIP core consists of 88 elements, whereas a 20/20 core giving the same core excess only requires 65 elements. This presents a significant problem for the OSTR because of potentially significant neutron flux loss in experimental facilities. Further analysis shows that to achieve a full size operational core of about 90 LEU elements the erbium content of the LEU fuel would need to be increased from 0.47wt% to about 0.85 wt%.

  1. Results of the MCNP analysis of 20/20 LEU fuel for the Oregon State University TRIGA reactor

    SciTech Connect

    Dodd, B.; Klein, A.C.; Lewis, B.R.; Merritt, P.A

    1994-07-01

    The Monte Carlo Neutron/Photon (MCNP) code has been used to perform the neutronics analysis required to support revision of the Oregon State University TRIGA Reactor (OSTR) Safety Analysis Report (SAR). The SAR revision is a necessary part of the preparation of the application for authorization to convert the OSTR core from High Enriched Uranium (HEU) FLIP fuel to a Low Enriched Uranium (LEU) fuel. Before MCNP was applied to LEU-fueled cores, it was first validated by comparing MCNP calculations on FLIP cores to historical, measured values for these cores. The LEU fuel considered was the 20 wt%, 20 % enriched (20/20) TRIGA fuel approved by the Nuclear Regulatory Commission (NRC) in NUREG 1282. The results show that the 20/20 fuel is much more reactive than FLIP fuel. A just-critical OSTR FLIP core contains 65 elements, while a just-critical 20/20 core only needs 51 elements. Similarly, the current operational FLIP core consists of 88 elements, whereas a 20/20 core giving the same core excess only requires 65 elements. This presents a significant problem for the OSTR because of potentially significant neutron flux loss in experimental facilities. Further analysis shows that to achieve a full size operational core of about 90 LEU elements the erbium content of the LEU fuel would need to be increased from 0.47 wt% to about 0.85 wt%. (author)

  2. Assemblies with both target and fuel pins in an isotope-production reactor

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

  3. Workshops on Enhancing the Impact of University Research on Critical Technologies. Final report

    SciTech Connect

    Eisenberger, Peter M.

    1998-12-01

    This proposal was designed to initiate a series of workshops in which participants from universities, government and industry would look strategically at selected areas of technological and societal importance. One of these workshops, dedicated to the discussion of how materials modeling can solve critical problems in industry, held at the Institute for Theoretical Physics, University of California Santa Barbara, January 7 through 11, 1996. ''Modeling of Industrial Materials: Connecting Atomistic and Continuum Scales'' was the first of a coordinated series of workshops, spanning over two years, being organized by a group of academic scientists at UCSB, MIT, and the Swiss Federal Institute of Technology Zurich (ETH). On the premise that materials research focused on technology also presents fundamental scientific challenges, the Workshop sought to identify specific areas of industrial needs and opportunities for materials modeling, interpreted as theory and simulation across all the relevant length scales, and to stimulate meaningful university-industry partnerships.

  4. UNFINISHED BUSINESS: The Economics of The Kyoto Protocol

    SciTech Connect

    JA Edmonds; CN MacCracken; RD Sands; SH Kim

    2000-07-06

    The Kyoto Protocol to the Framework Convention on Climate Change (FCCC) was completed on the morning of December 11, 1997, following over two years of negotiations. The product of these deliberations is a complex and incomplete document knitting together the diversity of interests and perspectives represented by the more than 150 delegations. Because the document is complex, its implications are not immediately obvious. If it enters into force, the Kyoto Protocol will have far-reaching implications for all nations--both nations with obligations under the Protocol and those without obligations. National energy systems, and the world's energy system, could be forever changed. In this paper the authors develop an assessment of the energy and economic implications of achieving the goals of the Kyoto Protocol. They find that many of the details of the Protocol that remain to be worked out introduce critical uncertainties affecting the cost of compliance. There are also a variety of uncertainties that further complicate the analysis. These include future non-CO{sub 2} greenhouse gas emissions and the cost of their mitigation. Other uncertainties include the resolution of negotiations to establish rules for determining and allocating land-use emissions rights, mechanisms for Annex 1 trading, and participation by non-Annex 1 members in the Clean Development Mechanism. In addition, there are economic uncertainties, such as the behavior of Eastern Europe and the former Soviet Union in supplying emissions credits under Annex 1 trading. These uncertainties in turn could affect private sector investments in anticipation of the Protocol's entrance into force. The longer the nature of future obligations remains unclear, the less able decision makers will be to incorporate these rules into their investment decisions. They find that the cost of implementing the Protocol in the US can vary by more than an order of magnitude. The marginal cost could be as low as $26 per tonne of

  5. Does Higher Education Foster Critical and Creative Learners? An Exploration of Two Universities in South Korea and the USA

    ERIC Educational Resources Information Center

    Lee, Hye-Jung; Lee, Jihyun; Makara, Kara A.; Fishman, Barry J.; Hong, Young-Il

    2015-01-01

    This paper describes two studies that explore students' beliefs about critical and creative learning at two universities, and considers the implications of those beliefs in comparison to the universities' stated education goals. One is a mixed method study of students at a top university in Korea, and the second is a comparative study…

  6. 239Pu Prompt Fission Neutron Spectra Impact on a Set of Criticality and Experimental Reactor Benchmarks

    NASA Astrophysics Data System (ADS)

    Peneliau, Y.; Litaize, O.; Archier, P.; De Saint Jean, C.

    2014-04-01

    A large set of nuclear data are investigated to improve the calculation predictions of the new neutron transport simulation codes. With the next generation of nuclear power plants (GEN IV projects), one expects to reduce the calculated uncertainties which are mainly coming from nuclear data and are still very important, before taking into account integral information in the adjustment process. In France, future nuclear power plant concepts will probably use MOX fuel, either in Sodium Fast Reactors or in Gas Cooled Fast Reactors. Consequently, the knowledge of 239Pu cross sections and other nuclear data is crucial issue in order to reduce these sources of uncertainty. The Prompt Fission Neutron Spectra (PFNS) for 239Pu are part of these relevant data (an IAEA working group is even dedicated to PFNS) and the work presented here deals with this particular topic. The main international data files (i.e. JEFF-3.1.1, ENDF/B-VII.0, JENDL-4.0, BRC-2009) have been considered and compared with two different spectra, coming from the works of Maslov and Kornilov respectively. The spectra are first compared by calculating their mathematical moments in order to characterize them. Then, a reference calculation using the whole JEFF-3.1.1 evaluation file is performed and compared with another calculation performed with a new evaluation file, in which the data block containing the fission spectra (MF=5, MT=18) is replaced by the investigated spectra (one for each evaluation). A set of benchmarks is used to analyze the effects of PFNS, covering criticality cases and mock-up cases in various neutron flux spectra (thermal, intermediate, and fast flux spectra). Data coming from many ICSBEP experiments are used (PU-SOL-THERM, PU-MET-FAST, PU-MET-INTER and PU-MET-MIXED) and French mock-up experiments are also investigated (EOLE for thermal neutron flux spectrum and MASURCA for fast neutron flux spectrum). This study shows that many experiments and neutron parameters are very sensitive to

  7. On the evaluation of pebble bed reactor critical experiments using the PEBBED code

    SciTech Connect

    Hans D. Gougar; R. Sonat Sen

    2001-10-01

    The PEBBED pebble bed reactor fuel management code under development at the Idaho National Laboratory is designed for rapid design and analysis of pebble bed high temperature reactors (PBRs). Embedded within the code are the THERMIX-KONVEK thermal fluid solver and the COMBINE-7 spectrum generation code for inline cross section homogenization. Because 1D symmetry can be found at each stage of core heterogeneity; spherical at TRISO and pebble levels, and cylindrical at the control rod and core levels, the 1-D transport capability of ANISN is assumed to be sufficient in most cases for generating flux solutions for cross section homogenization. Furthermore, it is fast enough to be executed during the analysis or the equilibrium core. Multi-group diffusion-based design codes such as PEBBED and VSOP are not expected to yield the accuracy and resolution of continuous energy Monte Carlo codes for evaluation of critical experiments. Nonetheless, if the preparation of multigroup cross sections can adequately capture the physics of the mixing of PBR fuel elements and leakage from the core, reasonable results may be obtained. In this paper, results of the application of PEBBED to two critical experiments (HTR Proteus and HTR-10) and associated computational models are presented. The embedded 1-D transport solver is shown to capture the double heterogeneity of the pebble fuel in unit cell calculations. Eigenvalue calculations of a whole core are more challenging, particularly if the boron concentration is uncertain. The sensitivity of major safety parameters to variations in modeling assumptions, however, is shown to be minimal. The embedded transport solver can also be used to obtain control rod worths but only with adjustment of the local spectrum. Results are compared to those of other codes as well as Core 4 of the HTR-Proteus experiment which contains partially inserted rods. They indicate the need for a reference solution to adjust the radius of the graphite in the

  8. Self-Assembled Superparamagnetic Iron Oxide Nanoclusters for Universal Cell Labeling and MRI.

    PubMed

    Chen, Shuzhen; Zhang, Jun; Jiang, Shengwei; Lin, Gan; Luo, Bing; Yao, Huan; Lin, Yuchun; He, Chengyong; Liu, Gang; Lin, Zhongning

    2016-12-01

    Superparamagnetic iron oxide (SPIO) nanoparticles have been widely used in a variety of biomedical applications, especially as contrast agents for magnetic resonance imaging (MRI) and cell labeling. In this study, SPIO nanoparticles were stabilized with amphiphilic low molecular weight polyethylenimine (PEI) in an aqueous phase to form monodispersed nanocomposites with a controlled clustering structure. The iron-based nanoclusters with a size of 115.3 ± 40.23 nm showed excellent performance on cellular uptake and cell labeling in different types of cells, moreover, which could be tracked by MRI with high sensitivity. The SPIO nanoclusters presented negligible cytotoxicity in various types of cells as detected using MTS, LDH, and flow cytometry assays. Significantly, we found that ferritin protein played an essential role in protecting stress from SPIO nanoclusters. Taken together, the self-assembly of SPIO nanoclusters with good magnetic properties provides a safe and efficient method for universal cell labeling with noninvasive MRI monitoring capability. PMID:27216601

  9. Self-assembled biosensor with universal reporter and dual-quenchers for detection of unlabelled nucleic acids.

    PubMed

    Huang, Liming; Aryal, Gyan H; Tam-Chang, Suk-Wah; Publicover, Nelson G; Hunter, Kenneth W

    2016-02-21

    A novel biosensor with universal reporter and dual quenchers was developed for rapid, sensitive, selective, and inexpensive detection of unlabelled nucleic acids. The biosensor is based on a single-strand DNA stem-loop motif with an extended universal reporter-binding region, a G-base rich stem region, and a universal address-binding region. The self-assembly of these stem-loop probes with fluorescence labeled universal reporter and a universal address region conjugated to gold nanoparticles forms the basis of a biosensor for DNA or microRNA targets in solution. The introduction of dual quenchers (G-base quenching and gold surface plasmon resonance-induced quenching) significantly reduces the fluorescence background to as low as 12% of its original fluorescence intensity and hence enhances the detection limit to 0.01 picomoles without signal ampilication. PMID:26757447

  10. Monte Carlo testing of new cross section data sets for thermal and intermediate highly enriched uranium critical assemblies

    SciTech Connect

    Weinman, J.P.

    1998-06-01

    The purpose of this study is to investigate the eigenvalue sensitivity to new {sup 235}U, hydrogen, and oxygen cross section data sets by comparing RACER Monte Carlo calculations for several thermal and intermediate spectrum critical experiments. The new {sup 235}U library (Version 107) was derived by L. Leal and H. Derrien by fitting differential experimental data for {sup 235}U while constraining the fit to match experimental capture and fission resonance integrals and Maxwellian averaged thermal K1 (v fission minus absorption). The new hydrogen library (Version 45) consists of the ENDF/B-VI release 3 data with a 332.0 mb 2,200 m/s cross section which replaces the value of 332.6 mb in the current library. The new oxygen library (Version 39) is based on a recent evaluation of {sup 16}O by E. Caro. Nineteen Oak Ridge and Rocky Flats thermal solution benchmark critical assemblies that span a range of hydrogen-to-{sup 235}U (H/U) concentrations (2,052 to 27.1) and above-thermal neutron leakage fractions (0.555 to 0.011) were analyzed. In addition, three intermediate spectrum critical assemblies (UH3-UR, UH3-NI, and HISS-HUG) were studied.

  11. The Intense Slow Positron Beam Facility at the NC State University PULSTAR Reactor

    SciTech Connect

    Hawari, Ayman I.; Moxom, Jeremy; Hathaway, Alfred G.; Brown, Benjamin; Xu, Jun

    2009-03-10

    An intense slow positron beam is in its early stages of operation at the 1-MW open-pool PULSTAR research reactor at North Carolina State University. The positron beam line is installed in a beam port that has a 30-cmx30-cm cross sectional view of the core. The positrons are created in a tungsten converter/moderator by pair-production using gamma rays produced in the reactor core and by neutron capture reactions in cadmium cladding surrounding the tungsten. Upon moderation, slow ({approx}3 eV) positrons that are emitted from the moderator are electrostatically extracted, focused and magnetically guided until they exit the reactor biological shield with 1-keV energy, approximately 3-cm beam diameter and an intensity exceeding 6x10{sup 8} positrons per second. A magnetic beam switch and transport system has been installed and tested that directs the beam into one of two spectrometers. The spectrometers are designed to implement state-of-the-art PALS and DBS techniques to perform positron and positronium annihilation studies of nanophases in matter.

