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Sample records for kyoto university critical assembly reactor

  1. Positron beam facility at Kyoto University Research Reactor

    NASA Astrophysics Data System (ADS)

    Xu, Q.; Sato, K.; Yoshiie, T.; Sano, T.; Kawabe, H.; Nagai, Y.; Nagumo, K.; Inoue, K.; Toyama, T.; Oshima, N.; Kinomura, A.; Shirai, Y.

    2014-04-01

    A positron beam facility is presently under construction at the Kyoto University Research Reactor (KUR), which is a light-water moderated tank-type reactor operated at a rated thermal power of 5 MW. A cadmium (Cd) - tungsten (W) source similar to that used in NEPOMUC was chosen in the KUR because Cd is very efficient at producing γ-rays when exposed to thermal neutron flux, and W is a widely used in converter and moderator materials. High-energy positrons are moderated by a W moderator with a mesh structure. Electrical lenses and a solenoid magnetic field are used to extract the moderated positrons and guide them to a platform outside of the reactor, respectively. Since Japan is an earthquake-prone country, a special attention is paid for the design of the in-pile positron source so as not to damage the reactor in the severe earthquake.

  2. Development of a mono-energetic positron beam line at the Kyoto University Research Reactor

    NASA Astrophysics Data System (ADS)

    Sato, K.; Xu, Q.; Yoshiie, T.; Sano, T.; Kawabe, H.; Nagai, Y.; Nagumo, K.; Inoue, K.; Toyama, T.; Oshima, N.; Kinomura, A.; Shirai, Y.

    2015-01-01

    Positron beam facilities are widely used for solid state physics and material science studies. A positron beam facility has been constructed at the Kyoto University Research Reactor (KUR) in order to expand its application range. The KUR is a light-water-moderated tank-type reactor operated at a rated thermal power of 5 MW. A positron beam has been transported successfully from the reactor to the irradiation chamber. The total moderated positron rate was greater than 1.4 × 106/s while the reactor operated at a reduced power of 1 MW. Special attention was paid for the design of the in-pile position source to prevent possible damage of the reactor in case of severe earthquakes.

  3. Nickel Mirror And Supermirror Neutron Guide Tubes At The Kyoto University Research Reactor

    NASA Astrophysics Data System (ADS)

    Ebisawa, Toru; Akiyoshi, Tsunekazu; Tasaki, Seiji; Kawai, Takeshi; Achiwa, Norio; Utsuro, Masahiko; Okamoto, Sunao

    1989-01-01

    We installed the first nickel mirror neutron guide tube with a characteristic wavelength of 2.85 Å at Kyoto University research reactor(KUR, 5MW, cooled and moderated by light water) in 1973 and a supermirror guide tube with a characteristic wavelength of 1.17 Å in 1984, in order to get more intense thermal neutron beam. Four guide tubes are under construction at a cold neutron source installed in 1986. Two of them are supermirror type with a characteristic wavelength of 3 Å and the others are supermirror and Ni-mirror type with characteristic wavelengths of 6 Å and 23 Å, respectively. Supermirrors are made by automatically controlled vacuum deposition of nickel and titanium metal with electron gun. Their averaging reflectivity for the first supermirror guide tube are the following: The apparent critical wavelength, λ/θ, of reflection is 240 Å in term of wavelength(λ/θ) corresponding to the component of wave number perpendicular to the mirror surface. The reflectivity is 0.65 at the apparent critical wavelength and becomes higher with increasing neutron wavelength up to nearly unity for wavelength longer than 500 Å. Supermirror guide tubes are featured by more available neutrons with larger divergent angles and shorter length of the guide tubes. These features would bring us significant advantages depending on experimental requirements.

  4. High contrast neutron radiography with optical devices in Kyoto University reactor

    NASA Astrophysics Data System (ADS)

    Kawabata, Y.; Nakano, T.; Hino, M.; Sunohara, H.; Matsushima, U.; Takenaka, N.

    2004-08-01

    The high-contrast neutron radiography has been performed at a VCN guide (VCN) and a supermirror cold neutron guide (CN-3) in Kyoto University Reactor. The large absorption cross-section of very low-energy neutrons can show a slight change of sample which thermal neutrons can not show. The effectiveness is shown in the fields of botany, agriculture and industrial researches. A new spectrum change option using high Qc supermirror ( m=4) is attached. It can change the upper limit of the energy of exposure neutrons by reflections, and gives a high flexibility of the experimental condition.

  5. 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.

  6. Reactor Dynamics Experiments with a Sub-Critical Assembly

    SciTech Connect

    Miley, G.H.; Yang, Y.; Wu, L.; Momota, H.

    2004-10-06

    A resurgence in use of nuclear power is now underway worldwide. However due to the shutdown of many university research reactors , student laboratories must rely more heavily on use of sub-critical assemblies. Here a driven sub-critical is described that uses a cylindrical Inertial Electrostatic Confinement (IEC) device to provide a fusion neutron source. The small IEC neutron source would be inserted in a fuel element position, with its power input controlled externally at a control panel. This feature opens the way to use of the critical assembly for a number of transient experiments such as sub-critical pulsing and neutron wave propagation. That in turn adds important new insights and excitement for the student teaching laboratory.

  7. 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.

  8. 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)

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

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

    DOE PAGES

    Bostelmann, Friederike; Hammer, Hans R.; Ortensi, Javier; ...

    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

  16. NEUTRONIC REACTOR BURIAL ASSEMBLY

    DOEpatents

    Treshow, M.

    1961-05-01

    A burial assembly is shown whereby an entire reactor core may be encased with lead shielding, withdrawn from the reactor site and buried. This is made possible by a five-piece interlocking arrangement that may be easily put together by remote control with no aligning of bolt holes or other such close adjustments being necessary.

  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. [Approach to Teaching Kampo Medicine at Kyoto Pharmaceutical University].

    PubMed

    Matsuda, Hisashi

    2016-01-01

    An approach to educating our pharmaceutical students about Kampo medicine in the six-year system of undergraduate pharmacy education at Kyoto Pharmaceutical University is introduced, including the author's opinions. Curriculum revisions have been made in our university for students entering after 2012. In teaching Kampo medicine at present, a medical doctor and an on-site pharmacist share information difficult to give in a lecture with the teaching staff in my laboratory. For example, before the curriculum revision, we conferred with a pharmacist and a doctor in the course "Kampo Medicine A, B" for 4th year students, in which students were presented a basic knowledge of Kampo medicine, the application of important Kampo medicines, combinations of crude drugs, etc. Further, in our "Introduction to Kampo Medicine" for 6th year students, presented after they have practiced in hospitals and community pharmacies, we again lecture on the pharmacological characteristics of Kampo medicines, on "pattern (Sho)", and on evidence-based medicine (EBM) and research studies of important Kampo medicines. After our curriculum revision, "Kampo Medicine A, B" was rearranged into the courses "Kampo and Pharmacognosy" and "Clinical Kampo Medicine". "Kampo and Pharmacognosy" is now provided in the second semester of the 3rd year, and in this course we lecture on the basic knowledge of Kampo medicine. An advanced lecture will be given on "Clinical Kampo Medicine" in the 6th year. We are searching for the best way to interest students in Kampo medicine, and to counteract any misunderstandings about Kampo medicine.

  19. University Reactor Sharing Program

    SciTech Connect

    Dr. W.D. Reece

    1999-09-01

    The University Reactor Sharing Program provides funding for reactor experimentation to institutions that do not normally have access to a research reactor. Research projects supported by the program include items such as dating geological material to producing high current super conducting magnets. The funding also gives small colleges and universities the opportunity to use the facility for teaching courses in nuclear processes; specifically neutron activation analysis and gamma spectroscopy.

  20. 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…

  1. 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…

  2. Derivation of criticality safety benchmarks from ZPR fast critical assemblies

    SciTech Connect

    Schaefer, R.W.; McKnight, R.D.

    1997-12-01

    Scores of critical assemblies were constructed, over a period of about three decades, at the Argonne National Laboratory ZPR-3, ZPR-6, ZPR-9, and zero power plutonium reactor (ZPPR) fast critical assembly facilities. Most of the assemblies were mock-ups of various liquid-metal fast breeder reactor designs. These tended to be complex, containing, for example, mock-ups of control rods and control rod positions. Some assemblies, however, were {open_quotes}physics benchmarks.{close_quotes} These relatively {open_quotes}clean{close_quotes} assemblies had uniform compositions and simple geometry and were designed to test fast reactor physics data and methods. Assemblies in this last category are well suited to form the basis for new criticality safety benchmarks. The purpose of this paper is to present an overview of some of these benchmark candidates and to describe the strategy being used to create the benchmarks.

  3. 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.

  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. Derivation of criticality safety benchmarks from ZPR fast critical assemblies

    SciTech Connect

    Schaefer, R.W.; McKnight, R.D.

    1997-09-01

    Scores of critical assemblies were constructed, over a period of about three decades, at the Argonne National Laboratory ZPR-3, ZPR-6, ZPR-9, and ZPPR fast critical assembly facilities. Most of the assemblies were mockups of various liquid-metal fast breeder reactor designs. These tended to be complex, containing, for example, mockups of control rods and control rod positions. Some assemblies, however, were `physics benchmarks`. These relatively `clean` assemblies had uniform compositions and simple geometry and were designed to test fast reactor physics data and methods. Assemblies in this last category are well suited to form the basis for new criticality safety benchmarks. The purpose of this paper is to present an overview of some of these benchmark candidates and to describe the strategy being used to create the benchmarks.

  6. 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)

  7. 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)

  8. 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

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

    SciTech Connect

    Kahler, A.; 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, Micheal W; Arcilla, r; Mughabghab, S F; Sublet, J C; Trkov, A.; Trumbull, T H; Dunn, Michael E

    2011-01-01

    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 [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 unrnoderated and uranium reflected (235)U and (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 selected actinide reaction rates such as (236)U; (238,242)Pu and (241,243)Am capture in fast systems. Other deficiencies, such as the overprediction of Pu solution system critical eigenvalues

  10. 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

  11. Preparation and planning for the replacement of the Oregon State University TRIGA reactor rotary specimen rack assembly

    SciTech Connect

    Anderson, T.V.; Dodd, B.; Johnson, A.G.; Carpenter, W.T.

    1984-07-01

    Recently there have been a number of indications that the rotating rack may be approaching the end of its useful life. In order to benefit from the experience of other reactors who have removed and replaced their rotating racks, General Atomic (GA) was contacted and previous TRIGA Conference proceedings were scanned. It was determined that a number of facilities, had experienced difficulties with their lazy susans and eventually had to replace them. However, most of the written descriptions of this project were not sufficiently detailed to be of great use. The purpose of this paper is to identify some of the more important questions related to the replacement of our rotating rack assembly and OSU's currently proposed solutions, with a view towards soliciting ideas from other members of the TRIGA reactor community.

  12. 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.

  13. 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.

  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. Effort of Course Evaluation Survey and Its Characteristics at Faculty of Engineering of Kyoto University

    NASA Astrophysics Data System (ADS)

    Otsuka, Yusaku; Matsushita, Kayo; Yuasa, Taiichi; Araki, Mituhiko

    At Faculty of Engineering of Kyoto University, the course evaluation survey was carried out to 120 lectures in the second semester of the 2004 fiscal year. In this paper, the structure of the course evaluation questionnaire was clarified, and the feature of the education of the faculty was described based on the results. Although the lectures seem generally to be well-organized and the students had attended the classes seriously, some problems also had been observed. For example, the wide variation of class means would suggest some lectures might not be so effective. Besides, the lack of active self-study of the students was shown. For improvement of engineering education in Kyoto University, it was suggested that a further follow-up survey from entrance to graduation should be carried out.

  16. University Reactor Sharing Program

    SciTech Connect

    W.D. Reese

    2004-02-24

    Research projects supported by the program include items such as dating geological material and producing high current super conducting magnets. The funding continues to give small colleges and universities the valuable opportunity to use the NSC for teaching courses in nuclear processes; specifically neutron activation analysis and gamma spectroscopy. The Reactor Sharing Program has supported the construction of a Fast Neutron Flux Irradiator for users at New Mexico Institute of Mining and Technology and the University of Houston. This device has been characterized and has been found to have near optimum neutron fluxes for A39/Ar 40 dating. Institution final reports and publications resulting from the use of these funds are on file at the Nuclear Science Center.

  17. [Dr. Michiharu Matsuoka, founder of the Department of Orthopaedic Surgery, Kyoto University, and his achievements. Part 4: Prof. M. Matsuoka's lecture to medical and civic communities].

    PubMed

    Hirotani, Hayato

    2010-03-01

    Dr. M. Matsuoka gave many lectures to physicians at the Postdoctoral Course Lectures sponsored by the Kyoto Eisei Kensasho (Kyoto Bacterial and Biochemical Laboratory) run by the Kyoto Medical Association, and the Postdoctoral Course Lectures of the Kyoto Medical School, Kyoto Imperial University. He was also invited to give lectures at several regional medical associations. He also was a speaker at the Kyoto Imperial University Extension course and he lectured at the Enryakuji Temple on Mt. Hiei, sponsored by a newspaper company. It is remarkable that these activities were carried out in addition to his other notable academic work previously reported.

  18. 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)

  19. 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.

  20. Scanning tunneling microscope assembly, reactor, and system

    SciTech Connect

    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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. A survey of attitudes toward clinical research among physicians at Kyoto University Hospital.

    PubMed

    Sumi, Eriko; Murayama, Toshinori; Yokode, Masayuki

    2009-12-22

    In Japan, only clinical research related to investigational new drug trials must be notified to regulatory bodies, and this lack of a uniform standard for clinical research has caused a number of difficulties. The objective of this study was to assess the willingness of physicians to participate in clinical research and to identify effective methods to promote and enhance clinical research. We conducted a cross-sectional survey by administrating questionnaires to physicians in 31 departments in Kyoto University Hospital from October through November 2007. A total of 51.5% (310 of 602) of physicians completed the questionnaire. More than two-thirds of them reported currently participating in clinical research, and nearly all believed that clinical research is necessary for physicians. Less than 20% of respondents had specific training regarding clinical research, and most reported a need to acquire concepts and skills regarding clinical research, especially those related to statistics. "Paperwork was complicated and onerous" was the most frequently cited obstacle in conducting clinical research, followed by "few eligible patients" and "lack of time". Previous participation in and prospective participation in clinical research, previous writing a research protocol were positively associated with current participation in clinical research. Physicians in university hospitals need more training regarding clinical research, particularly in biostatistics. They also require administrative assistance. Our findings indicate that the quality of clinical research could be improved if training in clinical research methodology and biostatistics were provided, and if greater assistance in the preparation of study documents requested by the institutional Independent Ethics Committee were available.

  6. A survey of attitudes toward clinical research among physicians at Kyoto University Hospital

    PubMed Central

    2009-01-01

    Background In Japan, only clinical research related to investigational new drug trials must be notified to regulatory bodies, and this lack of a uniform standard for clinical research has caused a number of difficulties. The objective of this study was to assess the willingness of physicians to participate in clinical research and to identify effective methods to promote and enhance clinical research. Methods We conducted a cross-sectional survey by administrating questionnaires to physicians in 31 departments in Kyoto University Hospital from October through November 2007. Results A total of 51.5% (310 of 602) of physicians completed the questionnaire. More than two-thirds of them reported currently participating in clinical research, and nearly all believed that clinical research is necessary for physicians. Less than 20% of respondents had specific training regarding clinical research, and most reported a need to acquire concepts and skills regarding clinical research, especially those related to statistics. "Paperwork was complicated and onerous" was the most frequently cited obstacle in conducting clinical research, followed by "few eligible patients" and "lack of time". Previous participation in and prospective participation in clinical research, previous writing a research protocol were positively associated with current participation in clinical research. Conclusions Physicians in university hospitals need more training regarding clinical research, particularly in biostatistics. They also require administrative assistance. Our findings indicate that the quality of clinical research could be improved if training in clinical research methodology and biostatistics were provided, and if greater assistance in the preparation of study documents requested by the institutional Independent Ethics Committee were available. PMID:20025782

  7. 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.

  8. [Experience of Collaborative Research through Department of Medical Instrumental Research and Technology in Kyoto Prefectural University of Medicine].

    PubMed

    Saitoh, Kensuke

    2016-01-01

    Both of Kyoto Prefectural University of Medicine which offers high, technical and safe medical treatment and Horiba, Ltd. which has small CBC analyzers in a core product established a joint research institute for development of advanced laboratory test analyzer from January 1, 2012 in Kyoto Prefectural University of Medicine as the "advanced treatment hospital" where the Ministry of Health, Labour and Welfare has got approved. Clinical needs about analyzer and reagent for a laboratory test are being investigated to the emergency medical care unit and the intensive care unit as well as the laboratory test part in the affiliated hospital and many medical departments of the pediatrics, the internal medicine and the surgery. Developing the new analyzer based on high technology, evaluating the performance of them and spreading them to a medical examination and treatment site is our main target.

  9. 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.

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

    PubMed

    Ito-Ihara, Toshiko; Hong, Jeong-Hwa; Kim, Ock-Joo; Sumi, Eriko; Kim, Soo-Youn; Tanaka, Shiro; Narita, Keiichi; Hatta, Taichi; Choi, Eun-Kyung; Choi, Kyu-Jin; Miyagawa, Takuya; Minami, Manabu; Murayama, Toshinori; Yokode, Masayuki

    2013-10-25

    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. 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. 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. 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 encourage doctors to participate in and

  11. 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

  12. Critical Casimir forces for colloidal assembly.

    PubMed

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

    2016-02-03

    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.

  13. Arkansas Tech University TRIGA nuclear reactor

    SciTech Connect

    Sankoorikal, J.; Culp, R.; Hamm, J.; Elliott, D.; Hodgson, L.; Apple, S.

    1990-07-01

    This paper describes the TRIGA nuclear reactor (ATUTR) proposed for construction on the campus of Arkansas Tech University in Russellville, Arkansas. The reactor will be part of the Center for Energy Studies located at Arkansas Tech University. The reactor has a steady state power level of 250 kW and can be pulsed with a maximum reactivity insertion of $2.0. Experience gained in dismantling and transporting some of the components from Michigan State University, and the storage of these components will be presented. The reactor will be used for education, training, and research. (author)

  14. 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.

  15. Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores

    SciTech Connect

    Krass, A.W.

    2005-12-19

    This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. The material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.

  16. 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.

  17. VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

    SciTech Connect

    Doshi, P.K.; DeMario, E.E.; Knott, R.P.

    1993-12-31

    Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR`s) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design.

  18. 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.

  19. 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.

  20. Fast critical experiment data for space reactors

    SciTech Connect

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

    1987-01-01

    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, the old experiments require recalculation with modern computational methods and nuclear cross section data before they can be applied to today's designs. Recently, we have calculated about 20 fast benchmark critical experiments 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.

  1. 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.

  2. 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

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

    SciTech Connect

    Hutter, Ernest

    1986-01-01

    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.

  4. 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.

  5. 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.

  6. Universal Fast Breeder Reactor Subassembly Counter manual

    SciTech Connect

    Menlove, H.O.; Eccleston, G.W.; Swansen, J.E.; Goris, P.; Abedin-Zadeh, R.; Ramalho, A.

    1984-08-01

    A neutron coincidence counter has been designed for the measurement of fast breeder reactor fuel assemblies. This assay system can accommodate the full range of geometries and masses found in fast breeder subassemblies under IAEA safeguards. The system's high-performance capability accommodates high plutonium loadings of up to 16 kg. This manual describes the system and its operation and gives performance and calibration parameters for typical applications.

  7. 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.

  8. 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.

  9. Lessons learned from applying VIM to fast reactor critical experiments

    SciTech Connect

    Schaefer, R.W.; McKnight, R.D.; Collins, P.J.

    1995-05-17

    VIM is a continuous energy Monte Carlo code first developed around 1970 for the analysis of plate-type, fast-neutron, zero-power critical assemblies. In most respects, VIM is functionally equivalent to the MCNP code but it has two features that make uniquely suited to the analysis of fast reactor critical experiments: (1) the plate lattice geometry option, which allows efficient description of and neutron tracking in the assembly geometry, and (2) a statistical treatment of neutron cross section data in the unresolved resonance range. Since its inception, VIM`s capabilities have expanded to include numerous features, such as thermal neutron cross sections, photon cross sections, and combinatorial and other geometry options, that have allowed its use in a wide range of neutral-particle transport problems. The earliest validation work at Argonne National Laboratory (ANL) focused on the validation of VIM itself. This work showed that, in order for VIM to be a ``rigorous`` tool, extreme detail in the pointwise Monte Carlo libraries was needed, and the required detail was added. The emphasis soon shifted to validating models, methods, data and codes against VIM. Most of this work was done in the context of analyzing critical experiments in zero power reactor (ZPR) assemblies. The purpose of this paper is to present some of the lessons learned from using VIM in ZPR analysis work. This involves such areas as uncovering problems in deterministic methods and models, pitfalls in using Monte Carlo codes, and improving predictions. The numerical illustrations included here were taken from the extensive documentation cited as references.

