Science.gov

Sample records for core neutronic design

  1. Advanced neutron source final preconceptual reference core design

    SciTech Connect

    Copeland, G.L.; Gambill, W.R.; Harrington, R.M.; Johnson, J.A.; Peretz, F.J.; Reutler, H.; Ryskamp, J.M.; Selby, D.L.; West, C.D.; Yoder, G.L.

    1989-08-01

    The preconceptual design phase of the Advanced Neutron Source (ANS) Project ended with the selection of a reference reactor core that will be used to begin conceptual design work. The new reference core consists of two involute fuel elements, of different diameters, aligned axially with a small axial gap between them. The use of different element diameters permits a separate flow of coolant to be provided for each one, thus enhancing the heat removal capability and increasing the thermal-hydraulic margins. The improved cooling allows the elements to be relatively long and thin, so self-shielding is reduced and an acceptable core life can be achieved with a relatively small loading of highly enriched uranium silicide fuel clad in aluminium. The new reference design has a fueled volume 67.4 L, each element having a heated length of 474 mm and a radial fuel thickness of 66 mm. The end-of-cycle peak thermal flux in the large heavy-water reflector tank around the core is estimated to be in the range of 0.8 to 1.0 /times/ 10/sup 20/ m/sup /minus/2/ /center dot/ s/sup /minus/1/. 7 refs., 23 figs., 15 tabs.

  2. McCARD for Neutronics Design and Analysis of Research Reactor Cores

    NASA Astrophysics Data System (ADS)

    Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang

    2014-06-01

    McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO, and YALINA.

  3. NEUTRONIC REACTOR CORE

    DOEpatents

    Thomson, W.B.; Corbin, A. Jr.

    1961-07-18

    An improved core for a gas-cooled power reactor which admits gas coolant at high temperatures while affording strong integral supporting structure and efficient moderation of neutrons is described. The multiplicities of fuel elements constituting the critical amassment of fissionable material are supported and confined by a matrix of metallic structure which is interspersed therebetween. Thermal insulation is interposed between substantially all of the metallic matrix and the fuel elements; the insulation then defines the principal conduit system for conducting the coolant gas in heat-transfer relationship with the fuel elements. The metallic matrix itseif comprises a system of ducts through which an externally-cooled hydrogeneous liquid, such as water, is circulated to serve as the principal neutron moderant for the core and conjointly as the principal coolant for the insulated metallic structure. In this way, use of substantially neutron transparent metals, such as aluminum, becomes possible for the supporting structure, despite the high temperatures of the proximate gas. The Aircraft Nuclear Propulsion program's "R-1" reactor design is a preferred embodiment.

  4. PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

    SciTech Connect

    Bi, G.; Liu, C.; Si, S.

    2012-07-01

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no

  5. NEUTRONIC REACTOR CORE INSTRUMENT

    DOEpatents

    Mims, L.S.

    1961-08-22

    A multi-purpose instrument for measuring neutron flux, coolant flow rate, and coolant temperature in a nuclear reactor is described. The device consists essentially of a hollow thimble containing a heat conducting element protruding from the inner wall, the element containing on its innermost end an amount of fissionsble materinl to function as a heat source when subjected to neutron flux irradiation. Thermocouple type temperature sensing means are placed on the heat conducting element adjacent the fissionable material and at a point spaced therefrom, and at a point on the thimble which is in contact with the coolant fluid. The temperature differentials measured between the thermocouples are determinative of the neutron flux, coolant flow, and temperature being measured. The device may be utilized as a probe or may be incorporated in a reactor core. (AE C)

  6. Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.; Harrington, R.M.

    1996-02-01

    In order to utilize reduced enrichment fuel, the three-element-core design for the Advanced Neutron Source has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. To assess the impact of changes in the core region configuration and the thermal-hydraulic steady-state conditions, the safety analysis has been updated. This report gives the safety margins for the loss-of-off-site power and pressure-boundary fault accidents based on the RELAP5 results. AU margins are greater for the three-element-core simulations than those calculated for the two-element core.

  7. Core Vessel Insert Handling Robot for the Spallation Neutron Source

    SciTech Connect

    Graves, Van B; Dayton, Michael J

    2011-01-01

    The Spallation Neutron Source provides the world's most intense pulsed neutron beams for scientific research and industrial development. Its eighteen neutron beam lines will eventually support up to twenty-four simultaneous experiments. Each beam line consists of various optical components which guide the neutrons to a particular instrument. The optical components nearest the neutron moderators are the core vessel inserts. Located approximately 9 m below the high bay floor, these inserts are bolted to the core vessel chamber and are part of the vacuum boundary. They are in a highly radioactive environment and must periodically be replaced. During initial SNS construction, four of the beam lines received Core Vessel Insert plugs rather than functional inserts. Remote replacement of the first Core Vessel Insert plug was recently completed using several pieces of custom-designed tooling, including a highly complicated Core Vessel Insert Robot. The design of this tool are discussed.

  8. NUCLEAR REACTOR CORE DESIGN

    DOEpatents

    Mahlmeister, J.E.; Peck, W.S.; Haberer, W.V.; Williams, A.C.

    1960-03-22

    An improved core design for a sodium-cooled, graphitemoderated nuclear reactor is described. The improved reactor core comprises a number of blocks of moderator material, each block being in the shape of a regular prism. A number of channels, extending the length of each block, are disposed around the periphery. When several blocks are placed in contact to form the reactor core, the channels in adjacent blocks correspond with each other to form closed conduits extending the length of the core. Fuel element clusters are disposed in these closed conduits, and liquid coolant is forced through the annulus between the fuel cluster and the inner surface of the conduit. In a preferred embodiment of the invention, the moderator blocks are in the form of hexagonal prisms with longitudinal channels cut into the corners of the hexagon. The main advantage of an "edge-loaded" moderator block is that fewer thermal neutrons are absorbed by the moderator cladding, as compared with a conventional centrally loaded moderator block.

  9. Modeling and design of a new core-moderator assembly and neutron beam ports for the Penn State Breazeale Nuclear Reactor (PSBR)

    NASA Astrophysics Data System (ADS)

    Ucar, Dundar

    This study is for modeling and designing a new reactor core-moderator assembly and new neutron beam ports that aimed to expand utilization of a new beam hall of the Penn State Breazeale Reactor (PSBR). The PSBR is a part of the Radiation Science and Engineering Facility (RSEC) and is a TRIGA MARK III type research reactor with a movable core placed in a large pool and is capable to produce 1MW output. This reactor is a pool-type reactor with pulsing capability up to 2000 MW for 10-20 msec. There are seven beam ports currently installed to the reactor. The PSBR's existing core design limits the experimental capability of the facility, as only two of the seven available neutron beam ports are usable. The finalized design features an optimized result in light of the data obtained from neutronic and thermal-hydraulics analyses as well as geometrical constraints. A new core-moderator assembly was introduced to overcome the limitations of the existing PSBR design, specifically maximizing number of available neutron beam ports and mitigating the hydrogen gamma contamination of the neutron beam channeled in the beam ports. A crescent-shaped moderator is favored in the new PSBR design since it enables simultaneous use of five new neutron beam ports in the facility. Furthermore, the crescent shape sanctions a coupling of the core and moderator, which reduces the hydrogen gamma contamination significantly in the new beam ports. A coupled MURE and MCNP5 code optimization analysis was performed to calculate the optimum design parameters for the new PSBR. Thermal-hydraulics analysis of the new design was achieved using ANSYS Fluent CFD code. In the current form, the PSBR is cooled by natural convection of the pool water. The driving force for the natural circulation of the fluid is the heat generation within the fuel rods. The convective heat data was generated at the reactor's different operating powers by using TRIGSIMS, the fuel management code of the PSBR core. In the CFD

  10. Neutronic Reactor Design to Reduce Neutron Loss

    DOEpatents

    Miles, F. T.

    1961-05-01

    A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall. The wall is surrounded by successive layers of pure fertile material and moderator containing fertile material. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. Since the steel has a smaller capture cross section for the fast neutrons, greater nunnbers of neutrons will pass into the blanket, thereby increasing the over-all efficiency of the reactor. (AEC)

  11. NEUTRONIC REACTOR DESIGN TO REDUCE NEUTRON LOSS

    DOEpatents

    Mills, F.T.

    1961-05-01

    A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall which is surrounded by successive layers of pure fertile material and fertile material having moderator. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. As the steel has a smaller capture cross-section for the fast neutrons, then greater numbers of the neutrons will pass into the blanket thereby increasing the over-all efficiency of the reactor.

  12. ATW neutronics design studies.

    SciTech Connect

    Wade, D. C.; Yang, W. S.; Khalil, H.

    2000-11-10

    The Accelerator Transmutation of Waste (ATW) concept has been proposed as a transuranics (TRU) (and long-lived fission product) incinerator for processing the 87,000 metric tonnes of Light Water Reactor used fuel which will have been generated by the time the currently deployed fleet of commercial reactors in the US reach the end of their licensed lifetime. The ATW is proposed to separate the uranium from the transuranics and fission products in the LWR used fuel, to fission the transuranics, to send the LWR and ATW generated fission products to the geologic repository and to send the uranium to either a low level waste disposal site or to save it for future use. The heat liberated in fissioning the transuranics would be converted to electricity and sold to partially offset the cost of ATW construction and operations. Options for incineration of long-lived fission products are under evaluation. A six-year science-based program of ATW trade and system studies was initiated in the US FY 2000 to achieve two main purposes: (1) ''to evaluate ATW within the framework of nonproliferation, waste management, and economic considerations,'' and (2) ''to evaluate the efficacy of the numerous technical options for ATW system configuration.'' This paper summarizes the results from neutronics and thermal/hydraulics trade studies which were completed at Argonne National Laboratory during the first year of the program. Core designs were developed for Pb-Bi cooled and Na cooled 840 MW{sub th} fast spectrum transmuter designs employing recycle. Additionally, neutronics analyses were performed at Argonne for a He cooled 600 MW{sub th} hybrid thermal and fast core design proposed by General Atomics Co. which runs critical for 3/4 and subcritical for 1/4 of its four year once-thin burn cycle. The mass flows and the ultimate loss of transuranic isotopes to the waste stream per unit of heat generated during transmutation have been calculated on a consistent basis and are compared. (Long

  13. Core Design Applications

    1995-07-12

    CORD-2 is intended for core desigh applications of pressurized water reactors. The main objective was to assemble a core design system which could be used for simple calculations (such as frequently required for fuel management) as well as for accurate calculations (for example, core design after refueling).

  14. Pharmacokinetics of core-polymerized, boron-conjugated micelles designed for boron neutron capture therapy for cancer.

    PubMed

    Sumitani, Shogo; Oishi, Motoi; Yaguchi, Tatsuya; Murotani, Hiroki; Horiguchi, Yukichi; Suzuki, Minoru; Ono, Koji; Yanagie, Hironobu; Nagasaki, Yukio

    2012-05-01

    Core-polymerized and boron-conjugated micelles (PM micelles) were prepared by free radical copolymerization of a PEG-b-PLA block copolymer bearing an acetal group and a methacryloyl group (acetal-PEG-b-PLA-MA), with 1-(4-vinylbenzyl)-closo-carborane (VB-carborane), and the utility of these micelles as a tumor-targeted boron delivery system was investigated for boron neutron capture therapy (BNCT). Non-polymerized micelles (NPM micelles) that incorporated VB-carborane physically showed significant leakage of VB-carborane (ca. 50%) after 12 h incubation with 10% fetal bovine serum (FBS) at 37 °C. On the other hand, no leakage from the PM micelles was observed even after 48 h of incubation. To clarify the pharmacokinetics of the micelles, (125)I (radioisotope)-labeled PM and NPM micelles were administered to colon-26 tumor-bearing BALB/c mice. The (125)I-labeled PM micelles showed prolonged blood circulation (area under the concentration curve (AUC): 943.4) than the (125)I-labeled NPM micelles (AUC: 495.1), whereas tumor accumulation was similar for both types of micelles (AUC(PM micelle): 249.6, AUC(NPM micelle): 201.1). In contrast, the tumor accumulation of boron species in the PM micelles (AUC: 268.6) was 7-fold higher than the NPM micelles (AUC: 37.1), determined by ICP-AES. Thermal neutron irradiation yielded tumor growth suppression in the tumor-bearing mice treated with the PM micelles without reduction in body weight. On the basis of these data, the PM micelles represent a promising approach to the creation of boron carrier for BNCT.

  15. FAST FOSSIL ROTATION OF NEUTRON STAR CORES

    SciTech Connect

    Melatos, A.

    2012-12-10

    It is argued that the superfluid core of a neutron star super-rotates relative to the crust, because stratification prevents the core from responding to the electromagnetic braking torque, until the relevant dissipative (viscous or Eddington-Sweet) timescale, which can exceed {approx}10{sup 3} yr and is much longer than the Ekman timescale, has elapsed. Hence, in some young pulsars, the rotation of the core today is a fossil record of its rotation at birth, provided that magnetic crust-core coupling is inhibited, e.g., by buoyancy, field-line topology, or the presence of uncondensed neutral components in the superfluid. Persistent core super-rotation alters our picture of neutron stars in several ways, allowing for magnetic field generation by ongoing dynamo action and enhanced gravitational wave emission from hydrodynamic instabilities.

  16. NEUTRONIC REACTOR OPERATIONAL METHOD AND CORE SYSTEM

    DOEpatents

    Winters, C.E.; Graham, C.B.; Culver, J.S.; Wilson, R.H.