  12. Micro-Pocket Fission Detectors (MPFD) For Fuel Assembly Analysis

    SciTech Connect

    Troy Unruh; Michael Reichenberger; Phillip Ugorowski

    2013-09-01

    Neutron sensors capable of real-time measurement of thermal flux, fast flux, and temperature in a single miniaturized probe are needed in irradiation tests required to demonstrate the performance of candidate new fuels, and cladding materials. In-core ceramic-based miniature neutron detectors or “Micro-Pocket Fission Detectors” (MPFDs) have been studied at Kansas State University (KSU). The first MPFD prototypes were tested in various neutron fields at the KSU TRIGA research reactor with successful results. Currently, a United States Department of Energy-sponsored joint KSU/Idaho National Laboratory (INL) effort is underway to develop a high-temperature, high-pressure version of the MPFD using radiation-resistant, high temperature materials, which would be capable of withstanding irradiation test conditions in high performance material and test reactors (MTRs). Ultimately, this more compact, more accurate, and longer lifetime flux sensor for critical mock-ups, existing and advanced reactor designs, high performance MTRs, and transient test reactors has the potential to lead to higher accuracy and resolution data from irradiation testing, more detailed core flux measurements and enhanced fuel assembly processing. Prior evaluations by KSU indicate that these sensors could also be used to monitor burn-up of nuclear fuel. If integrated into nuclear fuel assemblies, MPFDs offer several advantages to current spent fuel management systems.

  13. Conversion of the University of Virginia reactor to low-enrichment fuel

    SciTech Connect

    Rydin, R.A.; Freeman, D.W.; Fehr, M.K.

    1989-01-01

    The University of Virginia is using the LEOPARD-LINX-TWODB computational package to assist in converting the 2-MW(thermal) University of Virginia Reactor (UVAR) to low-enrichment uranium fuel (LEU). Recent efforts have been focused on designing an upgraded LEU replacement core. The UVAR is unique in that anywhere from 16 to 27 fuel elements have been used over the years in a variety of geometrical arrangements. However, a compact and fixed core arrangement offers significant operational and experimental advantages over past practice, including a higher and more temporally constant thermal flux. The authors have concentrated on relative burnup comparisons between representative 4 {times} 4, 4 {times} 5, and 5 {times} 5 cores using 18 curved-plant/element HEU (HEU-18) and 18 and 22 flat-plate/element LEU (LEU-18 and LEU-22) fuel. It is concluded that the UVAR should be supplied with LEU-22 fuel. The authors have also explored the reactivity and power-peaking effects of separately varying the reflector compositions on various reactor faces from light water to heavy water to graphite. Thermal-hydraulic behavior has also been analyzed with a combination of the PARET and THERHYD codes, which indicates the LEU cores will be acceptable with only small modifications in UVAR safety system settings.

  14. The Prompt Fission Neutron Spectrum: From Experiment to the Evaluated Data and its Impact on Critical Assemblies

    SciTech Connect

    Rising, Michael Evan

    2015-06-10

    After a brief introduction concerning nuclear data, prompt fission neutron spectrum (PFNS) evaluations and the limited PFNS covariance data in the ENDF/B-VII library, and the important fact that cross section uncertainties ~ PFNS uncertainties, the author presents background information on the PFNS (experimental data, theoretical models, data evaluation, uncertainty quantification) and discusses the impact on certain well-known critical assemblies with regard to integral quantities, sensitivity analysis, and uncertainty propagation. He sketches recent and ongoing research and concludes with some final thoughts.

  15. Critical heat flux in natural convection cooled TRIGA reactors with hexagonal bundle

    SciTech Connect

    Yang, J.; Avery, M.; De Angelis, M.; Anderson, M.; Corradini, M.; Feldman, E. E.; Dunn, F. E.; Matos, J. E.

    2012-07-01

    A three-rod bundle Critical Heat Flux (CHF) study at low flow, low pressure, and natural convection condition has been conducted, simulating TRIGA reactors with the hexagonally configured core. The test section is a custom-made trefoil shape tube with three identical fuel pin heater rods located symmetrically inside. The full scale fuel rod is electrically heated with a chopped-cosine axial power profile. CHF experiments were carried out with the following conditions: inlet water subcooling from 30 K to 95 K; pressure from 110 kPa to 230 kPa; mass flux up to 150 kg/m{sup 2}s. About 50 CHF data points were collected and compared with a few existing CHF correlations whose application ranges are close to the testing conditions. Some tests were performed with the forced convection to identify the potential difference between the CHF under the natural convection and forced convection. The relevance of the CHF to test parameters is investigated. (authors)

  16. Criticality accident dosimetry systems: an international intercomparison at the SILENE reactor in 2002.

    PubMed

    Médioni, R; Asselineau, B; Verrey, B; Trompier, F; Itié, C; Texier, C; Muller, H; Pelcot, G; Clairand, I; Jacquet, X; Pochat, J L

    2004-01-01

    In criticality accident dosimetry and more generally for high dose measurements, special techniques are used to measure separately the gamma ray and neutron components of the dose. To improve these techniques and to check their dosimetry systems (physical and/or biological), a total of 60 laboratories from 29 countries (America, Europe, Asia) participated in an international intercomparaison, which took place in France from 9 to 21 June 2002, at the SILENE reactor in Valduc and at a pure gamma source in Fontenay-aux-Roses. This intercomparison was jointly organised by the IRSN and the CEA with the help of the NEA/OCDE and was partly supported by the European Communities. This paper describes the aim of this intercomparison, the techniques used by the participants and the two radiation sources and their characteristics. The experimental arrangements of the dosemeters for the irradiations in free air or on phantoms are given. Then the dosimetric quantities measured and reported by the participants are summarised, analysed and compared with the reference values. The present paper concerns only the physical dosimetry and essentially experiments performed on the SILENE facility. The results obtained with the biological dosimetry are published in two other papers of this issue. PMID:15353686

  17. Electron versus proton accelerator driven sub-critical system performance using TRIGA reactors at power

    SciTech Connect

    Carta, M.; Burgio, N.; D'Angelo, A.; Santagata, A.; Petrovich, C.; Schikorr, M.; Beller, D.; Felice, L. S.; Imel, G.; Salvatores, M.

    2006-07-01

    This paper provides a comparison of the performance of an electron accelerator-driven experiment, under discussion within the Reactor Accelerator Coupling Experiments (RACE) Project, being conducted within the U.S. Dept. of Energy's Advanced Fuel Cycle Initiative (AFCI), and of the proton-driven experiment TRADE (TRIGA Accelerator Driven Experiment) originally planned at ENEA-Casaccia in Italy. Both experiments foresee the coupling to sub-critical TRIGA core configurations, and are aimed to investigate the relevant kinetic and dynamic accelerator-driven systems (ADS) core behavior characteristics in the presence of thermal reactivity feedback effects. TRADE was based on the coupling of an upgraded proton cyclotron, producing neutrons via spallation reactions on a tantalum (Ta) target, with the core driven at a maximum power around 200 kW. RACE is based on the coupling of an Electron Linac accelerator, producing neutrons via photoneutron reactions on a tungsten-copper (W-Cu) or uranium (U) target, with the core driven at a maximum power around 50 kW. The paper is focused on analysis of expected dynamic power response of the RACE core following reactivity and/or source transients. TRADE and RACE target-core power coupling coefficients are compared and discussed. (authors)

  18. Critical Heat Flux Experiments on the Reactor Vessel Wall Using 2-D Slice Test Section

    SciTech Connect

    Jeong, Yong Hoon; Chang, Soon Heung; Baek, Won-Pil

    2005-11-15

    The critical heat flux (CHF) on the reactor vessel outer wall was measured using the two-dimensional slice test section. The radius and the channel area of the test section were 2.5 m and 10 cm x 15 cm, respectively. The flow channel area and the heater width were smaller than those of the ULPU experiments, but the radius was greater than that of the ULPU. The CHF data under the inlet subcooling of 2 to 25 deg. C and the mass flux 0 to 300 kg/m{sup 2}.s had been acquired. The measured CHF value was generally slightly lower than that of the ULPU. The difference possibly comes from the difference of the test section material and the thickness. However, the general trend of CHF according to the mass flux was similar with that of the ULPU. The experimental CHF data were compared with the predicted values by SULTAN correlation. The SULTAN correlation predicted well this study's data only for the mass flux higher than 200 kg/m{sup 2}.s, and for the exit quality lower than 0.05. The local condition-based correlation was developed, and it showed good prediction capability for broad quality (-0.01 to 0.5) and mass flux (<300 kg/m{sup 2}.s) conditions with a root-mean-square error of 2.4%. There were increases in the CHF with trisodium phosphate-added water.

  19. Thermodynamic signature of quantum criticality: universally diverging Grüneisen ratio

    NASA Astrophysics Data System (ADS)

    Zhu, Lijun

    2005-03-01

    At a generic quantum critical point where pressure acts as (or couples to) the zero-temperature control parameter, the Grüneisen ratio γ (the ratio of thermal expansion to specific heat) is divergent[1]. This property provides a novel probe to quantum criticality from thermodynamics. When scaling applies, γ˜1/T^x at the critical pressure p=pc, where the exponent x measures the scaling dimension of the most singular operator coupled to pressure; in the alternative limit T ->0 and p !=pc, γ= Gr/(p-pc), where Gr is a universal combination of critical exponents. The predicted divergence has been observed near the quantum critical points of several heavy fermion metals[2]. Analyses based on specific models relevant to these experiments are also presented. [1] L. Zhu, M. Garst, A. Rosch, and Q. Si, Phys. Rev. Lett. 91, 066404 (2003). [2] R. Küchler et al., Phys. Rev. Lett. 91, 066405 (2003); ibid. 93, 096402 (2004).

  20. Thoughts on Sensitivity Analysis and Uncertainty Propagation Methods with Respect to the Prompt Fission Neutron Spectrum Impact on Critical Assemblies

    SciTech Connect

    Rising, M.E.

    2015-01-15

    The prompt fission neutron spectrum (PFNS) uncertainties in the n+{sup 239}Pu fission reaction are used to study the impact on several fast critical assemblies modeled in the MCNP6.1 code. The newly developed sensitivity capability in MCNP6.1 is used to compute the k{sub eff} sensitivity coefficients with respect to the PFNS. In comparison, the covariance matrix given in the ENDF/B-VII.1 library is decomposed and randomly sampled realizations of the PFNS are propagated through the criticality calculation, preserving the PFNS covariance matrix. The information gathered from both approaches, including the overall k{sub eff} uncertainty, is statistically analyzed. Overall, the forward and backward approaches agree as expected. The results from a new method appear to be limited by the process used to evaluate the PFNS and is not necessarily a flaw of the method itself. Final thoughts and directions for future work are suggested.

  1. The calculation of the YALINA BOOSTER zero power sub critical assembly driven by external neutron sources: Brazillian contribution

    NASA Astrophysics Data System (ADS)

    Carluccio, Thiago; Rossi, Pedro Carlos Russo; Maiorino, José Rubens

    2011-08-01

    The YALINA-Booster is an experimental zero power Accelerator Driven Reactor (ADS), which consists of a sub-critical assemby driven by external neutron sources. It has a fast spectrum booster zone in the center, surrounded by a thermal one. The sub-critical core is driven by external neutron sources. Several experiments have been proposed in the framework of IAEA Coordinated Reserch Project (CRP) on ADS. This work shows results obtained by IPEN modelling and simulating experiments proposed at CRP, using the MCNP code. The comparison among our results, the experimental one and the results obtained by other participants is being done by CRP coordinators. This coolaborative work has an important role in the qualification and improvement of calculational methodologies.

  2. U.S. Department of Energy University Reactor Sharing Program at the University of Florida. Final report for period August 15, 2000 - May 31, 2001

    SciTech Connect

    Vernetson, William G.

    2002-01-01

    Department of Energy Grant Number DE-FG02-96NE38152 was supplied to the University of Florida Training Reactor (UFTR) facility through the U.S. Department of Energy's University Reactor Sharing Program. The renewal proposal submitted in January 2000 originally requested over $73,000 to support various external educational institutions using the UFTR facilities in academic year 2000-01. The actual Reactor Sharing Grant was only in the amount of $40,000, all of which has been well used by the University of Florida as host institution to support various educational institutions in the use of our reactor and associated facilities as indicated in the proposal. These various educational institutions are located primarily within the State of Florida. However, when the 600-mile distance from Pensacola to Miami is considered, it is obvious that this Grant provides access to reactor utilization for a broad geographical region and a diverse set of user institutions serving over fourteen million inhabitants throughout the State of Florida and still others throughout the Southeast.

  3. History of critical experiments at Pajarito Site

    SciTech Connect

    Paxton, H.C.

    1983-03-01

    This account describes critical and subcritical assemblies operated remotely at the Pajarito Canyon Site at the Los Alamos National Laboratory. Earliest assemblies, directed exclusively toward the nuclear weapons program, were for safety tests. Other weapon-related assemblies provided neutronic information to check detailed weapon calculations. Topsy, the first of these critical assemblies, was followed by Lady Godiva, Jezebel, Flattop, and ultimately Big Ten. As reactor programs came to Los Alamos, design studies and mockups were tested at Pajarito Site. For example, nearly all 16 Rover reactors intended for Nevada tests were preceded by zero-power mockups and proof tests at Pajarito Site. Expanded interest and capability led to fast-pulse assemblies, culminating in Godiva IV and Skua, and to the Kinglet and Sheba solution assemblies.