  10. Pre-bomb marine reservoir ages in the western north Pacific: Preliminary result on Kyoto University collection

    NASA Astrophysics Data System (ADS)

    Yoneda, Minoru; Kitagawa, Hiroyuki; van der Plicht, Johannes; Uchida, Masao; Tanaka, Atsushi; Uehiro, Takashi; Shibata, Yasuyuki; Morita, Masatoshi; Ohno, Terufumi

    2000-10-01

    The calibration of radiocarbon dates on marine materials involves a global marine calibration with regional corrections. The marine reservoir ages in the Western North Pacific have not been discussed, while it is quite important to determine the timing of palaeo-environmental changes as well as archaeological interpretation around this region. The lack of adequate collection of the pre-bomb shell from western north Pacific was the biggest problem. Recently we had a chance to examine specimens from an old shell collection stored in Kyoto University, including shell specimens from Japan, Korea, Taiwan and the Micronesia of 1920s and 1930s. We explored the possibility for usage of specimen without clear evidence of live collection by measuring 30 apparent radiocarbon ages of pre-bomb mollusk shells from 18 sites in Western North Pacific. The preliminary results showed several discrepancies with previously reported results and with each other. We have to carefully select the shell specimen that has biological signs such as articulating fulcrum. In order to exploit this big resource of pre-bomb shell collection, the new technique to distinguish fossils from live collected samples should be developed by using chemical and physical methods.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. [Prof. Michiharu Matsuoka, founder of the Department of Orthopaedic Surgery, Kyoto University, and his achievements in orthopaedic surgery in the Meiji Era of Japan (part 1: establishment of the department)].

    PubMed

    Hirotani, Hayato

    2005-09-01

    The Department of Orthopaedic and Musculoskeletal Surgery, Graduate School of Medicine, Kyoto University (formerly the Department of Orthopaedic Surgery, Kyoto Medical School, Kyoto Imperial University) was founded by Imperial Ordinance, Article No. 89 issued on April 23, 1906. On May 4, 1906, Dr. Shinichiro Asahara, Assistant Professor of the Department of Surgery, was appointed as the first director of the Department of Orthopaedic Surgery, Kyoto Medical School, Kyoto Imperial University. Dr. Michiharu Matsuoka, Assistant Doctor of the Department of Surgery, Tokyo Medical School, Imperial University of Tokyo, was appointed Assistant Professor of Surgery, Kyoto Medical School, Kyoto Imperial University in March 1901. From August 1903 to May 1906, he studied orthopaedic surgery in Germany and returned on May 5, 1906. Dr. Matsuoka was appointed as the director and chief of the Department on May 13, 1906 and took over Dr. Asahara's position. On June 18, 1906, Dr. Matsuoka started his clinic and began giving lectures on orthopaedic surgery. This was the first department of orthopaedic surgery among the Japanese medical schools. Dr. Matsuoka was appointed as Professor in 1907. He had to overcome several obstacles to establish the medical department of a new discipline that had never existed in Japanese medical schools. This article discusses Dr. Matsuoka's contributions to establishing and developing orthopaedic surgery in Japan in the Meiji-era.

  20. Evaluated benchmark experiments at critical assemblies simulating features of an HTHR at the ASTRA facility

    SciTech Connect

    Garin, V. P.; Glushkov, A. E.; Glushkov, E. S.; Gomin, E. A.; Gurevich, M. I.; Zimin, A. A.; Kompaniets, G. V.; Kukharkin, N. E.; Lobyntsev, V. A.; Nosov, V. I.; Polyakov, D. N.; Ponomarev-Stepnoi, N. N.; Smirnov, O. N.; Tel'kovskaya, O. V.; Chunyaev, E. I.

    2010-12-15

    The design of the ASTRA facility and critical assemblies that simulate physics features of modular high-temperature reactors (HTHR-Ms) with a graphite moderator and reflectors loaded with fuel particles having multilayer ceramic coatings is described in detail. Geometrical dimensions of the main elements and regions of the critical assemblies, composition of the materials used, and experimental results for various configurations of the critical assemblies are presented. A detailed computational benchmark model allowing for the structural and compositional features of the critical assembly configurations in question is developed on the basis of all the above data. The results are to be used for verification of the neutronics codes used for calculations of high-temperature helium-cooled reactors.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Measuring the productivity of university research reactors

    SciTech Connect

    Voth, M.H.

    1989-11-01

    University Research Reactors (URRs) on 33 campuses in the United States provide valuable contributions to academic instruction and research programs. In most cases, there are no alternative diagnostic techniques to supplant the need for a reactor and associated facilities. Since URRs constitute a major financial commitment, it is important that they be operated in a productive manner. Productivity may be defined as the sum of new knowledge generated, existing knowledge transferred to others, and analytical services provided to assist in the generation of new knowledge; another definition of productivity is this sum expressed as a function of the cost incurred. In either case, a consistent measurement is difficult and more qualitative than quantitative. A uniform reporting system has been proposed that defines simplified categories through which meaningful comparisons can be performed.

  7. Halo assembly bias from Separate Universe simulations

    NASA Astrophysics Data System (ADS)

    Paranjape, Aseem; Padmanabhan, Nikhil

    2017-07-01

    We present a calibration of halo assembly bias using the Separate Universe technique. Specifically, we measure the response of halo abundances at fixed mass and concentration to the presence of an infinite-wavelength initial perturbation. We develop an analytical framework for describing the concentration dependence of this peak-background split halo bias - a measure of assembly bias - relying on the near-lognormal distribution of halo concentration at fixed halo mass. The combination of this analytical framework and the Separate Universe technique allows us to achieve very high precision in the calibration of the linear assembly bias b1, and qualitatively reproduces known trends such as the monotonic decrease (increase) of b1 with halo concentration at large (small) masses. The same framework extends to the concentration dependence of higher order bias parameters bn, and we present the first calibration of assembly bias in b2. Our calibrations are directly applicable in analytical Halo Model calculations that seek to robustly detect galaxy assembly bias in observational samples. We detect a non-universality in the b1-b2 relation arising from assembly bias, and suggest that simultaneous measurements of these bias parameters could be used to both detect the signature of assembly bias as well as mitigate its effects in cosmological analyses.

  8. 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.

  9. 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.

  10. Detecting pin diversion from pressurized water reactors spent fuel assemblies

    DOEpatents

    Ham, Young S.; Sitaraman, Shivakumar

    2017-01-10

    Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and taking the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.

  11. 76 FR 14436 - University of Wisconsin, University of Wisconsin Nuclear Reactor; Notice of Issuance of...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-16

    ... COMMISSION University of Wisconsin, University of Wisconsin Nuclear Reactor; Notice of Issuance of... which would authorize continued operation of the University of Wisconsin Nuclear Reactor. This action is... CONTACT: Geoffrey A. Wertz, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory...

  12. 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

  13. 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.

  14. Physical characteristics of GE (General Electric) BWR (boiling-water reactor) fuel assemblies

    SciTech Connect

    Moore, R.S.; Notz, K.J.

    1989-06-01

    The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs.

  15. 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.

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

    SciTech Connect

    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.

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

    DOE PAGES

    Vo, D.; Favalli, A.; Grogan, B.; ...

    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 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

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

    SciTech Connect

    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.

  19. 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.

  20. 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.

  1. 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

  2. Pressurized water reactor fuel assembly subchannel void fraction measurement

    SciTech Connect

    Akiyama, Yoshiei; Hori, Keiichi; Miyazaki, Keiji; Mishima, Kaichiro; Sugiyama, Shigekazu

    1995-12-01

    The void fraction measurement experiment of pressurized water reactor (PWR) fuel assemblies has been conducted since 1987 under the sponsorship of the Ministry of International Trade and Industry as a Japanese national project. Two types of test sections are used in this experiment. One is a 5 x 5 array rod bundle geometry, and the other is a single-channel geometry simulating one of the subchannels in the rod bundle. Wide gamma-ray beam scanners and narrow gamma-ray beam computed tomography scanners are used to measure the subchannel void fractions under various steady-state and transient conditions. The experimental data are expected to be used to develop a void fraction prediction model relevant to PWR fuel assemblies and also to verify or improve the subchannel analysis method. The first series of experiments was conducted in 1992, and a preliminary evaluation of the data has been performed. The preliminary results of these experiments are described.

  3. Decommissioning of a university research reactor.

    PubMed

    Abelquist, E W; Huda, A; State, S; Takahashi, J

    1994-07-01

    The objective of the UCLA Boelter Reactor Decommissioning Project was the release of the Boelter reactor facility for unrestricted use. The facility included a 100 kW Argonaut type research reactor that operated from 1963 to 1985, providing general reactor research. The decommissioning was planned as a two-phase program. Phase I involved removal of the reactor core structure for better access and assessment of the biological shield. Phase II decommissioning activities included structural steel removal, activated concrete removal, process equipment pit piping dismantlement, and clean concrete removal. The final release survey of the Boelter reactor facility demonstrated that all areas satisfied the project's release criteria. The total person-Sv for the project was 3.87 x 10(-2) (3.87 person-rem), most of which was received during the structural steel and activated concrete removal tasks and the disassembly of the reactor core.

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

    DOE PAGES

    Williams, M. L.; Wiarda, D.; Ilas, G.; ...

    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.

  5. 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.

  6. 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.

  7. U.S. Department of Energy University Reactor Instrumentation Program Final Report for 1992-94 Grant for the University of Florida Training Reactor

    SciTech Connect

    Vernetson, William G.

    1999-04-01

    Overall, the instrumentation obtained under the first year 1992-93 University Reactor Instrumentation Program grant assured that the goals of the program were well understood and met as well as possible at the level of support provided for the University of Florida Training Reactor facility. Though the initial grant support of $21,000 provided toward the purchase of $23,865 of proposed instrumentation certainly did not meet many of the facility's needs, the instrumentation items obtained and implemented did meet some critical needs and hence the goals of the Program to support modernization and improvement of reactor facilities such as the UFTR within the academic community. Similarly, the instrumentation obtained under the second year 1993-94 University Reactor Instrumentation Program grant again met some of the critical needs for instrumentation support at the UFTR facility. Again, though the grant support of $32,799 for proposed instrumentation at the same cost projection does not need all of the facility's needs, it does assure continued facility viability and improvement in operations. Certainly, reduction of forced unavailability of the reactor is the most obvious achievement of the University Reactor Instrumentation Program to date at the UFTR. Nevertheless, the ability to close out several expressed-inspection concerns of the Nuclear Regulatory Commission with acquisition of the low level survey meter and the area radiation monitoring system is also very important. Most importantly, with modest cost sharing the facility has been able to continue and even accelerate the improvement and modernization of a facility, especially in the Neutron Activation Analysis Laboratory, that is used by nearly every post-secondary school in the State of Florida and several in other states, by dozens of departments within the University of Florida, and by several dozen high schools around the State of Florida on a regular basis. Better, more reliable service to such a broad

  8. Performance Assessment of ISS Water Processor Assembly Reactor

    NASA Technical Reports Server (NTRS)

    Carter, Layne; Tatara, James; Mason, Rich; OConner, Ed; Bedard, John

    2004-01-01

    Due to modifications to the ISS waste water composition, the concentration of volatile organics has significantly increased in the feed to the Water Processor Assembly (WPA). In parallel, the oxygen supply pressure increased, resulting in a higher flow rate of oxygen to the WPA. Preliminary testing at Hamilton Sund strand indicated that the higher oxygen flow rate would increase the WPA capacity for volatile organics. Following an analysis of the expected waste water composition, personnel at NASA MSFC and Hamilton Sundstrand conducted a test of a flight-like reactor to assess its capacity for the higher organic loads. The results of this test indicate the WPA can accommodate the expected organic load in the ISS waste water with margin.

  9. Performance Assessment of ISS Water Processor Assembly Reactor

    NASA Technical Reports Server (NTRS)

    Carter, Layne; Tatara, James; Mason, Rich; OConner, Ed; Bedard, John

    2004-01-01

    Due to modifications to the ISS waste water composition, the concentration of volatile organics has significantly increased in the feed to the Water Processor Assembly (WPA). In parallel, the oxygen supply pressure increased, resulting in a higher flow rate of oxygen to the WPA. Preliminary testing at Hamilton Sund strand indicated that the higher oxygen flow rate would increase the WPA capacity for volatile organics. Following an analysis of the expected waste water composition, personnel at NASA MSFC and Hamilton Sundstrand conducted a test of a flight-like reactor to assess its capacity for the higher organic loads. The results of this test indicate the WPA can accommodate the expected organic load in the ISS waste water with margin.

  10. 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

  11. Depolymerization using sonochemical reactors: A critical review.

    PubMed

    Gogate, Parag R; Prajapat, Amrutlal L

    2015-11-01

    Ultrasonic irradiation has been proposed as a novel approach for degradation of polymer compounds, especially considering the fact that the reduction in the molecular weight (also the intrinsic viscosity) is simply by splitting the most susceptible chemical bond without causing any changes in the chemical nature. An overview of the application of ultrasound for the polymer degradation has been presented in this work, discussing the mechanism of degradation, kinetic modeling, effect of operating parameters and the type of reactors generally used for depolymerization. The effect of important operating parameters such as initial polymer concentration, presence of functional groups in the polymer chain, reaction volume, initial molecular weight, temperature, operating frequency, power dissipation and use of process intensifying additives have been discussed also giving guidelines about selection of the optimum parameters. It has been observed that the low concentrations and higher power dissipation (till an optimum) are favorable resulting in higher extents of degradation. Typically low frequency is recommended but for the case of water soluble polymers, higher frequencies would also give similar results due to the dominant action of chemical effects of cavitation. It has been demonstrated that the alkyl group substituent also affects the degradation rate of polymer. An overview of degradation using combined approach based on ultrasound and additives with comparison with individual approach has also been presented. It has been observed that the main contributing factor for the synergy of the combined approach is the selection of optimum loading of additives. Overall, it has been observed that efficient polymer degradation can be achieved using combined process based on the use of ultrasound. Copyright © 2015 Elsevier B.V. All rights reserved.

  12. Lessons learned from applying VIM to fast reactor critical experiments, summary

    SciTech Connect

    Schaefer, R.W.; McKnight, R.D.; Collins, P.J.

    1995-05-17

    VIM is a continuous energy Monte Carlo code first developed around 1970 for the analysis of plate-type, fast-neutron, zero-power critical assemblies. In most respects, VIM is functionally equivalent to the MCNP code but it has two features that make uniquely suited to the analysis of fast reactor critical experiments: (1) the place lattice geometry option, which allows efficient description of and neutron tracking in the assembly geometry, and (2) a statistical treatment of neutron cross section data in the unresolved resonance range. Since its inception, VIM`s capabilities have expanded to include numerous features, such as thermal neutron cross sections, photon cross sections, and combinatorial and other geometry options, that have allowed its use in a wide range of neutral-particle transport problems. The earliest validation work at Argonne National Laboratory (ANL) focused on the validation of VIM itself. This work showed that, in order for VIM to be a ``rigomus`` tool, extreme detail in the pointwise Monte Carlo libraries was needed, and the required detail was added. The emphasis soon shifted to validating models, methods, data and codes against VIM. Most of this work was done in the context of analyzing critical experiments in zero power reactor (ZPR) assemblies. The purpose of this paper is to present some of the lessons learned from using VIM in ZPR analysis work.

  13. 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.

  14. Summary of ORSphere critical and reactor physics measurements

    NASA Astrophysics Data System (ADS)

    Marshall, Margaret A.; Bess, John D.

    2017-09-01

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

  15. Summary of ORSphere Critical and Reactor Physics Measurements

    SciTech Connect

    Marshall, Margaret A.; Bess, John D.

    2016-09-01

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

  16. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 3, Calculated activity profiles of spent nuclear fuel assembly hardware for boiling water reactors

    SciTech Connect

    Short, S.M.; Luksic, A.T.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly that is also radioactive and required disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volume 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  17. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 2, Calculated activity profiles of spent nuclear fuel assembly hardware for pressurized water reactors

    SciTech Connect

    Short, S.M.; Luksic, A.T.; Lotz, T.L.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report present a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from Laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  18. 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…

  19. 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…

  20. 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…

  1. 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…

  2. 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.

  3. 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.

  4. 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.

  5. Use of research reactors in multidisciplinary education at Cornell University

    SciTech Connect

    Clark, D.D. )

    1992-01-01

    Multidisciplinary aspects of nuclear science and technology form a large part of the research and teaching activities of the Nuclear Science and Engineering (NS and E) Program at Cornell, and the two reactors housed in Ward Laboratory - a 500-kW TRIGA and a 100-W critical facility (zero-power reactor (ZPR))- play a central role in those activities. Several primarily educational and multidisciplinary features of the NS and E program are described in this paper.

  6. 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.

  7. Criticality safety benchmark experiments derived from ANL ZPR assemblies.

    SciTech Connect

    Schaefer, R. W.; Lell, R. M.; McKnight, R. D.

    2003-09-01

    Numerous criticality safety benchmarks have been, and continue to be, developed from experiments performed on Argonne National Laboratory's plate-type fast critical assemblies. The nature and scope of assemblies suitable for deriving these benchmarks are discussed. The benchmark derivation process, including full treatment of all significant uncertainties, is explained. Calculational results are presented that support the small uncertainty assigned to the key derivation step in which complex geometric detail is removed.

  8. Continuously Varying Critical Exponents Beyond Weak Universality

    PubMed Central

    Khan, N.; Sarkar, P.; Midya, A.; Mandal, P.; Mohanty, P. K.

    2017-01-01

    Renormalization group theory does not restrict the form of continuous variation of critical exponents which occurs in presence of a marginal operator. However, the continuous variation of critical exponents, observed in different contexts, usually follows a weak universality scenario where some of the exponents (e.g., β, γ, ν) vary keeping others (e.g., δ, η) fixed. Here we report ferromagnetic phase transition in (Sm1−yNdy)0.52Sr0.48MnO3 (0.5 ≤ y ≤ 1) single crystals where all three exponents β, γ, δ vary with Nd concentration y. Such a variation clearly violates both universality and weak universality hypothesis. We propose a new scaling theory that explains the present experimental results, reduces to the weak universality as a special case, and provides a generic route leading to continuous variation of critical exponents and multi-criticality. PMID:28327622

  9. Continuously Varying Critical Exponents Beyond Weak Universality

    NASA Astrophysics Data System (ADS)

    Khan, N.; Sarkar, P.; Midya, A.; Mandal, P.; Mohanty, P. K.

    2017-03-01

    Renormalization group theory does not restrict the form of continuous variation of critical exponents which occurs in presence of a marginal operator. However, the continuous variation of critical exponents, observed in different contexts, usually follows a weak universality scenario where some of the exponents (e.g., β, γ, ν) vary keeping others (e.g., δ, η) fixed. Here we report ferromagnetic phase transition in (Sm1‑yNdy)0.52Sr0.48MnO3 (0.5 ≤ y ≤ 1) single crystals where all three exponents β, γ, δ vary with Nd concentration y. Such a variation clearly violates both universality and weak universality hypothesis. We propose a new scaling theory that explains the present experimental results, reduces to the weak universality as a special case, and provides a generic route leading to continuous variation of critical exponents and multi-criticality.

  10. Analysis of experiments in the Phase III GCFR benchmark critical assembly

    SciTech Connect

    Hess, A.L.; Baylor, K.J.

    1980-04-01

    Experiments carried out in the third gas-cooled fast breeder reactor (GCFR) benchmark critical assembly on the Zero Power Reactor-9 at Argonne National Laboratory were analyzed using methods and computer codes employed routinely for design and performance evaluations on power-plant GCFR cores. The program for the Phase III GCFR assembly, with a 1900-liter, three-enrichment zone core, included measurements of reaction-rate profiles in a typical power-flattened design, studies of material reactivity coefficients, reaction ratio and breeding parameter determinations, and comparison of pin with plate fuel loadings. Calculated parameters to compare with all of the measured results were obtained using 10-group cross sections based on ENDF/B-4 and two-dimensional diffusion theory, with adjustments for fuel-cell heterogeneity and void-lattice streaming effects.

  11. 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.

  12. 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.

  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. Universal estimates for critical circle mappings.

    PubMed

    Khanin, K. M.

    1991-08-01

    A thermodynamic formalism is constructed for critical circle mappings. It is used to prove universal estimates for the asymptotic behavior of renormalized mappings. Certain applications of statistical mechanics to research on the ergodic properties of critical homeomorphisms of a circle are also discussed.

  15. 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

  16. 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.

  17. 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.

  18. The Assembly on University Goals and Governance. A First Report.

    ERIC Educational Resources Information Center

    American Academy of Arts and Sciences, Boston, MA.

    The Assembly on University Goals and Governance was established by the American Academy of Arts and Sciences in September 1969 to explore, develop, and help implement alternative approaches for resolving certain of the issues affecting colleges and universities today. This report presents the 85 theses that were developed by the Assembly. The…

  19. An ultracold neutron source at the NC State University PULSTAR reactor

    NASA Astrophysics Data System (ADS)

    Korobkina, E.; Wehring, B. W.; Hawari, A. I.; Young, A. R.; Huffman, P. R.; Golub, R.; Xu, Y.; Palmquist, G.

    2007-08-01

    Research and development is being completed for an ultracold neutron (UCN) source to be installed at the PULSTAR reactor on the campus of North Carolina State University (NCSU). The objective is to establish a university-based UCN facility with sufficient UCN intensity to allow world-class fundamental and applied research with UCN. To maximize the UCN yield, a solid ortho-D 2 converter will be implemented coupled to two moderators, D 2O at room temperature, to thermalize reactor neutrons, and solid CH 4, to moderate the thermal neutrons to cold-neutron energies. The source assembly will be located in a tank of D 2O in the space previously occupied by the thermal column of the PULSTAR reactor. Neutrons leaving a bare face of the reactor core enter the D 2O tank through a 45×45 cm cross-sectional area void between the reactor core and the D 2O tank. Liquid He will cool the disk-shaped UCN converter to below 5 K. Independently, He gas will cool the cup-shaped CH 4 cold-neutron moderator to an optimum temperature between 20 and 40 K. The UCN will be transported from the converter to experiments by a guide with an inside diameter of 16 cm. Research areas being considered for the PULSTAR UCN source include time-reversal violation in neutron beta decay, neutron lifetime determination, support measurements for a neutron electric-dipole-moment search, and nanoscience applications.

  20. 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.

  1. [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.