    1960-07-19

    Homogeneous neutronic reactor systems are described wherein an aqueous fuel solution is continuously circulated through a spherical core tank. The pumped fuel solution-is injected tangentially into the hollow spherical interior, thereby maintaining vigorous rotation of the solution within the tank in the form of a vortex; gaseous radiolytic decomposition products concentrate within the axial vortex cavity. The evolved gas is continuously discharged through a gas- outlet port registering with an extremity of the vortex cavity. and the solution stream is discharged through an annular liquid outlet port concentrically encircling the gas outlet by virtue of which the vortex and its cavity are maintained precisely axially aligned with the gas outlet. A primary heat exchanger extracts useful heat from the hot effluent fuel solution before its recirculation into the core tank. Hollow cylinders and other alternative core- tank configurations defining geometric volumes of revolution about a principal axis are also covered. AEC's Homogeneous Reactor Experiment No. 1 is a preferred embodiment.

  17. NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

    DOEpatents

    Moore, W.T.

    1958-09-01

    This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

  18. Dynamics of dissipative multifluid neutron star cores

    NASA Astrophysics Data System (ADS)

    Haskell, B.; Andersson, N.; Comer, G. L.

    2012-09-01

    We present a Newtonian multifluid formalism for superfluid neutron star cores, focusing on the additional dissipative terms which arise when one takes into account the individual dynamical degrees of freedom associated with the coupled “fluids.” The problem is of direct astrophysical interest as the nature of the dissipative terms can have significant impact on the damping of the various oscillation modes of the star and the associated gravitational-wave signatures. A particularly interesting application concerns the gravitational-wave driven instability of f- and r-modes. We apply the developed formalism to two specific three-fluid systems: (i) a hyperon core in which both Λ and Σ- hyperons are present and (ii) a core of deconfined quarks in the color-flavor-locked phase in which a population of neutral K0 kaons is present. The formalism is, however, general and can be applied to other problems in neutron-star dynamics (such as the effect of thermal excitations close to the superfluid transition temperature) as well as laboratory multifluid systems.

  19. Automated Core Design

    SciTech Connect

    Kobayashi, Yoko; Aiyoshi, Eitaro

    2005-07-15

    Multistate searching methods are a subfield of distributed artificial intelligence that aims to provide both principles for construction of complex systems involving multiple states and mechanisms for coordination of independent agents' actions. This paper proposes a multistate searching algorithm with reinforcement learning for the automatic core design of a boiling water reactor. The characteristics of this algorithm are that the coupling structure and the coupling operation suitable for the assigned problem are assumed and an optimal solution is obtained by mutual interference in multistate transitions using multiagents. Calculations in an actual plant confirmed that the proposed algorithm increased the convergence ability of the optimization process.

  20. Measurements of the thermal neutrons flux near the EAS core

    NASA Astrophysics Data System (ADS)

    Dzhappuev, D. D.; Volchenko, V. I.; Kudzhaev, A. U.; Mikhailova, O. I.; Petkov, V. B.; Stenkin, Yu V.; Tsyabuk, A. L.

    2013-02-01

    The characteristics of the thermal neutrons flux have been measured near the EAS core at the "Carpet-2" EAS array. The thermal neutron detectors were placed on the floor of the tunnel of the Muon Detector (MD) and showers with a core near the MD were selected. Thermal neutrons multiplicity spectrum has been obtained for these showers. Measurements of the lateral distribution function of thermal neutrons at distances of 1-16 m from the shower axis have been performed. The mean number of the recorded thermal neutrons as a function of the number of hadrons crossing the MD has been measured.

  1. Optical polarizing neutron devices designed for pulsed neutron sources

    SciTech Connect

    Takeda, M.; Kurahashi, K.; Endoh, Y.; Itoh, S.

    1997-09-01

    We have designed two polarizing neutron devices for pulsed cold neutrons. The devices have been tested at the pulsed neutron source at the Booster Synchrotron Utilization Facility of the National Laboratory for High Energy Physics. These two devices proved to have a practical use for experiments to investigate condensed matter physics using pulsed cold polarized neutrons.

  2. SmAHTR-CTC Neutronic Design

    SciTech Connect

    Ilas, Dan; Holcomb, David Eugene; Gehin, Jess C

    2014-01-01

    Building on prior experience for the 2010 initial SmAHTR neutronic design and on 2012 neutronic design for the advanced high temperature reactor (AHTR), this paper presents the main results of the neutronic design effort for the newly re-purposed SmAHTR-CTC reactor concept. The results are obtained based on full-core simulations performed with SCALE6.1. The dimensionality of the SmAHTR design space is reduced by using constraints originating in material fabricability, fuel licensing, molten salt chemistry, thermal-hydraulic and mechanical considerations. The new design represents in many regards a substantial improvement from the neutronic performance standpoint over the 2010 SmAHTR concept. Among other results, it is shown that fuel cycle length of over 2 years or discharged fuel burnup of 40GWd/MTHM are possible with a low, 8% fuel enrichment in a once-through fuel cycle, while 8-year once-through fuel cycle lengths are possible at higher fuel enrichments.

  3. Neutron flux and power in RTP core-15

    NASA Astrophysics Data System (ADS)

    Rabir, Mohamad Hairie; Zin, Muhammad Rawi Md; Usang, Mark Dennis; Bayar, Abi Muttaqin Jalal; Hamzah, Na'im Syauqi Bin

    2016-01-01

    PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core with literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.

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

    PubMed

    Lashkari, A; Khalafi, H; Kazeminejad, H

    2013-05-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

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

    PubMed

    Lashkari, A; Khalafi, H; Kazeminejad, H

    2013-05-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

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

    PubMed Central

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

    2013-01-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

  7. Simplified cut core inductor design

    NASA Technical Reports Server (NTRS)

    Mclyman, W. T.

    1974-01-01

    Although filter inductor designers have routinely tended to specify molypermalloy powder cores for use in high frequency power converters and pulse-width modulated switching regulators, there are sigificant advantages in specifying C cores and cut toroids fabricated from grain oriented silicon steels which should not be overlooked. Such steel cores can develop flux densities of 1.6 tesla, with useful linearity to 1.2 tesla, whereas molypermalloy cores carrying d.c. current have useful flux density capabilities only to about 0.3 tesla. The use of silicon steel cores thus makes it possible to design more compact cores, and therefore inductors of reduced volume, or conversely to provide greater load capacity in inductors of a given volume. Information is available which makes it possible to obtain quick and close approximations of significant parameters such as size, weight and temperature rise for silicon steel cores for breadboarding. Graphs, nomographs and tables are presented for this purpose, but more complete mathematical derivations of some of the important parameters are also included for a more rigorous treatment.

  8. Crust-core coupling in rotating neutron stars

    SciTech Connect

    Glampedakis, Kostas; Andersson, Nils

    2006-08-15

    Motivated by their gravitational wave driven instability, we investigate the influence of the crust on r-mode oscillations in a neutron star. Using a simplistic model of an elastic neutron star crust with constant shear modulus, we carry out an analytic calculation with the main objective of deriving an expression for the slippage between the core and the crust. Our analytic estimates support previous numerical results and provide useful insights into the details of the problem.

  9. Heat transfer and core neutronics considerations of the heat pipe cooled thermionic reactor

    NASA Astrophysics Data System (ADS)

    Determan, W. R.; Lewis, Brian

    The authors summarize the results of detailed neutronic and thermal-hydraulic evaluations of the heat pipe cooled thermionic (HPTI) reactor design, identify its key design attributes, and quantify its performance characteristics. The HPTI core uses modular, liquid-metal core heat transfer assemblies to replace the liquid-metal heat transport loop employed by in-core thermionic reactor designs of the past. The nuclear fuel, power conversion, heat transport, and heat rejection functions are all combined into a single modular unit. The reactor/converter assembly uses UN fuel pins to obtain a critical core configuration with in-core safety rods and reflector controls added to complete the subassembly. By thermally bonding the core heat transfer assemblies during the reactor core is coupled neutronically, thermally, and electrically into a modular assembly of individual power sources with cross-tied architecture. A forward-facing heat pipe radiator assembly extends from the reactor head in the shape of a frustum of a cone on the opposite side of the power system from the payload. Important virtues of the concept are the absence of any single-point failures and the ability of the core to effectively transfer the TFE waste heat load laterally to other in-core heat transfer assemblies in the event of multiple failures in either in-core and radiator heat pipes.

  10. Heat transfer and core neutronics considerations of the heat pipe cooled thermionic reactor

    NASA Technical Reports Server (NTRS)

    Determan, W. R.; Lewis, Brian

    1991-01-01

    The authors summarize the results of detailed neutronic and thermal-hydraulic evaluations of the heat pipe cooled thermionic (HPTI) reactor design, identify its key design attributes, and quantify its performance characteristics. The HPTI core uses modular, liquid-metal core heat transfer assemblies to replace the liquid-metal heat transport loop employed by in-core thermionic reactor designs of the past. The nuclear fuel, power conversion, heat transport, and heat rejection functions are all combined into a single modular unit. The reactor/converter assembly uses UN fuel pins to obtain a critical core configuration with in-core safety rods and reflector controls added to complete the subassembly. By thermally bonding the core heat transfer assemblies during the reactor core is coupled neutronically, thermally, and electrically into a modular assembly of individual power sources with cross-tied architecture. A forward-facing heat pipe radiator assembly extends from the reactor head in the shape of a frustum of a cone on the opposite side of the power system from the payload. Important virtues of the concept are the absence of any single-point failures and the ability of the core to effectively transfer the TFE waste heat load laterally to other in-core heat transfer assemblies in the event of multiple failures in either in-core and radiator heat pipes.

  11. MODULAR CORE UNITS FOR A NEUTRONIC REACTOR

    DOEpatents

    Gage, J.F. Jr.; Sherer, D.B.

    1964-04-01

    A modular core unit for use in a nuclear reactor is described. Many identical core modules can be placed next to each other to make up a complete core. Such a module includes a cylinder of moderator material surrounding a fuel- containing re-entrant coolant channel. The re-entrant channel provides for the circulation of coolant such as liquid sodium from one end of the core unit, through the fuel region, and back out through the same end as it entered. Thermal insulation surrounds the moderator exterior wall inducing heat to travel inwardly to the coolant channel. Spaces between units may be used to accommodate control rods and support structure, which may be cooled by a secondary gas coolant, independently of the main coolant. (AEC)

  12. Experiment Design and Analysis Guide - Neutronics & Physics

    SciTech Connect

    Misti A Lillo

    2014-06-01

    The purpose of this guide is to provide a consistent, standardized approach to performing neutronics/physics analysis for experiments inserted into the Advanced Test Reactor (ATR). This document provides neutronics/physics analysis guidance to support experiment design and analysis needs for experiments irradiated in the ATR. This guide addresses neutronics/physics analysis in support of experiment design, experiment safety, and experiment program objectives and goals. The intent of this guide is to provide a standardized approach for performing typical neutronics/physics analyses. Deviation from this guide is allowed provided that neutronics/physics analysis details are properly documented in an analysis report.

  13. Entrainment parameters in a cold superfluid neutron star core

    SciTech Connect

    Chamel, Nicolas; Haensel, Pawel

    2006-04-15

    Hydrodynamic simulations of neutron star cores that are based on a two-fluid description in terms of a neutron-proton superfluid mixture require the knowledge of the Andreev-Bashkin entrainment matrix which relates the momentum of one constituent to the currents of both constituents. This matrix is derived for arbitrary nuclear asymmetry at zero temperature and in the limits of small relative currents in the framework of the energy density functional theory. The Skyrme energy density functional is considered as a particular case. General analytic formulas for the entrainment parameters and various corresponding effective masses are obtained. These formulas are applied to the liquid core of a neutron star composed of homogeneous plasma of nucleons, electrons, and possibly muons in {beta} equilibrium.

  14. Magnetized neutron stars with superconducting cores: effect of entrainment

    NASA Astrophysics Data System (ADS)

    Palapanidis, K.; Stergioulas, N.; Lander, S. K.

    2015-09-01

    We construct equilibrium configurations of magnetized, two-fluid neutron stars using an iterative numerical method. Working in Newtonian framework we assume that the neutron star has two regions: the core, which is modelled as a two-component fluid consisting of type-II superconducting protons and superfluid neutrons, and the crust, a region composed of normal matter. Taking a new step towards more complete equilibrium models, we include the effect of entrainment, which implies that a magnetic force acts on neutrons, too. We consider purely poloidal field cases and present improvements to an earlier numerical scheme for solving equilibrium equations, by introducing new convergence criteria. We find that entrainment results in qualitative differences in the structure of field lines along the magnetic axis.

  15. Preliminary engineering design of sodium-cooled CANDLE core

    SciTech Connect

    Takaki, Naoyuki; Namekawa, Azuma; Yoda, Tomoyuki; Mizutani, Akihiko; Sekimoto, Hiroshi

    2012-06-06

    The CANDLE burning process is characterized by the autonomous shifting of burning region with constant reactivity and constant spacial power distribution. Evaluations of such critical burning process by using widely used neutron diffusion and burning codes under some realistic engineering constraints are valuable to confirm the technical feasibility of the CANDLE concept and to put the idea into concrete core design. In the first part of this paper, it is discussed that whether the sustainable and stable CANDLE burning process can be reproduced even by using conventional core analysis tools such as SLAROM and CITATION-FBR. As a result, it is certainly possible to demonstrate it if the proper core configuration and initial fuel composition required as CANDLE core are applied to the analysis. In the latter part, an example of a concrete image of sodium cooled, metal fuel, 2000MWt rating CANDLE core has been presented by assuming an emerging inevitable technology of recladding. The core satisfies engineering design criteria including cladding temperature, pressure drop, linear heat rate, and cumulative damage fraction (CDF) of cladding, fast neutron fluence and sodium void reactivity which are defined in the Japanese FBR design project. It can be concluded that it is feasible to design CANDLE core by using conventional codes while satisfying some realistic engineering design constraints assuming that recladding at certain time interval is technically feasible.