  4. Experimental studies of local coolant hydrodynamics using a scaled model of cassette-type fuel assembly of a KLT-40S reactor

    NASA Astrophysics Data System (ADS)

    Dmitriev, S. M.; Barinov, A. A.; Varentsov, A. V.; Doronkov, D. V.; Solntsev, D. N.; Khrobostov, A. E.

    2016-08-01

    The results of experimental studies of local hydrodynamic and mass exchange characteristics of the coolant flow behind the spacer grid in the fuel assembly of a KLT-40S reactor are presented. The experiments were aimed at the investigation of representative domains of the fuel assembly with three tracer injection regions. The studies were performed at the aerodynamic test facility using the tracer gas diffusion method. According to the theory of hydrodynamic similarity, the obtained experimental results can be transferred to full-scale coolant flow conditions in standard fuel assemblies. The analysis of the tracer concentration propagation made it possible to determine in detail the flow pattern and find the main regularities and specific features of the coolant flow behind the plate spacer grid of KLT-40S fuel assembly. The hydraulic resistance coefficient of the spacer grid was experimentally determined. The coefficients of mass exchange between cells for representative cells of the displacer region in the KLT-40S fuel assembly were calculated for the first time; these results are presented in the form of the "mixing matrix." The results of studies of local coolant flow hydrodynamics in the KLT-40S fuel assembly are used at AO Afrikantov OKBM for estimation of thermotechnical reliability of active cores for reactors of floating nuclear power stations. The experimental data on hydrodynamic and mass exchange characteristics are included in the database for verification of CDF codes and detailed cell-wise calculation of the active core for KLT-40S reactor installation. The results of these studies can be used at FSUE RFNC-VNIIEF for testing and verification of domestic three-dimensional hydrodynamic CFD codes ("Logos") that are applied for substantiation of newly designed reactor installations. Practical recommendations on the application of the obtained results in thermohydraulic calculations of the active core for the KLT-40S reactor will be worked out. Proposals

  5. Cosmological moduli and the post-inflationary universe: A critical review

    NASA Astrophysics Data System (ADS)

    Kane, Gordon; Sinha, Kuver; Watson, Scott

    2015-06-01

    We critically review the role of cosmological moduli in determining the post-inflationary history of the universe. Moduli are ubiquitous in string and M-theory constructions of beyond the Standard Model physics, where they parametrize the geometry of the compactification manifold. For those with masses determined by supersymmetry (SUSY) breaking this leads to their eventual decay slightly before Big Bang nucleosynthesis (BBN) (without spoiling its predictions). This results in a matter dominated phase shortly after inflation ends, which can influence baryon and dark matter genesis, as well as observations of the cosmic microwave background (CMB) and the growth of large-scale structure. Given progress within fundamental theory, and guidance from dark matter and collider experiments, nonthermal histories have emerged as a robust and theoretically well-motivated alternative to a strictly thermal one. We review this approach to the early universe and discuss both the theoretical challenges and the observational implications.

  6. Universal Pulse Oximetry Screening for Early Detection of Critical Congenital Heart Disease

    PubMed Central

    Kumar, Praveen

    2016-01-01

    Critical congenital heart disease (CCHD) is a major cause of infant death and morbidity worldwide. An early diagnosis and timely intervention can significantly reduce the likelihood of an adverse outcome. However, studies from the United States and other developed countries have shown that as many as 30%–50% of infants with CCHD are discharged after birth without being identified. This diagnostic gap is likely to be even higher in low-resource countries. Several large randomized trials have shown that the use of universal pulse-oximetry screening (POS) at the time of discharge from birth hospital can help in early diagnosis of these infants. The objective of this review is to share data to show that the use of POS for early detection of CCHD meets the criteria necessary for inclusion to the universal newborn screening panel and could be adopted worldwide. PMID:27279759

  7. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    SciTech Connect

    Broadhead, B.L.

    1991-08-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications.

  8. Universal conductivity in a two-dimensional superfluid-to-insulator quantum critical system.

    PubMed

    Chen, Kun; Liu, Longxiang; Deng, Youjin; Pollet, Lode; Prokof'ev, Nikolay

    2014-01-24

    We compute the universal conductivity of the (2+1)-dimensional XY universality class, which is realized for a superfluid-to-Mott insulator quantum phase transition at constant density. Based on large-scale Monte Carlo simulations of the classical (2+1)-dimensional J-current model and the two-dimensional Bose-Hubbard model, we can precisely determine the conductivity on the quantum critical plateau, σ(∞) = 0.359(4)σQ with σQ the conductivity quantum. The universal conductivity curve is the standard example with the lowest number of components where the bottoms-up AdS/CFT correspondence from string theory can be tested and made to use [R. C. Myers, S. Sachdev, and A. Singh, Phys. Rev. D 83, 066017 (2011)]. For the first time, the shape of the σ(iω(n)) - σ(∞) function in the Matsubara representation is accurate enough for a conclusive comparison and establishes the particlelike nature of charge transport. We find that the holographic gauge-gravity duality theory for transport properties can be made compatible with the data if temperature of the horizon of the black brane is different from the temperature of the conformal field theory. The requirements for measuring the universal conductivity in a cold gas experiment are also determined by our calculation. PMID:24484123

  9. Conformal perturbation of off-critical correlators in the 3D Ising universality class

    NASA Astrophysics Data System (ADS)

    Caselle, M.; Costagliola, G.; Magnoli, N.

    2016-07-01

    Thanks to the impressive progress of conformal bootstrap methods we have now very precise estimates of both scaling dimensions and operator product expansion coefficients for several 3D universality classes. We show how to use this information to obtain similarly precise estimates for off-critical correlators using conformal perturbation. We discuss in particular the ⟨σ (r )σ (0 )⟩ , ⟨ɛ (r )ɛ (0 )⟩ and ⟨σ (r )ɛ (0 )⟩ two-point functions in the high and low temperature regimes of the 3D Ising model and evaluate the leading and next to leading terms in the s =trΔt expansion, where t is the reduced temperature. Our results for ⟨σ (r )σ (0 )⟩ agree both with Monte Carlo simulations and with a set of experimental estimates of the critical scattering function.

  10. Maintaining a Critical Spectra within Monteburns for a Gas-Cooled Reactor Array by Way of Control Rod Manipulation

    DOE PAGESBeta

    Adigun, Babatunde John; Fensin, Michael Lorne; Galloway, Jack D.; Trellue, Holly Renee

    2016-06-07

    Our burnup study examined the effect of a predicted critical control rod position on the nuclide predictability of several axial and radial locations within a 4×4 graphite moderated gas cooled reactor fuel cluster geometry. To achieve this, a control rod position estimator (CRPE) tool was developed within the framework of the linkage code Monteburns between the transport code MCNP and depletion code CINDER90, and four methodologies were proposed within the tool for maintaining criticality. Two of the proposed methods used an inverse multiplication approach - where the amount of fissile material in a set configuration is slowly altered until criticalitymore » is attained - in estimating the critical control rod position. Another method carried out several MCNP criticality calculations at different control rod positions, then used a linear fit to estimate the critical rod position. The final method used a second-order polynomial fit of several MCNP criticality calculations at different control rod positions to guess the critical rod position. The results showed that consistency in prediction of power densities as well as uranium and plutonium isotopics was mutual among methods within the CRPE tool that predicted critical position consistently well. Finall, while the CRPE tool is currently limited to manipulating a single control rod, future work could be geared toward implementing additional criticality search methodologies along with additional features.« less

  11. The NTD-CTD intersubunit interface plays a critical role in assembly and stabilization of the HIV-1 capsid

    PubMed Central

    2013-01-01

    Background Lentiviruses exhibit a cone-shaped capsid composed of subunits of the viral CA protein. The intrinsic stability of the capsid is critical for HIV-1 infection, since both stabilizing and destabilizing mutations compromise viral infectivity. Structural studies have identified three intersubunit interfaces in the HIV-1 capsid, two of which have been previously studied by mutational analysis. In this present study we analyzed the role of a third interface, that which is formed between the amino terminal domain (NTD) and carboxyl terminal domain (CTD) of adjacent subunits. Results We provided evidence for the presence of the NTD-CTD interface in HIV-1 particles by engineering intersubunit NTD-CTD disulfide crosslinks, resulting in accumulation of disulfide-linked oligomers up to hexamers. We also generated and characterized a panel of HIV-1 mutants containing substitutions at this interface. Some mutants showed processing defects and altered morphology from that of wild type, indicating that the interface is important for capsid assembly. Analysis of these mutants by transmission electron microscopy corroborated the importance of this interface in assembly. Other mutants exhibited quantitative changes in capsid stability, many with unstable capsids, and one mutant with a hyperstable capsid. Analysis of the mutants for their capacity to saturate TRIMCyp-mediated restriction in trans confirmed that the unstable mutants undergo premature uncoating in target cells. All but one of the mutants were markedly attenuated in replication owing to impaired reverse transcription in target cells. Conclusions Our results demonstrate that the NTD-CTD intersubunit interface is present in the mature HIV-1 capsid and is critical for proper capsid assembly and stability. PMID:23497318

  12. The National Criticality Experiments Research Center at the Device Assembly Facility, Nevada National Security Site: Status and Capabilities, Summary Report

    SciTech Connect

    S. Bragg-Sitton; J. Bess; J. Werner

    2011-09-01

    The National Criticality Experiments Research Center (NCERC) was officially opened on August 29, 2011. Located within the Device Assembly Facility (DAF) at the Nevada National Security Site (NNSS), the NCERC has become a consolidation facility within the United States for critical configuration testing, particularly those involving highly enriched uranium (HEU). The DAF is a Department of Energy (DOE) owned facility that is operated by the National Nuclear Security Agency/Nevada Site Office (NNSA/NSO). User laboratories include the Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL). Personnel bring their home lab qualifications and procedures with them to the DAF, such that non-site specific training need not be repeated to conduct work at DAF. The NNSS Management and Operating contractor is National Security Technologies, LLC (NSTec) and the NNSS Safeguards and Security contractor is Wackenhut Services. The complete report provides an overview and status of the available laboratories and test bays at NCERC, available test materials and test support configurations, and test requirements and limitations for performing sub-critical and critical tests. The current summary provides a brief summary of the facility status and the method by which experiments may be introduced to NCERC.

  13. Low enrichment fuel conversion for Iowa State University

    SciTech Connect

    Rohach, A.F.

    1992-08-01

    This report discusses the UTR-10 reactor at Iowa State University which went critical on low enriched uranium (LEU) fuel on August 14, 1991. However, subsequent to the criticality experiments the fuel plates started to discolor. In addition, roll pins used to lift the fuel assemblies were discovered to be cracked. It was determined that these problems were due to chemical agents in the primary coolant water. The roll pins were replaced by solid stainless steel pins. The primary coolant was replaced and the reactor is currently in operation. Surveillance specimens will be used to monitor any possible future discoloration. The high enriched fuel (HEU) is being prepared for eventual shipment to a high enriched fuel receiving facility.

  14. Characterization of Carbon Particulates in the Exit Flow of a Plasma Pyrolysis Assembly (PPA) Reactor

    NASA Technical Reports Server (NTRS)

    Green, Robert D.; Meyer, Marit E.; Agui, Juan H.; Berger, Gordon M.; Vijayakumar, R.; Abney, Morgan B.; Greenwood, Zachary

    2015-01-01

    The ISS presently recovers oxygen from crew respiration via a Carbon Dioxide Reduction Assembly (CRA) that utilizes the Sabatier chemical process to reduce captured carbon dioxide to methane (CH4) and water. In order to recover more of the hydrogen from the methane and increase oxygen recovery, NASA Marshall Space Flight Center (MSFC) is investigating a technology, plasma pyrolysis, to convert the methane to acetylene. The Plasma Pyrolysis Assembly (or PPA), achieves 90% or greater conversion efficiency, but a small amount of solid carbon particulates are generated as a side product and must be filtered before the acetylene is removed and the hydrogen-rich gas stream is recycled back to the CRA. In this work, we present the experimental results of an initial characterization of the carbon particulates in the PPA exit gas stream. We also present several potential options to remove these carbon particulates via carbon traps and filters to minimize resupply mass and required downtime for regeneration.

  15. A kinetic model for impact/sliding wear of pressurized water reactor internal components: Application to rod cluster control assemblies

    SciTech Connect

    Zbinden, M.; Durbec, V.

    1996-12-01

    Certain internal components of Pressurized Water Reactors are damaged by wear when subjected to vibration induced by flow. In order to enable predictive calculation of such wear, one must have a model which takes account reliably of real damages. The modelling of wear represents a final link in a succession of numerical calculations which begins by the determination of hydraulic excitations induced by the flow. One proceeds, then, in the dynamic response calculation of the structure to finish up with an estimation of volumetric wear and of the depth of wear scars. A new concept of industrial wear model adapted to components of nuclear plants is proposed. Its originality is to be supported, on one hand, by experimental results obtained via wear machines of relatively short operational times, and, on the other hand, by the information obtained from the operating feedback over real wear kinetics of the reactors components. The proposed model is illustrated by an example which correspond to a specific real situation. The determination of the coefficients permitting to cover all assembly of configurations and the validation of the model in these configurations have been the object of the most recent work.