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

    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 critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations reported herein 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 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 three reactor critical configurations for the Sequoyah Unit 2 Cycle 3. This unit and cycle were chosen because of the relevance in spent fuel benchmark applications: (1) the unit had a significantly long downtime of 2.7 years during the middle of cycle (MOC) 3, and (2) the core consisted entirely of burned fuel at the MOC restart. The first benchmark critical calculation was the MOC restart at hot, full-power (HFP) critical conditions. The

  3. 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.

  4. Utilization of the Cornell University research reactors in support of the Nuclear Power Industry

    SciTech Connect

    Aderhold, A.C. )

    1993-01-01

    Cornell University is licensed to operate two research reactor facilities on its main campus in Ithaca, New York: a 500-kW pulsing TRIGA and a 100-W zero-power reactor (ZPR). The initial criticality of both reactors took place in 1962, and the utilization of each has been, and continues to be, dedicated to the teaching and research programs of Cornell's many academic departments. As the nation's nuclear power industry grew, the demand for services at research and test reactors increased. As a result, and in large part because of special design features of the TRIGA, Cornell responded to a few requests for reactor testing services while maintaining the policy that these services would not interfere with teaching and research programs. The frequency of service requests suddenly mushroomed in November of 1989, when the nation's major testing reactor was shut down for repairs. In spite of a small staff of two full-time reactor operators, a decision was made to support the nuclear industry to the fullest extent possible without jeopardizing Cornell's teaching and research programs. This turned into a monumental task of tight scheduling and meeting precise deadlines. It could only be accomplished by working late evenings and weekends and, on a number of occasions, staying at the facility for up to 5 days continuously.

  5. ANL Critical Assembly Covariance Matrix Generation - Addendum

    SciTech Connect

    McKnight, Richard D.; Grimm, Karl N.

    2014-01-13

    In March 2012, a report was issued on covariance matrices for Argonne National Laboratory (ANL) critical experiments. That report detailed the theory behind the calculation of covariance matrices and the methodology used to determine the matrices for a set of 33 ANL experimental set-ups. Since that time, three new experiments have been evaluated and approved. This report essentially updates the previous report by adding in these new experiments to the preceding covariance matrix structure.

  6. Characterization of a New Continuous Air Monitoring System For the University of Massachusetts Lowell Research Reactor

    NASA Astrophysics Data System (ADS)

    Alqahtani, Mohammad Saad

    A continuous air monitor (CAM) is a critical piece of equipment to support radiation safety in nuclear facilities where the generation of airborne radioactivity is a possibility for either normal operations or accident scenarios. The University of Massachusetts Lowell Research Reactor is planning to install a new CAM system manufactured by Canberra Industries for monitoring airborne radioactive particulates. In this study, the new CAM was evaluated to determine 1) baseline response, 2) response to high exposure rates, 3) appropriate background compensation, 4) detection limits, and 5) alarm settings. The results of this study will help to properly integrate the new CAM into the reactor radiation monitoring system.

  7. 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)

  8. Passive Gamma Analysis of the Boiling-Water-Reactor Assemblies

    SciTech Connect

    Vo, Duc Ta; Favalli, Andrea

    2016-03-31

    Passive gamma analysis can be used to determine BU and CT of BWR assembly. The analysis is somewhat more complicated and less effective than similar method for PWR assemblies. From the measurements along the lengths of the BWR1 and BWR9 assemblies, there are hints that we may be able to use their information to help improve the model functions for better results.

  9. 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

  10. 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

  11. Neutron collar calibration for assay of LWR (light-water reactor) fuel assemblies

    SciTech Connect

    Menlove, H.O.; Pieper, J.E.

    1987-03-01

    The neutron-coincidence collar is used for the verification of the uranium content in light-water reactor fuel assemblies. An AmLi neutron source is used to give an active interrogation of the fuel assembly to measure the /sup 235/U content, and the /sup 238/U content is verified from a passive neutron-coincidence measurement. This report gives the collar calibration data of pressurized-water reactor and boiling-water reactor fuel assemblies. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and different fuel assembly sizes. The data were collected at Exxon Nuclear, Franco-Belge de Fabrication de Combustibles, ASEA-Atom, and other nuclear fuel fabrication facilities.

  12. 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.

  13. 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

  14. 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…

  15. 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…

  16. 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.

  17. 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.

  18. 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).

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Neutron depth profiling at the University of Texas research reactor

    SciTech Connect

    Unlu, K.; Wehring, B.W. )

    1993-01-01

    A neutron depth profiling (NDP) facility has been developed at the University of Texas at Austin (UT) Nuclear Engineering Teaching Laboratory. The UT-NDP utilizes thermal neutrons from a tangential beam port of the 1-MW TRIGA Mark II research reactor. Aspects of the designs of the thermal neutron beam and target chamber for the UT-NDP facility are given in this paper. Also, a brief description of NDP and possible applications are included.

  5. Transient bowing of core assemblies in advanced liquid metal fast reactors

    SciTech Connect

    Kamal, S.A.; Orechwa, Y.

    1986-01-01

    Two alternative core restraint concepts are considered for a conceptual design of a 900 MWth liquid metal fast reactor core with a heterogeneous layout. The two concepts, known as limited free bowing and free flowering, are evaluated based on core bowing criteria that emphasize the enhancement of inherent reactor safety. The core reactivity change during a postulated loss of flow transient is calculated in terms of the lateral displacements and displacement-reactivity-worths of the individual assemblies. The NUBOW-3D computer code is utilized to determine the assembly deformations and interassembly forces that arise when the assemblies are subjected to temperature gradients and irradiation induced creep and swelling during the reactor operation. The assembly ducts are made of the ferritic steel HT-9 and remain in the reactor core for four-years at full power condition. Whereas both restraint systems meet the bowing criteria, a properly designed limited free bowing system appears to be more advantageous than a free flowering system from the point of view of enhancing the reactor inherent safety.

  6. 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.

  7. 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.

  8. Improvements to the Pool Critical Assembly Pressure Vessel Benchmark with 3-D Parallel SN PENTRAN

    NASA Astrophysics Data System (ADS)

    Edgar, Christopher A.; Sjoden, Glenn E.; Yi, Ce

    2014-06-01

    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. Improvements to the benchmark are proposed here through the application of more representative flux and volume weighted homogenized cross sections for the PCA reactor core, which were obtained from a rigorous heterogeneous modeling of all fuel assembly types in the core. A new source term definition is also proposed based on calculated relative power in each core fuel assembly with a spectrum based on the Uranium-235 fission spectra. This research focused on utilizing the BUGLE-96 cross section library and accompanying reaction rates, while also examining PENTRAN's adaptive differencing implemented on a coarse mesh basis, as well as fixed use of Directional Theta-Weighted (DTW) SN differencing scheme in order to compare the calculated PENTRAN results to measured data. The results show good comparison with the measured benchmark data, which suggests PENTRAN is a viable, reliable code system for calculation of light water reactor neutron shielding and pressure vessel dosimetry calculations. Furthermore, the improvements to the benchmark methodology resulting from this work provide a 6 percent increase in accuracy of the calculation (based on the average of all calculation points), when compared with experimentally measured results at the same spatial locations in the PCA pressure vessel simulator.

  9. 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.

  10. 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

  11. Neutronic performance of several LEU fuel assembly designs for the WWR-SM research reactor in Uzbekistan.

    SciTech Connect

    Bretscher, M. M.; Hanan, N. A.; Matos, J. E.; Yuldashev, B. S.; Baytelesov, S.; Rakhmanov, A.; Technology Development; Inst. of Nuclear Physics

    2002-01-01

    The 10 MW WWR-SM research reactor in Uzbekistan currently uses HEU (36%) IRT-3M 6-tube fuel assemblies manufactured by the Novosibirsk Chemical Concentrates Plant in Russia. Recent 4x4 core configurations reflected by beryllium have been operated at 8 MW. The Institute of Nuclear Physics plans to convert the reactor to LEU (19.7%) fuel as soon as a suitable LEU fuel assembly is qualified. This study compares the neutronic performance of the reactor and its experiments using LEU pin-type and LEU tube-type fuel assembly designs with the current HEU (36%) reference fuel assembly. Both 3D Monte Carlo and 3D diffusion theory calculations were performed to analyze a critical core configuration with partially-burned HEU fuel assemblies in order to establish the credibility of the analytical methods and computer models used to describe the reactor and its experiments. Results based on these techniques are in reasonable agreement with the measured data. An LEU pin-type design (164 pins, 4.5 g U/cm{sup 3}, 375g {sup 235}U) or an LEU tube-type design (IRT-3M, 6-tube, 5.4 gU/cm{sup 3}, and 364g {sup 235}U) with U9Mo-Al fuel meat could operate with about the same cycle length and experiment load as the reference HEU (36%) IRT-3M fuel. The annual fuel assembly consumption would be nearly the same in these HEU and LEU cores. For the LEU pin-type design, fast (thermal) fluxes would be reduced by 2.5% (14%) for experiments located at the center of the fuel assemblies and by 0.5% (4%) for experiments located in experiment channels in the beryllium reflector. For the LEU tube-type design, fast (thermal) fluxes would be reduced by 3.5% (15%) for experiments located at the center of the fuel assemblies and by 1.2% (5%) for experiments located in experiment channels in the beryllium reflector. If the {sup 235}U content of the LEU pin-type fuel assemblies were increased to 480g (using pins similar to those planned to be tested in the WWR-M reactor at Gatchina, Russia in 2003 and 2004

  12. Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

    SciTech Connect

    Sienicki, James J.; Grandy, Christopher

    2016-12-15

    A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

  13. 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.

  14. 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.

  15. Description and performance characteristics for the neutron Coincidence Collar for the verification of reactor fuel assemblies

    SciTech Connect

    Menlove, H.O.

    1981-08-01

    An active neutron interrogation method has been developed for the measurement of /sup 235/U content in fresh fuel assemblies. The neutron Coincidence Collar uses neutron interrogation with an AmLi neutron source and coincidence counting the induced fission reaction neutrons from the /sup 235/U. This manual describes the system components, operation, and performance characteristics. Applications of the Coincidence Collar to PWR and BWR types of reactor fuel assemblies are described.

  16. 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.

  17. 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.

  18. Analysis of assembly serial number usage in domestic light-water reactors

    SciTech Connect

    Reich, W.J. ); Moore, R.S. )

    1991-05-01

    Domestic light-water reactor (LWR) fuel assemblies are identified by a serial number that is placed on each assembly. These serial numbers are used as identifiers throughout the life of the fuel. The uniqueness of assembly serial numbers is important in determining their effectiveness as unambiguous identifiers. The purpose of this study is to determine what serial numbering schemes are used, the effectiveness of these schemes, and to quantify how many duplicate serial numbers occur on domestic LWR fuel assemblies. The serial numbering scheme adopted by the American National Standards Institute (ANSI) ensures uniqueness of assembly serial numbers. The latest numbering scheme adopted by General Electric (GE), was also found to be unique. Analysis of 70,971 fuel assembly serial numbers from permanently discharged fuel identified 11,948 serial number duplicates. Three duplicate serial numbers were found when analysis focused on duplication within the individual fuel inventory at each reactor site, but these were traced back to data entry errors and will be corrected by the Energy Information Administration (EIA). There were also three instances where the serial numbers used to identify assemblies used for hot cell studies differed from the serial numbers reported to the EIA. It is recommended that fuel fabricators and utilities adhere to the ANSI serial numbering scheme to ensure serial number uniqueness. In addition, organizations collecting serial number information, should request that all known serial numbers physically attached or associated with each assembly be reported and identified by the corresponding number scheme. 10 refs., 5 tabs.

  19. 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.

  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

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION University of Utah, University of Utah TRIGA Nuclear Reactor, Notice of Issuance of Renewed... University of Utah (UU, the licensee), which authorizes continued operation of the UU TRIGA Nuclear Reactor...

  1. Geometry and universality in self-assembly

    NASA Astrophysics Data System (ADS)

    Bowick, Mark

    2007-03-01

    I will discuss the use of ideas from geometry and topology in combination with the statistical mechanics of self-assembly to classify the possible types of mesoatoms that can constitute a library of raw materials for making mesomolecules and bulk materials. I will concentrate on mesoatoms made up of an ordered two-dimensional monolayer of particles on the surface of a liquid droplet. Both the shape and the topology of the two-dimensional surface can be varied as well as the architecture of the ordering particles. The topology of the surface and the symmetries of the two-dimensional order severely restrict the possible defect structure of the mesoatom, which in turn fixes its valency. Defective regions are natural places for biological activity, chemical linking, unusual elastic response and aggregation of disorder. Specific examples include crystalline and hexatic order of point particles on the sphere, paraboloid, torus and Gaussian bump, nematic order of nematogens on the sphere and torus, and vector order of polar units on the sphere. The Gaussian curvature of the underlying surface may also lead to new features in the ground state, such as extended defect arrays of various kinds and curvature-driven defect unbinding, all of which may be exploited via engineered or spontaneous self-assembly.

  2. 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

  3. 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…

  4. 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…

  5. 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.

  6. 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.

  7. 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

  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. Prediction of dryout performance for boiling water reactor fuel assemblies based on subchannel analysis with the RINGS code

    SciTech Connect

    Knabe, P.; Wehle, F.

    1995-12-01

    A fuel assembly with a large critical power margin introduces flexibility into reload fuel management. Therefore, optimization of the bundle and spacer geometry to maximize the bundle critical power is an important design objective. With a view to reducing the extent of the complex full-scale tests usually carried out to determine the thermal-hydraulic characteristics of various assembly geometries, the subchannel analysis method was further developed with the Siemens RINGS code. The annular flow code predicts dryout power and dryout location by calculating the conditions at which the liquid film flow rate is reduced to zero, allowing for evaporation, droplet entrainment, and droplet deposition. Appropriate attention is paid to the modeling of spacer effects. Comparison with experimental data of 3 x 3 and 4 x 4 tests shows the capability of RINGS to predict the flow quality and mass flux in subchannels under typical boiling water reactor operating conditions. By using the RINGS code, experimental critical power data for 3 x 3, 4 x 4, 5 x 5, 7 x 7, 8 x 8, 9 x 9, and 10 x 10 fuel assemblies were successfully postcalculated.

  10. Simplified failure sequence evaluation of reactor pressure vessel head corroding in-core instrumentation assembly

    SciTech Connect

    McVicker, J.P.; Conner, J.T.; Hasrouni, P.N.; Reizman, A.

    1995-11-01

    In-Core Instrumentation (ICI) assemblies located on a Reactor Pressure Vessel Head have a history of boric acid leakage. The acid tends to corrode the nuts and studs which fasten the flanges of the assembly, thereby compromising the assembly`s structural integrity. This paper provides a simplified practical approach in determining the likelihood of an undetected progressing assembly stud deterioration, which would lead to a catastrophic loss of reactor coolant. The structural behavior of the In-Core Instrumentation flanged assembly is modeled using an elastic composite section assumption, with the studs transmitting tension and the pressure sealing gasket experiencing compression. Using the above technique, one can calculate the flange relative deflection and the consequential coolant loss flow rate, as well as the stress in any stud. A solved real life example develops the expected failure sequence and discusses the exigency of leak detection for safe shutdown. In the particular case of Calvert Cliffs Nuclear Power Plant (CCNPP) it is concluded that leak detection occurs before catastrophic failure of the ICI flange assembly.

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

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-16

    .... 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 Laboratory...-ply, built-up surface roof. The Mechanical Engineering Building also contains classrooms, laboratories...

  14. 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.

  15. 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

  16. 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.

  17. Investigation of radiation fields outside the Sub-critical Assembly in Dubna.

    PubMed

    Seltbor, P; Lopatkin, A; Gudowski, W; Shvetsov, V; Polanski, A

    2005-01-01

    The radiation fields outside the planned experimental Sub-critical Assembly in Dubna (SAD) have been studied in order to provide a basis for the design of the concrete shielding that cover the reactor core. The effective doses around the reactor, induced by leakage of neutrons and photons through the shielding, have been determined for a shielding thickness varying from 100 to 200 cm. It was shown that the neutron flux and the effective dose is higher above the shielding than at the side of it, owing to the higher fraction of high-energy spallation neutrons emitted in the direction of the incident beam protons. At the top, the effective dose was found to be -150 microSv s(-1) for a concrete thickness of 100 cm, while -2.5 microSv s(-1) for a concrete thickness of 200 cm. It was also shown that the high-energy neutrons (> 10 MeV), which are created in the proton-induced spallation interactions in the target, contribute for the major part of the effective doses outside the reactor.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION License Amendment Request From The State University of New York, University of Buffalo Reactor Facility AGENCY: Nuclear Regulatory Commission. ACTION: Notice of decommissioning plan, proposed...

  6. URI Program Final Report FY 2001 Grant for the University of Florida Training Reactor

    SciTech Connect

    William G. Vernetson

    2004-03-15

    The purpose of the URI program is to upgrade and improve university nuclear research and training reactors and to contribute to strengthening the academic community's nuclear engineering infrastructure.

  7. 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).

  8. 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.

  9. Evaluation of Neutron Response of Criticality Accident Alarm System Detector to Quasi-Monoenergetic 24 keV Neutrons

    NASA Astrophysics Data System (ADS)

    Tsujimura, Norio; Yoshida, Tadayoshi; Yashima, Hiroshi

    The criticality accident alarm system (CAAS), which was recently developed and installed at the Japan Atomic Energy Agency's Tokai Reprocessing Plant, consists of a plastic scintillator combined with a cadmium-lined polyethylene moderator and thereby responds to both neutrons and gamma rays. To evaluate the neutron absorbed dose rate response of the CAAS detector, a 24 keV quasi-monoenergetic neutron irradiation experiment was performed at the B-1 facility of the Kyoto University Research Reactor. The detector's evaluated neutron response was confirmed to agree reasonably well with prior computer-predicted responses.

  10. 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.

  11. 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

  12. 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.

  13. Comparing the new generation accelerator driven subcritical reactor system (ADS) to traditional critical reactors

    NASA Astrophysics Data System (ADS)

    Kemah, Elif; Akkaya, Recep; Tokgöz, Seyit Rıza

    2017-02-01

    In recent years, the accelerator driven subcritical reactors have taken great interest worldwide. The Accelerator Driven System (ADS) has been used to produce neutron in subcritical state by the external proton beam source. These reactors, which are hybrid systems, are important in production of clean and safe energy and conversion of radioactive waste. The ADS with the selection of reliability and robust target materials have been the new generation of fission reactors. In addition, in the ADS Reactors the problems of long-lived radioactive fission products and waste actinides encountered in the fission process of the reactor during incineration can be solved, and ADS has come to the forefront of thorium as fuel for the reactors.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  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. Advanced fuel assembly characterization capabilities based on gamma tomography at the Halden boiling water reactor

    SciTech Connect

    Holcombe, S.; Eitrheim, K.; Svaerd, S. J.; Hallstadius, L.; Willman, C.

    2012-07-01

    Characterization of individual fuel rods using gamma spectroscopy is a standard part of the Post Irradiation Examinations performed on experimental fuel at the Halden Boiling Water Reactor. However, due to handling and radiological safety concerns, these measurements are presently carried out only at the end of life of the fuel, and not earlier than several days or weeks after its removal from the reactor core. In order to enhance the fuel characterization capabilities at the Halden facilities, a gamma tomography measurement system is now being constructed, capable of characterizing fuel assemblies on a rod-by-rod basis in a more timely and efficient manner. Gamma tomography for measuring nuclear fuel is based on gamma spectroscopy measurements and tomographic reconstruction techniques. The technique, previously demonstrated on irradiated commercial fuel assemblies, is capable of determining rod-by-rod information without the need to dismantle the fuel. The new gamma tomography system will be stationed close to the Halden reactor in order to limit the need for fuel transport, and it will significantly reduce the time required to perform fuel characterization measurements. Furthermore, it will allow rod-by-rod fuel characterization to occur between irradiation cycles, thus allowing for measurement of experimental fuel repeatedly during its irradiation lifetime. The development of the gamma tomography measurement system is a joint project between the Inst. for Energy Technology - OECD Halden Reactor Project, Westinghouse (Sweden), and Uppsala Univ.. (authors)

  20. Radionuclide characterization of reactor decommissioning waste and spent fuel assembly hardware

    SciTech Connect

    Robertson, D.E.; Thomas, C.W.; Wynhoff, N.L.; Hetzer, D.C. )

    1991-01-01

    This study is providing the NRC and licensees with a more comprehensive and defensible data base and regulatory assessment of the radiological factors associated with reactor decommissioning and disposal of wastes generated during these activities. The objectives of this study are being accomplished during a two-phase sampling, measurement, and assessment program involving the actual decommissioning of Shippingport Station and the detailed analysis of neutron-activated materials from commercial reactors. Radiological characterization studies at Shippingport have shown that neutron activation products, dominated by {sup 60}Co, comprised the residual radionuclide inventory. Fission products and transuranic radionuclides were essentially absent. Waste classification assessments have shown that all decommissioning materials (except reactor pressure vessel internals) could be disposed of as Class A waste. Measurements and assessments of spent fuel assembly hardware have shown that {sup 63}Ni, {sup 59}Ni, and {sup 94}Nb sometimes greatly exceed the 10CFR61 Class C limit for some components, and thus would require disposal in a high level waste repository. These measurements are providing the basis for an assessment of the disposal options for these types of highly radioactive materials. Comparisons of predicted (calculated) activation product concentrations with the empirical data are providing as assessment of the accuracy of calculational methods. Work is continuing on radiological characterization of spent PWR and BWR control rod assemblies. Additional work is planned on current issues/problems relating to reactor decommissioning. These efforts will be reported on in future supplements to this report. 20 refs., 23 figs., 34 tabs.

  1. 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)

  2. 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.