  16. Preliminary engineering design of sodium-cooled CANDLE core

    NASA Astrophysics Data System (ADS)

    Takaki, Naoyuki; Namekawa, Azuma; Yoda, Tomoyuki; Mizutani, Akihiko; Sekimoto, Hiroshi

    2012-06-01

    The CANDLE burning process is characterized by the autonomous shifting of burning region with constant reactivity and constant spacial power distribution. Evaluations of such critical burning process by using widely used neutron diffusion and burning codes under some realistic engineering constraints are valuable to confirm the technical feasibility of the CANDLE concept and to put the idea into concrete core design. In the first part of this paper, it is discussed that whether the sustainable and stable CANDLE burning process can be reproduced even by using conventional core analysis tools such as SLAROM and CITATION-FBR. As a result, it is certainly possible to demonstrate it if the proper core configuration and initial fuel composition required as CANDLE core are applied to the analysis. In the latter part, an example of a concrete image of sodium cooled, metal fuel, 2000MWt rating CANDLE core has been presented by assuming an emerging inevitable technology of recladding. The core satisfies engineering design criteria including cladding temperature, pressure drop, linear heat rate, and cumulative damage fraction (CDF) of cladding, fast neutron fluence and sodium void reactivity which are defined in the Japanese FBR design project. It can be concluded that it is feasible to design CADLE core by using conventional codes while satisfying some realistic engineering design constraints assuming that recladding at certain time interval is technically feasible.

  17. NASA'S Chandra Finds Superfluid in Neutron Star's Core

    NASA Astrophysics Data System (ADS)

    2011-02-01

    NASA's Chandra X-ray Observatory has discovered the first direct evidence for a superfluid, a bizarre, friction-free state of matter, at the core of a neutron star. Superfluids created in laboratories on Earth exhibit remarkable properties, such as the ability to climb upward and escape airtight containers. The finding has important implications for understanding nuclear interactions in matter at the highest known densities. Neutron stars contain the densest known matter that is directly observable. One teaspoon of neutron star material weighs six billion tons. The pressure in the star's core is so high that most of the charged particles, electrons and protons, merge resulting in a star composed mostly of uncharged particles called neutrons. Two independent research teams studied the supernova remnant Cassiopeia A, or Cas A for short, the remains of a massive star 11,000 light years away that would have appeared to explode about 330 years ago as observed from Earth. Chandra data found a rapid decline in the temperature of the ultra-dense neutron star that remained after the supernova, showing that it had cooled by about four percent over a 10-year period. "This drop in temperature, although it sounds small, was really dramatic and surprising to see," said Dany Page of the National Autonomous University in Mexico, leader of a team with a paper published in the February 25, 2011 issue of the journal Physical Review Letters. "This means that something unusual is happening within this neutron star." Superfluids containing charged particles are also superconductors, meaning they act as perfect electrical conductors and never lose energy. The new results strongly suggest that the remaining protons in the star's core are in a superfluid state and, because they carry a charge, also form a superconductor. "The rapid cooling in Cas A's neutron star, seen with Chandra, is the first direct evidence that the cores of these neutron stars are, in fact, made of superfluid and

  18. Radiation of Neutron Stars Produced by Superfluid Core

    NASA Astrophysics Data System (ADS)

    Svidzinsky, Anatoly A.

    2003-06-01

    We find a new mechanism of neutron star radiation wherein radiation is produced by the stellar interior. The main finding is that the neutron star interior is transparent for collisionless electron sound, the same way as it is transparent for neutrinos. In the presence of the magnetic field the electron sound is coupled with electromagnetic radiation; such collective excitation is known as a fast magnetosonic wave. At high densities such waves reduce to the zero sound in electron liquid, while near the stellar surface they are similar to electromagnetic waves in a medium. We find that zero sound is generated by superfluid vortices in the stellar core. Thermally excited helical vortex waves produce fast magnetosonic waves in the stellar crust that propagate toward the surface and transform into outgoing electromagnetic radiation. The magnetosonic waves are partially absorbed in a thin layer below the surface. The absorption is highly anisotropic; it is smaller for waves that in the absorbing layer propagate closer to the magnetic field direction. As a result, the vortex radiation is pulsed with the period of star rotation. The vortex radiation has the spectral index α~-0.45 and can explain nonthermal radiation of middle-aged pulsars observed in the infrared, optical, and hard X-ray bands. The radiation is produced in the star interior, rather than in the magnetosphere, which allows direct determination of the core temperature. Comparing the theory with available spectra observations, we find that the core temperature of the Vela pulsar is T~8×108 K, while the core temperature of PSR B0656+14 and Geminga exceeds 2×108 K. This is the first measurement of the temperature of a neutron star core. The temperature estimate rules out equations of state incorporating Bose condensations of pions or kaons and quark matter in these objects. The estimate also allows us to determine the critical temperature of triplet neutron superfluidity in the Vela core, Tc=(7.5+/-1.5)×109

  19. Design of multidirectional neutron beams for boron neutron capture synovectomy

    SciTech Connect

    Gierga, D.P.; Yanch, J.C.; Shefer, R.E.

    1997-12-01

    Boron neutron capture synovectomy (BNCS) is a potential application of the {sup 10}B(n, a) {sup 7}Li reaction for the treatment of rheumatoid arthritis. The target of therapy is the synovial membrane. Rheumatoid synovium is greatly inflamed and is the source of the discomfort and disability associated with the disease. The BNCS proposes to destroy the synovium by first injecting a boron-labeled compound into the joint space and then irradiating the joint with a neutron beam. This study discusses the design of a multidirectional neutron beam for BNCS.

  20. NASA'S Chandra Finds Superfluid in Neutron Star's Core

    NASA Astrophysics Data System (ADS)

    2011-02-01

    NASA's Chandra X-ray Observatory has discovered the first direct evidence for a superfluid, a bizarre, friction-free state of matter, at the core of a neutron star. Superfluids created in laboratories on Earth exhibit remarkable properties, such as the ability to climb upward and escape airtight containers. The finding has important implications for understanding nuclear interactions in matter at the highest known densities. Neutron stars contain the densest known matter that is directly observable. One teaspoon of neutron star material weighs six billion tons. The pressure in the star's core is so high that most of the charged particles, electrons and protons, merge resulting in a star composed mostly of uncharged particles called neutrons. Two independent research teams studied the supernova remnant Cassiopeia A, or Cas A for short, the remains of a massive star 11,000 light years away that would have appeared to explode about 330 years ago as observed from Earth. Chandra data found a rapid decline in the temperature of the ultra-dense neutron star that remained after the supernova, showing that it had cooled by about four percent over a 10-year period. "This drop in temperature, although it sounds small, was really dramatic and surprising to see," said Dany Page of the National Autonomous University in Mexico, leader of a team with a paper published in the February 25, 2011 issue of the journal Physical Review Letters. "This means that something unusual is happening within this neutron star." Superfluids containing charged particles are also superconductors, meaning they act as perfect electrical conductors and never lose energy. The new results strongly suggest that the remaining protons in the star's core are in a superfluid state and, because they carry a charge, also form a superconductor. "The rapid cooling in Cas A's neutron star, seen with Chandra, is the first direct evidence that the cores of these neutron stars are, in fact, made of superfluid and

  1. AHTR Mechanical, Structural, And Neutronic Preconceptual Design

    SciTech Connect

    Varma, Venugopal Koikal; Holcomb, David Eugene; Peretz, Fred J; Bradley, Eric Craig; Ilas, Dan; Qualls, A L; Zaharia, Nathaniel M

    2012-10-01

    This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming a commercial reactor class. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The AHTR employs plate type coated particle fuel assemblies with rapid, off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month 2-batch cycle with 9 weight-percent enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The present design intent is for used fuel to be stored inside of containment for at least 6 months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design concept incorporates multiple levels of radioactive material containment including fully passive responses to all identified design basis or non-very-low frequency beyond design basis accidents. Key building design elements include: 1) below grade siting to minimize vulnerability to aircraft impact, 2) multiple natural circulation decay heat rejection chimneys, 3) seismic

  2. AHTR Mechanical, Structural, and Neutronic Preconceptual Design

    SciTech Connect

    Varma, V.K.; Holcomb, D.E.; Peretz, F.J.; Bradley, E.C.; Ilas, D.; Qualls, A.L.; Zaharia, N.M.

    2012-09-15

    This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming an option for commercial reactor deployment. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The reactor concept development remains at a preconceptual level of maturity. While the overall appearance of an AHTR design is anticipated to be similar to the current concept, optimized dimensions will differ from those presented here. The AHTR employs plate type coated particle fuel assemblies with rapid, off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month two-batch cycle with 9 wt. % enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The report includes a preconceptual design of the manipulators, the fuel transfer system, and the used fuel storage system. The present design intent is for used fuel to be stored inside of containment for at least six months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design

  3. Neutron tube design study for boron neutron capture therapy application

    SciTech Connect

    Verbeke, J.M.; Lee, Y.; Leung, K.N.; Vujic, J.; Williams, M.D.; Wu, L.K.; Zahir, N.

    1999-05-06

    Radio-frequency (RF) driven ion sources are being developed in Lawrence Berkeley National Laboratory (LBNL) for sealed-accelerator-tube neutron generator application. By using a 5-cm-diameter RF-driven multicusp source H{sup +} yields over 95% have been achieved. These experimental findings will enable one to develop compact neutron generators based on the D-D or D-T fusion reactions. In this new neutron generator, the ion source, the accelerator and the target are all housed in a sealed metal container without external pumping. Recent moderator design simulation studies have shown that 14 MeV neutrons could be moderated to therapeutically useful energy ranges for boron neutron capture therapy (BNCT). The dose near the center of the brain with optimized moderators is about 65% higher than the dose obtained from a typical neutron spectrum produced by the Brookhaven Medical Research Reactor (BMRR), and is comparable to the dose obtained by other accelerator-based neutron sources. With a 120 keV and 1 A deuteron beam, a treatment time of {approx}35 minutes is estimated for BNCT.

  4. NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE

    SciTech Connect

    John D. Bess

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately +/-1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  5. Instability of superfluid flow in the neutron star core

    NASA Astrophysics Data System (ADS)

    Link, B.

    2012-04-01

    Pinning of superfluid vortices to magnetic flux tubes in the outer core of a neutron star supports a velocity difference of ˜105 cm s-1 between the neutron superfluid and the proton-electron fluid as the star spins down. Under the Magnus force that arises on the vortex array, vortices undergo vortex creep through thermal activation or quantum tunnelling. We examine the hydrodynamic stability of this situation. Vortex creep introduces two low-frequency modes, one of which is unstable above a critical wavenumber for any non-zero flow velocity of the neutron superfluid with respect to the charged fluid. For typical pinning parameters of the outer core, the superfluid flow is unstable over wavelengths λ≲ 10 m and over time-scales of ˜(λ/1 m)1/2 yr down to ˜1 d. The vortex lattice could degenerate into a tangle, and the superfluid flow would become turbulent. We suggest that superfluid turbulence could be responsible for the red timing noise seen in many neutron stars, and find a predicted spectrum that is generally consistent with observations.

  6. DANDE: a linked code system for core neutronics/depletion analysis

    SciTech Connect

    LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

    1985-06-01

    This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is made clear in this report by following a sample problem.

  7. Energy-Deposition and Damage Calculations in Core-Vessel Inserts at the Spallation Neutron Source

    SciTech Connect

    Murphy, B.D.

    2002-06-25

    Heat-deposition and damage calculations are described for core-vessel inserts in the target area of the Spallation Neutron Source. Two separate designs for these inserts (or neutron beam tubes) were studied; a single-unit insert and a multi-unit insert. The single unit contains a neutron guide; the multi unit does not. Both units are constructed of stainless steel. For the single unit, separate studies were carried out with the guide composed of stainless steel, glass, and aluminum. Results are also reported for an aluminum window on the front of the insert, a layer of nickel on the guide, a cadmium shield surrounding the guide, and a stainless steel plug in the beam-tube opening. The locations of both inserts were the most forward positions to be occupied by each design respectively thus ensuring that the calculations are conservative.

  8. Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor

    NASA Astrophysics Data System (ADS)

    Parma, Edward J.; Naranjo, Gerald E.; Lippert, Lance L.; Vehar, David W.