  16. Monte Carlo cross section testing for thermal and intermediate {sup 235}U/{sup 238}U critical assemblies, ENDF/B-V vs ENDF/B-VI

    SciTech Connect

    Weinman, J.P.

    1997-06-01

    The purpose of this study is to investigate the eigenvalue sensitivity to changes in ENDF/B-V and ENDF/B-VI cross section data sets by comparing RACER vectorized Monte Carlo calculations for several thermal and intermediate spectrum critical experiments. Nineteen Oak Ridge and Rocky Flats thermal solution benchmark critical assemblies that span a range of hydrogen-to-{sup 235}U (H/U) concentrations (2052 to 27.1) and above-thermal neutron leakage fractions (0.555 to 0.011) were analyzed. In addition, three intermediate spectrum critical assemblies (UH3-UR, UH3-NI, and HISS-HUG) were studied.

  17. Critical Issues Facing America's Community Colleges: A Summary of the Community College Futures Assembly 2008

    ERIC Educational Resources Information Center

    Basham, Matthew J.; Campbell, Dale F.; Mendoza, Pilar

    2008-01-01

    Three focus groups consisting of board of trustee members, community college presidents, senior administrators, administrators, and faculty members developed critical issues facing community colleges with respect to instructional planning and services; planning, governance, and finance; and workforce development. Thereafter, the delegation of more…

  18. Critical Issues Facing America's Community Colleges: A Summary of the Community Colleges Futures Assembly 2006

    ERIC Educational Resources Information Center

    Campbell, Dale F.; Basham, Matthew J.

    2007-01-01

    Three focus groups consisting of 42 board of trustee members, community college presidents, senior administrators, and faculty members developed critical issues facing community colleges with respect to instructional planning and services; planning, governance, finance; and workforce development. Thereafter, the delegation of more than 200 voted…

  19. DOE/NE University Program in robotics for advanced reactors research

    SciTech Connect

    Trivedi, M.M.

    1990-01-01

    The document presents the bimonthly progress reports published during 1990 regarding the US Department of Energy/NE-sponsored research at the University of Tennessee Knoxville under the DOE Robitics for Advanced Reactors Research Grant. Significant accomplishments are noted in the following areas: development of edge-segment based stereo matching algorithm; vision system integration in the CESAR laboratory; evaluation of algorithms for surface characterization from range data; comparative study of data fusion techniques; development of architectural framework, software, and graphics environment for sensor-based robots; algorithms for acquiring tactile images from planer surfaces; investigations in geometric model-based robotic manipulation; investigations of non-deterministic approaches to sensor fusion; and evaluation of sensor calibration techniques. (MB)

  20. LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    SciTech Connect

    Bennett, W D; Goodman, R L; Heaberlin, S W; Hesson, G M; Nealley, C; Kirg, L L; Marshall, R K; McNair, G W; Meitzler, W D; Neally, G W; Parchen, L J; Pilger, J P; Rausch, W N; Russcher, G E; Schreiber, R E; Wildung, N J; Wilson, C L

    1981-02-01

    Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.

  1. Unusual corrections to scaling and convergence of universal Renyi properties at quantum critical points

    NASA Astrophysics Data System (ADS)

    Sahoo, Sharmistha; Stoudenmire, E. Miles; Stéphan, Jean-Marie; Devakul, Trithep; Singh, Rajiv R. P.; Melko, Roger G.

    2016-02-01

    At a quantum critical point, bipartite entanglement entropies have universal quantities which are subleading to the ubiquitous area law. For Renyi entropies, these terms are known to be similar to the von Neumann entropy, while being much more amenable to numerical and even experimental measurement. We show here that when calculating universal properties of Renyi entropies, it is important to account for unusual corrections to scaling that arise from relevant local operators present at the conical singularity in the multisheeted Riemann surface. These corrections grow in importance with increasing Renyi index. We present studies of Renyi correlation functions in the 1 +1 transverse-field Ising model (TFIM) using conformal field theory, mapping to free fermions, and series expansions, and the logarithmic entropy singularity at a corner in 2 +1 for both free bosonic field theory and the TFIM, using numerical linked cluster expansions. In all numerical studies, accurate results are only obtained when unusual corrections to scaling are taken into account. In the worst case, an analysis ignoring these corrections can get qualitatively incorrect answers, such as predicting a decrease in critical exponents with the Renyi index, when they are actually increasing. We discuss a two-step extrapolation procedure that can be used to account for the unusual corrections to scaling.

  2. The Caenorhabditis elegans protein SAS-5 forms large oligomeric assemblies critical for centriole formation

    PubMed Central

    Rogala, Kacper B; Dynes, Nicola J; Hatzopoulos, Georgios N; Yan, Jun; Pong, Sheng Kai; Robinson, Carol V; Deane, Charlotte M; Gönczy, Pierre; Vakonakis, Ioannis

    2015-01-01

    Centrioles are microtubule-based organelles crucial for cell division, sensing and motility. In Caenorhabditis elegans, the onset of centriole formation requires notably the proteins SAS-5 and SAS-6, which have functional equivalents across eukaryotic evolution. Whereas the molecular architecture of SAS-6 and its role in initiating centriole formation are well understood, the mechanisms by which SAS-5 and its relatives function is unclear. Here, we combine biophysical and structural analysis to uncover the architecture of SAS-5 and examine its functional implications in vivo. Our work reveals that two distinct self-associating domains are necessary to form higher-order oligomers of SAS-5: a trimeric coiled coil and a novel globular dimeric Implico domain. Disruption of either domain leads to centriole duplication failure in worm embryos, indicating that large SAS-5 assemblies are necessary for function in vivo. DOI: http://dx.doi.org/10.7554/eLife.07410.001 PMID:26023830

  3. Secondary Structure Transition and Critical Stress for a Model of Spider Silk Assembly.

    PubMed

    Giesa, Tristan; Perry, Carole C; Buehler, Markus J

    2016-02-01

    Spiders spin their silk from an aqueous solution to a solid fiber in ambient conditions. However, to date, the assembly mechanism in the spider silk gland has not been satisfactorily explained. In this paper, we use molecular dynamics simulations to model Nephila clavipes MaSp1 dragline silk formation under shear flow and determine the secondary structure transitions leading to the experimentally observed fiber structures. While no experiments are performed on the silk fiber itself, insights from this polypeptide model can be transferred to the fiber scale. The novelty of this study lies in the calculation of the shear stress (300-700 MPa) required for fiber formation and identification of the amino acid residues involved in the transition. This is the first time that the shear stress has been quantified in connection with a secondary structure transition. By study of molecules containing varying numbers of contiguous MaSp1 repeats, we determine that the smallest molecule size giving rise to a "silk-like" structure contains six polyalanine repeats. Through a probability analysis of the secondary structure, we identify specific amino acids that transition from α-helix to β-sheet. In addition to portions of the polyalanine section, these amino acids include glycine, leucine, and glutamine. The stability of β-sheet structures appears to arise from a close proximity in space of helices in the initial spidroin state. Our results are in agreement with the forces exerted by spiders in the silking process and the experimentally determined global secondary structure of spidroin and pulled MaSp1 silk. Our study emphasizes the role of shear in the assembly process of silk and can guide the design of microfluidic devices that attempt to mimic the natural spinning process and predict molecular requirements for the next generation of silk-based functional materials. PMID:26669270

  4. BOILING REACTORS

    DOEpatents

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  5. Criticality safety considerations in the geologic disposal of spent nuclear fuel assemblies

    SciTech Connect

    Gore, B.F.; McNair, G.W.; Heaberlin, S.W.

    1980-05-01

    Features of geologic disposal which hamper the demonstration that criticality cannot occur therein include possible changes of shape and form, intrusion of water as a neutron moderator, and selective leaching of spent fuel constituents. If the criticality safety of spent fuel disposal depends on burnup, independent measurements verifying the burnup should be performed prior to disposal. The status of nondestructive analysis method which might provide such verification is discussed. Calculations were performed to assess the potential for increasing the allowed size of a spent fuel disposal canister if potential water intrusion were limited by close-packing the enclosed rods. Several factors were identified which severely limited the potential of this application. The theoretical limit of hexagonal close-packing cannot be achieved due to fuel rod bowing. It is concluded that disposal canisters should be sized on the basis of assumed optimum moderation. Several topics for additional research were identified during this limited study.

  6. Critical island-size, stability and island morphology in nanoparticle island self-assembly

    NASA Astrophysics Data System (ADS)

    Amar, Jacques; Hubartt, Bradley

    2015-03-01

    The critical island-size, stability, and morphology of 2D colloidal Au nanoparticle (NP) islands formed at the toluene-air interface during drop-drying are studied using molecular dynamics and energetics calculations. Our calculations were carried out using an empirical potential which takes into account interactions between the dodecanethiol ligands and the toluene solvent, ligand-ligand interactions, and the van der Waals interaction between the Au cores. Good agreement with experimental results is obtained for the dependence of the critical island-size on NP diameter. Our results for the critical length-scale for smoothing via edge-diffusion are also consistent with the limited facet size and island-relaxation observed in experiments. The relatively high rate of NP diffusion on an island obtained in our simulations as well as the low calculated activation barrier for interlayer diffusion are also consistent with experimental observations that second-layer growth does not occur until after the first layer is complete. Supported by NSF CHE-1012896 and DMR-1410840

  7. Developing a University Learning Community of Critical Readers and Writers: The Story of a Liberal Arts and IEP Partnership

    ERIC Educational Resources Information Center

    Ernst, Beth Kozbial; Wonder, Kelly; Adler, Julie

    2016-01-01

    Integrating English language learners into the academic mainstream is a critically important goal. For students who are learning content in their second or third language as well as negotiating the university's social context, integrating into the mainstream academic environment can be challenging. Instructors at a public university intensive…

  8. Neutron Activation and Thermoluminescent Detector Responses to a Bare Pulse of the CEA Valduc SILENE Critical Assembly

    SciTech Connect

    Miller, Thomas Martin; Celik, Cihangir; McMahan, Kimberly L.; Lee, Yi-kang; Gagnier, Emmanuel; Authier, Nicolas; Piot, Jerome; Jacquet, Xavier; Rousseau, Guillaume; Reynolds, Kevin H.

    2015-09-01

    This benchmark experiment was conducted as a joint venture between the US Department of Energy (DOE) and the French Commissariat à l'Energie Atomique (CEA). Staff at the Oak Ridge National Laboratory (ORNL) in the US and the Centre de Valduc in France planned this experiment. The experiment was conducted on October 11, 2010 in the SILENE critical assembly facility at Valduc. Several other organizations contributed to this experiment and the subsequent evaluation, including CEA Saclay, Lawrence Livermore National Laboratory (LLNL), the Y-12 National Security Complex (NSC), Babcock International Group in the United Kingdom, and Los Alamos National Laboratory (LANL). The goal of this experiment was to measure neutron activation and thermoluminescent dosimeter (TLD) doses from a source similar to a fissile solution critical excursion. The resulting benchmark can be used for validation of computer codes and nuclear data libraries as required when performing analysis of criticality accident alarm systems (CAASs). A secondary goal of this experiment was to qualitatively test performance of two CAAS detectors similar to those currently and formerly in use in some US DOE facilities. The detectors tested were the CIDAS MkX and the Rocky Flats NCD-91. These detectors were being evaluated to determine whether they would alarm, so they were not expected to generate benchmark quality data.

  9. ExSPAnder: a universal repeat resolver for DNA fragment assembly

    PubMed Central

    Prjibelski, Andrey D.; Vasilinetc, Irina; Bankevich, Anton; Gurevich, Alexey; Krivosheeva, Tatiana; Nurk, Sergey; Pham, Son; Korobeynikov, Anton; Lapidus, Alla; Pevzner, Pavel A.

    2014-01-01

    Next-generation sequencing (NGS) technologies have raised a challenging de novo genome assembly problem that is further amplified in recently emerged single-cell sequencing projects. While various NGS assemblers can use information from several libraries of read-pairs, most of them were originally developed for a single library and do not fully benefit from multiple libraries. Moreover, most assemblers assume uniform read coverage, condition that does not hold for single-cell projects where utilization of read-pairs is even more challenging. We have developed an exSPAnder algorithm that accurately resolves repeats in the case of both single and multiple libraries of read-pairs in both standard and single-cell assembly projects. Availability and implementation: http://bioinf.spbau.ru/en/spades Contact: ap@bioinf.spbau.ru PMID:24931996

  10. Universal Scaling in the Fan of an Unconventional Quantum Critical Point

    SciTech Connect

    Melko, Roger G; Kaul, Ribhu

    2008-01-01

    We present the results of extensive finite-temperature Quantum Monte Carlo simulati ons on a SU(2) symmetric, $S=1/2$ quantum antiferromagnet with a frustrating four-s pin interaction -- the so-called 'JQ' model~[Sandvik, Phys. Rev. Lett. {\\bf 98}, 22 7202 (2007)]. Our simulations, which are unbiased, free of the sign-problem and car ried out on lattice sizes containing in excess of $1.6\\times 10^4$ spins, indicate that N\\'eel order is destroyed through a continuous quantum transition at a critica l value of the frustrating interaction. At larger values of this coupling the param agnetic state obtained has valence-bond solid order. The scaling behavior in the 'q uantum critical fan' above the putative critical point confirms a $z=1$ quantum pha se transition that is not in the conventional $O(3)$ universality class. Our result s are consistent with the predictions of the 'deconfined quantum criticality' scena rio.