  3. The Conceptual Design for a Fuel Assembly of a New Research Reactor

    SciTech Connect

    Ryu, J-S.; Cho, Y-G.; Yoon, D-B.; Dan, H-J.; Chae, H-T.; Park, C.

    2004-10-06

    A new Research Reactor (ARR) has been under design by KAERI since 2002. In this work, as a first step for the design of the fuel assembly of the ARR, the conceptual design has been carried out. The vibration characteristics of the tubular fuel model and the locking performance of the preliminary designed locking devices were investigated. In order to investigate the effects of the stiffener on the vibration characteristics of the tubular fuel, a modal analysis was performed for the finite element models of the tubular fuels with stiffeners and without stiffeners. The analysis results show that the vibration characteristics of the tubular fuel with stiffeners are better than those of the tubular fuel without stiffeners. To investigate the locking performance of the preliminary designed locking devices for the fuel assembly of the ARR, the elements of the locking devices were fabricated. Then the torsional resistance, fixing status and vibration characteristics of the locking devices were tested. The test results show that using the locking device with fins on the bottom guide can prevent the torsional motion of the fuel assembly, and that additional springs or guides on the top of the fuel assembly are needed to suppress the lateral motion of the fuel assembly. Based on the modal analysis and experimental results, the fuel assembly and locking devices of the ARR were designed and its prototype was fabricated. The locking performance, pressure drop characteristics and vibration characteristics of the newly designed fuel assembly will be tested in the near future.

  4. 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.

  5. 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.

  6. 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

  7. 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.

  8. 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.

  9. 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.

  10. Assembly-level calculations for transuranics recycling in pressurized water reactors

    SciTech Connect

    Lee, J.C.; Du, J. )

    1993-01-01

    In this paper, we present fuel depletion calculations evaluating the feasibility and efficiency of transmuting transuranics (TRUs) from spent nuclear fuel in pressurized water reactors (PWRs). In contrast to previous studies, which focused on plutonium recycling, we consider recycling all TRUS, including neptunium, americium, and curium isotopes. Single-assembly and (2 x 2) color-set calculations have been performed with the CASMO assembly-level collision probability code to model equilibrium PWR configurations. There has been renewed interest in the separation and transmutation of TRUs for reducing the long-term radioactivity in spent-fuel repositories. With low capture-to-fission ratios for TRUs in a fast spectrum, liquid-metal reactors (LMRS) are often considered favorable for transmutation of TRUS. The TRU fission cross sections are, however, larger in a thermal spectrum, which offers the potential for thermal transmutors to operate with a smaller TRU inventory, resulting in a higher fractional depletion rate and a higher TRU inventory reduction (TIR) factor. The TIR factor is defined as the ratio of the TRU inventory, without the TRU transmutor, to the inventory that would accumulate with TRU reprocessing and recycling. We present a preliminary study on how standard PWR designs could realize these potential advantages of thermal spectrum reactors.

  11. STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

    SciTech Connect

    Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart

    2016-05-01

    With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

  12. 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

  13. 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.

  14. 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.

  15. Critical Incidents of Student Satisfaction at German Universities

    ERIC Educational Resources Information Center

    Vianden, Jörg; Yakaboski, Tamara

    Purpose: The purpose of this paper is to advance a classification of satisfactory and unsatisfactory critical incidents of student-university relationships at German universities. Design/methodology/approach: Using the Critical Incident Technique (CIT), this paper reports the results of an exploratory study of 15 tertiary education students at…

  16. 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.

  17. 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.

  18. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-08-01

    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.

  20. 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…

  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. 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.

  3. 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.

  4. 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).

  5. 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.

  6. 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…

  7. Evaluation of the Pool Critical Assembly Benchmark with Explicitly-Modeled Geometry using MCNP6

    DOE PAGES

    Kulesza, Joel A.; Martz, Roger Lee

    2017-03-01

    Despite being one of the most widely used benchmarks for qualifying light water reactor (LWR) radiation transport methods and data, no benchmark calculation of the Oak Ridge National Laboratory (ORNL) Pool Critical Assembly (PCA) pressure vessel wall benchmark facility (PVWBF) using MCNP6 with explicitly modeled core geometry exists. As such, this paper provides results for such an analysis. First, a criticality calculation is used to construct the fixed source term. Next, ADVANTG-generated variance reduction parameters are used within the final MCNP6 fixed source calculations. These calculations provide unadjusted dosimetry results using three sets of dosimetry reaction cross sections of varyingmore » ages (those packaged with MCNP6, from the IRDF-2002 multi-group library, and from the ACE-formatted IRDFF v1.05 library). These results are then compared to two different sets of measured reaction rates. The comparison agrees in an overall sense within 2% and on a specific reaction- and dosimetry location-basis within 5%. Except for the neptunium dosimetry, the individual foil raw calculation-to-experiment comparisons usually agree within 10% but is typically greater than unity. Finally, in the course of developing these calculations, geometry that has previously not been completely specified is provided herein for the convenience of future analysts.« less

  8. 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…

  9. Enhanced protein digestion through the confinement of nanozeolite-assembled microchip reactors.

    PubMed

    Ji, Ji; Zhang, Yahong; Zhou, Xiaoqin; Kong, Jilie; Tang, Yi; Liu, Baohong

    2008-04-01

    An on-chip microreactor was proposed toward the acceleration of protein digestion through the construction of a nanozeolite-assembled network. The nanozeolite microstructure was assembled using a layer-by-layer technique based on poly(diallyldimethylammonium chloride) and zeolite nanocrystals. The adsorption of trypsin in the nanozeolite network was theoretically studied based on the Langmuir adsorption isotherm model. It was found that the controlled trypsin-containing nanozeolite networks assembled within a microchannel could act as a stationary phase with a large surface-to-volume ratio for the highly efficient proteolysis of both proteins at low levels and with complex extracts. The maximum proteolytic rate of the adsorbed trypsin was measured to be 350 mM min-1 microg-1, much faster than that in solution. Moreover, due the large surface-to-volume ratio and biocompatible microenvironment provided by the nanozeolite-assembled films as well as the microfluidic confinement effect, the low-level proteins down to 16 fmol per analysis were confidently identified using the as-prepared microreactor within a very short residence time coupled to matrix-assisted laser desorption-time-of-flight mass spectrometry. The on-chip approach was further demonstrated in the identification of the complex extracts from mouse macrophages integrated with two-dimensional liquid chromatography-electrospray ionization-tandem mass spectrometry. This microchip reactor is promising for the development of a facile means for protein identification.

  10. Radioisotope production at the University of Missouri Research Reactor: Past and present

    SciTech Connect

    Ehrhardt, G.J.; Ketring, A.R.; Gunn, S.L.

    1993-12-31

    Isotope production for industrial, medical, and specialty research use has been a major effort at the University of Missouri Research Reactor (MURR) since its initial critically in 1966, due primarily to the MURR`s high thermal neutron flux and ongoing commitment to reliability for its customers and research users. The history of this effort will be described, from the early supply of (n,gamma) Mo-99 for Tc-99m generators to the current program of production of high specific activity isotopes such as Re-186, Sm-153, Ho-166, Lu-177, and Re-188 which support the renaissance now taking place in nuclear medicine therapy. Using {open_quotes}guidance{close_quotes} methods as varied as seed implantation, microsphere entrapment, and chemical or immunochemical uptake, this resurgence of internal radioisotope therapy techniques depends, particularly in its more elegant forms, on just such high specific activity, beta-emitting isotopes.

  11. 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)

  12. 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)

  13. 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.

  14. Gold nanoparticle assembly microfluidic reactor for efficient on-line proteolysis.

    PubMed

    Liu, Yun; Xue, Yan; Ji, Ji; Chen, Xian; Kong, Jilie; Yang, Pengyuan; Girault, Hubert H; Liu, Baohong

    2007-08-01

    A microchip reactor coated with a gold nanoparticle network entrapping trypsin was designed for the efficient on-line proteolysis of low level proteins and complex extracts originating from mouse macrophages. The nanostructured surface coating was assembled via a layer-by-layer electrostatic binding of poly(diallyldimethylammonium chloride) and gold nanoparticles. The assembly process was monitored by UV-visible spectroscopy, atomic force microscopy, and quartz crystal microbalance. The controlled adsorption of trypsin was theoretically studied on the basis of the Langmuir isotherm model, and the fitted Gamma(max) and K values were estimated to be 1.2 x 10(-7) mol/m(2) and 4.1 x 10(5) M(-1), respectively. An enzymatic kinetics assay confirmed that trypsin, which was entrapped in the biocompatible gold nanoparticle network with a high loading capacity, preserved its bioactivity. The maximum proteolytic rate of the adsorbed trypsin was 400 mM/(min.microg). Trace amounts of proteins down to femtomole per analysis were digested using the microchip reactor, and the resulting tryptic products were identified by MALDI-TOF MS/MS. The protein mixtures extracted from the mouse macrophages were efficiently identified by on-line digestion and LC-ESI-MS/MS analysis.

  15. Critical Schwinger pair production. II. Universality in the deeply critical regime

    NASA Astrophysics Data System (ADS)

    Gies, Holger; Torgrimsson, Greger

    2017-01-01

    We study electron-positron pair production by spatially inhomogeneous electric fields. Depending on the localization of the field, a critical point (critical surface) exists in the space of field configurations where the pair production probability vanishes. Near criticality, pair production exhibits universal properties similar to those of continuous phase transitions. We extend results previously obtained in the semiclassical (weak-field) critical regime to the deeply critical regime for arbitrary peak field strength. In this regime, we find an enhanced universality, featuring a unique critical exponent β =3 for all sufficiently localized fields. For a large class of field profiles, we also compute the nonuniversal amplitudes.

  16. 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

  17. Analysis and evaluation of ZPPR (Zero Power Physics Reactor) critical experiments for a 100 kilowatt-electric space reactor

    SciTech Connect

    McFarlane, H.F.; Collins, P.J.; Carpenter, S.G.; Olsen, D.N.; Smith, D.M.; Schaefer, R.W. ); Doncals, R.A.; Andre, S.V.; Porter, C.A. ); Cowan, C.L; Stewart, S.L.; Protsik, R. . Astro Space Div.)

    1990-01-01

    ZPPR critical experiments were used for physics testing the reactor design of the SP-100, a 100-kW thermoelectric LMR that is being developed to provide electrical power for space applications. These tests validated all key physics characteristics of the design, including the ultimate safety in the event of a launch or re-entry accident. Both the experiments and the analysis required the use of techniques not previously applied to fast reactor designs. A few significant discrepancies between the experimental and calculated results leave opportunities for further optimization. An initial investigation has been made into application of the ZPPR-20 results, along with those of other relevant integral data, to the SP-100 design. 13 refs., 5 figs., 7 tabs.

  18. 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.

  19. 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)

  20. 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).

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

    SciTech Connect

    Chodak, III, Paul

    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 UO2 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 239Pu and ≥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.

  2. 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.

  3. 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.

  4. Can specific transcriptional regulators assemble a universal cancer signature?

    NASA Astrophysics Data System (ADS)

    Roy, Janine; Isik, Zerrin; Pilarsky, Christian; Schroeder, Michael

    2013-10-01

    Recently, there is a lot of interest in using biomarker signatures derived from gene expression data to predict cancer progression. We assembled signatures of 25 published datasets covering 13 types of cancers. How do these signatures compare with each other? On one hand signatures answering the same biological question should overlap, whereas signatures predicting different cancer types should differ. On the other hand, there could also be a Universal Cancer Signature that is predictive independently of the cancer type. Initially, we generate signatures for all datasets using classical approaches such as t-test and fold change and then, we explore signatures resulting from a network-based method, that applies the random surfer model of Google's PageRank algorithm. We show that the signatures as published by the authors and the signatures generated with classical methods do not overlap - not even for the same cancer type - whereas the network-based signatures strongly overlap. Selecting 10 out of 37 universal cancer genes gives the optimal prediction for all cancers thus taking a first step towards a Universal Cancer Signature. We furthermore analyze and discuss the involved genes in terms of the Hallmarks of cancer and in particular single out SP1, JUN/FOS and NFKB1 and examine their specific role in cancer progression.

  5. Research reactor usage at the Idaho National Engineering Laboratory in support of university research and education

    SciTech Connect

    Woodall, D.M.; Dolan, T.J.; Stephens, A.G. )

    1990-01-01

    The Idaho National Engineering Laboratory is a US Department of Energy laboratory which has a substantial history of research and development in nuclear reactor technologies. There are a number of available nuclear reactor facilities which have been incorporated into the research and training needs of university nuclear engineering programs. This paper addresses the utilization of the Advanced Reactivity Measurement Facility (ARMF) and the Coupled Fast Reactivity Measurement Facility (CFRMF) for thesis and dissertation research in the PhD program in Nuclear Science and Engineering by the University of Idaho and Idaho State University. Other reactors at the INEL are also being used by various members of the academic community for thesis and dissertation research, as well as for research to advance the state of knowledge in innovative nuclear technologies, with the EBR-II facility playing an essential role in liquid metal breeder reactor research. 3 refs.

  6. 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)

  7. 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)

  8. Universal short-time quantum critical dynamics in imaginary time

    NASA Astrophysics Data System (ADS)

    Yin, Shuai; Mai, Peizhi; Zhong, Fan

    2014-04-01

    We propose a scaling theory for the universal imaginary-time quantum critical dynamics for both short and long times. We discover that there exists a universal critical initial slip related to a small initial order parameter M0. In this stage, the order parameter M increases with the imaginary time τ as M ∝M0τθ with a universal initial-slip exponent θ. For the one-dimensional transverse-field Ising model, we estimate θ to be 0.373, which is markedly distinct from its classical counterpart. Apart from the local order parameter, we also show that the entanglement entropy exhibits universal behavior in the short-time region. As the critical exponents in the early stage and in equilibrium are identical, we apply the short-time dynamics method to determine quantum critical properties. The method is generally applicable in both the Landau-Ginzburg-Wilson paradigm and topological phase transitions.

  9. 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.

  10. [Analyzing the attributes of surgeons and working environment required for a successful career path and work-life balance: results of a survey administered to doctors working at Kyoto University Hospital].

    PubMed

    Okoshi, Kae; Tanabe, Tomoko; Hisamoto, Norio; Sakai, Yoshiharu

    2012-05-01

    We conducted a survey in March 2010 of all physicians at Kyoto University Hospital on working environments, levels of satisfaction, and level of exhaustion. A comparison of surgeons with other physicians showed tendencies among surgeons toward longer working hours and lower income. The findings indicated that surgeons experienced satisfaction from teamwork with fellow physicians, opportunities to manage interesting cases, and patient gratitude. Surgeons tended to have low fatigue level and were satisfied with their working environments, despite their low wages and long working hours. Although surgical treatment is currently built upon the feelings of accomplishment and satisfaction of individual surgeons, there is always a limit to his/her psychological strength. Indeed, the number of young surgeons is not increasing. In the future, efforts must be taken to prevent the departure of currently practicing surgeons. Consideration must also be given to reducing nonsurgical duties by increasing the numbers of medical staff, and making work conditions more appealing to young surgeons by guaranteeing income and prohibiting long working hours, particularly consecutive working hours.

  11. 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.

  12. 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.

  13. 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)

  14. Universal Postquench Prethermalization at a Quantum Critical Point.

    PubMed

    Gagel, Pia; Orth, Peter P; Schmalian, Jörg

    2014-11-28

    We consider an open system near a quantum critical point that is suddenly moved towards the critical point. The bath-dominated diffusive nonequilibrium dynamics after the quench is shown to follow scaling behavior, governed by a critical exponent that emerges in addition to the known equilibrium critical exponents. We determine this exponent and show that it describes universal prethermalized coarsening dynamics of the order parameter in an intermediate time regime. Implications of this quantum critical prethermalization are: (i) a power law rise of order and correlations after an initial collapse of the equilibrium state and (ii) a crossover to thermalization that occurs arbitrarily late for sufficiently shallow quenches.

  15. 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.

  16. 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…

  17. 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…

  18. Connecting University Supervision and Critical Reflection: Mentoring and Modeling

    ERIC Educational Resources Information Center

    Bates, Alisa J.; Ramirez, Laurie; Drits, Dina

    2009-01-01

    This study reports on the experiences of supervisors in a university teacher preparation program regarding their critical reflection on their practice. This has an impact on the learning opportunities available for their student teachers. Findings include: (a) an understanding of critical reflection is something that builds over time for student…

  19. 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…

  20. 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…

  1. 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.

  2. Off-line life tests of Topaz-2 system reactor unit assembly units

    NASA Astrophysics Data System (ADS)

    Oglobin, Boris G.; Proshin, Yuriy F.; Shalaev, Anatoliy I.

    1995-01-01

    Experimental trials of the Topaz-2 nuclear power system (NPS) have been carried out according to the Integrated Experimental Trials Program (IETP). This program involved ground off-line tests of the reactor unit assembly units and their integrated tests as components of the prototype systems. Scopes of tests performed and major results of integrated and off-line tests were presented in the paper by Vladimir P. Nikitin et al. (1993). This paper considers major results of off-line life tests of the cesium unit, automatic control drive, ionization chamber suspension and dry friction pairs. These tests have been carried out in the Central Design Bureau for Machine Building (CDBMB) between 1982 and April, 1994.

  3. 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.

  4. Novel zero-valent iron-assembled reactor for strengthening anammox performance under low temperature.

    PubMed

    Ren, Long-Fei; Lv, Lu; Zhang, Jian; Gao, Baoyu; Ni, Shou-Qing; Yang, Ning; Zhou, Qingxin; Liu, Xiaoyong

    2016-10-01

    To further expand the application of anammox biotechnology, a novel zero-valent iron-assembled upflow anaerobic sludge bed reactor was employed to strengthen anammox performance under low temperature and shock load. Packed with sponge iron and polyester sponge, this novel reactor could speed up the recovery of anammox activity in 12 days and improve the adaptability of anammox bacteria at the temperature of 10-15 °C. The high nitrogen loading rate of 1109.2 mg N/L/day could be adapted in 27 days and the new nitrogen pathway under the effect of sponge iron was clarified by batch experiment. Moreover, the real-time quantitative PCR analysis and Illumina MiSeq sequencing verified the dominant status of Candidatus Kuenenia stuttgartiensis and planctomycete KSU-1, as well as demonstrated the positive role of sponge iron on anammox microorganisms' proliferation. The findings might be beneficial to popularize anammox-related processes in municipal and industrial wastewater engineering.

  5. CSRL-V ENDF/B-V library and thermal reactor and criticality safety benchmarks

    SciTech Connect

    Ford, W.E. III; Diggs, B.R.; Knight, J.R.; Greene, N.M.; Petrie, L.M.; Williams, M.L.

    1982-01-01

    CSRL-V, an ENDF/B-V 227-group neutron cross-section library which has recently been expanded to include Bondarenko factor data for unresolved resonance processing, was used to calculate performance parameters for a series of thermal reactor and criticality safety benchmarks. Among the thermal benchmarks calculated were the Babcock and Wilcox lattice critical experiments B and W-XIII and B and W-XX. These two slightly-enriched (2.46%) UO/sub 2/, water-moderated, tight-pitch lattice experiments were chosen because (a) they have similar U/sup 238/ resonance shielding characteristics as power reactor cores, and (b) they provide benchmark results representative of high-leakage and low-leakage lattices, respectively. Among the criticality safety benchmarks calculated were homogeneous, highly-enriched (93.2%) uranyl fluoride spheres with hydrogen-to-uranium ratios varying from 76 to 972.

  6. 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. Copyright © 2015 Elsevier Ltd. All rights reserved.

  7. 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

  8. Applicability of ZPR critical experiment data to criticality safety

    SciTech Connect

    Schaefer, R.W.; Aumeier, S.E.; McFarlane, H.F.

    1995-12-31

    More than a hundred zero power reactor (ZPR) critical assemblies were constructed, over a period of about three decades, at the Argonne National Laboratory ZPR-3, ZPR-6, ZPR-9 and ZPPR fast critical assembly facilities. To be sure, the original reason for performing these critical experiments was to support fast reactor development. Nevertheless, data from some of the assemblies are well suited to form the basis for valuable, new criticality safety benchmarks. The purpose of this paper is to describe the ZPR data that would be of benefit to the criticality safety community and to explain how these data could be developed into practical criticality safety benchmarks.

  9. 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)

  10. 76 FR 11291 - University of New Mexico AGN-201M Reactor Notice of Issuance of Renewed Facility Operating...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-01

    ..., New Mexico. The UNMR is a solid homogeneous core research reactor licensed to operate at a steady... Nuclear Regulatory Commission. Jessie Quichocho, Chief, Research and Test Reactors Licensing Branch... COMMISSION University of New Mexico AGN-201M Reactor Notice of Issuance of Renewed Facility Operating...

  11. Spectroscopic critical dimension technology (SCD) for directed self assembly

    NASA Astrophysics Data System (ADS)

    Nishibe, Senichi; Dziura, Thaddeus; Nagaswami, Venkat; Gronheid, Roel

    2014-04-01

    Directed self-assembly (DSA) is being actively investigated as a potential patterning solution for future generation devices. While SEM based CD measurement is currently used in research and development, scatterometry-based techniques like spectroscopic CD (SCD) are preferred for high volume manufacturing. SCD can offer information about sub-surface features that are not available from CD-SEM measurement. Besides, SCD is a non-destructive, high throughput technique already adopted in HVM in several advanced nodes. The directed self assembly CD measurement can be challenging because of small dimensions and extremely thin layers in the DSA stack. In this study, the SCD technology was investigated for a 14 nm resolution PS-b-PMMA chemical epitaxy UW process optimized by imec. The DSA stack involves new materials such as cross-linkable polysterene (XPS) of thickness approximately 5 nm, ArF immersion resist (subsequently removed), -OH terminated neutral brush layer, and BCP material (Polystyrene-blockmethyl methacrylate of thickness roughly 20 to 30 nm). The mask contains a large CD and pitch matrix, for studying the quality of self-assembly as a function of the guide pattern dimensions. We report on the ability of SCD to characterize the dimensional variation in these targets and hence provide a viable process control solution.