    2016-02-01

    Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR) is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity "bucket" environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters. Work supported by the United States Department of Energy at Sandia National Laboratories. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned

  9. Combined analysis of neutron and photon flux measurements for the Jules Horowitz reactor core mapping

    SciTech Connect

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

    2011-07-01

    We study the combined analysis of nuclear measurements to improve the knowledge of the irradiation conditions in the experimental locations of the future Jules Horowitz Reactor (JHR). The goal of the present work is to measure more accurately neutron flux, photon flux and nuclear heating in the reactor. In a Material Testing Reactor (MTR), nuclear heating is a crucial parameter to design the experimental devices to be irradiated in harsh nuclear conditions. This parameter drives the temperature of the devices and of the samples. The numerical codes can predict this parameter but in-situ measurements are necessary to reach the expected accuracy. For this reason, one objective of the IN-CORE program [1] is to study the combined measurements of neutron and photon flux and their cross advanced interpretation. It should be reminded that both neutron and photon sensors are not totally selective as their signals are due to neutron and photon interactions. We intend to measure the neutron flux by three different kinds of sensors (Uranium Fission chamber, Plutonium Fission chamber and Self Powered Neutron Detector), the photon flux by two different sensors (Ionization chamber and Self Powered Gamma Detector) and the nuclear heating by two different ones (Differential calorimeter and Gamma Thermometer). For the same parameter, we expect that the use of different kinds of sensors will allow a better estimation of the aimed parameter by mixing different spectrum responses and different neutron and gamma contributions. An experimental test called CARMEN-1 is scheduled in OSIRIS reactor (CEA Saclay - France) at the end of 2011, with the goal to map irradiation locations in the reactor reflector to get a first validation of the analysis model. This article focuses on the sensor selection for CARMEN-1 experiment and to the way to link neutron and photon flux measurements in view to reduce their uncertainties but also to better assess the neutron and photon contributions to nuclear

  10. One pass core design of a super fast reactor

    SciTech Connect

    Liu, Qingjie; Oka, Yoshiaki

    2013-07-01

    One pass core design for Supercritical-pressure light water-cooled fast reactor (Super FR) is proposed. The whole core is cooled with upward flow in one through flow pattern like PWR. Compared with the previous two pass core design; this new flow pattern can significantly simplify the core concept. Upper core structure, coolant flow scheme as well as refueling procedure are as simple as in PWR. In one pass core design, supercritical-pressure water is at approximately 25.0 MPa and enters the core at 280 C. degrees and is heated up in one through flow pattern upwardly to the average outlet temperature of 500 C. degrees. Great density change in vertical direction can cause significant axial power offset during the cycle. Meanwhile, Pu accumulated in the UO{sub 2} fuel blanket assemblies also introduces great power increase during cycle, which requires large amount of flow for heat removal and makes the outlet temperature of blanket low at the beginning of equilibrium cycle (BOEC). To deal with these issues, some MOX fuel is applied in the bottom region of the blanket assembly. This can help to mitigate the power change in blanket due to Pu accumulation and to increase the outlet temperature of the blanket during cycle. Neutron transport and thermohydraulics coupled calculation shows that this design can satisfy the requirement in the Super FR principle for both 500 C. degrees outlet temperature and negative coolant void reactivity. (authors)

  11. r-Modes of Neutron Stars with Superfluid Cores

    NASA Astrophysics Data System (ADS)

    Lee, Umin; Yoshida, Shijun

    2003-03-01

    We investigate the modal properties of the r-modes of rotating neutron stars with the core filled with neutron and proton superfluids, taking account of entrainment effects between the superfluids. The stability of the r-modes against gravitational radiation reaction is also examined considering viscous dissipation due to shear and a damping mechanism called ``mutual friction'' between the superfluids in the core. We find that the r-modes in the superfluid core are split into ordinary r-modes and superfluid r-modes, which we call, respectively, ro- and rs-modes. The two superfluids in the core flow together for the ro-modes, while they countermove for the rs-modes. For the ro-modes, the coefficient κ0≡limΩ-->0ω/Ω is equal to 2m/[l'(l'+1)], almost independent of the parameter η that parameterizes the entrainment effects between the superfluids, where Ω is the angular frequency of rotation, ω is the oscillation frequency observed in the corotating frame of the star, and l' and m are the indices of the spherical harmonic function representing the angular dependence of the r-modes. For the rs-modes, on the other hand, κ0 is equal to 2m/[l'(l'+1)] at η=0 (no entrainment), and it almost linearly increases as η is increased from η=0. The ro-modes, for which w'≡v'p- v'n~Ω3, correspond to the r-modes discussed by L. Lindblom & G. Mendell, where v'n and v'p are the Eulerian velocity perturbations of the neutron and proton superfluids, respectively. The mutual friction in the superfluid core is found ineffective to stabilize the r-mode instability caused by the ro-mode except in a few narrow regions of η. The r-mode instability caused by the rs-modes, on the other hand, is extremely weak and easily damped by dissipative processes in the star.

  12. A Methodology for the Neutronics Design of Space Nuclear Reactors

    SciTech Connect

    King, Jeffrey C.; El-Genk, Mohamed S.

    2004-02-04

    A methodology for the neutronics design of space power reactors is presented. This methodology involves balancing the competing requirements of having sufficient excess reactivity for the desired lifetime, keeping the reactor subcritical at launch and during submersion accidents, and providing sufficient control over the lifetime of the reactor. These requirements are addressed by three reactivity values for a given reactor design: the excess reactivity at beginning of mission, the negative reactivity at shutdown, and the negative reactivity margin in submersion accidents. These reactivity values define the control worth and the safety worth in submersion accidents, used for evaluating the merit of a proposed reactor type and design. The Heat Pipe-Segmented Thermoelectric Module Converters space reactor core design is evaluated and modified based on the proposed methodology. The final reactor core design has sufficient excess reactivity for 10 years of nominal operation at 1.82 MW of fission power and is subcritical at launch and in all water submersion accidents.

  13. A Methodology for the Neutronics Design of Space Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2004-02-01

    A methodology for the neutronics design of space power reactors is presented. This methodology involves balancing the competing requirements of having sufficient excess reactivity for the desired lifetime, keeping the reactor subcritical at launch and during submersion accidents, and providing sufficient control over the lifetime of the reactor. These requirements are addressed by three reactivity values for a given reactor design: the excess reactivity at beginning of mission, the negative reactivity at shutdown, and the negative reactivity margin in submersion accidents. These reactivity values define the control worth and the safety worth in submersion accidents, used for evaluating the merit of a proposed reactor type and design. The Heat Pipe-Segmented Thermoelectric Module Converters space reactor core design is evaluated and modified based on the proposed methodology. The final reactor core design has sufficient excess reactivity for 10 years of nominal operation at 1.82 MW of fission power and is subcritical at launch and in all water submersion accidents.

  14. Automated Design and Optimization of Pebble-bed Reactor Cores

    SciTech Connect

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2010-07-01

    We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.

  15. LETTER TO THE EDITOR: The halo structure of neutron-drip line nuclei: (neutron) cluster-core model

    NASA Astrophysics Data System (ADS)

    Gupta, Raj K.; Balasubramaniam, M.; Puri, Rajeev K.; Scheid, Werner

    2000-02-01

    Nuclei at the neutron-drip line are studied. In particular, we analyse the binding energies with the effects of Coulomb repulsion, nuclear attraction due to proximity and rotational energy due to angular momentum. This includes the study of the `neutron-halo structure' of some 18 light neutron-rich nuclei. In terms of the neutron-cluster + core picture, the halo structures of these nuclei are identified which agree with the hypothesis of one/two neutron separation energies and are supported by available experimental information. All the studied nuclei possess 1n- or 2n-halo structure. It is further shown that these nuclei prefer stable cores with not only the neutron number N = 2Z but also with 2Z +/-2.

  16. Stellar encounters involving neutron stars in globular cluster cores

    NASA Technical Reports Server (NTRS)

    Davies, M. B.; Benz, W.; Hills, J. G.

    1992-01-01

    Encounters between a 1.4 solar mass neutron star and a 0.8 solar mass red giant (RG) and between a 1.4 solar mass neutron star (NS) and an 0.8 solar mass main-sequence (MS) star have been successfully simulated. In the case of encounters involving an RG, bound systems are produced when the separation at periastron passage R(MIN) is less than about 2.5 R(RG). At least 70 percent of these bound systems are composed of the RG core and NS forming a binary engulfed in a common envelope of what remains of the former RG envelope. Once the envelope is ejected, a tight white dwarf-NS binary remains. For MS stars, encounters with NSs will produce bound systems when R(MIN) is less than about 3.5 R(MS). Some 50 percent of these systems will be single objects with the NS engulfed in a thick disk of gas almost as massive as the original MS star. The ultimate fate of such systems is unclear.

  17. High-burnup core design using minor actinide-containing metal fuel

    SciTech Connect

    Ohta, Hirokazu; Ogata, Takanari; Obara, T.

    2013-07-01

    A neutronic design study of metal fuel fast reactor (FR) cores is conducted on the basis of an innovative fuel design concept to achieve an extremely high burnup and realize an efficient fuel cycle system. Since it is expected that the burnup reactivity swing will become extremely large in an unprecedented high burnup core, minor actinides (MAs) from light water reactors (LWRs) are added to fresh fuel to improve the core internal conversion. Core neutronic analysis revealed that high burnups of about 200 MWd/kg for a small-scale core and about 300 MWd/kg for a large-scale core can be attained while suppressing the burnup reactivity swing to almost the same level as that of conventional cores with normal burnup. An actinide burnup analysis has shown that the MA consumption ratio is improved to about 60% and that the accumulated MAs originating from LWRs can be efficiently consumed by the high-burnup metal fuel FR. (authors)

  18. A High Temperature-Tolerant and Radiation-Resistant In-Core Neutron Sensor for Advanced Reactors. Final report

    SciTech Connect

    Cao, Lei; Miller, Don

    2015-01-23

    The objectives of this project are to develop a small and reliable gallium nitride (GaN) neutron sensor that is capable of withstanding high neutron fluence and high temperature, isolating gamma background, and operating in a wide dynamic range. The first objective will be the understanding of the fundamental materials properties and electronic response of a GaN semiconductor materials and device in an environment of high temperature and intense neutron field. To achieve such goal, an in-situ study of electronic properties of GaN device such as I-V, leakage current, and charge collection efficiency (CCE) in high temperature using an external neutron beam will be designed and implemented. We will also perform in-core irradiation of GaN up to the highest yet fast neutron fluence and an off-line performance evaluation.

  19. Advanced Neutron Sources: Plant Design Requirements

    SciTech Connect

    Not Available

    1990-07-01

    The Advanced Neutron Source (ANS) is a new, world class facility for research using hot, thermal, cold, and ultra-cold neutrons. At the heart of the facility is a 350-MW{sub th}, heavy water cooled and moderated reactor. The reactor is housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides fans out into a large guide hall, housing about 30 neutron research stations. Office, laboratory, and shop facilities are included to provide a complete users facility. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory at the end of the decade. This Plant Design Requirements document defines the plant-level requirements for the design, construction, and operation of the ANS. This document also defines and provides input to the individual System Design Description (SDD) documents. Together, this Plant Design Requirements document and the set of SDD documents will define and control the baseline configuration of the ANS.

  20. Neutronics Analyses of the Minimum Original HEU TREAT Core

    SciTech Connect

    Kontogeorgakos, D.; Connaway, H.; Yesilyurt, G.; Wright, A.

    2014-04-01

    This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to validate the MCNP model of the TREAT reactor with the well-documented measurements which were taken during the start-up and early operation of TREAT. Furthermore, the effect of carbon graphitization was also addressed. The graphitization level was assumed to be 100% (ANL/GTRI/TM-13/4). For this purpose, a set of experiments was chosen to validate the TREAT MCNP model, involving the approach to criticality procedure, in-core neutron flux measurements with foils, and isothermal temperature coefficient and temperature distribution measurements. The results of this study extended the knowledge base for the TREAT MCNP calculations and established the credibility of the MCNP model to be used in the core conversion feasibility analysis.

  1. Advanced Neutron Source: Plant Design Requirements

    SciTech Connect

    Not Available

    1990-07-01

    The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS.

  2. A new fuel loading design for the Advanced Neutron Source

    SciTech Connect

    Gehin, J.C.; Renier, J.P.; Worley, B.A.

    1994-06-01

    A new fuel loading design has been developed for the Advanced Neutron Source Reactor. In this reactor the combination of a small core volume and high power results in a very high power density. Using a direct optimization procedure the thermal-hydraulic margins for oxide temperature drop, centerline temperature and incipient boiling (and thus critical heat flux) were maximized to increase the limiting thermal power from 298 MW to 346 MW compared to the previous fuel grading, while maintaining the desired peak reflector thermal flux.

  3. Recent Problems of Transformer Core Design

    NASA Astrophysics Data System (ADS)

    Valkovic, Z.

    1988-01-01

    The paper describes the result of the investigations of the efficiency of power loss reduction in transformer cores made with high-permeability (HGO) and laser scribed (LS) grain-oriented electrical steels, and also the phenomena in three-limb three-phase cores with the so-called staggered T-joint design. The efficiency of the HGO material depends on core form and core induction. The efficiency is better for single-phase than for three-phase cores and also for higher induction. The localised efficiency of HGO material is not uniform and it is significantly lower in the yoke than in other parts. The efficiency of LS material (grade ZDKH) is better than that of the HGO material and also somewhat higher for single-phase than for three-phase cores. The localised flux distribution in the central limb of the core with staggered T-joint is more uniform and the content of higher harmonics is smaller than in the core with conventional V-45° T-joint. This results in a 13% loss reduction in the central limb and in a 4-5% reduction of total core loss.

  4. Global shielding analysis for the three-element core advanced neutron source reactor under normal operating conditions

    SciTech Connect

    Slater, C.O.; Bucholz, J.A.

    1995-08-01

    Two-dimensional discrete ordinates radiation transport calculations were performed for a model of the three-element core Advanced Neutron Source reactor design under normal operating conditions. The core consists of two concentric upper elements and a lower element radially centered in the annulus between the upper elements. The initial radiation transport calculations were performed with the DORT two-dimensional discrete ordinates radiation transport code using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub 6} quadrature, and a P{sub 1} Legendre polynomial expansion of the cross sections to determine the fission neutron source distribution in the core fuel elements. These calculations were limited to neutron groups only. The final radiation transport calculations, also performed with DORT using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub l0} quadrature, and a P{sub 3} Legendre polynomial expansion of the cross sections, produced neutron and gamma-ray fluxes over the full extent of the geometry model. Responses (or activities) at various locations in the model were then obtained by folding the appropriate response functions with the fluxes at those locations. Some comparisons were made with VENTURE-calculated (diffusion theory) 20-group neutron fluxes that were summed into four broad groups. Tne results were in reasonably good agreement when the effects of photoneutrons were not included, thus verifying the physics model upon which the shielding model was based. Photoneutrons increased the fast-neutron flux levels deep within the D{sub 2}0 several orders of magnitude. Results are presented as tables of activity values for selected radial and axial traverses, plots of the radial and axial traverse data, and activity contours superimposed on the calculational geometry model.

  5. The long-term rotation dynamics of neutron stars with differentially rotating unmagnetized core

    NASA Astrophysics Data System (ADS)

    Barsukov, D. P.; Goglichidze, O. A.; Tsygan, A. I.