  11. Universal self-field critical current for thin-film superconductors

    PubMed Central

    Talantsev, E. F.; Tallon, J. L.

    2015-01-01

    For any practical superconductor the magnitude of the critical current density, Jc, is crucially important. It sets the upper limit for current in the conductor. Usually Jc falls rapidly with increasing external magnetic field, but even in zero external field the current flowing in the conductor generates a self-field that limits Jc. Here we show for thin films of thickness less than the London penetration depth, λ, this limiting Jc adopts a universal value for all superconductors—metals, oxides, cuprates, pnictides, borocarbides and heavy Fermions. For type-I superconductors, it is Hc/λ where Hc is the thermodynamic critical field. But surprisingly for type-II superconductors, we find the self-field Jc is Hc1/λ where Hc1 is the lower critical field. Jc is thus fundamentally determined and this provides a simple means to extract absolute values of λ(T) and, from its temperature dependence, the symmetry and magnitude of the superconducting gap. PMID:26240014

  12. Correlation between Knowledge, Experience and Common Sense, with Critical Thinking Capability of Medical Faculty's Students at Indonesia Christian University

    ERIC Educational Resources Information Center

    Nadeak, Bernadetha

    2015-01-01

    This research discusses correlation between knowledge, experience and common sense with critical thinking of Medical Faculty's Student. As to the objective of this research is to find the correlation between knowledge, experience and common sense with critical thinking of Medical Faculty's Students at Christian University of Indonesia. It is…

  13. PREFACE: MEM05: The 3rd International Workshop on Mechano-Electromagnetic Properties of Composite Superconductors (Kyoto, Japan, 17 20 July 2005)

    NASA Astrophysics Data System (ADS)

    Osamura, Kozo; Hampshire, Damian

    2005-12-01

    One of the important challenges facing the international scientific community at the beginning of the third millennium is how to manage the world's energy resources properly. Superconductivity will provide one of the strategies employed to avoid an energy crisis. Of course the ITER Fusion Tokomak that is to be built in France provides an exciting focus for the whole superconductivity community. In parallel, we can expect that other key technologies for superconductivity such as large capacity transmission cables, energy storage systems, and generators and motors will have a real impact in technologically advanced countries. There is broadly a consensus that the prototype stage for high-current high-field superconducting applications is largely completed, and the required performance has been demonstrated. However, before we move to full industrialization of large-scale superconducting technologies, feasibility studies suggest there are two types of problem that remain. The first is the development of high performance and low cost materials which are fully optimized in terms of critical current, low ac loss and high strength. The second is the establishment of optimal procedures for system design accompanying scale up. As the system design is dependent on material development, there is a critical need to study the key issues for developing high performance superconducting materials. Under the activities of the NEDO Grant Project (Applied Superconductivity), MEM05 was organized by Professor Osamura (Kyoto University), Professor Itoh (NIMS), Professor Hojo (Kyoto University) and Professor Matsumoto (Kyoto University) and held in Kyoto, Japan. The focus for the workshop was the elimination of grain boundary weak links, the creation of strong flux pinning sites, the optimal arrangement of filaments and barriers for reducing ac losses, and the design of high strength strain tolerant composite conductors. Five subsessions were held at MEM05. • Mechanical properties of

  14. Radioisotope radiotherapy research and achievements at the University of Missouri Research Reactor

    NASA Astrophysics Data System (ADS)

    Ehrhardt, G. J.; Ketring, A. R.; Cutler, C. S.

    2003-01-01

    The University of Missouri Research Reactor (MURR) in collaboration with faculty in other departments at the University of Missouri has been involved in developing new means of internal radioisotopic therapy for cancer for many years. These efforts have centered on methods of targeting radioisotopes such as brachytherapy, embolisation of liver tumors with radioactive microspheres, small-molecule-labelled chelates for the treatment of bone cancer, and various means of radioimmunotherapy or labelled receptor agent targeting. This work has produced two radioactive agents, Sm-153 Quadramet™ and Y-90 TheraSphere™, which have U.S. Food and Drug Administration approval for the palliation of bone cancer pain and treatment of inoperable liver cancer, respectively. MURR has also pioneered development of other beta-emitting isotopes for internal radiotherapy such as Re-186, Re-188, Rh-105, Ho-166, Lu-177, and Pm-149, many of which are in research and clinical trials throughout the U.S. and the world. This important work has been made possible by the very high neutron flux available at MURR combined with MURR's outstanding reliability of operation and flexibility in meeting the needs of researchers and the radiopharmaceutical industry.

  15. Final Site Specific Decommissioning Inspection Report #2 for the University of Washington Research and Test Reactor, Seattle, Washington

    SciTech Connect

    S.J. Roberts

    2007-03-20

    During the period of August through November 2006, ORISE performed a comprehensive IV at the University of Washington Research and Test Reactor Facility. The objective of the ORISE IV was to validate the licensee’s final status survey processes and data, and to assure the requirements of the DP and FSSP were met.

  16. Study of neutron physics: conversion of the University of Missouri-Rolla reactor to low-enriched fuel

    SciTech Connect

    Straka, M.; Covington, L.

    1987-01-01

    A detailed study of a fuel conversion (using LEU) has been undertaken for the University of Missouri-Rolla reactor. Results achieved with the available code package have been compared with the measured data whenever possible. The neutronic codes LEOPARD and 2DB-UM provided adequate results in most cases examined.

  17. Final report on the University of Florida U.S. Department of Energy 1995--96 Reactor Sharing Program

    SciTech Connect

    Vernetson, W.G.

    1996-11-01

    Grant support has been well used by the University of Florida as host institution to support various educational institutions in the use of the reactor and associated facilities as indicated in the proposal. These various educational institutions are located primarily within Florida. However, when the 600-mile distance from Pensacola to Miami is considered, it is obvious that this Grant provides access to reactor utilization for a broad geographical region and a diverse set of user institutions serving over twelve million inhabitants throughout the State of Florida and still others throughout the nation. All users and uses were carefully screened to assure the usage was for educational institutions eligible for participation in the Reactor Sharing Program; where research activities were involved, care was taken to assure the research activities were not funded by grants for contract funding from outside sources. In some cases external grant funding is limited or is used up, in which case the Reactor Sharing Grant and frequent cost sharing by the UFTR facility and the University of Florida provide the necessary support to complete a project or to provide more results to make a complete project even better. In some cases this latter usage has aided renewal of external funding. The role of the Reactor Sharing Program, though relatively small in dollars, has been the single most important occurrence in assuring the rebirth and continued high utilization of the UFTR in a time when many better equipped and better placed facilities have ceased operations. Through dedicated and effective advertising efforts, the UFTR has seen nearly every four-year college and university in Florida make substantive use of the facility under the Reactor Sharing Program with many now regular users. Some have even been able to support usage from outside grants where the Reactor Sharing Grant has served as seed money; still others have been assisted when external grants were depleted.

  18. Cardiolipin, a critical determinant of mitochondrial carrier protein assembly and function

    PubMed Central

    Claypool, Steven M.

    2009-01-01

    The ability of phospholipids to act as determinants of membrane protein structure and function is probably best exemplified by cardiolipin (CL), the signature phospholipid of mitochondria. Early efforts to reconstitute individual respiratory complexes and members of the mitochondrial carrier family, most notably the ADP/ATP carrier (AAC), often demonstrated the importance of CL. Over the past decade, the significance of CL in the organization of components of the electron transport chain into higher order assemblies, termed respiratory supercomplexes, has been established. Another protein required for oxidative phosphorylation, AAC, has received comparatively little attention likely stemming from the fact that AACs were thought to function in isolation as either homodimers or monomers. Recently however, AACs have been demonstrated to interact with the respiratory supercomplex, other members of the mitochondrial carrier family, and the TIM23 translocon. Interestingly, many if not all of these interactions depend on CL. As the paradigm for the mitochondrial carrier family, these discoveries with AAC suggest that other members of this large group of important proteins may be more gregarious than anticipated. Moreover, it is proposed that AAC and perhaps additional members of the mitochondrial carrier family might represent downstream targets of pathological states involving alterations in CL. PMID:19422785

  19. Integral Reactor Physics Benchmarks - the International Criticality Safety Benchmark Evaluation Project (icsbep) and the International Reactor Physics Experiment Evaluation Project (irphep)

    NASA Astrophysics Data System (ADS)

    Briggs, J. Blair; Nigg, David W.; Sartori, Enrico

    2006-04-01

    Since the beginning of the nuclear industry, thousands of integral experiments related to reactor physics and criticality safety have been performed. Many of these experiments can be used as benchmarks for validation of calculational techniques and improvements to nuclear data. However, many were performed in direct support of operations and thus were not performed with a high degree of quality assurance and were not well documented. For years, common validation practice included the tedious process of researching integral experiment data scattered throughout journals, transactions, reports, and logbooks. Two projects have been established to help streamline the validation process and preserve valuable integral data: the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP). The two projects are closely coordinated to avoid duplication of effort and to leverage limited resources to achieve a common goal. A short history of these two projects and their common purpose are discussed in this paper. Accomplishments of the ICSBEP are highlighted and the future of the two projects outlined.

  20. Summer Freezing Resistance: A Critical Filter for Plant Community Assemblies in Mediterranean High Mountains

    PubMed Central

    Pescador, David S.; Sierra-Almeida, Ángela; Torres, Pablo J.; Escudero, Adrián

    2016-01-01

    Assessing freezing community response and whether freezing resistance is related to other functional traits is essential for understanding alpine community assemblages, particularly in Mediterranean environments where plants are exposed to freezing temperatures and summer droughts. Thus, we characterized the leaf freezing resistance of 42 plant species in 38 plots at Sierra de Guadarrama (Spain) by measuring their ice nucleation temperature, freezing point (FP), and low-temperature damage (LT50), as well as determining their freezing resistance mechanisms (i.e., tolerance or avoidance). The community response to freezing was estimated for each plot as community weighted means (CWMs) and functional diversity (FD), and we assessed their relative importance with altitude. We established the relationships between freezing resistance, growth forms, and four key plant functional traits (i.e., plant height, specific leaf area, leaf dry matter content (LDMC), and seed mass). There was a wide range of freezing resistance responses and more than in other alpine habitats. At the community level, the CWMs of FP and LT50 responded negatively to altitude, whereas the FD of both traits increased with altitude. The proportion of freezing-tolerant species also increased with altitude. The ranges of FP and LT50 varied among growth forms, and only leaf dry matter content was negatively correlated with freezing-resistance traits. Summer freezing events represent important abiotic filters for assemblies of Mediterranean high mountain communities, as suggested by the CWMs. However, a concomitant summer drought constraint may also explain the high freezing resistance of species that thrive in these areas and the lower FD of freezing resistance traits at lower altitudes. Leaves with high dry matter contents may maintain turgor at lower water potential and enhance drought tolerance in parallel to freezing resistance. This adaptation to drought seems to be a general prerequisite for plants

  1. Summer Freezing Resistance: A Critical Filter for Plant Community Assemblies in Mediterranean High Mountains.

    PubMed

    Pescador, David S; Sierra-Almeida, Ángela; Torres, Pablo J; Escudero, Adrián

    2016-01-01

    Assessing freezing community response and whether freezing resistance is related to other functional traits is essential for understanding alpine community assemblages, particularly in Mediterranean environments where plants are exposed to freezing temperatures and summer droughts. Thus, we characterized the leaf freezing resistance of 42 plant species in 38 plots at Sierra de Guadarrama (Spain) by measuring their ice nucleation temperature, freezing point (FP), and low-temperature damage (LT50), as well as determining their freezing resistance mechanisms (i.e., tolerance or avoidance). The community response to freezing was estimated for each plot as community weighted means (CWMs) and functional diversity (FD), and we assessed their relative importance with altitude. We established the relationships between freezing resistance, growth forms, and four key plant functional traits (i.e., plant height, specific leaf area, leaf dry matter content (LDMC), and seed mass). There was a wide range of freezing resistance responses and more than in other alpine habitats. At the community level, the CWMs of FP and LT50 responded negatively to altitude, whereas the FD of both traits increased with altitude. The proportion of freezing-tolerant species also increased with altitude. The ranges of FP and LT50 varied among growth forms, and only leaf dry matter content was negatively correlated with freezing-resistance traits. Summer freezing events represent important abiotic filters for assemblies of Mediterranean high mountain communities, as suggested by the CWMs. However, a concomitant summer drought constraint may also explain the high freezing resistance of species that thrive in these areas and the lower FD of freezing resistance traits at lower altitudes. Leaves with high dry matter contents may maintain turgor at lower water potential and enhance drought tolerance in parallel to freezing resistance. This adaptation to drought seems to be a general prerequisite for plants

  2. Reaction kinetic analysis of reactor surveillance data

    NASA Astrophysics Data System (ADS)

    Yoshiie, T.; Sato, K.; Xu, Q.; Nagai, Y.

    2015-06-01

    In reactor pressure vessel surveillance data, it was found that the concentration of matrix defects was very low even after nearly 40 years of operation, though a large number of precipitates existed. In this paper, defect structures obtained from surveillance data of A533B (high Cu concentration) were simulated using reaction kinetic analysis with 11 rate equations. The coefficients used in the equations were quite different from those obtained by fitting a Fe-0.6 wt%Cu alloy irradiated by the Kyoto University Reactor. The difference was mainly caused by alloying elements in A533B, and the effect of alloying elements was extracted. The same code was applied to low-Cu A533B irradiated with high irradiation damage rate, and the formation of voids was correctly simulated.