  12. 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.

  13. 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)

  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. Analysis on Reactor Criticality Condition and Fuel Conversion Capability Based on Different Loaded Plutonium Composition in FBR Core

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Saputra, Geby; Suzuki, Mitsutoshi; Saito, Masaki

    2017-01-01

    Reactor criticality condition and fuel conversion capability are depending on the fuel arrangement schemes, reactor core geometry and fuel burnup process as well as the effect of different fuel cycle and fuel composition. Criticality condition of reactor core and breeding ratio capability have been investigated in this present study based on fast breeder reactor (FBR) type for different loaded fuel compositions of plutonium in the fuel core regions. Loaded fuel of Plutonium compositions are based on spent nuclear fuel (SNF) of light water reactor (LWR) for different fuel burnup process and cooling time conditions of the reactors. Obtained results show that different initial fuels of plutonium gives a significant chance in criticality conditions and fuel conversion capability. Loaded plutonium based on higher burnup process gives a reduction value of criticality condition or less excess reactivity. It also obtains more fuel breeding ratio capability or more breeding gain. Some loaded plutonium based on longer cooling time of LWR gives less excess reactivity and in the same time, it gives higher breeding ratio capability of the reactors. More composition of even mass plutonium isotopes gives more absorption neutron which affects to decresing criticality or less excess reactivity in the core. Similar condition that more absorption neutron by fertile material or even mass plutonium will produce more fissile material or odd mass plutonium isotopes to increase the breeding gain of the reactor.

  16. Results of Coupling a Thermal-Hydraulic Test Loop and University Research Reactor

    SciTech Connect

    Cetiner, Sacit M.; Edwards, Robert M.

    2002-07-01

    The coupling of a university thermal-hydraulic test loop and a simulated reactor is presented. The thermal-hydraulic test loop used in this work is a one-half height scaled version of General Electric's Simplified Boiling Water Reactor (SBWR). The digitally simulated reactor exploits modal neutron kinetics equations up to the first harmonic, and governing equations are not linearized. The preserved nonlinearity makes the simulated reactor behave more realistically, and eigenfunction expansion to the first order lets half of the core be represented independently. A series of experiments are performed with the hybrid system including simulated control rod reactivity insertion/withdrawal, cross-mode interaction, etc. The experimental results are compared with the theoretical expectations. (authors)

  17. 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.

  18. 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

  19. Fast Reactors

    NASA Astrophysics Data System (ADS)

    Esposito, S.; Pisanti, O.

    The following sections are included: * Elementary Considerations * The Integral Equation to the Neutron Distribution * The Critical Size for a Fast Reactor * Supercritical Reactors * Problems and Exercises

  20. Power reactor and critical experiment heterogeneity effects assessment for bias factors definition

    SciTech Connect

    Salvatores, M.; Soule, R.; Carta, M.

    1988-09-01

    Heterogeneity effects are compared in a power reactor subassembly of the Superphenix type and in the lattices of the critical experiments performed in the Masurca critical facility. Both the fuel in heterogeneity and the structure tube heterogeneity are evaluated with a two-step method based on the subgroup technique for self-shielding effect evaluation and on the Benoist method for streaming effect evaluation (the DHARMA method). Besides validation with reference calculations for simple geometries, experimental evidence confirms the good performance of the method proposed.

  1. The Primordial Black Hole Criticality of the Universe

    NASA Astrophysics Data System (ADS)

    Long, K. F.

    We consider the speculative suggestion that copious primordial black holes may have formed in over dense regions at early epochs with the final singularities tunnelling to big bang initial singularities. A critical parameter is introduced as a boundary between those primordial black holes that result in a baby universe contributing towards the nth generation and those that fail to produce such a progeny.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Strategies for bringing stem cell-derived dopamine neurons to the clinic: The Kyoto trial.

    PubMed

    Takahashi, Jun

    2017-01-01

    Concerted efforts are realizing cell-based therapy for Parkinson's disease (PD). In this chapter, I describe efforts at the Center for iPS Cell Research and Application (CiRA), Kyoto University. These efforts use induced pluripotent stem cells (iPSCs) as donor cells. The iPSCs were established as human leukocyte antigen homozygous at CiRA and are intended for allogeneic transplantation. Our manufacturing protocol includes a feeder-free cell culture with laminin fragment LM511-E8 and the sorting of CORIN(+) cells. Animal experiments, including those with monkey PD models, proved that the grafted cells survive and function as dopaminergic neurons in the brain without forming any tumors. Furthermore, I emphasize that not only the donor cells but also the host brain environment is critical for successful transplantation. To achieve optimization of the host environment, drug administration, gene modification, and rehabilitation are recommended. Based on these results, researchers plan to start a clinical trial at Kyoto University Hospital in the near future. © 2017 Elsevier B.V. All rights reserved.

  7. [Climate change and Kyoto protocol].

    PubMed

    Ergasti, G; Pippia, V; Murzilli, G; De Luca D'Alessandro, E

    2009-01-01

    Due to industrial revolution and the heavy use of fossil fuels, the concentration of greenhouse gases in the atmosphere has increased dramatically during the last hundred years, and this has lead to an increase in mean global temperature. The environmental consequences of this are: the melting of the ice caps, an increase in mean sea-levels, catastrophic events such as floodings, hurricanes and earthquakes, changes to the animal and vegetable kingdoms, a growth in vectors and bacteria in water thus increasing the risk of infectious diseases and damage to agriculture. The toxic effects of the pollution on human health are both acute and chronic. The Kyoto Protocol is an important step in the campaign against climatic changes but it is not sufficient. A possible solution might be for the States which produce the most of pollution to adopt a better political stance for the environment and to use renewable resources for the production of energy.

  8. 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.

  9. Assembly for facilitating inservice inspection of a reactor coolant pump rotor

    DOEpatents

    Veronesi, Luciano

    1990-01-01

    A reactor coolant pump has an outer casing with an internal cavity holding a coolant and a rotor rotatably mounted in the cavity within the coolant. An assembly for permitting inservice inspection of the pump rotor without first draining the coolant from the casing cavity is attached to an end of the pump. A cylindrical bore is defined through the casing in axial alignment with an end of pump rotor and opening into the internal cavity. An extension attached on the rotor end and rotatable therewith has a cylindrical coupler member extending into the bore. An outer end of the coupler member has an element configured to receive a tool for performance of inservice rotor inspection. A hollow cylindrical member is disposed in the bore and surrounds the coupler member. The cylindrical member is slidably movable relative to the coupler member along the bore between a retracted position wherein the cylindrical member is stored for normal pump operation and an extended position wherein the cylindrical member is extended for permitting inservice rotor inspection. A cover member is detachably and sealably attached to the casing across the bore for closing the bore and retaining the cylindrical member at its retracted position for normal pump operation. Upon detachment of the cover member, the cylindrical member can be extended to permit inservice rotor inspection.

  10. 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)

  11. 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.

  12. 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.

  13. Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core

    NASA Astrophysics Data System (ADS)

    Rochman, D.; Leray, O.; Hursin, M.; Ferroukhi, H.; Vasiliev, A.; Aures, A.; Bostelmann, F.; Zwermann, W.; Cabellos, O.; Diez, C. J.; Dyrda, J.; Garcia-Herranz, N.; Castro, E.; van der Marck, S.; Sjöstrand, H.; Hernandez, A.; Fleming, M.; Sublet, J.-Ch.; Fiorito, L.

    2017-01-01

    The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

  14. 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.

  15. 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.

  16. 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.

  17. Effective delayed neutron fraction measurement in the critical VENUS-F reactor using noise techniques

    SciTech Connect

    Doligez, X.

    2015-07-01

    This paper present the measurements of VENUS-F kinetic parameters using the Rossi-Alpha methods. The VENUS-F reactor is a zero-power reactor based in Mol, Belgium at SCK-CEN where the fuel is made of metallic enriched uranium with pure lead in order to simulate the behavior of lead fast reactor. The reactor can be operated under critical when it is coupled with GENEPI-3C. At the beginning of 2014, a measurement campaign was performed in order to estimate the kinetics parameters. In this paper, two measurements are analyzed at two different powers (approximately 2 W and 30 W) with 7 different fission chambers (with a 235-U deposit that varies from 1 g to 10 mg). All the correlation functions needed for the Rossi-Alpha method have been built for each possible set of two detectors in each configuration and values obtained from those functions for the effective delayed neutron fraction are then compared. The absolute necessity to operate at very low power is presented. The final value for the effective delayed neutron fraction is finally estimated to be 730 pcm ± 11 pcm and the prompt neutron generation time is estimated to be equal to 0,041 μseconds ± 0.04 μsec. (authors)

  18. 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.

  19. 77 FR 26050 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-02

    ... criticality safety analyses of pressurized water reactor spent nuclear fuel (SNF) in transportation packages... Doc No: 2012-10618] NUCLEAR REGULATORY COMMISSION [NRC-2012-0100] Burnup Credit in the Criticality... the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.'' This...

  20. Physics analyses of an accelerator-driven sub-critical assembly

    NASA Astrophysics Data System (ADS)

    Naberezhnev, Dmitry G.; Gohar, Yousry; Bailey, James; Belch, Henry

    2006-06-01

    Physics analyses have been performed for an accelerator-driven sub-critical assembly as a part of the Argonne National Laboratory activity in preparation for a joint conceptual design with the Kharkov Institute of Physics and Technology (KIPT) of Ukraine. KIPT has a plan to construct an accelerator-driven sub-critical assembly targeted towards the medical isotope production and the support of the Ukraine nuclear industry. The external neutron source is produced either through photonuclear reactions in tungsten or uranium targets, or deuteron reactions in a beryllium target. KIPT intends using the high-enriched uranium (HEU) for the fuel of the sub-critical assembly. The main objective of this paper is to study the possibility of utilizing low-enriched uranium (LEU) fuel instead of HEU fuel without penalizing the sub-critical assembly performance, in particular the neutron flux level. In the course of this activity, several studies have been carried out to investigate the main choices for the system's parameters. The external neutron source has been characterized and a pre-conceptual target design has been developed. Several sub-critical configurations with different fuel enrichments and densities have been considered. Based on our analysis, it was shown that the performance of the LEU fuel is comparable with that of the HEU fuel. The LEU fuel sub-critical assembly with 200-MeV electron energy and 100-kW electron beam power has an average total flux of ˜2.50×10 13 n/s cm 2 in the irradiation channels. The corresponding total facility power is ˜204 kW divided into 91 and 113 kW deposited in the target and sub-critical assemblies, respectively.

  1. 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.

  2. 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.

  3. Predicting growth of deposits on fuel assemblies of VVER-440 reactors

    NASA Astrophysics Data System (ADS)

    Kritskii, V. G.; Rodionov, Yu. A.; Berezina, I. G.; Fedorov, A. I.; Vitkovskii, S. L.; Shchedrin, M. G.; Galanin, A. V.; Slobodov, A. A.

    2009-05-01

    We discuss issues of simulating the growth of deposits on the surface of the core of VVER-440 reactors, which result in a higher pressure difference of coolant across the reactor and in the need to reduce the reactor power. We also present results of model calculations carried out for one fuel campaign of Unit 4 at the Novovoronezh nuclear power station.

  4. Development of neutron beam projects at the University of Texas TRIGA Mark II Reactor

    SciTech Connect

    Unlu, Kenan; Bauer, Thomas L.; Wehring, Bernard W.

    1992-07-01

    Recently, the UT-TRIGA research reactor was licensed and has become fully operational. This reactor, the first new US university reactor in 17 years, is the focus of a new reactor laboratory facility which is located on the Balcones Research Center at The University of Texas at Austin. The TRIGA Mark II reactor is licensed for 1.1 MW steady power operation, 3 dollar pulsing, and includes five beam ports. Various neutron beam-line projects have been assigned to each beam port. Neutron Depth Profiling (NDP) and the Texas Cold Neutron Source (TCNS) are close to completion and will be operational in the near future. The design of the NDP instrument has been completed, a target chamber has been built, and the thermal neutron collimator, detectors, data acquisition electronics, and data processing computers have been acquired. The target chamber accommodates wafers up to 12'' in diameter and provides remote positioning of these wafers. The design and construction of the TCNS has been completed. The TCNS consists of a moderator (mesitylene), a neon heat pipe, a cryogenic refrigerator, and neutron guide tubes. In addition, fission-fragment research (HIAWATHA), Neutron Capture Therapy, and Neutron Radiography are being pursued as projects for the other three beam ports. (author)

  5. 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

  6. The Kyoto Protocol and forestry practices in the United States

    Treesearch

    Bov B. Eav; Richard A. Birdsey; Linda S. Heath

    2000-01-01

    Forestry may play an important if not critical role in the ability of the U.S. to meet its greenhouse gas emissions target under the terms of the Kyoto Protocol. Given the low rate of change in the U.S. forest land area, the major anthropogenic influences on the current net forest carbon flux are forest management and protection activities that have resulted in...

  7. 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.

  8. 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.

  9. 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.

  10. Critical exponents and universality in pinned charge density waves

    NASA Astrophysics Data System (ADS)

    Celino, M.; Corberi, F.

    1992-07-01

    We study analytically and numerically the behaviour of charge density waves above the depinning transition. The system behaves as a complex self-organized object which reacts to the external stimulations by adapting itself in a marginally stable state, which gives rise to a power law distribution of avalanches. Critical exponents and current power spectrum are evaluated analytically in any dimension by means of a simple coarse grained model. Extensive simulations are performed in one-dimensional systems below and above threshold, to critically verify the question of universality of coarse grained models with respect to the original Hamiltonian.

  11. 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.

  12. 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.

  13. 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.

  14. 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

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

    NASA Astrophysics Data System (ADS)

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

    2014-12-01

    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.

  16. 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.

  17. A simple calculation of control assembly effectiveness in a liquid-metal fast breeder reactor by a transport-diffusion equivalence method

    SciTech Connect

    Benoist, P. ); Carta, M. ); Palmiotti, G. ); Salvatores, M. )

    1989-11-01

    A method to calculate the effectiveness of the control assembly in a fast neutron reactor is proposed. For each type of heterogeneous assembly (control or follower), a polar parameter, taking into account the assembly absorption and the axial leakage of neutrons inside the assembly, is defined. In a similar way, a bipolar parameter, taking into account the reaction of the assembly to a transverse flux gradient, is also defined. These two parameters, deduced from transport theory, are used to determine the absorption cross section and the diffusion coefficient of an equivalent homogeneous control or follower assembly. These new parameters are introduced in a one-group diffusion code, calculating the reactor as a whole with any number of control and follower assemblies. An approximate generalization to multigroup theory is proposed. Numerical comparisons show that this equivalent diffusion method gives results that are much closer to transport results than those obtained by the classical diffusion theory.

  18. Industry-University SBIR NDT Projects — A Critical Assessment

    NASA Astrophysics Data System (ADS)

    Reinhart, Eugene R.

    2007-03-01

    The Small Business Innovative Research (SBIR) program, funded by various United States government agencies (DOD, DOE, NSF, etc.), provides funds for Research and Development (R&D) of nondestructive testing (NDT) techniques and equipment, thereby supplying valuable money for NDT development by small businesses and stimulating cooperative university programs. A review and critical assessment of the SBIR program as related to NDT is presented and should provide insight into reasons for or against pursuing this source of R&D funding.

  19. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    SciTech Connect

    Montierth, Leland M.

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  20. Crossover of critical Casimir forces between different surface universality classes.

    PubMed

    Mohry, T F; Maciołek, A; Dietrich, S

    2010-06-01

    In confined systems near a continuous phase transition the long-ranged fluctuations of the corresponding order parameter are subject to boundary conditions. These constraints result in so-called critical Casimir forces acting as effective forces on the confining surfaces. For systems belonging to the Ising bulk universality class corresponding to a scalar order parameter the critical Casimir force is studied for the film geometry in the crossover regime characterized by different surface fields at the two surfaces. The scaling function of the critical Casimir force is calculated within mean-field theory. Within our approach, the scaling functions of the critical Casimir force and of the order parameter profile for finite surface fields can be mapped by rescaling, except for a narrow crossover regime, onto the corresponding scaling function of the so-called normal fixed point of strong surface fields. In the crossover regime, the critical Casimir force as function of temperature exhibits more than one extremum and for certain ranges of surface field strengths it changes sign twice upon varying temperature. Monte Carlo simulation data obtained for a three-dimensional Ising film show similar trends. The sign of the critical Casimir force can be inferred from the comparison of the order parameter profiles in the film and in the semi-infinite geometry.

  1. 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.

  2. Experimental studies of heat exchange for sodium boiling in the fuel assembly model: Safety substantiation of a promising fast reactor

    NASA Astrophysics Data System (ADS)

    Khafizov, R. R.; Poplavskii, V. M.; Rachkov, V. I.; Sorokin, A. P.; Trufanov, A. A.; Ashurko, Yu. M.; Volkov, A. V.; Ivanov, E. F.; Privezentsev, V. V.

    2017-01-01

    Numerical simulation of the ULOF-type accident development in a fast reactor with sodium coolant performed using the COREMELT code indicates that sodium boiling in the active core takes place. The boiling is accompanied by oscillations of the technological parameters of the reactor installation; these oscillations can go on during several tens of seconds. In this case, it is possible that a stable regime of removal of heat from residual energy release is implemented. The model of the two-phase coolant flow applied in the code has an important effect on the numerical results; that is why this model needs experimental verification. For eliminating the development of an accident resulting in destruction of the active core elements, a structural solution is proposed; the essence of it is the application of the sodium void above the reactor active core. The experimental installation was developed and the heat exchange at sodium boiling in the model fuel assembly of the fast reactor in the regimes of natural and forced circulation in the presence of the sodium void and the top end shield was studied. It was demonstrated that, in the presence of the sodium void, it is possible to provide long-term cooling of the fuel assembly for a thermal flux density on the fuel element simulator surface of up to 140 and 170 kW/m2 in the natural and forced circulation modes, respectively. The obtained data are used for more precise determination of the numerical model of sodium boiling in the fuel assembly and verification of the COREMELT code.

  3. 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)

  4. A Set of Critical Loading Configurations of the IPEN/MB-01 Reactor

    SciTech Connect

    Santos, Adimir dos; Fanaro, Leda C.C.B.; Siqueira, Pauloj D.; Jerez, Rogerio; Fuga, Rinaldo

    2005-05-24

    A new set of critical loading configurations has been performed at the IPEN/MB-01 research reactor facility. The configurations were submitted to ICSBEP and have been approved for publication in the coming September edition, 2004. The final keff uncertainty is judged to be 100 pcm for all configurations and suitable for a benchmark experiment. The keff results using the currently released nuclear data libraries ENDF/B-VI.8 and JEF3.0 are in very good agreement with the experimental values. JENDL3.3 overestimates keff by around 200 pcm.

  5. 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

  6. 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.

  7. 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)

  8. 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

    2016-03-28

    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(C)(S) governs the strength of the assembly as predicted by the shear-lag model. The intermediate critical length L(C)(P) 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(C)(T) 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.

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. 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

  14. 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.

  15. Recent accomplishments in neutron beam projects at the University of Texas Research Reactor

    SciTech Connect

    Uenlue, K.; Wehring, B.W.

    1994-12-31

    The design of a cold neutron source facility at the University of Texas TRIGA research reactor is described. The UT-TRIGA has 5 neutron beam ports. Because of the different characteristics of the ports, various research projects are being pursued. Among these projects, The Texas cold neutron source and neutron depth profiling are operational; neutron focusing, prompt gamma activation analysis, and neutron capture therapy research are progressing.

  16. 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.

  17. Void fraction distribution in a boiling water reactor fuel assembly and the evaluation of subchannel analysis codes

    SciTech Connect

    Inoue, Akira; Futakuchi, Masanobu; Yagi, Makoto; Mitsutake, Toru; Morooka, Shinichi

    1995-12-01

    Void fraction measurement tests for boiling water reactor (BWR) simulated nuclear fuel assemblies have been conducted using an X-ray computed tomography scanner.there are two types of fuel assemblies concerning water rods. One fuel assembly has two water rods; the other has one large water rod. The effects of the water rods on radial void fraction distributions are measured within the fuel assemblies. The results show that the water rod effect does not make a large difference in void fraction distribution. The subchannel analysis codes COBRA/BWR and THERMIT-2 were compared with subchannel-averaged void fractions. The prediction accuracy of COBRA/BWR and THERMIT-2 for the subchannel-averaged void fraction was {Delta}{alpha} = {minus}3.6%, {sigma} = 4.8% and {Delta}{alpha} = {minus}4.1%, {sigma} = 4.5%, respectively, where {Delta}{alpha} is the average of the difference measured and calculated values. The subchannel analysis codes are highly applicable for the prediction of a two-phase flow distribution within BWR fuel assemblies.

  18. 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.