    2014-10-01

    We consider the pulsar long-term rotation dynamics taking into account the non-rigidity of neutron star rotation. We restrict our attention to the models with two essential assumptions: (1) crust-core interaction occurs via the viscosity (magnetic coupling is not important); (2) neutron star shape is symmetrical over the magnetic axis. The neutron star core is described by linearized quasi-stationary Newtonian hydrodynamical equations in one-fluid and two-fluid (neutron superfluidity) approximations. It is shown that in this case the pulsar inclination angle evolves to 0° or 90° very quickly. Since such fast evolution seems to contradict the observation data, either neutron stars are triaxial or the magnetic field plays the leading role in crust-core coupling.

  6. /sup 18/O as a core plus two valence neutrons: A three-body Faddeev calculation

    SciTech Connect

    Ueta, K.; Miyake, H.; Mizukami, A.

    1983-01-01

    The nucleus /sup 18/O is studied assuming a three-body model: two neutrons outside an inert core of /sup 16/O: and solving the Faddeev equations. The calculated spectrum is in good agreement with experiment.

  7. Characterization of the fast neutron irradiation facility of the Portuguese Research Reactor after core conversion.

    PubMed

    Marques, J G; Sousa, M; Santos, J P; Fernandes, A C

    2011-08-01

    The fast neutron irradiation facility of the Portuguese Research Reactor was characterized after the reduction in uranium enrichment and rearrangement of the core configuration. In this work we report on the determination of the hardness parameter and the 1MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard.

  8. Calculated Neutron and Gamma-ray Spectra across the Prismatic Very High Temperature Reactor Core

    SciTech Connect

    James W. Sterbentz

    2008-05-01

    Neutron and gamma-ray flux spectra are calculated using the MCNP5 computer code and a one-sixth core model of a prismatic Very High Temperature Reactor based on the General Atomics Gas Turbine-Modular Helium Reactor. Spectra are calculated in the five inner reflector graphite block rings, three annular active core fuel rings, three outer graphite reflector block rings, and the core barrel. The neutron spectra are block and fuel pin averages and are calculated as a function of temperature and burnup. Also provided are the total, fast, and thermal radial profile fluxes and core barrel dpa rates.

  9. MPACT Fast Neutron Multiplicity System Design Concepts

    SciTech Connect

    D. L. Chichester; S. A. Pozzi; J. L. Dolan; M. T. Kinlaw; A. C. Kaplan; M. Flaska; A. Enqvist; J. T. Johnsom; S. M. Watson

    2012-10-01

    This report documents work performed by Idaho National Laboratory and the University of Michigan in fiscal year (FY) 2012 to examine design parameters related to the use of fast-neutron multiplicity counting for assaying plutonium for materials protection, accountancy, and control purposes. This project seeks to develop a new type of neutron-measurement-based plutonium assay instrument suited for assaying advanced fuel cycle materials. Some current-concept advanced fuels contain high concentrations of plutonium; some of these concept fuels also contain other fissionable actinides besides plutonium. Because of these attributes the neutron emission rates of these new fuels may be much higher, and more difficult to interpret, than measurements made of plutonium-only materials. Fast neutron multiplicity analysis is one approach for assaying these advanced nuclear fuels. Studies have been performed to assess the conceptual performance capabilities of a fast-neutron multiplicity counter for assaying plutonium. Comparisons have been made to evaluate the potential improvements and benefits of fast-neutron multiplicity analyses versus traditional thermal-neutron counting systems. Fast-neutron instrumentation, using for example an array of liquid scintillators such as EJ-309, have the potential to either a) significantly reduce assay measurement times versus traditional approaches, for comparable measurement precision values, b) significantly improve assay precision values, for measurement durations comparable to current-generation technology, or c) moderating improve both measurement precision and measurement durations versus current-generation technology. Using the MCNPX-PoliMi Monte Carlo simulation code, studies have been performed to assess the doubles-detection efficiency for a variety of counter layouts of cylindrical liquid scintillator detector cells over one, two, and three rows. Ignoring other considerations, the best detector design is the one with the most

  10. The Spallation Neutron Source accelerator system design

    NASA Astrophysics Data System (ADS)

    Henderson, S.; Abraham, W.; Aleksandrov, A.; Allen, C.; Alonso, J.; Anderson, D.; Arenius, D.; Arthur, T.; Assadi, S.; Ayers, J.; Bach, P.; Badea, V.; Battle, R.; Beebe-Wang, J.; Bergmann, B.; Bernardin, J.; Bhatia, T.; Billen, J.; Birke, T.; Bjorklund, E.; Blaskiewicz, M.; Blind, B.; Blokland, W.; Bookwalter, V.; Borovina, D.; Bowling, S.; Bradley, J.; Brantley, C.; Brennan, J.; Brodowski, J.; Brown, S.; Brown, R.; Bruce, D.; Bultman, N.; Cameron, P.; Campisi, I.; Casagrande, F.; Catalan-Lasheras, N.; Champion, M.; Champion, M.; Chen, Z.; Cheng, D.; Cho, Y.; Christensen, K.; Chu, C.; Cleaves, J.; Connolly, R.; Cote, T.; Cousineau, S.; Crandall, K.; Creel, J.; Crofford, M.; Cull, P.; Cutler, R.; Dabney, R.; Dalesio, L.; Daly, E.; Damm, R.; Danilov, V.; Davino, D.; Davis, K.; Dawson, C.; Day, L.; Deibele, C.; Delayen, J.; DeLong, J.; Demello, A.; DeVan, W.; Digennaro, R.; Dixon, K.; Dodson, G.; Doleans, M.; Doolittle, L.; Doss, J.; Drury, M.; Elliot, T.; Ellis, S.; Error, J.; Fazekas, J.; Fedotov, A.; Feng, P.; Fischer, J.; Fox, W.; Fuja, R.; Funk, W.; Galambos, J.; Ganni, V.; Garnett, R.; Geng, X.; Gentzlinger, R.; Giannella, M.; Gibson, P.; Gillis, R.; Gioia, J.; Gordon, J.; Gough, R.; Greer, J.; Gregory, W.; Gribble, R.; Grice, W.; Gurd, D.; Gurd, P.; Guthrie, A.; Hahn, H.; Hardek, T.; Hardekopf, R.; Harrison, J.; Hatfield, D.; He, P.; Hechler, M.; Heistermann, F.; Helus, S.; Hiatt, T.; Hicks, S.; Hill, J.; Hill, J.; Hoff, L.; Hoff, M.; Hogan, J.; Holding, M.; Holik, P.; Holmes, J.; Holtkamp, N.; Hovater, C.; Howell, M.; Hseuh, H.; Huhn, A.; Hunter, T.; Ilg, T.; Jackson, J.; Jain, A.; Jason, A.; Jeon, D.; Johnson, G.; Jones, A.; Joseph, S.; Justice, A.; Kang, Y.; Kasemir, K.; Keller, R.; Kersevan, R.; Kerstiens, D.; Kesselman, M.; Kim, S.; Kneisel, P.; Kravchuk, L.; Kuneli, T.; Kurennoy, S.; Kustom, R.; Kwon, S.; Ladd, P.; Lambiase, R.; Lee, Y. Y.; Leitner, M.; Leung, K.-N.; Lewis, S.; Liaw, C.; Lionberger, C.; Lo, C. C.; Long, C.; Ludewig, H.; Ludvig, J.; Luft, P.; Lynch, M.; Ma, H.; MacGill, R.; Macha, K.; Madre, B.; Mahler, G.; Mahoney, K.; Maines, J.; Mammosser, J.; Mann, T.; Marneris, I.; Marroquin, P.; Martineau, R.; Matsumoto, K.; McCarthy, M.; McChesney, C.; McGahern, W.; McGehee, P.; Meng, W.; Merz, B.; Meyer, R.; Meyer, R.; Miller, B.; Mitchell, R.; Mize, J.; Monroy, M.; Munro, J.; Murdoch, G.; Musson, J.; Nath, S.; Nelson, R.; Nelson, R.; O`Hara, J.; Olsen, D.; Oren, W.; Oshatz, D.; Owens, T.; Pai, C.; Papaphilippou, I.; Patterson, N.; Patterson, J.; Pearson, C.; Pelaia, T.; Pieck, M.; Piller, C.; Plawski, T.; Plum, M.; Pogge, J.; Power, J.; Powers, T.; Preble, J.; Prokop, M.; Pruyn, J.; Purcell, D.; Rank, J.; Raparia, D.; Ratti, A.; Reass, W.; Reece, K.; Rees, D.; Regan, A.; Regis, M.; Reijonen, J.; Rej, D.; Richards, D.; Richied, D.; Rode, C.; Rodriguez, W.; Rodriguez, M.; Rohlev, A.; Rose, C.; Roseberry, T.; Rowton, L.; Roybal, W.; Rust, K.; Salazer, G.; Sandberg, J.; Saunders, J.; Schenkel, T.; Schneider, W.; Schrage, D.; Schubert, J.; Severino, F.; Shafer, R.; Shea, T.; Shishlo, A.; Shoaee, H.; Sibley, C.; Sims, J.; Smee, S.; Smith, J.; Smith, K.; Spitz, R.; Staples, J.; Stein, P.; Stettler, M.; Stirbet, M.; Stockli, M.; Stone, W.; Stout, D.; Stovall, J.; Strelo, W.; Strong, H.; Sundelin, R.; Syversrud, D.; Szajbler, M.; Takeda, H.; Tallerico, P.; Tang, J.; Tanke, E.; Tepikian, S.; Thomae, R.; Thompson, D.; Thomson, D.; Thuot, M.; Treml, C.; Tsoupas, N.; Tuozzolo, J.; Tuzel, W.; Vassioutchenko, A.; Virostek, S.; Wallig, J.; Wanderer, P.; Wang, Y.; Wang, J. G.; Wangler, T.; Warren, D.; Wei, J.; Weiss, D.; Welton, R.; Weng, J.; Weng, W.-T.; Wezensky, M.; White, M.; Whitlatch, T.; Williams, D.; Williams, E.; Wilson, K.; Wiseman, M.; Wood, R.; Wright, P.; Wu, A.; Ybarrolaza, N.; Young, K.; Young, L.; Yourd, R.; Zachoszcz, A.; Zaltsman, A.; Zhang, S.; Zhang, W.; Zhang, Y.; Zhukov, A.

    2014-11-01

    The Spallation Neutron Source (SNS) was designed and constructed by a collaboration of six U.S. Department of Energy national laboratories. The SNS accelerator system consists of a 1 GeV linear accelerator and an accumulator ring providing 1.4 MW of proton beam power in microsecond-long beam pulses to a liquid mercury target for neutron production. The accelerator complex consists of a front-end negative hydrogen-ion injector system, an 87 MeV drift tube linear accelerator, a 186 MeV side-coupled linear accelerator, a 1 GeV superconducting linear accelerator, a 248-m circumference accumulator ring and associated beam transport lines. The accelerator complex is supported by ~100 high-power RF power systems, a 2 K cryogenic plant, ~400 DC and pulsed power supply systems, ~400 beam diagnostic devices and a distributed control system handling ~100,000 I/O signals. The beam dynamics design of the SNS accelerator is presented, as is the engineering design of the major accelerator subsystems.

  11. Wireless, in-vessel neutron monitor for initial core-loading of advanced breeder reactors

    NASA Technical Reports Server (NTRS)

    Delorenzo, J. T.; Kennedy, E. J.; Blalock, T. V.; Rochelle, J. M.; Chiles, M. M.; Valentine, K. H.

    1981-01-01

    An experimental wireless, in-vessel neutron monitor was developed to measure the reactivity of an advanced breeder reactor as the core is loaded for the first time to preclude an accidental critically incident. The environment is liquid sodium at a temperature of approx. 220 C, with negligible gamma or neutron radiation. With ultrasonic transmission of neutron data, no fundamental limitation was observed after tests at 230 C for 2000 h. The neutron sensitivity was approx. 1 count/s-nv, and the potential data transmission rate was approx. 10,000 counts/s.

  12. Measuring Neutron-Star Spins via Burst Oscillations (core Program)

    NASA Astrophysics Data System (ADS)

    Measuring the spin of neutron stars in low-mass X-ray binaries is one of the great strengths and highest priorities for RXTE. We propose targeted observations of known thermonuclear burst sources which do not have confirmed burst oscillations, as well as previously unknown sources, in order to detect new examples of burst oscillations and thus add to the sample of neutron star spins. We will target sources in states of frequent, bright bursts by triggering on the detection of bursts by INTEGRAL and/or Swift. Detection of neutron stars spinning beyond the present maximum will allow us to significantly constrain the neutron-star equation of state, presently an area of major uncertainty.

  13. Determination of the kinetic parameters of the CALIBAN metallic core reactor from stochastic neutron measurements

    SciTech Connect

    Casoli, P.; Authier, N.; Chapelle, A.

    2012-07-01

    Several experimental devices are operated by the Criticality and Neutron Science Research Dept. of the CEA Valduc Laboratory. One of these is the Caliban metallic core reactor. The purpose of this study is to develop and perform experiments allowing to determinate some of fundamental kinetic parameters of the reactor. The prompt neutron decay constant and particularly its value at criticality can be measured with reactor noise techniques such as Rossi-{alpha} and Feynman variance-to-mean methods. Subcritical, critical, and even supercritical experiments were performed. Fission chambers detectors were put nearby the core and measurements were analyzed with the Rossi-{alpha} technique. A new value of the prompt neutron decay constant at criticality was determined, which allows, using the Nelson number method, new evaluations of the effective delayed neutron fraction and the in core neutron lifetime. As an introduction of this paper, some motivations of this work are given in part 1. In part 2, principles of the noise measurements experiments performed at the CEA Valduc Laboratory are reminded. The Caliban reactor is described in part 3. Stochastic neutron measurements analysis techniques used in this study are then presented in part 4. Results of fission chamber experiments are summarized in part 5. Part 6 is devoted to the current work, improvement of the experimental device using He 3 neutron detectors and first results obtained with it. Finally, conclusions and perspectives are given in part 7. (authors)

  14. Monte Carlo code for neutron scattering instrumentation design and analysis

    SciTech Connect

    Daemen, L.; Fitzsimmons, M.; Hjelm, R.; Olah, G.; Roberts, J.; Seeger, P.; Smith, G.; Thelliez, T.