  3. Higher Criticism and Higher Education at the University of Chicago: William Rainey Harper's Vision of Religion in the Research University

    ERIC Educational Resources Information Center

    Lee, Michael

    2008-01-01

    In the 1890s, the Board of Trustees of the not-yet-built University of Chicago had just elected Rainey Harper to be its first president, and later, he would formally accept the position. Harper left a secure position at Yale University to accept the presidency of a university that was nothing more than an idea, a board of trustees, and the…

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

    SciTech Connect

    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.

  5. Optimum design and criticality safety of a beam-shaping assembly with an accelerator-driven subcritical neutron multiplier for boron neutron capture therapies.

    PubMed

    Hiraga, F

    2015-12-01

    The beam-shaping assembly for boron neutron capture therapies with a compact accelerator-driven subcritical neutron multiplier was designed so that an epithermal neutron flux of 1.9×10(9) cm(-2) s(-1) at the treatment position was generated by 5 MeV protons in a beam current of 2 mA. Changes in the atomic density of (135)Xe in the nuclear fuel due to the operation of the beam-shaping assembly were estimated. The criticality safety of the beam-shaping assembly in terms of Xe poisoning is discussed. PMID:26235186

  6. Critical Exponents of Dynamical Conductivity in 2D Percolative Superconductor-Insulator Transitions: Three Universality Classes

    NASA Astrophysics Data System (ADS)

    Karki, Pragalv; Loh, Yen Lee

    We simulate three types of random inductor-capacitor (LC) networks on 4000x4000 lattices. We calculate the dynamical conductivity using an equation-of-motion method in which timestep error is eliminated and windowing error is minimized. We extract the critical exponent a such that σ (ω) ~ω-a at low frequencies. The results suggest that there are three different universality classes. The LijCi model, with capacitances from each site to ground, has a = 0 . 32 . The LijCij model, with capacitances along bonds, has a = 0 . The LijCiCij model, with both types of capacitances, has a = 0 . 30 . This implies that classical percolative 2D superconductor-insulator transitions (SITs) generically have σ (ω) --> ∞ as ω --> 0 . Therefore, experiments that give a constant conductivity as ω --> 0 must be explained in terms of quantum effects.

  7. Universal properties of the Higgs resonance in (2+1)-dimensional U(1) critical systems.

    PubMed

    Chen, Kun; Liu, Longxiang; Deng, Youjin; Pollet, Lode; Prokof'ev, Nikolay

    2013-04-26

    We present spectral functions for the magnitude squared of the order parameter in the scaling limit of the two-dimensional superfluid to Mott insulator quantum phase transition at constant density, which has emergent particle-hole symmetry and Lorentz invariance. The universal functions for the superfluid, Mott insulator, and normal liquid phases reveal a low-frequency resonance which is relatively sharp and is followed by a damped oscillation (in the first two phases only) before saturating to the quantum critical plateau. The counterintuitive resonance feature in the insulating and normal phases calls for deeper understanding of collective modes in the strongly coupled (2+1)-dimensional relativistic field theory. Our results are derived from analytically continued correlation functions obtained from path-integral Monte Carlo simulations of the Bose-Hubbard model. PMID:23679688

  8. Evaluated Iridium, Yttrium, and Thulium Cross Sections and Integral Validation Against Critical Assembly and Bethe Sphere Measurements

    SciTech Connect

    Chadwick, M.B. Frankle, S.; Trellue, H.; Talou, P.; Kawano, T.; Young, P.G.; MacFarlane, R.E.; Wilkerson, C.W.

    2007-12-15

    We describe new dosimetry (radiochemical) ENDF evaluations for yttrium, iridium, and thulium. These LANL2006 evaluations were based upon measured data and on nuclear model cross section calculations. In the case of iridium and yttrium, new measurements using the GEANIE gamma-ray detector at LANSCE were used to infer (n,xn) cross sections, the measurements being augmented by nuclear model calculations using the GNASH code. The thulium isotope evaluations were based on GNASH calculations and older measurements. The evaluated cross section data are tested through comparisons of simulations with measurements of reaction rates in critical assemblies and in Bethe sphere (sometimes called Wyman sphere) integral experiments. Two types of Bethe sphere experiments were studied - a LiD experiment that had a significant component of 14 MeV neutrons, and a LiD-U experiment that additionally had varying amounts of fission neutrons depending upon the location. These simulations were performed with the MCNP code using continuous energy Monte Carlo, and because the neutron fluences can be modeled fairly accurately by MCNP at different locations in these assemblies, the comparisons provide a valuable validation test of the accuracy of the evaluated cross sections and their energy dependencies. The MCNP integral reaction rate validation testing for the three detectors yttrium, iridium, and thulium, in the LANL2006 database is summarized as follows: (1) (n,2n)near 14 MeV: In 14 MeV-dominated locations (the LiD Bethe spheres and the outer regions of the LiD-U Bethe spheres), the (n,2n) products are modeled very well for all three detectors, suggesting that the evaluated {sup 89}Y(n,2n), {sup 191}Ir(n,2n), and {sup 169}Tm(n,2n) cross sections are accurate to better than about 5% near 14 MeV; (2) (n,2n)near threshold: In locations that have a significant number of fission spectrum neutrons or downscattered neutrons from 14 MeV inelastic scattering (the central regions of the Li

  9. High-precision estimate of the critical exponents for the directed Ising universality class

    NASA Astrophysics Data System (ADS)

    Park, Su-Chan

    2013-02-01

    With extensive Monte Carlo simulations, we present high-precision estimates of the critical exponents of branching annihilating random walks with two offspring, a prototypical model of the directed Ising universality class in one dimension. To estimate the exponents accurately, we propose a systematic method to find corrections to scaling whose leading behavior is supposed to take the form t -χ in the long-time limit at the critical point. Our study shows that χ ≈ 0.75 for the number of particles in defect simulations and χ ≈ 0.5 for other measured quantities, which should be compared with the widely used value of χ = 1. Using χ so obtained, we analyze the effective exponents to find that β/ν ‖ = 0.2872(2), z = 1.7415(5), η = 0.0000(2), and accordingly, β/ν ⊥ = 0.5000(6). Our numerical results for β/ν ‖ and z are clearly different from the conjectured rational numbers β/ν ‖ = tfrac{2} {7} ≈ 0.2857, z = tfrac{7} {4} = 1.75 by Jensen [Phys. Rev. E, 50, 3623 (1994)]. Our result for β/ν ⊥, however, is consistent with tfrac{1} {2} , which is believed to be exact.

  10. Universal critical phenomena of the cloud --> crystal phase transition in the Paul trap: Powerlaws

    NASA Astrophysics Data System (ADS)

    Weiss, Daniel; Nam, Yunseong; Blümel, Reinhold

    N charged particles, simultaneously stored in a radio-frequency (rf) Paul trap, exhibit deterministic heating. Depending on the damping (γ) imparted to the system, these particles can exist in multiple phases, the most commonly found being the cloud and crystal phases. With a small γ, the particles exhibit gas-like behavior, where the heating and cooling equilibrate and a stable cloud results. For larger γ, the damping overcomes the heating and the particles are forced into the crystalline state. We explore the cloud --> crystal transition as a critical phenomenon. We find that the transition occurs at a critical value γc of the damping constant γ. We find that as a function of N, γc scales approximately like an iterated log law. We also present a universal power law, τm ~(γ -γc) - β , γ >γc , β > 0 , independent of both N and the Paul trap parameter a, depending only on the Paul trap parameter q, that describes the number of cycles necessary for the system to crystallize as a function of γ -γc .

  11. A high converter concept for fuel management with blanket fuel assemblies in boiling water reactors

    SciTech Connect

    Martinez-Frances, N.; Timm, W.; Rossbach, D.

    2012-07-01

    Studies on the natural Uranium saving and waste reduction potential of a multiple-plant BWR system were performed. The BWR High Converter system should enable a multiple recycling of MOX fuel in current BWR plants by introducing blanket fuel assemblies and burning Uranium and MOX fuel separately. The feasibility of Uranium cores with blankets and full-MOX cores with Plutonium qualities as low as 40% were studied. The power concentration due to blanket insertion is manageable with modern fuel and acceptable values for the thermal limits and reactivity coefficients were obtained. While challenges remain, full-MOX cores also complied with the main design criteria. The combination of Uranium and Plutonium burners in appropriate proportions could enable obtaining as much as 40% more energy out of Uranium ore. Moreover, a proper adjustment of blanket average stay and Plutonium qualities could lead to a system with nearly no Plutonium left for final disposal. The achievement of such goals with current light water technology makes the BWR HC concept an attractive option to improve the fuel cycle until Gen-IV designs are mature. (authors)

  12. EVALUATION OF ACOUSTICAL HOLOGRAPHY FOR THE INSPECTION OF LIGHT WATER REACTOR WELD ASSEMBLIES

    SciTech Connect

    Collins, H. D.; Gribble, R. P.

    1982-06-01

    The primary objective of this program was the evaluation of acoustical holography techniques for characterization of the light water reactor weld surface signatures in the nuclear safeguards program. The accurate characterization of weld surface irregulari ties and vertical deviations was achieved using acoustical holographic interferometric techniques. Preselected weld surfaces were inspected and the vertical deviations characterized by phase measurements or fringe densities in the image. Experimental results on Sandia samples verify depth deviation sensitivities of 0.11 {micro}m to 0.16 {micro}m. The two point interferogram technique is recommended for surveillance of the weld surface associated wi th fuel rod removal in the nuclear safeguard program. The use of this unique holographic signal processing provides essentially a fail-safe method for surveillance of clandestine fuel rod removal. Statistical analysis indicates 99.99% (weld surface deviation) confidence interval between 2~m and 3~m can be achieved. These results illustrate the extremely high resolution capabilities of the surveillance technique employing coherent signal processing.

  13. The World Health Assembly resolutions on eHealth: eHealth in support of universal health coverage.

    PubMed

    Al-Shorbaji, N

    2013-01-01

    The World Health Assembly (WHA) of the World Health Organization (WHO) and three of the six WHO Regional Committees adopted a number of resolutions on eHealth: the use of information and communication technology for health. These resolutions have given legitimacy to eHealth as an area of work for WHO and its member states. The implementation of these resolutions will contribute to the achievement of the Millennium Development Goals (MDGs) and the Universal Health Coverage. eHealth has been perceived as reducing the cost of healthcare, improving quality and equitable access to health services. PMID:24310395

  14. Single and Coupled Electrochemical Processes and Reactors for the Abatement of Organic Water Pollutants: A Critical Review.

    PubMed

    Martínez-Huitle, Carlos A; Rodrigo, Manuel A; Sirés, Ignasi; Scialdone, Onofrio

    2015-12-23

    Traditional physicochemical and biological techniques, as well as advanced oxidation processes (AOPs), are often inadequate, ineffective, or expensive for industrial water reclamation. Within this context, the electrochemical technologies have found a niche where they can become dominant in the near future, especially for the abatement of biorefractory substances. In this critical review, some of the most promising electrochemical tools for the treatment of wastewater contaminated by organic pollutants are discussed in detail with the following goals: (1) to present the fundamental aspects of the selected processes; (2) to discuss the effect of both the main operating parameters and the reactor design on their performance; (3) to critically evaluate their advantages and disadvantages; and (4) to forecast the prospect of their utilization on an applicable scale by identifying the key points to be further investigated. The review is focused on the direct electrochemical oxidation, the indirect electrochemical oxidation mediated by electrogenerated active chlorine, and the coupling between anodic and cathodic processes. The last part of the review is devoted to the critical assessment of the reactors that can be used to put these technologies into practice. PMID:26654466

  15. Critical review of the reactor-safety study radiological health effects model. Final report

    SciTech Connect

    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.

  16. Sensitivity Analysis of Fuel Centerline Temperatures in SuperCritical Water-cooled Reactors (SCWRs)

    NASA Astrophysics Data System (ADS)

    Abdalla, Ayman

    SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today's conventional NPPs, which have thermal efficiencies between 30 - 35%, SCW NPPs will have thermal efficiencies within a range of 45 - 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current fuel bundles with UO2 fuel at the high operating parameters of SCWRs may cause high fuel centerline temperatures, which could lead to fuel failure and fission gas release. Studies have shown that when the Variant-20 (43-element) fuel bundle was examined at SCW conditions, the fuel centerline temperature industry limit of 1850°C for UO2 and the sheath temperature design limit of 850°C might be exceeded. Therefore, new fuel-bundle designs, which comply with the design requirements, are required for future use in SCWRs. The main objective of this study to conduct a sensitivity analysis in order to identify the main factors that leads to fuel centerline temperature reduction. Therefore, a 54-element fuel bundle with smaller diameter of fuel elements compared to that of the 43-element bundle was designed and various nuclear fuels are examined for future use in a generic Pressure Tube (PT) SCWR. The 54-element bundle consists of 53 heated fuel elements with an outer diameter of 9.5 mm and one central unheated element of 20-mm outer diameter which contains burnable poison. The 54-element fuel bundle has an outer diameter of 103.45 mm, which is the same as the outer diameter of the 43-element fuel bundle. After developing the 54-element fuel bundle, one-dimensional heat-transfer analysis was conducted using MATLAB and NIST REFPROP programs. As a result, the Heat Transfer

  17. Advances in Computational Fluid Dynamics Modeling of Two Phase Flow in a Boiling Water Reactor Fuel Assembly

    SciTech Connect

    Tentner, Adrian; Lo, Simon; Ioilev, Andrey; Melnikov, Vladimir; Samigulin, Maskhud; Ustinenko, Vasily; Kozlov, Valentin

    2006-07-01

    A new code, CFD-BWR, is being developed for the simulation of two-phase flow phenomena inside a BWR fuel bundle. These phenomena include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. CFD-BWR is a specialized module built on the foundation of the commercial CFD code STAR-CD which provides general two-phase flow modeling capabilities. New models describing the inter-phase mass, momentum, and energy transfer phenomena specific for BWRs have been developed and implemented in the CFD-BWR module. A set of experiments focused on two-phase flow and phase-change phenomena has been identified for the validation of the CFD-BWR code and results of two experiment analyses focused on the radial void distribution are presented. The close agreement between the computed results, the measured data and the correlation results provides confidence in the accuracy of the models. (authors)

  18. Filtered fast neutron irradiation system using Texas A&M University Nuclear Science Center Reactor

    NASA Astrophysics Data System (ADS)

    Jang, S. Y.; Kim, C. H.; Reece, W. D.; Braby, L. A.