  19. Design of a weapons-grade plutonium assembly for optimal burnup in a standard pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Alonso-Vargas, Gustavo

    We created a new MOX fuel assembly design that can be used in standard Westinghouse pressurized water reactors (PWR) to maximize the plutonium throughput while introducing the lowest perturbation possible to the control and safety systems of the reactor. Our assembly design, which is called MIX-33, appears to be a good option for the disposition of weapons-grade plutonium (WG-Pu), increasing the plutonium disposition rate by 8% compared to a previous Westinghouse design. It is based in two novel ideas: the use of both uranium and plutonium fuel pins in the same assembly, and the increase of the moderation ratio of the assembly. We replaced 8 fuel pins by water holes to increase the moderation ratio. We can transition smoothly from a full LEU core to a full MIX-33 core meeting the operational and safety regulations of a standard PWR. Given a MOX supply interruption scenario we can transition smoothly to full LEU meeting the safety regulations and using standard LEU assemblies with uniform enriched pin-wise distribution. If the MOX supply is interrupted for only one cycle, we are able to transition back to full MIX-33 core. However, in this case we probably need to de-rate the power by a few percent for a few weeks at the beginning of the cycle (BOC) to accommodate high peaking. For comparison we created another assembly design without extra water holes, which we called "MIX-25". It behaves in all the conditions analyzed in a similar way to the MIX-33 but it does present minor control problems. These can be solved by making small modifications to the control and safety systems, namely by enriching the boron-10 content of some boron absorbers. Thus, the addition of water holes replacing fuel pins helps to improve the MIX-33 performance and eliminate the difficulties seen in the MIX-25 design. We also performed a benchmarking analysis to test the code CASMO-3 to analyze WG-Pu assemblies, using the code MCNP-4A to compare. We found good agreement between CASMO-3 and

  20. Criticality and Universality in Healthy Heart Rate Dynamics

    NASA Astrophysics Data System (ADS)

    Struzik, Zbigniew R.; Kiyono, Ken; Hayano, Junichiro; Sakata, Seiichiro; Kwak, Shin; Yamamoto, Yoshiharu

    2005-08-01

    Methodologies originally developed in the field of statistical physics of complex phenomena have been proven to provide new insights into the modeling, description and understanding of the human heart rate regulatory system. Recent studies have shown the heart rate control system to maintain universality properties characteristic of physical systems exhibiting far-from-equilibrium, critical state-like dynamics. Simultaneously, heart rate regulation has been shown to display correlation properties of antagonist dynamics involving antagonist actors, pertinent to some far-from-equilibrium systems. We discuss the range of validity and breakdown scenarios of the universal properties in heart rate regulation leading to the diagnostic capability and also to new challenges for both analysis methods and up-to-date simulation models.

  1. Forest fire spread with non-universal critical behavior

    NASA Astrophysics Data System (ADS)

    Khelloufi, K.; Baara, Y.; Clerc, J. P.; Porterie, B.; Zekri, N.

    2013-10-01

    The critical behavior of spread dynamics is examined using a forest fire model. This model is characterized by long-range interactions due to flame radiation and a weighting process induced by the combustibles’ ignition energy and the flame residence time. Unlike magnetic systems, this model exhibits a non-universal phase transition. The critical exponents of the rate of spread depend both on the local interaction and on weighting. Near the transition, the exponent x of rate of spread is found to be equivalent to that of correlation time. The weighting process exhibits a new phase transition related to the heating process. This transition is analogous to the gelation transition in spin glasses.

  2. Critical mingling and universal correlations in model binary active liquids

    PubMed Central

    Bain, Nicolas; Bartolo, Denis

    2017-01-01

    Ensembles of driven or motile bodies moving along opposite directions are generically reported to self-organize into strongly anisotropic lanes. Here, building on a minimal model of self-propelled bodies targeting opposite directions, we first evidence a critical phase transition between a mingled state and a phase-separated lane state specific to active particles. We then demonstrate that the mingled state displays algebraic structural correlations also found in driven binary mixtures. Finally, constructing a hydrodynamic theory, we single out the physical mechanisms responsible for these universal long-range correlations typical of ensembles of oppositely moving bodies. PMID:28656963

  3. Critical mingling and universal correlations in model binary active liquids

    NASA Astrophysics Data System (ADS)

    Bain, Nicolas; Bartolo, Denis

    2017-06-01

    Ensembles of driven or motile bodies moving along opposite directions are generically reported to self-organize into strongly anisotropic lanes. Here, building on a minimal model of self-propelled bodies targeting opposite directions, we first evidence a critical phase transition between a mingled state and a phase-separated lane state specific to active particles. We then demonstrate that the mingled state displays algebraic structural correlations also found in driven binary mixtures. Finally, constructing a hydrodynamic theory, we single out the physical mechanisms responsible for these universal long-range correlations typical of ensembles of oppositely moving bodies.

  4. 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

  5. Kinetic stability analysis of protein assembly on the center manifold around the critical point.

    PubMed

    Tsuruyama, Tatsuaki

    2017-02-02

    Non-linear kinetic analysis is a useful method for illustration of the dynamic behavior of cellular biological systems. To date, center manifold theory (CMT) has not been sufficiently applied for stability analysis of biological systems. The aim of this study is to demonstrate the application of CMT to kinetic analysis of protein assembly and disassembly, and to propose a novel framework for nonlinear multi-parametric analysis. We propose a protein assembly model with nonlinear kinetics provided by the fluctuation in monomer concentrations during their diffusion. When the diffusion process of a monomer is self-limited to give kinetics non-linearity, numerical simulations suggest the probability that the assembly and disassembly oscillate near the critical point. We applied CMT to kinetic analysis of the center manifold around the critical point in detail, and successfully demonstrated bifurcation around the critical point, which explained the observed oscillation. The stability kinetics of the present model based on CMT illustrates a unique feature of protein assembly, namely non-linear behavior. Our findings are expected to provide methodology for analysis of biological systems.

  6. 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].

  7. A critical assembly designed to measure neutronic benchmarks in support of the space nuclear thermal propulsion program

    NASA Astrophysics Data System (ADS)

    Parma, Edward J.; Ball, Russell M.; Hoovler, Gary S.; Selcow, Elizabeth C.; Cerbone, Ralph J.

    1993-01-01

    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 rector-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 overall description of the reactor is presented along with key design features and safety-related aspects.

  8. 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

  9. 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.

  10. 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.

  11. [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.

  12. Achieving collaborative workplace learning in a university critical care course.

    PubMed

    Manias, Elizabeth; Aitken, Robyn

    2003-02-01

    The purpose of this paper is to describe the development, implementation and evaluation of a new critical care curriculum based on the tenets of collaborative workplace learning. It also examines lecturers' and clinical educators' issues, and explores students' evaluations of the old curriculum compared with those of the new curriculum. Three data collection methods were used for this study. Comprehensive notes were maintained of the meetings conducted with lecturers, clinical educators and representative students during the development and implementation of the course. Three focus group interviews were conducted with students before the introduction of the new curriculum and three focus group interviews were conducted during first semester following implementation of the new curriculum. Quality-of-teaching surveys were also completed by two groups of critical care course students: one group before and one group following the introduction of the new curriculum. Major findings in this study included: developing a sense of ownership of the curriculum for clinical educators, clinical educators' difficulties with addressing their responsibilities, amalgamating theoretical learning with clinical practice, and tackling students' workload. This paper demonstrates the value of using the collaborative workplace learning approach in strategically addressing the challenges of developing and conducting a university critical care course.

  13. 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)

  14. 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.

  15. Benchmarking of HEU mental annuli critical assemblies with internally reflected graphite cylinder

    NASA Astrophysics Data System (ADS)

    Xiaobo, Liu; Bess, John D.; Marshall, Margaret A.

    2017-09-01

    Three experimental configurations of critical assemblies, performed in 1963 at the Oak Ridge Critical Experiment Facility, which are assembled using three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches) metal annuli with internally reflected graphite cylinder are evaluated and benchmarked. The experimental uncertainties which are 0.00057, 0.00058 and 0.00057 respectively, and biases to the benchmark models which are - 0.00286, - 0.00242 and - 0.00168 respectively, were determined, and the experimental benchmark keff results were obtained for both detailed and simplified models. The calculation results for both detailed and simplified models using MCNP6-1.0 and ENDF/B-VII.1 agree well to the benchmark experimental results within difference less than 0.2%. The benchmarking results were accepted for the inclusion of ICSBEP Handbook.

  16. Benchmarking of HEU Mental Annuli Critical Assemblies with Internally Reflected Graphite Cylinder

    SciTech Connect

    Xiaobo, Liu; Bess, John D.; Marshall, Margaret A.

    2016-09-01

    Three experimental configurations of critical assemblies, performed in 1963 at the Oak Ridge Critical Experiment Facility, which are assembled using three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches) metal annuli with internally reflected graphite cylinder are evaluated and benchmarked. The experimental uncertainties which are 0.00055, 0.00055 and 0.00055 respectively, and biases to the detailed benchmark models which are -0.00179, -0.00189 and -0.00114 respectively, were determined, and the experimental benchmark keff results were obtained for both detailed and simplified model. The calculation results for both detailed and simplified models using MCNP6-1.0 and ENDF VII.1 agree well to the benchmark experimental results with a difference of less than 0.2%. These are acceptable benchmark experiments for inclusion in the ICSBEP Handbook.

  17. Self-assembly in chains, rings, and branches: a single component system with two critical points.

    PubMed

    Rovigatti, Lorenzo; Tavares, José Maria; Sciortino, Francesco

    2013-10-18

    We study the interplay between phase separation and self-assembly in chains, rings, and branched structures in a model of particles with dissimilar patches. We extend Wertheim's first order perturbation theory to include the effects of ring formation and to theoretically investigate the thermodynamics of the model. We find a peculiar shape for the vapor-liquid coexistence, featuring reentrant behavior in both phases and two critical points, despite the single-component nature of the system. The emergence of the lower critical point is caused by the self-assembly of rings taking place in the vapor, generating a phase with lower energy and lower entropy than the liquid. Monte Carlo simulations of the same model fully support these unconventional theoretical predictions.

  18. Fission products measured from highly-enriched uranium irradiated under 10B4C in a research reactor

    SciTech Connect

    Metz, Lori A.; Friese, Judah I.; Finn, Erin C.; Greenwood, Lawrence R.; Hines, Corey C.; King, Matthew D.; Wall, Donald E.

    2016-03-01

    Critical assemblies provide one method of achieving a fast neutron spectrum that is close to a 235U fission-energy neutron spectrum for nuclear data measurements. Previous work has demonstrated the use of a natural boron carbide capsule for spectral-tailoring in a mixed spectrum reactor as an alternate and complementary method for performing fission-energy neutron experiments. Previous fission products measurements showed that the neutron spectrum achievable with natural boron carbide was not as hard as what can be achieved with critical assemblies. New measurements performed with the Washington State University TRIGA reactor using a boron carbide capsule 96% enriched in 10B for irradiations resulted in a neutron spectrum very similar to a critical assembly and a pure 235U fission spectrum. The current work describes an experiment involving a highly-enriched uranium target irradiated under the new 10B4C capsule. Fission product yields were measured following radiochemical separations and are presented here. Reactor dosimetry measurements for characterizing neutron spectra and fluence for the enriched boron carbide capsule and critical assemblies are also discussed.

  19. Analysis of pulsed neutron measurements on the fuel pebble assembly during the approach to critical mass

    SciTech Connect

    Brodkin, E.; Lebedev, G.

    1995-12-31

    The two-dimensional cylindrical model of HTR-ASTRA fuel pebble bed assembly was used in the transport calculations of k{sub eff} and corresponding Rossi-{alpha} for interpretation of pulsed neutron measurements which have been carrying out during approach to critical mass. This analysis demonstrates possibility to evaluate k{sub eff} above 0.9 using {alpha}-prompt decay constant measured during core loading by fuel balls and to extrapolate these data for determination of critical mass similar to inverse counting technique.

  20. Thermal-Hydraulic Simulations of Single Pin and Assembly Sector for IVG- 1M Reactor

    SciTech Connect

    Kraus, A.; Garner, P.; Hanan, N.

    2015-01-15

    Thermal-hydraulic simulations have been performed using computational fluid dynamics (CFD) for the highly-enriched uranium (HEU) design of the IVG.1M reactor at the Institute of Atomic Energy (IAE) at the National Nuclear Center (NNC) in the Republic of Kazakhstan. Steady-state simulations were performed for both types of fuel assembly (FA), i.e. the FA in rows 1 & 2 and the FA in row 3, as well as for single pins in those FA (600 mm and 800 mm pins). Both single pin calculations and bundle sectors have been simulated for the most conservative operating conditions corresponding to the 10 MW output power, which corresponds to a pin unit cell Reynolds number of only about 7500. Simulations were performed using the commercial code STAR-CCM+ for the actual twisted pin geometry as well as a straight-pin approximation. Various Reynolds-Averaged Navier-Stokes (RANS) turbulence models gave different results, and so some validation runs with a higher-fidelity Large Eddy Simulation (LES) code were performed given the lack of experimental data. These singled out the Realizable Two-Layer k-ε as the most accurate turbulence model for estimating surface temperature. Single-pin results for the twisted case, based on the average flow rate per pin and peak pin power, were conservative for peak clad surface temperature compared to the bundle results. Also the straight-pin calculations were conservative as compared to the twisted pin simulations, as expected, but the single-pin straight case was not always conservative with regard to the straight-pin bundle. This was due to the straight-pin temperature distribution being strongly influenced by the pin orientation, particularly near the outer boundary. The straight-pin case also predicted the peak temperature to be in a different location than the twisted-pin case. This is a limitation of the straight-pin approach. The peak temperature pin was in a different location from the peak power pin in every case simulated, and occurred at an

  1. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

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

  2. A catalytic chiral gel microfluidic reactor assembled via dynamic covalent chemistry.

    PubMed

    Liu, Haoliang; Feng, Juan; Zhang, Jianyong; Miller, Philip W; Chen, Liuping; Su, Cheng-Yong

    2015-04-16

    A novel dynamic covalent gel strategy is reported to immobilize an asymmetric catalyst within the channels of a microfluidic flow reactor. A layer of a catalytically active Mn-salen dynamic covalent imine gel matrix was coated onto a functionalized capillary. Mn-salen active moiety was incorporated into dynamic covalent imine gel matrix via the reaction of a chiral Mn-salen dialdehyde unit with a tetraamine linker. The catalytic activity of the capillary reactor has been demonstrated in enantioselective kinetic resolution of secondary alcohols.

  3. 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.

  4. Microstructural examination of fatigue accumulation in critical LWR (light water reactor) components: Final report

    SciTech Connect

    Allen, A.J.; Buttle, D.J.; Coleman, C.F.; Smith, F.A.; Smith, R.L.

    1988-01-01

    This report describes a morphological study of the feasibility of measuring the fatigue damage accumulation state of critical light water reactor (LWR) components by microstructural examination. The changes in microstructure associated with fatigue processes are first discussed so that relevant NDE measurement parameters can be identified. (The creep regime is not considered in this report). The candidate NDE techniques are then reviewed in detail under the following headings: positron annihilation, x-ray diffraction, magnetic techniques, the magnetic Barkhausen effect, the magneto acoustic technique, acoustic emission, ultrasonic techniques and finally other miscellaneous techniques applicable to fatigue damage assessment. All the feasible techniques are summarised and rated in a set of comparison tables. A possible programme for the immediate development of the positron annihilation lineshape technique is proposed. It is concluded that the most successful method of measuring the fatigue accumulation in LWR critical components in a way which relates to the intent of the ASME pressure vessel codes, is likely to be the use of several techniques together and the cross-relation of the results obtained by each. Five techniques are highlighted for immediate possible development: 'etching and surface replication', 'positron annihilation lineshapes', 'x-ray diffraction residual stress', 'acoustic emission' and 'ultrasonic surface acoustic waves'.

  5. 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.

  6. 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.

  7. Diversity of critical behavior within a universality class.

    PubMed

    Dohm, Volker

    2008-06-01

    We study spatial anisotropy effects on the bulk and finite-size critical behavior of the O(n) symmetric anisotropic phi;{4} lattice model with periodic boundary conditions in a d -dimensional hypercubic geometry above, at, and below Tc. The absence of two-scale factor universality is discussed for the bulk order-parameter correlation function, the bulk scattering intensity, and for several universal bulk amplitude relations. The anisotropy parameters are observable by scattering experiments at Tc. For the confined system, renormalization-group theory within the minimal subtraction scheme at fixed dimension d for 2universality. The tails of the large- L

  8. 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.

  9. 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.

  10. 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

  11. 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

  12. 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

  13. Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

    NASA Astrophysics Data System (ADS)

    Krása, Antonín; Kochetkov, Anatoly; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente

    2017-09-01

    VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented.

  14. A Neutronic Program for Critical and Nonequilibrium Study of Mobile Fuel Reactors: The Cinsf1D Code

    SciTech Connect

    Lecarpentier, David; Carpentier, Vincent

    2003-01-15

    Molten salt reactors (MSRs) have the distinction of having a liquid fuel that is also the coolant. The transport of delayed-neutron precursors by the fuel modifies the precursors' equation. As a consequence, it is necessary to adapt the methods currently used for solid fuel reactors to achieve critical or kinetics calculations for an MSR. A program is presented for which this adaptation has been carried out within the framework of the two-energy-group diffusion theory with one dimension of space. This program has been called Cinsf1D (Cinetique pour reacteur a sels fondus 1D)

  15. A setup for active neutron analysis of the fissile material content in fuel assemblies of nuclear reactors

    NASA Astrophysics Data System (ADS)

    Bushuev, A. V.; Kozhin, A. F.; Aleeva, T. B.; Zubarev, V. N.; Petrova, E. V.; Smirnov, V. E.

    2016-12-01

    An active neutron method for measuring the residual mass of 235U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual 235U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of 238U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be used in experiments where the studied object needs to be irradiated with a uniform fluence.

  16. A setup for active neutron analysis of the fissile material content in fuel assemblies of nuclear reactors

    SciTech Connect

    Bushuev, A. V.; Kozhin, A. F. Aleeva, T. B.; Zubarev, V. N.; Petrova, E. V.; Smirnov, V. E.

    2016-12-15

    An active neutron method for measuring the residual mass of {sup 235}U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual {sup 235}U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of {sup 238}U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be used in experiments where the studied object needs to be irradiated with a uniform fluence.

  17. 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:…

  18. 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:…

  19. Universality of P - V criticality in horizon thermodynamics

    NASA Astrophysics Data System (ADS)

    Hansen, Devin; Kubizňák, David; Mann, Robert B.

    2017-01-01

    We study P - V criticality of black holes in Lovelock gravities in the context of horizon thermodynamics. The corresponding first law of horizon thermodynamics emerges as one of the Einstein-Lovelock equations and assumes the universal (independent of matter content) form δ E = T δ S - P δ V , where P is identified with the total pressure of all matter in the spacetime (including a cosmological constant Λ if present). We compare this approach to recent advances in extended phase space thermodynamics of asymptotically AdS black holes where the `standard' first law of black hole thermodynamics is extended to include a pressure-volume term, where the pressure is entirely due to the (variable) cosmological constant. We show that both approaches are quite different in interpretation. Provided there is sufficient non-linearity in the gravitational sector, we find that horizon thermodynamics admits the same interesting black hole phase behaviour seen in the extended case, such as a Hawking-Page transition, Van der Waals like behaviour, and the presence of a triple point. We also formulate the Smarr formula in horizon thermodynamics and discuss the interpretation of the quantity E appearing in the horizon first law.

  20. 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…

  1. 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.

  2. 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

  3. 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)

  4. 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.

  5. Critical role of wettability in assembly of zirconia nanoparticles on a self-assembled monolayer-patterned substrate

    NASA Astrophysics Data System (ADS)

    Yang, Mi-Sun; Lee, Seung-Hoon; Moon, Byung Kee; Yoo, Seung Ryul; Hwang, Seongpil; Jang, Jae-Won

    2016-08-01

    This study investigated which factors decisively influence colloidal nanoparticle (NP) assembly on a self-assembled monolayer (SAM)-patterned substrate. Zirconia (ZrO2) NP assembly on a poly(dimethylsiloxane) (PDMS)-stamped SAM-patterned Au substrate was carried out while the size and surface charge state of the NPs and the substrate wettability were altered. ZrO2 particles with diameters of 350 nm, 560 nm, and 1100 nm were employed to examine the effect of NP size on the assembly. Bare ZrO2 NPs with a negatively charged surface and ZrO2 NPs with a positively charged surface through 3-aminopropyltriethoxysilane encapsulation were prepared for the NP assembly. Moreover, the substrate wettability effect on the NP assembly was evaluated by comparing the assembly on substrates with the PDMS-patterned SAMs of thiols with polar and non-polar functional groups. From the characterization of the number of NPs in a pattern and the effective area of assembled NPs (Aeff), positively charged ZrO2 NP assembly on negatively charged patterns showed the highest number density of particles in a pattern compared with the other combinations in both 350-nm and 560-nm ZrO2 NPs. This observation can be attributed to negatively charged 16-mercaptohexadecanoic acid SAMs having greater polarity (more polar groups) than positively charged 11-amino-1-undecanethiol SAMs within the condition of the colloidal ZrO2 NP assembly.

  6. 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...

  7. 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...

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. Exploratory Learning through Critical Inquiry: Survey of Critical Inquiry Programs at Mid-Sized U.S. Universities

    ERIC Educational Resources Information Center

    Foote, Stephanie M.; Harrison, David S.; Ritchie, C. Michael; Dyer, Andrew

    2012-01-01

    Many alternatives exist for setting the delivery, content, direction, tone, and priorities for a Critical Inquiry/Thinking general education program. Review of our university's overall general education program indicated the need, and overwhelming faculty approval, for a program to improve critical thinking skills, to specifically include…

  14. 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.

  15. 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.