    1996-09-01

    This is the final report of a one-year, Laboratory-Directed Research and Development (LDRD) at the Los Alamos National Laboratory (LANL). The development of next generation, accelerator based neutron sources calls for the design of new instruments for neutron scattering studies of materials. It will be necessary, in the near future, to evaluate accurately and rapidly the performance of new and traditional neutron instruments at short- and long-pulse spallation neutron sources, as well as continuous sources. We have developed a code that is a design tool to assist the instrument designer model new or existing instruments, test their performance, and optimize their most important features.

  15. Layered shielding design for an active neutron interrogation system

    NASA Astrophysics Data System (ADS)

    Whetstone, Zachary D.; Kearfott, Kimberlee J.

    2016-08-01

    The use of source and detector shields in active neutron interrogation can improve detector signal. In simulations, a shielded detector with a source rotated π/3 rad relative to the opening decreased neutron flux roughly three orders of magnitude. Several realistic source and detector shield configurations were simulated. A layered design reduced neutron and secondary photon flux in the detector by approximately one order of magnitude for a deuterium-tritium source. The shield arrangement can be adapted for a portable, modular design.

  16. Advanced Neutron Source radiological design criteria

    SciTech Connect

    Westbrook, J.L.

    1995-08-01

    The operation of the proposed Advanced Neutron Source (ANS) facility will present a variety of radiological protection problems. Because it is desired to design and operate the ANS according to the applicable licensing standards of the Nuclear Regulatory Commission (NRC), it must be demonstrated that the ANS radiological design basis is consistent not only with state and Department of Energy (DOE) and other usual federal regulations, but also, so far as is practicable, with NRC regulations and with recommendations of such organizations as the Institute of Nuclear Power Operations (INPO) and the Electric Power Research Institute (EPRI). Also, the ANS radiological design basis is in general to be consistent with the recommendations of authoritative professional and scientific organizations, specifically the National Council on Radiation Protection and Measurements (NCRP) and the International Commission on Radiological Protection (ICRP). As regards radiological protection, the principal goals of DOE regulations and guidance are to keep occupational doses ALARA [as low as (is) reasonably achievable], given the current state of technology, costs, and operations requirements; to control and monitor contained and released radioactivity during normal operation to keep public doses and releases to the environment ALARA; and to limit doses to workers and the public during accident conditions. Meeting these general design objectives requires that principles of dose reduction and of radioactivity control by employed in the design, operation, modification, and decommissioning of the ANS. The purpose of this document is to provide basic radiological criteria for incorporating these principles into the design of the ANS. Operations, modification, and decommissioning will be covered only as they are affected by design.

  17. Conceptual design of an RFQ accelerator-based neutron source for boron neutron-capture therapy

    SciTech Connect

    Wangler, T.P.; Stovall, J.E.; Bhatia, T.S.; Wang, C.K.; Blue, T.E.; Gahbauer, R.A.

    1989-01-01

    We present a conceptual design of a low-energy neutron generator for treatment of brain tumors by boron neutron capture theory (BNCT). The concept is based on a 2.5-MeV proton beam from a radio-frequency quadrupole (RFQ) linac, and the neutrons are produced by the /sup 7/Li(p,n)/sup 7/Be reaction. A liquid lithium target and modulator assembly are designed to provide a high flux of epithermal neutrons. The patient is administered a tumor-specific /sup 10/Be-enriched compound and is irradiated by the neutrons to create a highly localized dose from the reaction /sup 10/B(n,..cap alpha..)/sup 7/Li. An RFQ accelerator-based neutron source for BNCT is compact, which makes it practical to site the facility within a hospital. 11 refs., 5 figs., 1 tab.

  18. Sensitive and transportable gadolinium-core plastic scintillator sphere for neutron detection and counting

    NASA Astrophysics Data System (ADS)

    Dumazert, Jonathan; Coulon, Romain; Carrel, Frédérick; Corre, Gwenolé; Normand, Stéphane; Méchin, Laurence; Hamel, Matthieu

    2016-08-01

    Neutron detection forms a critical branch of nuclear-related issues, currently driven by the search for competitive alternative technologies to neutron counters based on the helium-3 isotope. The deployment of plastic scintillators shows a high potential for efficient detectors, safer and more reliable than liquids, more easily scalable and cost-effective than inorganic. In the meantime, natural gadolinium, through its 155 and mostly 157 isotopes, presents an exceptionally high interaction probability with thermal neutrons. This paper introduces a dual system including a metal gadolinium core inserted at the center of a high-scale plastic scintillator sphere. Incident fast neutrons are thermalized by the scintillator shell and then may be captured with a significant probability by gadolinium 155 and 157 nuclei in the core. The deposition of a sufficient fraction of the capture high-energy prompt gamma signature inside the scintillator shell will then allow discrimination from background radiations by energy threshold, and therefore neutron detection. The scaling of the system with the Monte Carlo MCNPX2.7 code was carried out according to a tradeoff between the moderation of incident fast neutrons and the probability of slow neutron capture by a moderate-cost metal gadolinium core. Based on the parameters extracted from simulation, a first laboratory prototype for the assessment of the detection method principle has been synthetized. The robustness and sensitivity of the neutron detection principle are then assessed by counting measurement experiments. Experimental results confirm the potential for a stable, highly sensitive, transportable and cost-efficient neutron detector and orientate future investigation toward promising axes.

  19. Characterization of the Annular Core Research Reactor (ACRR) Neutron Radiography System Imaging Plane

    NASA Astrophysics Data System (ADS)

    Kaiser, Krista; Chantel Nowlen, K.; DePriest, K. Russell

    2016-02-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1) available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  20. Design Analyses and Shielding of HFIR Cold Neutron Scattering Instruments

    SciTech Connect

    Gallmeier, F.X.; Selby, D.L.; Winn, B.; Stoica, D.; Jones, A.B.; Crow, L.

    2011-07-01

    Research reactor geometries and special characteristics present unique dosimetry analysis and measurement issues. The introduction of a cold neutron moderator and the production of cold neutron beams at the Oak Ridge National Laboratory High Flux Isotope Reactor have created the need for modified methods and devices for analyzing and measuring low energy neutron fields (0.01 to 100 meV). These methods include modifications to an MCNPX version to provide modeling of neutron mirror reflection capability. This code has been used to analyze the HFIR cold neutron beams and to design new instrument equipment that will use the beams. Calculations have been compared with time-of-flight measurements performed at the start of the neutron guides and at the end of one of the guides. The results indicate that we have a good tool for analyzing the transport of these low energy beams through neutron mirror and guide systems for distance up to 60 meters from the reactor. (authors)

  1. Reactor physics analyses of the advanced neutron source three-element core

    SciTech Connect

    Gehin, J.C.

    1995-08-01

    A reactor physics analysis was performed for the Advanced Neutron Source reactor with a three-element core configuration. The analysis was performed with a two-dimensional r-z 20-energy-group finite-difference diffusion theory model of the 17-d fuel cycle. The model included equivalent r-z geometry representations of the central control rods, the irradiation and production targets, and reflector components. Calculated quantities include fuel cycle parameters, fuel element power distributions, unperturbed neutron fluxes in the reflector and target regions, reactivity perturbations, and neutron kinetics parameters.

  2. Neutron Dosimetry on the Full-Core First Generation VVER-440 Aimed at Reactor Support Structure Load Evaluation

    NASA Astrophysics Data System (ADS)

    Borodkin, P.; Borodkin, G.; Khrennikov, N.; Konheiser, J.; Noack, K.

    2009-08-01

    Reactor support structures (RSS), especially the ferritic steel wall of the water tank, of first-generation VVER-440 are non-restorable reactor equipment, and their lifetime may restrict plant-life. All operated Russian first generation VVER-440 have a reduced core with dummy assemblies except Unit 4 of Novovoronezh nuclear power plant (NPP). In comparison with other reactors, the full-core loading scheme of this reactor provides the highest neutron fluence on the reactor pressure vessel (RPV) and RSS accumulated over design service-life and its prolongation. The radiation load parameters on the RPV and RSS that have resulted from this core loading scheme should be evaluated by means of precise calculations and validated by ex-vessel neutron dosimetry to provide the reliable assessment of embrittlement parameters of these reactor components. The results of different types of calculations and their comparison with measured data have been analyzed in this paper. The calculational analysis of RSS fluence rate variation in dependence on the core loading scheme, including the standard and low leakage core as well as the introduction of dummy assemblies, is presented in this paper.

  3. ARCADIA{sup R} - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

    SciTech Connect

    Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas; Thareau, Sebastien

    2007-07-01

    Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code system ARCADIA{sup R} and concludes on customer benefits. ARCADIA{sup R} is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA{sup R} system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)

  4. Hans A. Bethe Prize: Neutron Stars and Core-Collapse Supernovae

    NASA Astrophysics Data System (ADS)

    Lattimer, James

    2015-04-01

    Core-collapse supernovae lead to the formation of neutron stars, and both are sensitive to the dense matter equation of state. Hans Bethe first recognized that the matter in the collapsing core of a massive star has a relatively low entropy which prevents nuclear dissociation until nuclei merge near the nuclear saturation density. This recognition means that collapse continues until the core exceeds the saturation density. This prediction forms the foundation for modern simulations of supernovae. These supernovae sample matter up to about twice nuclear saturation density, but neutron stars are sensitive to the equation of state both near the saturation density and at several times higher densities. Two important recent developments are the discovery of two-solar mass neutron stars and refined experimental determinations of the behavior of the symmetry energy of nuclear matter near the saturation density. Combined with the assumption of causality, they imply that the radii of observed neutron stars are largely independent of their mass, and that this radius is in the range of 11 to 13 km. These theoretical results are not only consistent with expectations from theoretical studies of pure neutron matter, but also accumulated observations of both bursting and cooling neutron stars. In the near future, new pulsar timing data, which could lead to larger measured masses as well as measurements of moments of inertia, X-ray observations, such as from NICER, of bursting and other sources, and gravitational wave observations of neutron stars in merging compact binaries, will provide important new constraints on neutron stars and the dense matter equation of state. DOE DE-FG02-87ER-40317.

  5. The new Cold Neutron Chopper Spectrometer at the Spallation Neutron Source -- Design and Performance

    SciTech Connect

    Ehlers, Georg; Podlesnyak, Andrey A.; Niedziela, Jennifer L.; Iverson, Erik B.; Sokol, Paul E.

    2011-01-01

    The design and performance of the new cold neutron chopper spectrometer (CNCS) at the Spallation Neutron Source in Oak Ridge are described. CNCS is a direct-geometry inelastic time-of-flight spectrometer, designed essentially to cover the same energy and momentum transfer ranges as IN5 at ILL, LET at ISIS, DCS at NIST, TOFTOF at FRM-II, AMATERAS at J-PARC, PHAROS at LANSCE, and NEAT at HZB, at similar energy resolution. Measured values of key figures such as neutron flux at sample position and energy resolution are compared between measurements and ray tracing Monte Carlo simulations, and good agreement (better than 20% of absolute numbers) has been achieved. The instrument performs very well in the cold and thermal neutron energy ranges, and promises to become a workhorse for the neutron scattering community for quasielastic and inelastic scattering experiments.

  6. The new cold neutron chopper spectrometer at the Spallation Neutron Source: Design and performance

    SciTech Connect

    Ehlers, G.; Podlesnyak, A. A.; Niedziela, J. L.; Iverson, E. B.; Sokol, P. E.

    2011-08-15

    The design and performance of the new cold neutron chopper spectrometer (CNCS) at the Spallation Neutron Source in Oak Ridge are described. CNCS is a direct-geometry inelastic time-of-flight spectrometer, designed essentially to cover the same energy and momentum transfer ranges as IN5 at ILL, LET at ISIS, DCS at NIST, TOFTOF at FRM-II, AMATERAS at J-PARC, PHAROS at LANSCE, and NEAT at HZB, at similar energy resolution. Measured values of key figures such as neutron flux at sample position and energy resolution are compared between measurements and ray tracing Monte Carlo simulations, and good agreement (better than 20% of absolute numbers) has been achieved. The instrument performs very well in the cold and thermal neutron energy ranges, and promises to become a workhorse for the neutron scattering community for quasielastic and inelastic scattering experiments.

  7. Neutron porosity logging and core porosity measurements in the Beauvoir granite, Massif Central Range, France

    NASA Astrophysics Data System (ADS)

    Gallé, C.