    2004-09-01

    A heavily filtered fast neutron irradiation system (FNIS) was developed for a variety of applications, including the study of long-term health effects of fast neutrons by evaluating the biological mechanisms of damage in cultured cells and living animals such as rats or mice. This irradiation system includes an exposure cave made with a lead-bismuth alloy, a cave positioning system, a gamma and neutron monitoring system, a sample transfer system, and interchangeable filters. This system was installed in the irradiation cell of the Texas A&M University Nuclear Science Center Reactor (NSCR). For a realistic modeling of the NSCR, the irradiation cell, and the FNIS, this study used the Monte Carlo N-Particle (MCNP) code and a set of high-temperature ENDF/B-VI continuous neutron cross-section data. Sensitivity analysis was performed to find the characteristics of the FNIS as a function of the thickness of the lead-bismuth alloy. A paired ion chamber system was constructed with a tissue-equivalent plastic (A-150) and propane gas for total dose monitoring and with graphite and argon for gamma dose monitoring. This study, in addition, tested the Monte Carlo modeling of the FNIS system, as well as the performance of the system by comparing the calculated results with experimental measurements using activation foils and paired ion chambers.

  19. Production of 37Ar in The University of Texas TRIGA reactor facility

    SciTech Connect

    Egnatuk, Christine M.; Lowrey, Justin; Biegalski, S.; Bowyer, Ted W.; Haas, Derek A.; Orrell, John L.; Woods, Vincent T.; Keillor, Martin E.

    2011-06-19

    The detection of {sup 37}Ar is important for on-site inspections for the Comprehensive Nuclear-Test-Ban Treaty monitoring. In an underground nuclear explosion this radionuclide is produced by {sup 40}Ca(n,{alpha}){sup 37}Ar reaction in surrounding soil and rock. With a half-life of 35 days, {sup 37}Ar provides a signal useful for confirming the location of an underground nuclear event. An ultra-low-background proportional counter developed by Pacific Northwest National Laboratory is used to detect {sup 37}Ar, which decays via electron capture. The irradiation of Ar gas at natural enrichment in the 3L facility within the Mark II TRIGA reactor facility at The University of Texas at Austin provides a source of {sup 37}Ar for the calibration of the detector. The {sup 41}Ar activity is measured by the gamma activity using an HPGe detector after the sample is removed from the core. Using the {sup 41}Ar/{sup 37}Ar production ratio and the {sup 41}Ar activity, the amount of {sup 37}Ar created is calculated. The {sup 41}Ar decays quickly (half-life of 109.34 minutes) leaving a radioactive sample of high purity {sup 37}Ar and only trace levels of {sup 39}Ar.

  20. Optimization study for an epithermal neutron beam for boron neutron capture therapy at the University of Virginia Research Reactor

    SciTech Connect

    Burns, T.D. Jr.

    1995-05-01

    The non-surgical brain cancer treatment modality, Boron Neutron Capture Therapy (BNCT), requires the use of an epithermal neutron beam. This purpose of this thesis was to design an epithermal neutron beam at the University of Virginia Research Reactor (UVAR) suitable for BNCT applications. A suitable epithermal neutron beam for BNCT must have minimal fast neutron and gamma radiation contamination, and yet retain an appreciable intensity. The low power of the UVAR core makes reaching a balance between beam quality and intensity a very challenging design endeavor. The MCNP monte carlo neutron transport code was used to develop an equivalent core radiation source, and to perform the subsequent neutron transport calculations necessary for beam model analysis and development. The code accuracy was validated by benchmarking output against experimental criticality measurements. An epithermal beam was designed for the UVAR, with performance characteristics comparable to beams at facilities with cores of higher power. The epithermal neutron intensity of this beam is 2.2 {times} 10{sup 8} n/cm{sup 2} {center_dot} s. The fast neutron and gamma radiation KERMA factors are 10 {times} 10{sup {minus}11}cGy{center_dot}cm{sup 2}/n{sub epi} and 20 {times} 10{sup {minus}11} cGy{center_dot}cm{sup 2}/n{sub epi}, respectively, and the current-to-flux ratio is 0.85. This thesis has shown that the UVAR has the capability to provide BNCT treatments, however the performance characteristics of the final beam of this study were limited by the low core power.

  1. Conversion Analyses for the VR-1 Reactor, part I and II.

    SciTech Connect

    Hannan, N. A.; Matos, J. E.; Stillman, J. A.; Olson, A. P.; Garner, P.L.

    2005-11-14

    At the request of the Czech Technical University (CTU) in Prague, ANL has performed independent verification calculations using the MCNP Monte Carlo code for three core configurations of the VR-1 reactor: a current core configuration B1 with HEU (36%) IRT-3M fuel assemblies and planned core configurations C1 and C2 with LEU (19.7%) IRT-4M fuel assemblies. Details of these configurations were provided to ANL by CTU. For core configuration B1, criticality calculations were performed for two sets of control rod positions provided to ANL by CTU. Fore core configurations C1 and C2, criticality calculations were done for cases with all control rods at the top positions, all control rods at the bottom positions, and two critical states of the reactor for different control rod positions. In addition, sensitivity studies for variation of the {sup 235}U mass in each fuel assembly and variation of the fuel meat and cladding thicknesses in each of the fuel tubes were doe for the C1 core configuration. The reactivity worth of the individual control rods was calculated for the B1, C1, and C2 core configurations. Finally, the reactivity feedback coefficients, the prompt neutron lifetime, and the total effective delay neutron fraction were calculated for each of the three cores.

  2. Noninvasive Reactor Imaging Using Cosmic-Ray Muons

    NASA Astrophysics Data System (ADS)

    Miyadera, H.; Fujita, K.; Karino, Y.; Kume, N.; Nakayama, K.; Sano, Y.; Sugita, T.; Yoshioka, K.; Morris, C. L.; Bacon, J. D.; Borozdin, K. N.; Perry, J. O.; Mizokami, S.; Otsuka, Y.; Yamada, D.

    2015-10-01

    Cosmic-ray-muon imaging is proposed to assess the damages to the Fukushima Daiichi reactors. Simulation studies showed capability of muon imaging to reveal the core conditions.The muon-imaging technique was demonstrated at Toshiba Nuclear Critical Assembly, where the uranium-dioxide fuel assembly was imaged with 3-cm spatial resolution after 1 month of measurement.

  3. Avalanches in self-organized critical neural networks: a minimal model for the neural SOC universality class.

    PubMed

    Rybarsch, Matthias; Bornholdt, Stefan

    2014-01-01

    The brain keeps its overall dynamics in a corridor of intermediate activity and it has been a long standing question what possible mechanism could achieve this task. Mechanisms from the field of statistical physics have long been suggesting that this homeostasis of brain activity could occur even without a central regulator, via self-organization on the level of neurons and their interactions, alone. Such physical mechanisms from the class of self-organized criticality exhibit characteristic dynamical signatures, similar to seismic activity related to earthquakes. Measurements of cortex rest activity showed first signs of dynamical signatures potentially pointing to self-organized critical dynamics in the brain. Indeed, recent more accurate measurements allowed for a detailed comparison with scaling theory of non-equilibrium critical phenomena, proving the existence of criticality in cortex dynamics. We here compare this new evaluation of cortex activity data to the predictions of the earliest physics spin model of self-organized critical neural networks. We find that the model matches with the recent experimental data and its interpretation in terms of dynamical signatures for criticality in the brain. The combination of signatures for criticality, power law distributions of avalanche sizes and durations, as well as a specific scaling relationship between anomalous exponents, defines a universality class characteristic of the particular critical phenomenon observed in the neural experiments. Thus the model is a candidate for a minimal model of a self-organized critical adaptive network for the universality class of neural criticality. As a prototype model, it provides the background for models that may include more biological details, yet share the same universality class characteristic of the homeostasis of activity in the brain. PMID:24743324

  4. IVA cloning: A single-tube universal cloning system exploiting bacterial In Vivo Assembly

    PubMed Central

    García-Nafría, Javier; Watson, Jake F.; Greger, Ingo H.

    2016-01-01

    In vivo homologous recombination holds the potential for optimal molecular cloning, however, current strategies require specialised bacterial strains or laborious protocols. Here, we exploit a recA-independent recombination pathway, present in widespread laboratory E.coli strains, to develop IVA (In Vivo Assembly) cloning. This system eliminates the need for enzymatic assembly and reduces all molecular cloning procedures to a single-tube, single-step PCR, performed in <2 hours from setup to transformation. Unlike other methods, IVA is a complete system, and offers significant advantages over alternative methods for all cloning procedures (insertions, deletions, site-directed mutagenesis and sub-cloning). Significantly, IVA allows unprecedented simplification of complex cloning procedures: five simultaneous modifications of any kind, multi-fragment assembly and library construction are performed in approximately half the time of current protocols, still in a single-step fashion. This system is efficient, seamless and sequence-independent, and requires no special kits, enzymes or proprietary bacteria, which will allow its immediate adoption by the academic and industrial molecular biology community. PMID:27264908

  5. IVA cloning: A single-tube universal cloning system exploiting bacterial In Vivo Assembly.

    PubMed

    García-Nafría, Javier; Watson, Jake F; Greger, Ingo H

    2016-01-01

    In vivo homologous recombination holds the potential for optimal molecular cloning, however, current strategies require specialised bacterial strains or laborious protocols. Here, we exploit a recA-independent recombination pathway, present in widespread laboratory E.coli strains, to develop IVA (In Vivo Assembly) cloning. This system eliminates the need for enzymatic assembly and reduces all molecular cloning procedures to a single-tube, single-step PCR, performed in <2 hours from setup to transformation. Unlike other methods, IVA is a complete system, and offers significant advantages over alternative methods for all cloning procedures (insertions, deletions, site-directed mutagenesis and sub-cloning). Significantly, IVA allows unprecedented simplification of complex cloning procedures: five simultaneous modifications of any kind, multi-fragment assembly and library construction are performed in approximately half the time of current protocols, still in a single-step fashion. This system is efficient, seamless and sequence-independent, and requires no special kits, enzymes or proprietary bacteria, which will allow its immediate adoption by the academic and industrial molecular biology community. PMID:27264908

  6. The University's Role in Assembling Resources to Establish and Develop a Science Park.

    ERIC Educational Resources Information Center

    Rowe, David

    1987-01-01

    The strategies used by British universities to raise the capital necessary for the launching and development of science parks are outlined, and important considerations in undertaking such a venture are discussed. (Author/MSE)

  7. The ENEA criticality accident dosimetry system: a contribution to the 2002 international intercomparison at the SILENE reactor.