  16. Conceptions of Critical Thinking from University EFL Teachers

    ERIC Educational Resources Information Center

    Marin, Matias A.; de la Pava, Luisa

    2017-01-01

    Critical Thinking has become an educational and social ideal. English as a Foreign Language (EFL) teaching has not been apart from the discussion on the importance of implementing Critical Thinking into the educational process. However, research on Critical Thinking has broadly been carried out in other fields of knowledge rather than in EFL.…

  17. 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)

  18. Radiation and criticality safety analyses for the highly-enriched uranium core removal from a research reactor.

    PubMed

    Dennis, Haile; Grant, Charles; Preston, John

    2017-11-01

    Analysis was performed to estimate radiation levels during removal and packaging of the highly-enriched uranium core of the JM-1 SLOWPOKE-2 research reactor. Due to severe limitations of space in and around the reactor pool, the core could not be removed in the conventional manner as was done for previous SLOWPOKE defuelling operations. A transfer shield, with a balance between shielding efficacy, volume and weight was designed. Fuel depletion, Monte Carlo shielding and criticality calculations were performed. Comparisons of measured and calculated dose rates as well as results of the criticality safety assessment are presented. The designed transfer shield reduced the calculated unshielded dose rate from 29Sv/h to 8mSv/h. The maximum calculated effective neutron multiplication factor of approximately 0.89 was below the 0.91 upper subricital limit. Copyright © 2017 Elsevier Ltd. All rights reserved.

  19. Universal relations in the self-assembly of proteins and RNA

    NASA Astrophysics Data System (ADS)

    Thirumalai, D.

    2014-10-01

    Concepts rooted in physics are becoming increasingly important in biology as we transition to an era in which quantitative descriptions of all processes from molecular to cellular level are needed. In this perspective I discuss two unexpected findings of universal behavior, uncommon in biology, in the self-assembly of proteins and RNA. These findings, which are surprising, reveal that physics ideas applied to biological problems, ranging from folding to gene expression to cellular movement and communication between cells, might lead to discovery of universal principles operating in adoptable living systems.

  20. Universal relations in the self-assembly of proteins and RNA.

    PubMed

    Thirumalai, D

    2014-10-08

    Concepts rooted in physics are becoming increasingly important in biology as we transition to an era in which quantitative descriptions of all processes from molecular to cellular level are needed. In this perspective I discuss two unexpected findings of universal behavior, uncommon in biology, in the self-assembly of proteins and RNA. These findings, which are surprising, reveal that physics ideas applied to biological problems, ranging from folding to gene expression to cellular movement and communication between cells, might lead to discovery of universal principles operating in adoptable living systems.

  1. Particulate Filtration from Emissions of a Plasma Pyrolysis Assembly Reactor Using Regenerable Porous Metal Filters

    NASA Technical Reports Server (NTRS)

    Agui, Juan H.; Abney, Morgan; Greenwood, Zachary; West, Philip; Mitchell, Karen; Vijayakumar, R.; Berger, Gordon M.

    2017-01-01

    Microwave-based plasma pyrolysis technology is being studied as a means of supporting oxygen recovery in future spacecraft life support systems. The process involves the conversion of methane produced from a Sabatier reactor to acetylene and hydrogen, with a small amount of solid carbon particulates generated as a side product. The particles must be filtered before the acetylene is removed and the hydrogen-rich gas stream is recycled back to the CRA. We discuss developmental work on porous metal media filters for removing the carbon particulate emissions from the PPA exit gas stream and to provide in situ media regeneration capability. Because of the high temperatures involved in oxidizing the deposited carbon during regeneration, there was particular focus in this development on the materials that could be used, the housing design, and heating methods. This paper describes the design and operation of the filter and characterizes their performance from integrated testing at the Environmental Chamber (E-Chamber) at MSFC.

  2. Particulate filtration from emissions of a plasma pyrolysis assembly reactor using regenerable porous metal filters

    NASA Technical Reports Server (NTRS)

    Berger, Gordon M.; Agui, Juan H.; Vijayakumar, R.; Abney, Morgan B.; Greenwood, Zachary W.; West, Philip J.; Mitchell, Karen O.

    2017-01-01

    Microwave-based plasma pyrolysis technology is being studied as a means of supporting oxygen recovery in future spacecraft life support systems. The process involves the conversion of methane produced from a Sabatier reactor to acetylene and hydrogen, with a small amount of solid carbon particulates generated as a side product. The particles must be filtered before the acetylene is removed and the hydrogen-rich gas stream is recycled back to the CRA. We discuss developmental work on porous metal media filters for removing the carbon particulate emissions from the PPA exit gas stream and to provide in situ media regeneration capability. Because of the high temperatures involved in oxidizing the deposited carbon during regeneration, there was particular focus in this development on the materials that could be used, the housing design, and heating methods. This paper describes the design and operation of the filter and characterizes their performance from integrated testing at the Environmental Chamber (E-Chamber) at MSFC.

  3. Machine for removing in-core instrument assemblies from a nuclear reactor

    SciTech Connect

    Klumb, R.H.; Margotta, K.V.; Shendy, D.S.

    1982-02-02

    A machine for smoothly and controllably winding or unwinding a stiff in-core-instrument tube onto and off of a reel during ythe refueling of a nuclear reactor. The machine includes a frame and a circular reel having a substantially continuous helical groove extending around the circumference of the reel. The groove is adapted to receive the tube. A plurality of cam rollers are carried by the frame and closely spaced around the circumference of the reel. The rollers keep the tube in the groove whereby the tube may be more easily wound onto or off of the reel. In the preferred embodiment, the reel carries a disposable cartridge in which the grooves are formed.

  4. 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.

  5. Universal Off-Equilibrium Scaling of Critical Cumulants in the QCD Phase Diagram

    DOE PAGES

    Mukherjee, Swagato; Venugopalan, Raju; Yin, Yi

    2016-11-23

    Exploiting the universality between the QCD critical point and the three-dimensional Ising model, closed form expressions derived for nonequilibrium critical cumulants on the crossover side of the critical point reveal that they can differ in both magnitude and sign from equilibrium expectations. Here, we demonstrate here that key elements of the Kibble-Zurek framework of nonequilibrium phase transitions can be employed to describe the dynamics of these critical cumulants. Lastly, our results suggest that observables sensitive to critical dynamics in heavy-ion collisions should be expressible as universal scaling functions, thereby providing powerful model-independent guidance in searches for the QCD critical point.

  6. Universal Off-Equilibrium Scaling of Critical Cumulants in the QCD Phase Diagram.

    PubMed

    Mukherjee, Swagato; Venugopalan, Raju; Yin, Yi

    2016-11-25

    Exploiting the universality between the QCD critical point and the three-dimensional Ising model, closed form expressions derived for nonequilibrium critical cumulants on the crossover side of the critical point reveal that they can differ in both magnitude and sign from equilibrium expectations. We demonstrate here that key elements of the Kibble-Zurek framework of nonequilibrium phase transitions can be employed to describe the dynamics of these critical cumulants. Our results suggest that observables sensitive to critical dynamics in heavy-ion collisions should be expressible as universal scaling functions, thereby providing powerful model-independent guidance in searches for the QCD critical point.

  7. Universal Off-Equilibrium Scaling of Critical Cumulants in the QCD Phase Diagram

    NASA Astrophysics Data System (ADS)

    Mukherjee, Swagato; Venugopalan, Raju; Yin, Yi

    2016-11-01

    Exploiting the universality between the QCD critical point and the three-dimensional Ising model, closed form expressions derived for nonequilibrium critical cumulants on the crossover side of the critical point reveal that they can differ in both magnitude and sign from equilibrium expectations. We demonstrate here that key elements of the Kibble-Zurek framework of nonequilibrium phase transitions can be employed to describe the dynamics of these critical cumulants. Our results suggest that observables sensitive to critical dynamics in heavy-ion collisions should be expressible as universal scaling functions, thereby providing powerful model-independent guidance in searches for the QCD critical point.

  8. Maintaining a Critical Spectra within Monteburns for a Gas-Cooled Reactor Array by Way of Control Rod Manipulation

    SciTech Connect

    Adigun, Babatunde John; Fensin, Michael Lorne; Galloway, Jack D.; Trellue, Holly Renee

    2016-10-01

    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 criticality 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.

  9. Maintaining a Critical Spectra within Monteburns for a Gas-Cooled Reactor Array by Way of Control Rod Manipulation

    SciTech Connect

    Adigun, Babatunde John; Fensin, Michael Lorne; Galloway, Jack D.; Trellue, Holly Renee

    2016-10-01

    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 criticality 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.

  10. Review of the state of criticality of the Three Mile Island Unit 2 core and reactor vessel

    SciTech Connect

    Stratton, W.R. )

    1987-04-15

    The events during the early hours of the Three Mile Island Unit 2 (TMI-2) accident on March 28, 1979 caused the fuel in the reactor core to crumble or disintegrate, and then subside into a rubble structure more compact that its normal configuration. The present height of the core is about seven feet, five feet less than its normal configuration of 12 feet. With the same boron content and some or all of the control rod and burnable poison rod material as the normal core configuration, the collapsed structure is calculated to be more reactive. However, the reactor is assuredly subcritical at present because of the extraordinarily high boron concentration maintained in the coolant water. Four additional and different physical models are discussed briefly in the report to illustrate the margin of subcriticality, to provide a better estimate of the neutron multiplication factor, and to provide some understanding of the criticality effects of the important parameters. Two different finite, cylindrical models of a collapsed core are also presented in this report. The conclusion of this review is that the reactor is now very far subcritical with a boron concentration of 4350 ppM or more, and no conceivable rearrangement of fuel can create a critical state. Careful administrative control to maintain the boron concentration of the reactor coolant close to 5000 ppM, and controls to rigorously exclude addition of unborated water to the primary system, provide additional assurance that subcriticality will be maintained. The immediate corollary is that the defueling of the reactor vessel can proceed as planned, with complete confidence that such operations will remain subcritical. 20 refs.

  11. 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

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Notice of License Termination for University of Illinois Advanced TRIGA Reactor, License No. R-115 The U.S. Nuclear Regulatory Commission (NRC) is noticing the termination of Facility Operating...

  12. 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.

  13. Formation of clusters composed of C60 molecules via self-assembly in critical fluids

    NASA Astrophysics Data System (ADS)

    Fukuda, Takahiro; Ishii, Koji; Kurosu, Shunji; Whitby, Raymond; Maekawa, Toru

    2007-04-01

    Fullerenes are promising candidates for intelligent, functional nanomaterials because of their unique mechanical, electronic and chemical properties. However, it is necessary to invent some efficient but relatively simple methods of producing structures composed of fullerenes for the development of nanomechatronic, nanoelectronic and biochemical devices and sensors. In this paper, we show that various structures such as straight fibres, networks formed by fibres, wide sheets and helical structures, which are composed of C60 molecules, are created by placing C60-crystals in critical ethane, carbon dioxide and xenon even though C60 molecules do not dissolve or disperse in the above fluids. It is supposed, judging by the intermolecular potentials between C60 and C60, between C60 and ethane, and between ethane and ethane, that C60-clusters grow with the assistance of solvent molecules, which are trapped between C60 molecules under critical conditions. This room-temperature self-assembly cluster growth process in critical fluids may open up a new methodology of forming structures built up with fullerenes without the need for any ultra-fine processing technologies.

  14. 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.

  15. Maintaining a Critical Spectra within Monteburns for a Gas-Cooled Reactor Array by Way of Control Rod Manipulation

    DOE PAGES

    Adigun, Babatunde John; Fensin, Michael Lorne; Galloway, Jack D.; ...

    2016-10-01

    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

  16. DOE/NE robotics for advanced reactors

    SciTech Connect

    Not Available

    1991-01-01

    This bimonthly progress report for February through March, 1991, for the US Department of Energy Robotics for Advanced Reactors program contains information on research efforts in the following areas at the noted facilities: modeling of the Fuel Cycle Facility of the Integral Fast Reactor (IFR) program (University of Florida), assembly of infrared range sensors (University of Michigan), demonstration of the autonomous surveillance of a waste storage container for radioactive surface contamination (Oak Ridge National Laboratory), development of a simulation and animation environment for sensor-based robots (University of Tennessee), and machining and installation of the actuator test stand and associated fixtures and adapters (University of Texas). (MHB)

  17. 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…

  18. 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.

  19. Critical Influence of Cosolutes and Surfaces on the Assembly of Serpin-Derived Amyloid Fibrils.

    PubMed

    Risør, Michael W; Juhl, Dennis W; Bjerring, Morten; Mathiesen, Joachim; Enghild, Jan J; Nielsen, Niels C; Otzen, Daniel E

    2017-08-08

    Many proteins and peptides self-associate into highly ordered and structurally similar amyloid cross-β aggregates. This fibrillation is critically dependent on properties of the protein and the surrounding environment that alter kinetic and thermodynamic equilibria. Here, we report on dominating surface and solution effects on the fibrillogenic behavior and amyloid assembly of the C-36 peptide, a circulating bioactive peptide from the α1-antitrypsin serine protease inhibitor. C-36 converts from an unstructured peptide to mature amyloid twisted-ribbon fibrils over a few hours when incubated on polystyrene plates under physiological conditions through a pathway dominated by surface-enhanced nucleation. In contrast, in plates with nonbinding surfaces, slow bulk nucleation takes precedence over surface catalysis and leads to fibrillar polymorphism. Fibrillation is strongly ion-sensitive, underlining the interplay between hydrophilic and hydrophobic forces in molecular self-assembly. The addition of exogenous surfaces in the form of silica glass beads and polyanionic heparin molecules potently seeds the amyloid conversion process. In particular, heparin acts as an interacting template that rapidly forces β-sheet aggregation of C-36 to distinct amyloid species within minutes and leads to a more homogeneous fibril population according to solid-state NMR analysis. Heparin's template effect highlights its role in amyloid seeding and homogeneous self-assembly, which applies both in vitro and in vivo, where glycosaminoglycans are strongly associated with amyloid deposits. Our study illustrates the versatile thermodynamic landscape of amyloid formation and highlights how different experimental conditions direct C-36 into distinct macromolecular structures. Copyright © 2017 Biophysical Society. Published by Elsevier Inc. All rights reserved.

  20. ENDF/B-VII.0 Data Testing for Three Fast Critical Assemblies

    SciTech Connect

    Cullen, D E; Blomquist, R N; Brown, P N; Dean, C J; Dunn, M E; Lee, Y; Lent, E; MacFarlane, R; McKinley, S; Plechaty, E F; Sublet, J C

    2007-07-27

    In this report we consider three fast critical assemblies, each assembly is dominated by a different nuclear fuel: Godiva (U235), Jezebel (Pu239) and Jezebel23 (U233) [1]. We first show the improvement in results when using the new ENDF/B-VII.0 data [2], rather than the older, now frozen, ENDF/B-VI.8 data [3]. We do this using what we call a one code/ multiple library approach, where results from one code (MCNP) are compared using two different data libraries (ENDF/B-VII.0 and VI.8). Next we show that MCNP results are not specific to this one code by using what we call a one data library/multiple code approach; for this purpose we invited many codes to submit results using the ENDF/B-VII.0 data; the most detailed results presented in this report compare MCNP and TART. The bottom line is that we have shown that using the new ENDF/B-VII.0 data library with a variety of transport codes, for the first time we are able to reproduce the expected K-eff values for all three assemblies to within the quoted accuracy of the models, namely 1.0 +/- 0.001. This is a BIG improvement compared to the results obtained using the older ENDF/B-VI.8 data library. Another important result of this study is that we have demonstrated that currently there are many computer codes that can accurately use the new ENDF/B-VII.0 data.

  1. 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.

  2. 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.

  3. Studying the vibration and random hydrodynamic loads on the fuel rods bundles in the fuel assemblies of the reactor installations used at nuclear power stations equipped with VVER reactors

    NASA Astrophysics Data System (ADS)

    Solonin, V. I.; Perevezentsev, V. V.

    2012-05-01

    Random hydrodynamic loads causing vibration of fuel rod bundles in a turbulent flow of coolant are obtained from the results of pressure pulsation measurements carried out over the perimeter of the external row of fuel rods in the bundle of a full-scale mockup of a fuel assembly used in a second-generation VVER-440 reactor. It is shown that the turbulent flow structure is a factor determining the parameters of random hydrodynamic loads and the vibration of fuel rod bundles excited by these loads. The results from a calculation of random hydrodynamic loads are used for estimating the vibration levels of fuel rod bundles used in prospective designs of fuel assemblies for VVER reactors.

  4. 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.

  5. 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.

  6. 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

  7. In-cell reaction rate distributions and cell-average reaction rates in fast critical assemblies

    SciTech Connect

    Brumbach, S.B.; Gasidlo, J.M.

    1985-08-01

    Measurements are described for determining average values of fission rates in /sup 235/U, /sup 238/U and /sup 239/Pu and capture rates in /sup 238/U for heterogeneous cells used to construct fast critical assemblies. The measurements are based on irradiations of foils of /sup 238/U, /sup 235/U and /sup 239/Pu with counting of fission and capture products using gamma-ray spectroscopy. Both plate and pin cells are considered. Procedures are described for inferring cell-average reaction rate values from a single foil location based on a cell using a quantity called a cell factor. Cell factors are determined from special measurements in which several foils are irradiated within a cell. Comparisons are presented between cell factors determined by measurements and by Monte Carlo calculations which lend credibility to the measurement procedures.

  8. License renewal and power upgrade of the Cornell University TRIGA reactor

    SciTech Connect

    Aderhold, Howard C.

    1984-07-01

    The Cornell Mark II TRIGA reactor has been a principal facility for instruction and research in nuclear science and engineering at Cornell, and it has been extensively used by other departments at Cornell and by nearby universities and industries. Initially the fuel was low hydride, 8.5w/o 19%-enriched, aluminum clad; in 1974 it was changed to high-hydride, stainless-steel-clad. The maximum power has been 100 kW, with pulses to $2, and operation has been on a one-shift demand basis. Annual energy generation of 50 MWH has been typical. Standard features include a 4-inch tangential port and our 6-inch radial ports, a thermal column with hohlraum and vertical access, a central thimble, a 'rabbit', and a set of dry irradiation tubes, replacing the 'Lazy Susan'. The license was renewed and amended in November 1983; the new limits are 500 kW and $3 pulses. Physical changes to the facility included addition of a water-to-water heat exchanger and of a diffuser at the water outlet {approx} 60 cm above the core. The flow rate is 300 liters per minute in the primary (reactor) side of the heat exchanger. The temperature of the chilled water entering the secondary of the exchanger is {approx} 12?C; its flow rate is adjusted by a servo-controlled by-pass valve to maintain the desired range of pool water temperature. Steps taken to go to higher power included rearrangement of fuel elements to increase excess reactivity, recalibration of control rods, and power vs ion chamber current calibrations at successively higher power by comparing the rate of rise of pool temperature with a known rate using electrical heating elements. Steady-state operation has been done up to 480 kW (nominal) but pulsing at the newly allowed higher levels has not been tested as yet.

  9. 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…

  10. 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…

  11. 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…

  12. Nuclear reactors built, being built, or planned, 1994

    SciTech Connect

    1995-07-01

    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, 1994. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company -- working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an 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).

  13. Nuclear reactors built, being built, or planned: 1995

    SciTech Connect

    1996-08-01

    This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an 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).

  14. 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.

  15. In Brief: Kyoto Protocol moves forward

    NASA Astrophysics Data System (ADS)

    Showstack, Randy

    2004-10-01

    The Russian cabinet's 30 September endorsement of the Kyoto Protocol to the United Nations Framework Convention on Climate Change (UNFCCC) likely clears the way for the treaty's ratification by that country's parliament and for its entry into force. The protocol enters into force when not less than 55 Parties to the Convention, including industrialized countries (so called ``Annex I Parties'') which accounted in total for at least 55 % of the total carbon dioxide emissions for 1990 from that group, officially have agreed to the treaty.

  16. Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

    NASA Astrophysics Data System (ADS)

    Stacey, W. M.

    2009-09-01

    The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

  17. International ranking systems for universities and institutions: a critical appraisal

    PubMed Central

    Ioannidis, John PA; Patsopoulos, Nikolaos A; Kavvoura, Fotini K; Tatsioni, Athina; Evangelou, Evangelos; Kouri, Ioanna; Contopoulos-Ioannidis, Despina G; Liberopoulos, George

    2007-01-01

    Background Ranking of universities and institutions has attracted wide attention recently. Several systems have been proposed that attempt to rank academic institutions worldwide. Methods We review the two most publicly visible ranking systems, the Shanghai Jiao Tong University 'Academic Ranking of World Universities' and the Times Higher Education Supplement 'World University Rankings' and also briefly review other ranking systems that use different criteria. We assess the construct validity for educational and research excellence and the measurement validity of each of the proposed ranking criteria, and try to identify generic challenges in international ranking of universities and institutions. Results None of the reviewed criteria for international ranking seems to have very good construct validity for both educational and research excellence, and most don't have very good construct validity even for just one of these two aspects of excellence. Measurement error for many items is also considerable or is not possible to determine due to lack of publication of the relevant data and methodology details. The concordance between the 2006 rankings by Shanghai and Times is modest at best, with only 133 universities shared in their top 200 lists. The examination of the existing international ranking systems suggests that generic challenges include adjustment for institutional size, definition of institutions, implications of average measurements of excellence versus measurements of extremes, adjustments for scientific field, time frame of measurement and allocation of credit for excellence. Conclusion Naïve lists of international institutional rankings that do not address these fundamental challenges with transparent methods are misleading and should be abandoned. We make some suggestions on how focused and standardized evaluations of excellence could be improved and placed in proper context. PMID:17961208

  18. 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.

  19. 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.