    1994-08-01

    A large suite of geophysical logs have been run in the Beauvoir granite. The drillhole (900 m deep), first target of the French Deep Geology programme, is located in the Hercynian bedrock of Echassières in central France (Massif Central Range). After geochemical and petrological studies, the batholith was used for experiments pertaining to the storage of radioactive wastes. With its low porosity, its weak fracturing and its high homogeneity, the Beauvoir granite was chosen for the analysis of the relationship between logged data and the properties measured in the core. The study focused on neutron porosity and core water porosity. The Beauvoir granite has a total free water porosity of around 2% (average value of 54 core samples of rock mass) and an average neutron porosity of around 10%. We show that the origin of this significant difference is related to the neutron matrix effect of the granite. This phenomenon is partly due to the slowing-down effect of the combined water of clays and micas but also to the neutronic capture effect linked with the relatively high lepidolite (lithium mica) content of the granite. The Li 2O content controls 85% of the granite macroscopic capture cross-section. These two factors represent around 75% of the global neutron porosity of the Beauvoir granite. They have to be taken into consideration to get representative water contents of a low-porosity igneous rock from a neutron porosity log. Further investigations also demonstrated the necessity of choosing a better adapted neutron tool calibration for crystalline rocks. Instead of a standard calibration with limestone blocks, a calibration in granite blocks was simulated in order to obtain a better evaluation of the global neutron response of the granite. Then, by correcting this new neutron porosity for the matrix effect, it has been possible to determine water contents in accordance with laboratory water porosity values measured on core samples (2% average porosity). This

  8. Implementing a New Ion Chamber Design for Neutron Spin Rotation

    NASA Astrophysics Data System (ADS)

    Gardiner, Hannah; Anderson, Eamon; Fry, Jason; Holley, Adam; Snow, Mike

    2012-10-01

    The quark-quark weak interaction is difficult to measure due to the presence of the strong force. However, low energy neutrons passing through liquid Helium-4 can be used to probe the nucleon-nucleon weak interaction, which is induced by the quark-quark weak interaction. The neutron spin rotation experiment seeks to measure the spin rotation angle of neutrons due to their weak interaction with Helium-4 nuclei. This rotation angle is translated into a neutron flux asymmetry with a neutron polarizer/analyzer pair. A segmented Helium-3 gas ionization chamber was developed to measure the resultant neutron flux. We report on the design and initial tests of that ionization chamber. This work is supported by the National Science Foundation REU program and NSF grant #PHY-0969490.

  9. Verification of SMART Neutronics Design Methodology by the MCNAP Monte Carlo Code

    SciTech Connect

    Jong Sung Chung; Kyung Jin Shim; Chang Hyo Kim; Chungchan Lee; Sung Quun Zee

    2000-11-12

    SMART is a small advanced integral pressurized water reactor (PWR) of 330 MW(thermal) designed for both electricity generation and seawater desalinization. The CASMO-3/MASTER nuclear analysis system, a design-basis of Korean PWR plants, has been employed for the SMART core nuclear design and analysis because the fuel assembly (FA) characteristics and reactor operating conditions in temperature and pressure are similar to those of PWR plants. However, the SMART FAs are highly poisoned with more than 20 Al{sub 2}O{sub 3}-B{sub 4}C plus additional Gd{sub 2}O{sub 3}/UO{sub 2} BPRs each FA. The reactor is operated with control rods inserted. Therefore, the flux and power distribution may become more distorted than those of commercial PWR plants. In addition, SMART should produce power from room temperature to hot-power operating condition because it employs nuclear heating from room temperature. This demands reliable predictions of core criticality, shutdown margin, control rod worth, power distributions, and reactivity coefficients at both room temperature and hot operating condition, yet no such data are available to verify the CASMO-3/MASTER (hereafter MASTER) code system. In the absence of experimental verification data for the SMART neutronics design, the Monte Carlo depletion analysis program MCNAP is adopted as near-term alternatives for qualifying MASTER neutronics design calculations. The MCNAP is a personal computer-based continuous energy Monte Carlo neutronics analysis program written in C++ language. We established its qualification by presenting its prediction accuracy on measurements of Venus critical facilities and core neutronics analysis of a PWR plant in operation, and depletion characteristics of integral burnable absorber FAs of the current PWR. Here, we present a comparison of MASTER and MCNAP neutronics design calculations for SMART and establish the qualification of the MASTER system.

  10. From the crust to the core of neutron stars on a microscopic basis

    NASA Astrophysics Data System (ADS)

    Baldo, M.; Burgio, G. F.; Centelles, M.; Sharma, B. K.; Viñas, X.

    2014-09-01

    Within a microscopic approach the structure of Neutron Stars is usually studied by modelling the homogeneous nuclear matter of the core by a suitable Equation of State, based on a many-body theory, and the crust by a functional based on a more phenomenological approach. We present the first calculation of Neutron Star overall structure by adopting for the core an Equation of State derived from the Brueckner-Hartree-Fock theory and for the crust, including the pasta phase, an Energy Density Functional based on the same Equation of State, and which is able to describe accurately the binding energy of nuclei throughout the mass table. Comparison with other approaches is discussed. The relevance of the crust Equation of State for the Neutron Star radius is particularly emphasised.

  11. Energy Efficient Engine core design and performance report

    NASA Technical Reports Server (NTRS)

    Stearns, E. Marshall

    1982-01-01

    The Energy Efficient Engine (E3) is a NASA program to develop fuel saving technology for future large transport aircraft engines. Testing of the General Electric E3 core showed that the core component performance and core system performance necessary to meet the program goals can be achieved. The E3 core design and test results are described.

  12. Design of a transportable high efficiency fast neutron spectrometer

    NASA Astrophysics Data System (ADS)

    Roecker, C.; Bernstein, A.; Bowden, N. S.; Cabrera-Palmer, B.; Dazeley, S.; Gerling, M.; Marleau, P.; Sweany, M. D.; Vetter, K.

    2016-08-01

    A transportable fast neutron detection system has been designed and constructed for measuring neutron energy spectra and flux ranging from tens to hundreds of MeV. The transportability of the spectrometer reduces the detector-related systematic bias between different neutron spectra and flux measurements, which allows for the comparison of measurements above or below ground. The spectrometer will measure neutron fluxes that are of prohibitively low intensity compared to the site-specific background rates targeted by other transportable fast neutron detection systems. To measure low intensity high-energy neutron fluxes, a conventional capture-gating technique is used for measuring neutron energies above 20 MeV and a novel multiplicity technique is used for measuring neutron energies above 100 MeV. The spectrometer is composed of two Gd containing plastic scintillator detectors arranged around a lead spallation target. To calibrate and characterize the position dependent response of the spectrometer, a Monte Carlo model was developed and used in conjunction with experimental data from gamma ray sources. Multiplicity event identification algorithms were developed and used with a Cf-252 neutron multiplicity source to validate the Monte Carlo model Gd concentration and secondary neutron capture efficiency. The validated Monte Carlo model was used to predict an effective area for the multiplicity and capture gating analyses. For incident neutron energies between 100 MeV and 1000 MeV with an isotropic angular distribution, the multiplicity analysis predicted an effective area of 500 cm2 rising to 5000 cm2. For neutron energies above 20 MeV, the capture-gating analysis predicted an effective area between 1800 cm2 and 2500 cm2. The multiplicity mode was found to be sensitive to the incident neutron angular distribution.

  13. Characterization and quantification of an in-core neutron irradiation facility at a TRIGA II research reactor

    NASA Astrophysics Data System (ADS)

    Aghara, Sukesh; Charlton, William

    2006-07-01

    Experiments have been performed to characterize the neutron environment at an in-core TRIGA type nuclear research reactor. Steady-state thermal and epithermal neutron environment testing is important for many applications including, materials, electronics and biological cells. A well characterized neutron environment at a research reactor, including energy spectrum and spatial distribution, can be useful to many research communities and for educational research. This paper describes the characterization process and an application of exposing electronics to high neutron fluence.

  14. Maximum mass of neutron stars with quark matter core

    SciTech Connect

    Takatsuka, Tatsuyuki; Hatsuda, Tetsuo; Masuda, Kota

    2012-11-12

    We propose a new strategy to construct the equation of state (EOS) for neutron stars (NSs) with hadron-quark (H-Q) phase transition, by considering three density-regions. We supplement the EOS at H-Q region, very uncertain due to the confinement-deconfinement problems, by sandwitching in between and matching to the relatively 'well known' EOSs, i.e., the EOS at lower densities (H-phase up to several times nuclear density, calculated from a G-matrix approach) and that at ultra high densities (Q-phase, form a view of asymptotic freedom). Here, as a first step, we try a simple case and discuss the maximum mass of NSs.

  15. Learn from the Core--Design from the Core

    ERIC Educational Resources Information Center

    Ockerse, Thomas

    2012-01-01

    The current objective, object-oriented approach to design is questioned along with design education viewed as a job-oriented endeavor. Instead relational knowledge and experience in a holistic sense, both tacit and explicit, are valued along with an appreciation of the unique character of the student. A new paradigm for design education is…

  16. Characterization of Neutron Fields in the Experimental Fast Reactor Joyo Mk-Iii Core

    NASA Astrophysics Data System (ADS)

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

    2009-08-01

    In 2003, Joyo MK-III core was upgraded to increase the irradiation testing capability. This paper describes the details of distributions of neutron flux and reaction rate in the MK-III core that was measured by characterization tests during the first two operating cycles. The calculation accuracy of the core management codes HESTIA, TORT and MCNP, was also evaluated by the measured data. The calculated fission rates of 235U by HESTIA agreed well with the measured one within approximately 4% in the fuel region. MCNP could simulate within 6% in the central non-fuel irradiation test subassembly and the radial reflector region, while large discrepancies were obtained in TORT results. Hence, the precise geometry model was effective in evaluating the neutron spectrum and the flux at such locations.

  17. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    SciTech Connect

    Schulz, K.C.; Yahr, G.T.

    1995-08-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in K{sub Q} due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail.

  18. CORE-COLLAPSE SUPERNOVA EQUATIONS OF STATE BASED ON NEUTRON STAR OBSERVATIONS

    SciTech Connect

    Steiner, A. W.; Hempel, M.; Fischer, T.

    2013-09-01

    Many of the currently available equations of state for core-collapse supernova simulations give large neutron star radii and do not provide large enough neutron star masses, both of which are inconsistent with some recent neutron star observations. In addition, one of the critical uncertainties in the nucleon-nucleon interaction, the nuclear symmetry energy, is not fully explored by the currently available equations of state. In this article, we construct two new equations of state which match recent neutron star observations and provide more flexibility in studying the dependence on nuclear matter properties. The equations of state are also provided in tabular form, covering a wide range in density, temperature, and asymmetry, suitable for astrophysical simulations. These new equations of state are implemented into our spherically symmetric core-collapse supernova model, which is based on general relativistic radiation hydrodynamics with three-flavor Boltzmann neutrino transport. The results are compared with commonly used equations of state in supernova simulations of 11.2 and 40 M{sub Sun} progenitors. We consider only equations of state which are fitted to nuclear binding energies and other experimental and observational constraints. We find that central densities at bounce are weakly correlated with L and that there is a moderate influence of the symmetry energy on the evolution of the electron fraction. The new models also obey the previously observed correlation between the time to black hole formation and the maximum mass of an s = 4 neutron star.

  19. Core refueling subsystem design description. Revision 1

    SciTech Connect

    Anderson, J.K.; Harvey, E.C.

    1987-07-01

    The Core Refueling Subsystem of the Fuel Handling and Storage System provides the mechanisms and tools necessary for the removal and replacement of the hexagonal elements which comprise the reactor core. The Core Refueling Subsystem is not "safety-related." The Core Refueling Subsystem equipment is used to prepare the plant for element removal and replacement, install the machines which handle the elements, maintain control of air inleakage and radiation release, transport the elements between the core and storage, and control the automatic and manual operations of the machines. Much of the element handling is performed inside the vessel, and the entire exchange of elements between storage and core is performed with the elements in a helium atmosphere. The core refueling operations are conducted with the reactor module shutdown and the primary coolant pressure slightly subatmospheric. The subsystem is capable of accomplishing the refueling in a reliable manner commensurate with the plant availability requirements.

  20. Microscopic vortex velocity in the inner crust and outer core of neutron stars

    NASA Astrophysics Data System (ADS)

    Gügercinoğlu, Erbil; Alpar, M. Ali

    2016-10-01

    Treatment of the vortex motion in the superfluids of the inner crust and the outer core of neutron stars is a key ingredient in modelling a number of pulsar phenomena, including glitches and magnetic field evolution. After recalculating the microscopic vortex velocity in the inner crust, we evaluate the velocity for the vortices in the outer core for the first time. The vortex motion between pinning sites is found to be substantially faster in the inner crust than in the outer core, v_0^crust ˜ 107{ cm s^{-1}} ≫ v_0^core ˜ 1{ cm s^{-1}}. One immediate result is that vortex creep is always in the nonlinear regime in the outer core in contrast to the inner crust, where both nonlinear and linear regimes of vortex creep are possible. Other implications for pulsar glitches and magnetic field evolution are also presented.

  1. A demonstration of a whole core neutron transport method in a gas cooled reactor

    SciTech Connect

    Connolly, K. J.; Rahnema, F.

    2013-07-01

    This paper illustrates a capability of the whole core transport method COMET. Building on previous works which demonstrated the accuracy of the method, this work serves to emphasize the robust capability of the method while also accentuating its efficiency. A set of core configurations is presented based on an operating gas-cooled thermal reactor, Japan's HTTR, and COMET determines the eigenvalue and fission density profile throughout each core configuration. Results for core multiplication factors are compared to MCNP for accuracy and also to compare runtimes. In all cases, the values given by COMET differ by those given by MCNP by less than the uncertainty inherent in the stochastic solution procedure, however, COMET requires runtimes shorter on the order of a few hundred. Figures are provided illustrating the whole core fission density profile, with segments of pins explicitly modeled individually, so that pin-level neutron flux behavior can be seen without any approximation due to simplification strategies such as homogenization. (authors)

  2. Optimization of a neutron detector design using adjoint transport simulation

    SciTech Connect

    Yi, C.; Manalo, K.; Huang, M.; Chin, M.; Edgar, C.; Applegate, S.; Sjoden, G.