    PubMed

    Gualdrini, G; Bedogni, R; Fantuzzi, E; Mariotti, F

    2004-01-01

    The present paper summarises the activity carried out at the ENEA Radiation Protection Institute for updating the methodologies employed for the evaluation of the neutron and photon dose to the exposed workers in case of a criticality accident, in the framework of the 'International Intercomparison of Criticality Accident Dosimetry Systems' (Silène reactor, IRSN-CEA-Valduc June 2002). The evaluation of the neutron spectra and the neutron dosimetric quantities relies on activation detectors and on unfolding algorithms. Thermoluminescent detectors are employed for the gamma dose measurement. The work is aimed at accurately characterising the measurement system and, at the same time, testing the algorithms. Useful spectral information were included, based on Monte Carlo simulations, to take into account the potential accident scenarios of practical interest. All along this exercise intercomparison a particular attention was devoted to the 'traceability' of all the experimental and computational parameters and therefore, aimed at an easy treatment by the user. PMID:15353692

  8. A Temporospatial Map That Defines Specific Steps at Which Critical Surfaces in the Gag MA and CA Domains Act during Immature HIV-1 Capsid Assembly in Cells

    PubMed Central

    Robinson, Bridget A.; Reed, Jonathan C.; Geary, Clair D.; Swain, J. Victor

    2014-01-01

    ABSTRACT During HIV-1 assembly, Gag polypeptides target to the plasma membrane, where they multimerize to form immature capsids that undergo budding and maturation. Previous mutational analyses identified residues within the Gag matrix (MA) and capsid (CA) domains that are required for immature capsid assembly, and structural studies showed that these residues are clustered on four exposed surfaces in Gag. Exactly when and where the three critical surfaces in CA function during assembly are not known. Here, we analyzed how mutations in these four critical surfaces affect the formation and stability of assembly intermediates in cells expressing the HIV-1 provirus. The resulting temporospatial map reveals that critical MA residues act during membrane targeting, residues in the C-terminal CA subdomain (CA-CTD) dimer interface are needed for the stability of the first membrane-bound assembly intermediate, CA-CTD base residues are necessary for progression past the first membrane-bound intermediate, and residues in the N-terminal CA subdomain (CA-NTD) stabilize the last membrane-bound intermediate. Importantly, we found that all four critical surfaces act while Gag is associated with the cellular facilitators of assembly ABCE1 and DDX6. When correlated with existing structural data, our findings suggest the following model: Gag dimerizes via the CA-CTD dimer interface just before or during membrane targeting, individual CA-CTD hexamers form soon after membrane targeting, and the CA-NTD hexameric lattice forms just prior to capsid release. This model adds an important new dimension to current structural models by proposing the potential order in which key contacts within the immature capsid lattice are made during assembly in cells. IMPORTANCE While much is known about the structure of the completed HIV-1 immature capsid and domains of its component Gag proteins, less is known about the sequence of events leading to formation of the HIV-1 immature capsid. Here we used

  9. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  10. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  11. Investigation of applications for high-power, self-critical fissioning uranium plasma reactors

    NASA Technical Reports Server (NTRS)

    Rodgers, R. J.; Latham, T. S.; Krascella, N. L.

    1976-01-01

    Analytical studies were conducted to investigate potentially attractive applications for gaseous nuclear cavity reactors fueled by uranium hexafluoride and its decomposition products at temperatures of 2000 to 6000 K and total pressures of a few hundred atmospheres. Approximate operating conditions and performance levels for a class of nuclear reactors in which fission energy removal is accomplished principally by radiant heat transfer from the high temperature gaseous nuclear fuel to surrounding absorbing media were determined. The results show the radiant energy deposited in the absorbing media may be efficiently utilized in energy conversion system applications which include (1) a primary energy source for high thrust, high specific impulse space propulsion, (2) an energy source for highly efficient generation of electricity, and (3) a source of high intensity photon flux for heating working fluid gases for hydrogen production or MHD power extraction.

  12. LBE water interaction in sub-critical reactors: First experimental and modelling results

    NASA Astrophysics Data System (ADS)

    Ciampichetti, A.; Agostini, P.; Benamati, G.; Bandini, G.; Pellini, D.; Forgione, N.; Oriolo, F.; Ambrosini, W.

    2008-06-01

    This paper concerns the study of the phenomena involved in the interaction between LBE and pressurised water which could occur in some hypothetical accidents in accelerator driven system type reactors. The LIFUS 5 facility was designed and built at ENEA-Brasimone to reproduce this kind of interaction in a wide range of conditions. The first test of the experimental program was carried out injecting water at 70 bar and 235 °C in a reaction vessel containing LBE at 1 bar and 350 °C. A pressurisation up to 80 bar was observed in the test section during the considered transient. The SIMMER III code was used to simulate the performed test. The calculated data agree in a satisfactory way with the experimental results giving confidence in the possibility to use this code for safety analyses of heavy liquid metal cooled reactors.

  13. Syngap1 haploinsufficiency damages a postnatal critical period of pyramidal cell structural maturation linked to cortical circuit assembly

    PubMed Central

    Aceti, Massimiliano; Creson, Thomas K.; Vaissiere, Thomas; Rojas, Camilo; Huang, Wen-Chin; Wang, Ya-Xian; Petralia, Ronald S.; Page, Damon T.; Miller, Courtney A.; Rumbaugh, Gavin

    2014-01-01

    Background Genetic haploinsufficiency of Syngap1 commonly occurs in developmental brain disorders, such as intellectual disability (ID), epilepsy, schizophrenia (SCZ), and autism spectrum (ASD) disorder. Thus, studying mouse models of Syngap1 haploinsufficiency may uncover pathological developmental processes common among distinct brain disorders. Methods A Syngap1 haploinsufficiency model was used to explore the relationship between critical period dendritic spine damage, cortical circuit assembly and the window for genetic rescue in order to understand how damaging mutations disrupt key substrates of mouse brain development. Results Syngap1 mutations broadly disrupted a developmentally sensitive period that corresponded to the period of heightened postnatal cortical synaptogenesis. Pathogenic Syngap1 mutations caused a coordinated acceleration of dendrite elongation and spine morphogenesis, and pruning of these structures in neonatal cortical pyramidal neurons. These mutations also prevented a form of developmental structural plasticity associated with experience-dependent reorganization of brain circuits. Consistent with these findings, Syngap1 mutant mice displayed an altered pattern of long-distance synaptic inputs into a cortical area important for cognition. Interestingly, the ability to genetically improve the behavioral endophenotype of Syngap1 mice decreased slowly over postnatal development and mapped onto the developmental period of coordinated dendritic insults. Conclusions Pathogenic Syngap1 mutations have a profound impact on the dynamics and structural integrity of pyramidal cell postsynaptic structures known to guide the de novo wiring of nascent cortical circuits. These findings support the idea that disrupted critical periods of dendritic growth and spine plasticity may be a common pathological process in developmental brain disorders. PMID:25444158

  14. Critical experiments with mixed oxide fuel

    SciTech Connect

    Harris, D.R.

    1997-06-01

    This paper very briefly outlines technical considerations in performing critical experiments on weapons-grade plutonium mixed oxide fuel assemblies. The experiments proposed would use weapons-grade plutonium and Er{sub 2}O{sub 3} at various dissolved boron levels, and for specific fuel assemblies such as the ABBCE fuel assembly with five large water holes. Technical considerations described include the core, the measurements, safety, security, radiological matters, and licensing. It is concluded that the experiments are feasible at the Rensselaer Polytechnic Institute Reactor Critical Facility. 9 refs.

  15. Critical role of Rab11a-mediated recycling endosomes in the assembly of type I parainfluenza viruses.

    PubMed

    Stone, Raychel; Hayashi, Tsuyoshi; Bajimaya, Shringkhala; Hodges, Erin; Takimoto, Toru

    2016-01-01

    Paramyxoviruses replicate in the cytoplasm of infected cells and newly synthesized viral nucleocapsids (vRNPs) are transported to the plasma membrane to be incorporated into progeny virions. In this study, we analyzed the impact of the Rab11-mediated recycling pathway in Sendai virus (SeV) and human parainfluenza virus type 1 (hPIV1) vRNP transport. We found that suppression of Rab11 expression caused vRNP aggregation in the cytoplasm and reduced progeny virion formation. Overexpression of constitutively active Rab11Q70L, but not dominant negative Rab11S25N co-localized with vRNP, showing that vRNP specifically recognizes the GTP-bound active form of Rab11. Moreover, Rab11Q70L co-localized with the dominant negative tails of all three subtypes of myosins, Va, Vb, and Vc, while SeV and hPIV1 vRNPs co-localized with only myosin Vb and Vc. These results highlight the critical role of Rab11 in vRNP trafficking, and suggest a specificity in the recycling endosomes parainfluenza viruses utilize for virus assembly. PMID:26484934

  16. Osteogenesis of peripheral blood mesenchymal stem cells in self assembling peptide nanofiber for healing critical size calvarial bony defect

    PubMed Central

    Wu, Guofeng; Pan, Mengjie; Wang, Xianghai; Wen, Jinkun; Cao, Shangtao; Li, Zhenlin; Li, Yuanyuan; Qian, Changhui; Liu, Zhongying; Wu, Wutian; Zhu, Lixin; Guo, Jiasong

    2015-01-01

    Peripheral blood mesenchymal stem cells (PBMSCs) may be easily harvested from patients, permitting autologous grafts for bone tissue engineering in the future. However, the PBMSC’s capabilities of survival, osteogenesis and production of new bone matrix in the defect area are still unclear. Herein, PBMSCs were seeded into a nanofiber scaffold of self-assembling peptide (SAP) and cultured in osteogenic medium. The results indicated SAP can serve as a promising scaffold for PBMSCs survival and osteogenic differentiation in 3D conditions. Furthermore, the SAP seeded with the induced PBMSCs was splinted by two membranes of poly(lactic)-glycolic acid (PLGA) to fabricate a composited scaffold which was then used to repair a critical-size calvarial bone defect model in rat. Twelve weeks later the defect healing and mineralization were assessed by H&E staining and microcomputerized tomography (micro-CT). The osteogenesis and new bone formation of grafted cells in the scaffold were evaluated by immunohistochemistry. To our knowledge this is the first report with solid evidence demonstrating PBMSCs can survive in the bone defect area and directly contribute to new bone formation. Moreover, the present data also indicated the tissue engineering with PBMSCs/SAP/PLGA scaffold can serve as a novel prospective strategy for healing large size cranial defects. PMID:26568114

  17. Carbon Sequestered, Carbon Displaced and the Kyoto Context

    SciTech Connect

    Marland, G.; Schlamadinger, B.

    1999-04-18

    The integrated system that embraces forest management, forest products, and land-use change impacts the global carbon cycle - and hence the net emission of the greenhouse gas carbon dioxide - in four fundamental ways. Carbon is stored in living and dead biomass, carbon is stored in wood products and landfills, forest products substitute in the market place for products made from other materials, and forest harvests can be used wholly or partially to displace fossil fuels in the energy sector. Implementation of the Kyoto Protocol to the United Nations Framework Convention on Climate Change would result in the creation of international markets for carbon dioxide emissions credits, but the current Kyoto text does not treat all carbon identically. We have developed a carbon accounting model, GORCAM, to examine a variety of scenarios for land management and the production of forest products. In this paper we explore, for two simple scenarios of forest management, the carbon flows that occur and how these might be accounted for under the Kyoto text. The Kyoto protocol raises questions about what activities can result in emissions credits, which carbon reservoirs will be counted, who will receive the credits, and how much credit will be available? The Kyoto Protocol would sometimes give credits for carbon sequestered, but it would always give credits when fossil-fuel carbon dioxide emissions are displaced.

  18. Universal Behavior of the BEC Critical Temperature for a Multi-slab Ideal Bose Gas

    NASA Astrophysics Data System (ADS)

    Rodríguez, O. A.; Solís, M. A.

    2016-05-01

    For an ideal Bose gas within a multi-slab periodic structure, we discuss the effect of the spatial distribution of the gas on its Bose-Einstein condensation critical temperature T_c, as well as on the origin of its dimensional crossover observed in the specific heat. The multi-slabs structure is generated by applying a Kronig-Penney potential to the gas in the perpendicular direction to the slabs of width b and separated by a distance a, and allowing the particles to move freely in the other two directions. We found that T_c decreases continuously as the potential barrier height increases, becoming inversely proportional to the square root of the barrier height when it is large enough. This behavior is universal as it is independent of the width and spacing of the barriers. The specific heat at constant volume shows a crossover from 3D to 2D when the height of the potential or the barrier width increases, in addition to the well-known peak related to the Bose-Einstein condensation. These features are due to the trapping of the bosons by the potential barriers and can be characterized by the energy difference between the energy bands below the potential height.

  19. Histidine 114 Is Critical for ATP Hydrolysis by the Universally Conserved ATPase YchF.

    PubMed

    Rosler, Kirsten S; Mercier, Evan; Andrews, Ian C; Wieden, Hans-Joachim

    2015-07-24

    GTPases perform a wide range of functions, ranging from protein synthesis to cell signaling. Of all known GTPases, only eight are conserved across all three domains of life. YchF is one of these eight universally conserved GTPases; however, its cellular function and enzymatic properties are poorly understood. YchF differs from the classical GTPases in that it has a higher affinity for ATP than for GTP and is a functional ATPase. As a hydrophobic amino acid-substituted ATPase, YchF does not possess the canonical catalytic Gln required for nucleotide hydrolysis. To elucidate the catalytic mechanism of ATP hydrolysis by YchF, we have taken a two-pronged approach combining classical biochemical and in silico techniques. The use of molecular dynamics simulations allowed us to complement our biochemical findings with information about the structural dynamics of YchF. We have thereby identified the highly conserved His-114 as critical for the ATPase activity of YchF from Escherichia coli. His-114 is located in a flexible loop of the G-domain, which undergoes nucleotide-dependent conformational changes. The use of a catalytic His is also observed in the hydrophobic amino acid-substituted GTPase RbgA and is an identifier of the translational GTPase family. PMID:26018081

  20. Low enrichment fuel conversion for Iowa State University. Progress report, August 1, 1991--July 31, 1992

    SciTech Connect

    Rohach, A.F.

    1992-08-01

    This report discusses the UTR-10 reactor at Iowa State University which went critical on low enriched uranium (LEU) fuel on August 14, 1991. However, subsequent to the criticality experiments the fuel plates started to discolor. In addition, roll pins used to lift the fuel assemblies were discovered to be cracked. It was determined that these problems were due to chemical agents in the primary coolant water. The roll pins were replaced by solid stainless steel pins. The primary coolant was replaced and the reactor is currently in operation. Surveillance specimens will be used to monitor any possible future discoloration. The high enriched fuel (HEU) is being prepared for eventual shipment to a high enriched fuel receiving facility.