  20. Reactor performances and microbial communities of biogas reactors: effects of inoculum sources.

    PubMed

    Han, Sheng; Liu, Yafeng; Zhang, Shicheng; Luo, Gang

    2016-01-01

    Anaerobic digestion is a very complex process that is mediated by various microorganisms, and the understanding of the microbial community assembly and its corresponding function is critical in order to better control the anaerobic process. The present study investigated the effect of different inocula on the microbial community assembly in biogas reactors treating cellulose with various inocula, and three parallel biogas reactors with the same inoculum were also operated in order to reveal the reproducibility of both microbial communities and functions of the biogas reactors. The results showed that the biogas production, volatile fatty acid (VFA) concentrations, and pH were different for the biogas reactors with different inocula, and different steady-state microbial community patterns were also obtained in different biogas reactors as reflected by Bray-Curtis similarity matrices and taxonomic classification. It indicated that inoculum played an important role in shaping the microbial communities of biogas reactor in the present study, and the microbial community assembly in biogas reactor did not follow the niche-based ecology theory. Furthermore, it was found that the microbial communities and reactor performances of parallel biogas reactors with the same inoculum were different, which could be explained by the neutral-based ecology theory and stochastic factors should played important roles in the microbial community assembly in the biogas reactors. The Bray-Curtis similarity matrices analysis suggested that inoculum affected more on the microbial community assembly compared to stochastic factors, since the samples with different inocula had lower similarity (10-20 %) compared to the samples from the parallel biogas reactors (30 %).

  1. Comparison of measured and calculated concrete and rebar specific activity during decommissioning of the Dalhousie SLOWPOKE-2 reactor.

    PubMed

    Smith, Philip G; Everall, Dave; Ariani, Imelda; Tsang, Kwok

    2013-07-01

    Following the defuelling and dismantling of the Dalhousie University SLOWPOKE-2 Reactor (DUSR) in 2011, the reactor pool concrete and rebar were sampled to support the unconditional free release of the material such that the facility could be classified for unrestricted use. A detailed MCNP5 model of the critical core assembly was simulated to calculate the thermal, intermediate and fast neutron flux profile below the reactor pool floor. The neutron fluxes were used to calculate the specific activity of significant radionuclides in the concrete and rebar. The calculated specific activity and consequently the calculated neutron fluxes were validated at a number of sample locations. The calculated concrete and rebar specific activity were found to be in good agreement with the measured specific activity at the sample locations. The unrestricted use of the facility was granted through the approval of the licence to abandon the facility in August 2011.

  2. Universal crossover from ground-state to excited-state quantum criticality

    NASA Astrophysics Data System (ADS)

    Kang, Byungmin; Potter, Andrew C.; Vasseur, Romain

    2017-01-01

    We study the nonequilibrium properties of a nonergodic random quantum chain in which highly excited eigenstates exhibit critical properties usually associated with quantum critical ground states. The ground state and excited states of this system belong to different universality classes, characterized by infinite-randomness quantum critical behavior. Using strong-disorder renormalization group techniques, we show that the crossover between the zero and finite energy density regimes is universal. We analytically derive a flow equation describing the unitary dynamics of this isolated system at finite energy density from which we obtain universal scaling functions along the crossover.

  3. 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…

  4. 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…

  5. 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…

  6. 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…

  7. A Theory of Argumentation, Rhetorical Criticism and the Universal Audience.

    ERIC Educational Resources Information Center

    Rohrer, Daniel Morgan

    Emphasizing the need for sound logic in the decision making and policy making process, this paper equates the concept of rationality with the universal audience as a means of analyzing argument, evaluating rhetoric, and persuading audiences. The paper argues that the policy systems paradigm most approximates this objective within the context of…

  8. 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…

  9. 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…

  10. 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…

  11. 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…

  12. Review of critical incidents in a university department of anaesthesia.

    PubMed

    Saito, T; Wong, Z W; Thinn, K K; Poon, K H; Liu, E

    2015-03-01

    In 2011, our hospital started a new system of 100% procedural audit of anaesthesia work, in which we incorporated the reporting of critical incidents. This monitoring of critical incidents has enabled identification of the spectrum of incidents and risk factors and helped in the education of trainees and specialists. In this review, we analyse 379 incidents that had been reported among 44,915 anaesthetics administered in a two-year period. The risk of incidents was higher in patients of lower American Society of Anesthesiologists physical status, anaesthesia of long duration and anaesthesia carried out after-hours. The most common incidents were airway problems and drug administration problems. Fifty-nine percent of incidents were evaluated to be preventable and adverse outcomes occurred in 48% of cases. Human factors were the major contributors to incidents. We suggest that incorporating critical incident reporting as part of a 100% procedural audit facilitated, rather than discouraged, the reporting of critical incidents, even though reporting was not anonymous. The rate of incident reporting increased from 0.37% to 0.84%.

  13. Monitoring and Control Research Using a University Reactor and SBWR Test-Loop

    SciTech Connect

    Robert M. Edwards

    2003-09-28

    The existing hybrid simulation capability of the Penn State Breazeale nuclear reactor was expanded to conduct research for monitoring, operations and control. Hybrid simulation in this context refers to the use of the physical time response of the research reactor as an input signal to a real-time simulation of power-reactor thermal-hydraulics which in-turn provides a feedback signal to the reactor through positioning of an experimental changeable reactivity device. An ECRD is an aluminum tube containing an absorber material that is positioned in the central themble of the reactor kinetics were used to expand the hybrid reactor simulation (HRS) capability to include out-of-phase stability characteristics observed in operating BWRs.

  14. Studying critical values: adverse event identification following a critical laboratory values study at the ohio state university medical center.

    PubMed

    Jenkins, James J; Mac Crawford, J; Bissell, Michael G

    2007-10-01

    No study to date has used laboratory critical values to evaluate variations in patient adverse events. We retrospectively analyzed a database of critical values to determine their distribution by hospital unit over time. The data were drawn from the Ohio State University Medical Center Information Warehouse (Columbus) for a 58-month period. Critical values were plotted over time on statistical control charts and analyzed for unusual peaks in monthly occurrence rates. Chart review of individual patient results yielded several predictor variables for the unusual peaks. Of these, occurrence of patient adverse events was the most relevant independent predictor variable for a month with an unusual number of critical values vs a normal month. This result epidemiologically confirms the basic premise of critical value reporting and suggests that the control-chart method of this type could be a new statistical tool to compare clinical activity of different hospital locations at different times.

  15. Development and implementation of an automatic control algorithm for the University of Utah nuclear reactor

    SciTech Connect

    Crawford, Kevan C.; Sandquist, Gary M.

    1990-07-01

    The emphasis of this work is the development and implementation of an automatic control philosophy which uses the classical operational philosophies as a foundation. Three control algorithms were derived based on various simplifying assumptions. Two of the algorithms were tested in computer simulations. After realizing the insensitivity of the system to the simplifications, the most reduced form of the algorithms was implemented on the computer control system at the University of Utah (UNEL). Since the operational philosophies have a higher priority than automatic control, they determine when automatic control may be utilized. Unlike the operational philosophies, automatic control is not concerned with component failures. The object of this philosophy is the movement of absorber rods to produce a requested power. When the current power level is compared to the requested power level, an error may be detected which will require the movement of a control rod to correct the error. The automatic control philosophy adds another dimension to the classical operational philosophies. Using this philosophy, normal operator interactions with the computer would be limited only to run parameters such as power, period, and run time. This eliminates subjective judgements, objective judgements under pressure, and distractions to the operator and insures the reactor will be operated in a safe and controlled manner as well as providing reproducible operations.

  16. 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.

  17. Argon-41 production and evolution at the Oregon State University TRIGA Reactor (OSTR)

    SciTech Connect

    Anellis, L.G.; Johnson, A.G.; Higginbotham, J.F.

    1988-07-01

    In this study, argon-41 concentrations were measured at various locations within the reactor facility to assess the accuracy of models used to predict argon-41 evolution from the reactor tank, and to determine the relationship between argon gas evolution from the tank and subsequent argon-41 concentrations throughout the reactor room. In particular, argon-41 was measured directly above the reactor tank with the reactor tank lids closed, at other accessible locations on the reactor top with the tank lids both closed and open, and at several locations on the first floor of the reactor room. These measured concentrations were then compared to values calculated using a modified argon-41 production and evolution model for TRIGA reactor tanks and ventilation values applicable to the OSTR facility. The modified model was based in part on earlier TRIGA models for argon-41 production and release, but added features which improved the agreement between predicted and measured values. The approximate dose equivalent rate due to the presence of argon-41 in reactor room air was calculated for several different locations inside the OSTR facility. These dose rates were determined using the argon-41 concentration measured at each specific location, and were subsequently converted to a predicted quarterly dose equivalent for each location based on the reactor's operating history. The predicted quarterly dose equivalent values were then compared to quarterly doses measured by film badges deployed as dose-integrating area radiation monitors at the locations of interest. The results indicate that the modified production and evolution model is able to predict argon-41 concentrations to within a factor of ten when compared to the measured data. Quarterly dose equivalents calculated from the measured argon-41 concentrations and the reactor's operating history seemed consistent with results obtained from the integrating area radiation monitors. Given the argon-41 concentrations measured

  18. 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

  19. NUCLEAR REACTORS

    DOEpatents

    Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

    1961-12-01

    An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

  20. Critical limits (alert values) for physician notification: universal or medical center specific limits?

    PubMed

    Lum, G

    1998-01-01

    The concept of critical limits (alert values), defined as an imminent life threatening laboratory result requiring immediate physician notification, has been widely adopted as a standard of good laboratory practice. Although virtually all laboratories have tests with critical limits, surveys have shown that there is no universal alert value list. Recently, nine VA medical centers in the New England region, which now constitute one consolidated entity, were surveyed with the objective of summarizing critical limits. Universal (100 percent) critical limit tests for clinical chemistry were: Calcium; mean low/high, 6.5/12.4 mg/dL: Glucose 48/432 mg/dL: Potassium 2.8/6.1 mmol/L: Sodium 121/159 mmol/L. Universal hematology tests included: Hematocrit 22.2/59.7 percent: Platelet count 61K/983K: white blood count 1.9K/29K. Although there was universal agreement that abnormal coagulation tests (PT, PTT) should be included on the hematology critical limit list, there was wide variation in the reporting of coagulation tests (seconds and INR) and patient therapeutic status (anticoagulant or no-anticoagulant). Universal alert values for microbiology were: Positive blood culture: Positive cerebral spinal fluid (CSF) culture: Positive CSF Gram stain. There was no universal agreement regarding critically high (potentially toxic) therapeutic drugs, with two medical centers declining to notify physicians of any abnormally high therapeutic drug level. No other qualitative critical limits for other laboratory sections, such as physician notification of an unexpected malignancy (surgical pathology) were universal. Medical center specific critical limits, designed to meet the clinical needs of each facility, are the norm in the nine medical centers. Laboratories do need periodically to review their critical limit lists with appropriate clinical input to avoid including critical limits for laboratory tests not required for urgent physician notification and patient evaluation and treatment.

  1. Behavior of universal critical parameters in the QCD phase diagram

    NASA Astrophysics Data System (ADS)

    Bluhm, Marcus; Nahrgang, Marlene; Bass, Steffen A.; Schäfer, Thomas

    2017-01-01

    We determine the dependence of important parameters for critical fluctuations on temperature and baryon chemical potential in the QCD phase diagram. The analysis is based on an identification of the fluctuations of the order parameter obtained from the Ising model equation of state and the Ginzburg-Landau effective potential approach. The impact of the mapping from Ising model variables to QCD thermodynamics is discussed.

  2. Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics

    SciTech Connect

    Stillman, J.; Feldman, E.; Foyto, L; Kutikkad, K; McKibben, J C; Peters, N.; Stevens, J.

    2012-09-01

    This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.

  3. Optimization study for an epithermal neutron beam for boron neutron capture therapy at the University of Virginia Research Reactor

    SciTech Connect

    Burns, Jr., Thomas Dean

    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 x 108 n/cm2 • s. The fast neutron and gamma radiation KERMA factors are 10 x 10-11cGy•cm2/nepi and 20 x 10-11 cGy•cm2/nepi , 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.

  4. Impact of selected non-steroidal anti-inflammatory pharmaceuticals on microbial community assembly and activity in sequencing batch reactors.

    PubMed

    Jiang, Cong; Geng, Jinju; Hu, Haidong; Ma, Haijun; Gao, Xingsheng; Ren, Hongqiang

    2017-01-01

    This study covers three widely detected non-steroidal anti-inflammatory pharmaceuticals (NSAIDs), diclofenac (DCF), ibuprofen (IBP) and naproxen (NPX), as NSAIDs pollutants. The objective is to evaluate the impact of NSAIDs at their environmental concentrations on microbial community assembly and activity. The exposure experiments were conducted under three conditions (5 μg L-1 DCF, 5 μg L-1 DCF+5 μg L-1 IBP and 5 μg L-1 DCF+5 μg L-1 IBP+ 5 μg L-1 NPX) in sequencing batch reactors (SBRs) for 130 days. Removals of COD and NH4+-N were not affected but total nitrogen (TN) removal decreased. IBP and NPX had the high removal efficiencies (79.96% to 85.64%), whereas DCF was more persistent (57.24% to 64.12%). In addition, the decreased removals of TN remained the same under the three conditions (p > 0.05). The results of oxidizing enzyme activities, live cell percentages and extracellular polymeric substances (EPS) indicated that NSAIDs damaged the cell walls or microorganisms and the mixtures of the three NSAIDs increased the toxicity. The increased Shannon-Wiener diversity index suggested that bacterial diversity was increased with the addition of selected NSAIDs. Bacterial ribosomal RNA small subunit (16S) gene sequencing results indicated that Actinobacteria and Bacteroidetes were enriched, while Micropruina and Nakamurella decreased with the addition of NSAIDs. The enrichment of Actinobacteria and Bacteroidetes indicated that both of them might have the ability to degrade NSAIDs and thereby could adapt well with the presence of NSAIDs.

  5. Impact of selected non-steroidal anti-inflammatory pharmaceuticals on microbial community assembly and activity in sequencing batch reactors

    PubMed Central

    Jiang, Cong; Hu, Haidong; Ma, Haijun; Gao, Xingsheng; Ren, Hongqiang

    2017-01-01

    This study covers three widely detected non-steroidal anti-inflammatory pharmaceuticals (NSAIDs), diclofenac (DCF), ibuprofen (IBP) and naproxen (NPX), as NSAIDs pollutants. The objective is to evaluate the impact of NSAIDs at their environmental concentrations on microbial community assembly and activity. The exposure experiments were conducted under three conditions (5 μg L-1 DCF, 5 μg L-1 DCF+5 μg L-1 IBP and 5 μg L-1 DCF+5 μg L-1 IBP+ 5 μg L-1 NPX) in sequencing batch reactors (SBRs) for 130 days. Removals of COD and NH4+-N were not affected but total nitrogen (TN) removal decreased. IBP and NPX had the high removal efficiencies (79.96% to 85.64%), whereas DCF was more persistent (57.24% to 64.12%). In addition, the decreased removals of TN remained the same under the three conditions (p > 0.05). The results of oxidizing enzyme activities, live cell percentages and extracellular polymeric substances (EPS) indicated that NSAIDs damaged the cell walls or microorganisms and the mixtures of the three NSAIDs increased the toxicity. The increased Shannon-Wiener diversity index suggested that bacterial diversity was increased with the addition of selected NSAIDs. Bacterial ribosomal RNA small subunit (16S) gene sequencing results indicated that Actinobacteria and Bacteroidetes were enriched, while Micropruina and Nakamurella decreased with the addition of NSAIDs. The enrichment of Actinobacteria and Bacteroidetes indicated that both of them might have the ability to degrade NSAIDs and thereby could adapt well with the presence of NSAIDs. PMID:28640897

  6. Criticality in a Vlasov-Poisson system: a fermioniclike universality class.

    PubMed

    Ivanov, A V; Vladimirov, S V; Robinson, P A

    2005-05-01

    A model Vlasov-Poisson system is simulated close to the point of marginal stability, thus assuming only the wave-particle resonant interactions are responsible for saturation, and shown to obey the power-law scaling of a second-order phase transition. The set of critical exponents analogous to those of the Ising universality class is calculated and shown to obey the Widom and Rushbrooke scaling and Josephson's hyperscaling relations at the formal dimensionality d=5 below the critical point at nonzero order parameter. However, the two-point correlation function does not correspond to the propagator of Euclidean quantum field theory, which is the Gaussian model for the Ising universality class. Instead, it corresponds to the propagator for the fermionic vector field and to the upper critical dimensionality d(c) = 2. This suggests criticality of collisionless Vlasov-Poisson systems corresponds to a universality class analogous to that of critical phenomena of a fermionic quantum field description.

  7. 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.

  8. 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.

  9. Critical evaluation of molybdenum and its alloys for use in space reactor core heat pipes

    SciTech Connect

    Lundberg, L.B.

    1981-01-01

    The choice of pure molybdenum as the prime candidate material for space reactor core heat pipes is examined, and the advantages and disadvantages of this material are brought into focus. Even though pure molybdenum heat pipes have been built and tested, this metal's high ductile-brittle transition temperature and modest creep strength place significant design restrictions on a core heat pipe made from it. Molybdenum alloys are examined with regard to their promise as potential replacements for pure molybdenum. The properties of TZM and molybdenum-rhenium alloys are examined, and it appears that Mo-Re alloys with 10 to 15 wt % rhenium offer the most advantage as an alternative to pure molybdenum in space reactor core heat pipes.

  10. 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.

  11. Proper ciliary assembly is critical for restricting Hedgehog signaling during early eye development in mice.

    PubMed

    Burnett, Jacob B; Lupu, Floria I; Eggenschwiler, Jonathan T

    2017-10-01

    Patterning of the vertebrate eye into optic stalk, retinal pigment epithelium (RPE) and neural retina (NR) territories relies on a number of signaling pathways, but how these signals are interpreted by optic progenitors is not well understood. The primary cilium is a microtubule-based organelle that is essential for Hedgehog (Hh) signaling, but it has also been implicated in the regulation of other signaling pathways. Here, we show that the optic primordium is ciliated during early eye development and that ciliogenesis is essential for proper patterning and morphogenesis of the mouse eye. Ift172 mutants fail to generate primary cilia and exhibit patterning defects that resemble those of Gli3 mutants, suggesting that cilia are required to restrict Hh activity during eye formation. Ift122 mutants, which produce cilia with abnormal morphology, generate optic vesicles that fail to invaginate to produce the optic cup. These mutants also lack formation of the lens, RPE and NR. Such phenotypic features are accompanied by strong, ectopic Hh pathway activity, evidenced by altered gene expression patterns. Removal of GLI2 from Ift122 mutants rescued several aspects of optic cup and lens morphogenesis as well as RPE and NR specification. Collectively, our data suggest that proper assembly of primary cilia is critical for restricting the Hedgehog pathway during eye formation in the mouse. Copyright © 2017 Elsevier Inc. All rights reserved.

  12. Fast critical assembly safeguards: NDA methods for highly enriched uranium. Summary report, October 1978-September 1979

    SciTech Connect

    Bellinger, F.O.; Winslow, G.H.

    1980-12-01

    Nondestructive assay (NDA) methods, principally passive gamma measurements and active neutron interrogation, have been studied for their safeguards effectiveness and programmatic impact as tools for making inventories of highly enriched uranium fast critical assembly fuel plates. It was concluded that no NDA method is the sole answer to the safeguards problem, that each of those emphasized here has its place in an integrated safeguards system, and that each has minimum facility impact. It was found that the 185-keV area, as determined with a NaI detector, was independent of highly-enriched uranium (HEU) plate irradiation history, though the random neutron driver methods used here did not permit accurate assay of irradiated plates. Containment procedures most effective for accurate assaying were considered, and a particular geometry is recommended for active interrogation by a random driver. A model, pertinent to that geometry, which relates the effects of multiplication and self-absorption, is described. Probabilities of failing to detect that plates are missing are examined.

  13. Irradiation campaign in the EOLE critical facility of fiber optic Bragg gratings dedicated to the online temperature measurement in zero power research reactors

    SciTech Connect

    Mellier, Frederic; Cheymol, Guy; Destouches, Christophe; Di Salvo, Jacques; Laffont, Guillaume; Morana, Adriana; Girard, Sylvain; Marin, Emmanuel

    2015-07-01

    The control of temperature during operation of zero power research reactors participates to the overall control of experimentation conditions and reveals itself of a major importance more especially when measuring small multiplication factor variations. Within the framework of the refurbishment of the MASURCA facility, the development of a new temperature measurement system based on the optical fiber Bragg grating (FBG) technology is under consideration. In a first step, a series of FBGs is irradiated in the EOLE critical facility with the aim to select the most appropriate. Online temperature measurements are performed during a set of irradiations that should allow reaching a fast neutron fluence of some 10{sup 14} n.cm{sup -2}. The results obtained, more especially the Bragg wavelength shifts during the irradiation campaign, are discussed in this paper and compared to data from standard PT100 temperature sensors to highlight possible radiation effects on sensor performances. Work to be conducted during the second step of the project, aiming to a feasibility demonstration using a MASURCA assembly, is also presented. (authors)

  14. 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.

  15. CSRL-V ENDF/B-V 227-group neutron cross-section library and its application to thermal-reactor and criticality safety benchmarks

    SciTech Connect

    Ford, W.E. III; Diggs, B.R.; Knight, J.R.; Greene, N.M.; Petrie, L.M.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.; Williams, M.L.

    1982-01-01

    Characteristics and contents of the CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data) 227-neutron-group AMPX master and pointwise cross-section libraries are described. Results obtained in using CSRL-V to calculate performance parameters of selected thermal reactor and criticality safety benchmarks are discussed.

  16. Nuclear reactor

    DOEpatents

    Pennell, William E.; Rowan, William J.

    1977-01-01

    A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

  17. 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…

  18. 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…

  19. Critical Thinking, Community Service, and Participatory Research: Restructuring the American University for a Framework of Learning.

    ERIC Educational Resources Information Center

    Sapp, David Alan

    2002-01-01

    Slowly, some universities are beginning to shift towards a new paradigm, the "learning paradigm," that proposes that institutions of higher learning exist not to provide instruction but instead to produce learning. The learning paradigm focuses on teaching critical thinking skills as life skills. If a university follows the learning…

  20. 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…