    2012-07-01

    A synthetic aperture approach has been developed and investigated for Special Nuclear Materials (SNM) detection in vehicles passing a checkpoint at highway speeds. SNM is postulated to be stored in a moving vehicle and detector assemblies are placed on the road-side or in chambers embedded below the road surface. Neutron and gamma spectral awareness is important for the detector assembly design besides high efficiencies, so that different SNMs can be detected and identified with various possible shielding settings. The detector assembly design is composed of a CsI gamma-ray detector block and five neutron detector blocks, with peak efficiencies targeting different energy ranges determined by adjoint simulations. In this study, formulations are derived using adjoint transport simulations to estimate detector efficiencies. The formulations is applied to investigate several neutron detector designs for Block IV, which has its peak efficiency in the thermal range, and Block V, designed to maximize the total neutron counts over the entire energy spectrum. Other Blocks detect different neutron energies. All five neutron detector blocks and the gamma-ray block are assembled in both MCNP and deterministic simulation models, with detector responses calculated to validate the fully assembled design using a 30-group library. The simulation results show that the 30-group library, collapsed from an 80-group library using an adjoint-weighting approach with the YGROUP code, significantly reduced the computational cost while maintaining accuracy. (authors)

  3. Advanced BWR core component designs and the implications for SFD analysis

    SciTech Connect

    Ott, L.J.

    1997-02-01

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B{sub 4}C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities.

  4. System Design for a Nuclear Electric Spacecraft Utilizing Out-of-core Thermionic Conversion

    NASA Technical Reports Server (NTRS)

    Estabrook, W. C.; Phillips, W. M.; Hsieh, T.

    1976-01-01

    Basic guidelines are presented for a nuclear space power system which utilizes heat pipes to transport thermal power from a fast nuclear reactor to an out of core thermionic converter array. Design parameters are discussed for the nuclear reactor, heat pipes, thermionic converters, shields (neutron and gamma), waste heat rejection systems, and the electrical bus bar-cable system required to transport the high current/low voltage power to the processing equipment. Dimensions are compatible with shuttle payload bay constraints.

  5. How useful is neutron diffusion theory for nuclear rocket engine design

    SciTech Connect

    Hilsmeier, T.A.; Aithal, S.M.; Aldemir, T. )

    1992-01-01

    Correct modeling of neutron leakage and geometry effects is important in the design of a nuclear rocket engine because of the need for small reactor cores in space applications. In principle, there are generalized procedures that can account for these effects in a reliable manner (e.g., a three-dimensional, continuous-energy Monte Carlo calculation with all core components explicitly modeled). However, these generalized procedures are not usually suitable for parametric design studies because of the long computational times required, and the feasibility of using faster running, more approrimate neutronic modeling approaches needs to be investigated. Faster running neutronic models are also needed for simulator development to assess the engine performance during startup and power level changes. This paper investigates the potential of the few-group diffusion approach for nuclear rocket engine core design and optimization by comparing the k[sub eff] and power distributions obtained by the MCNP code against those obtained from the LEOPARD and 2DB codes for the particle bed reactor (PBR) concept described. The PBRs have been identified as one of the two near-term options for nuclear thermal propulsion by the joint National Aeronautics and Space Administration (NASA)/US Department of Energy/US Department of Defense program that was recently set up at the NASA Lewis Research Center to develop a flight-rated nuclear rocket engine by the 2020s.

  6. Advanced Microstructured Semiconductor Neutron Detectors: Design, Fabrication, and Performance

    NASA Astrophysics Data System (ADS)

    Bellinger, Steven Lawrence

    The microstructured semiconductor neutron detector (MSND) was investigated and previous designs were improved and optimized. In the present work, fabrication techniques have been refined and improved to produce three-dimensional microstructured semiconductor neutron detectors with reduced leakage current, reduced capacitance, highly anisotropic deep etched trenches, and increased signal-to-noise ratios. As a result of these improvements, new MSND detection systems function with better gamma-ray discrimination and are easier to fabricate than previous designs. In addition to the microstructured diode fabrication improvement, a superior batch processing backfill-method for 6LiF neutron reactive material, resulting in a nearly-solid backfill, was developed. This method incorporates a LiF nano-sizing process and a centrifugal batch process for backfilling the nanoparticle LiF material. To better transition the MSND detector to commercialization, the fabrication process was studied and enhanced to better facilitate low cost and batch process MSND production. The research and development of the MSND technology described in this work includes fabrication of variant microstructured diode designs, which have been simulated through MSND physics models to predict performance and neutron detection efficiency, and testing the operational performance of these designs in regards to neutron detection efficiency, gamma-ray rejection, and silicon fabrication methodology. The highest thermal-neutron detection efficiency reported to date for a solid-state semiconductor detector is presented in this work. MSNDs show excellent neutron to gamma-ray (n/γ) rejection ratios, which are on the order of 106, without significant loss in thermal-neutron detection efficiency. Individually, the MSND is intrinsically highly sensitive to thermal neutrons, but not extrinsically sensitive because of their small size. To improve upon this, individual MSNDs were tiled together into a 6x6-element array

  7. Core-coupled states and split proton-neutron quasiparticle multiplets in 122-126Ag

    NASA Astrophysics Data System (ADS)

    Lalkovski, S.; Bruce, A. M.; Jungclaus, A.; Górska, M.; Pfützner, M.; Cáceres, L.; Naqvi, F.; Pietri, S.; Podolyák, Zs.; Simpson, G. S.; Andgren, K.; Bednarczyk, P.; Beck, T.; Benlliure, J.; Benzoni, G.; Casarejos, E.; Cederwall, B.; Crespi, F. C. L.; Cuenca-García, J. J.; Cullen, I. J.; Denis Bacelar, A. M.; Detistov, P.; Doornenbal, P.; Farrelly, G. F.; Garnsworthy, A. B.; Geissel, H.; Gelletly, W.; Gerl, J.; Grebosz, J.; Hadinia, B.; Hellström, M.; Hinke, C.; Hoischen, R.; Ilie, G.; Jaworski, G.; Jolie, J.; Khaplanov, A.; Kisyov, S.; Kmiecik, M.; Kojouharov, I.; Kumar, R.; Kurz, N.; Maj, A.; Mandal, S.; Modamio, V.; Montes, F.; Myalski, S.; Palacz, M.; Prokopowicz, W.; Reiter, P.; Regan, P. H.; Rudolph, D.; Schaffner, H.; Sohler, D.; Steer, S. J.; Tashenov, S.; Walker, J.; Walker, P. M.; Weick, H.; Werner-Malento, E.; Wieland, O.; Wollersheim, H. J.; Zhekova, M.

    2013-03-01

    Neutron-rich silver isotopes were populated in the fragmentation of a 136Xe beam and the relativistic fission of 238U. The fragments were mass analyzed with the GSI Fragment Separator and subsequently implanted into a passive stopper. Isomeric transitions were detected by 105 high-purity germanium detectors. Eight isomeric states were observed in 122-126Ag nuclei. The level schemes of 122,123,125Ag were revised and extended with isomeric transitions being observed for the first time. The excited states in the odd-mass silver isotopes are interpreted as core-coupled states. The isomeric states in the even-mass silver isotopes are discussed in the framework of the proton-neutron split multiplets. The results of shell-model calculations, performed for the most neutron-rich silver nuclei are compared to the experimental data.

  8. Neutron Tube Design Study for Boron Neutron Capture TherapyApplication

    SciTech Connect

    Verbeke, J.M.; Lee, Y.; Leung, K.N.; Vujic, J.; Williams, M.D.; Wu, L.K.; Zahir, N.

    1998-01-04

    Radio-frequency (RF) driven ion sources are being developed in Lawrence Berkeley National Laboratory (LBNL) for sealed-accelerator-tube neutron generator application. By using a 5-cm-diameter RF-driven multicusp source H{sup +} yields over 95% have been achieved. These experimental findings will enable one to develop compact neutron generators based on the D-D or D-T fusion reactions. In this new neutron generator, the ion source, the accelerator and the target are all housed in a sealed metal container without external pumping. Recent moderator design simulation studies have shown that 14 MeV neutrons could be moderated to therapeutically useful energy ranges for boron neutron capture therapy (BNCT). The dose near the center of the brain with optimized moderators is about 65% higher than the dose obtained from a typical neutron spectrum produced by the Brookhaven Medical Research Reactor (BMRR), and is comparable to the dose obtained by other accelerator-based neutron sources. With a 120 keV and 1 A deuteron beam, a treatment time of {approx}35 minutes is estimated for BNCT.

  9. Code System for 2-Group, 3D Neutronic Kinetics Calculations Coupled to Core Thermal Hydraulics.

    2000-05-12

    Version 00 QUARK is a combined computer program comprising a revised version of the QUANDRY three-dimensional, two-group neutron kinetics code and an upgraded version of the COBRA transient core analysis code (COBRA-EN). Starting from either a critical steady-state (k-effective or critical dilute Boron problem) or a subcritical steady-state (fixed source problem) in a PWR plant, the code allows one to simulate the neutronic and thermal-hydraulic core transient response to reactivity accidents initiated both inside themore » vessel (such as a control rod ejection) and outside the vessel (such as the sudden change of the Boron concentration in the coolant). QUARK output can be used as input to PSR-470/NORMA-FP to perform a subchannel analysis from converged coarse-mesh nodal solutions.« less

  10. Introduction to Neutron Coincidence Counter Design Based on Boron-10

    SciTech Connect

    Kouzes, Richard T.; Ely, James H.; Lintereur, Azaree T.; Siciliano, Edward R.

    2012-01-22

    The Department of Energy Office of Nonproliferation Policy (NA-241) is supporting the project 'Coincidence Counting With Boron-Based Alternative Neutron Detection Technology' at Pacific Northwest National Laboratory (PNNL) for development of an alternative neutron coincidence counter. The goal of this project is ultimately to design, build and demonstrate a boron-lined proportional tube based alternative system in the configuration of a coincidence counter. This report, providing background information for this project, is the deliverable under Task 1 of the project.

  11. Influence of the N=50 neutron core on dipole excitations in 87Rb

    NASA Astrophysics Data System (ADS)

    Käubler, L.; Schilling, K. D.; Schwengner, R.; Dönau, F.; Grosse, E.; Belic, D.; von Brentano, P.; Bubner, M.; Fransen, C.; Grinberg, M.; Kneissl, U.; Kohstall, C.; Linnemann, A.; Matschinsky, P.; Nord, A.; Pietralla, N.; Pitz, H. H.; Scheck, M.; Stedile, F.; Werner, V.

    2002-05-01

    Dipole excitations in the semimagic N=50 nucleus 87Rb were investigated at the Stuttgart Dynamitron facility using bremsstrahlung with an end-point energy of 4.0 MeV. The widths Γ or the reduced excitation probabilities B(Π1)↑ of 18 states were determined for the first time. The magnetic dipole excitations are well reproduced in the framework of the shell model, however, these calculations cannot describe the observed electric dipole excitations. The 1/2+ state at 3060 keV is proposed to be the weak coupling of an f5/2 proton hole to the 3- octupole vibrational state in the N=50 core 88Sr. The relatively strong E1 transition from that state to the ground state is explained as mainly the neutron h11/2-->g9/2 transition. The breakup of the N=50 core and neutron excitations into the h11/2 shell are essential to describe electric dipole excitations, but neutron-core excitations do not play an important role for the structure of magnetic dipole excitations.

  12. Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core

    SciTech Connect

    Sterbentz, James W

    2007-05-01

    A new non-TRISO fuel and clad design concept is proposed for the prismatic, heliumcooled Very High Temperature Reactor core. The new concept could substantially reduce the current 10-20 wt% TRISO uranium enrichments down to 4-6 wt% for both initial and reload cores. The proposed fuel form would be a high-temperature, high-density uranium ceramic, for example UO2, configured into very small diameter cylindrical rods. The small diameter fuel rods significantly increase core reactivity through improved neutron moderation and fuel lumping. Although a high-temperature clad system for the concept remains to be developed, recent success in tube fabrication and preliminary irradiation testing of silicon carbide (SiC) cladding for light water reactor applications offers good potential for this application, and for future development of other carbide clad designs. A high-temperature ceramic fuel, together with a high-temperature clad material, could also lead to higher thermal safety margins during both normal and transient reactor conditions relative to TRISO fuel. The calculated neutronic results show that the lowenrichment, small diameter fuel rods and low thermal neutron absorbing clad retain the strong negative Doppler fuel temperature coefficient of reactivity that ensures inherent safe operation of the VHTR, and depletion studies demonstrate that an 18-month power cycle can be achieved with the lower enrichment fuel.

  13. Accelerator shield design of KIPT neutron source facility

    SciTech Connect

    Zhong, Z.; Gohar, Y.

    2013-07-01

    Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the design development of a neutron source facility at KIPT utilizing an electron-accelerator-driven subcritical assembly. Electron beam power is 100 kW, using 100 MeV electrons. The facility is designed to perform basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The biological shield of the accelerator building is designed to reduce the biological dose to less than 0.5-mrem/hr during operation. The main source of the biological dose is the photons and the neutrons generated by interactions of leaked electrons from the electron gun and accelerator sections with the surrounding concrete and accelerator materials. The Monte Carlo code MCNPX serves as the calculation tool for the shield design, due to its capability to transport electrons, photons, and neutrons coupled problems. The direct photon dose can be tallied by MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is less than 0.01 neutron per electron. This causes difficulties for Monte Carlo analyses and consumes tremendous computation time for tallying with acceptable statistics the neutron dose outside the shield boundary. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were developed for the study. The generated neutrons are banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron and secondary photon doses. The weight windows variance reduction technique is utilized for both neutron and photon dose calculations. Two shielding materials, i.e., heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total

  14. Design and simulations of the neutron dump for the back-streaming white neutron beam at CSNS

    NASA Astrophysics Data System (ADS)

    Zhang, L. Y.; Jing, H. T.; Tang, J. Y.; Wang, X. Q.

    2016-10-01

    For nuclear data measurements with a white neutron source, to control the background at the detector is a key issue. The neutron dump which locates at the end of the white neutron beam line at CSNS has a very important impact to the neutron and gamma backgrounds in the endstation. A sophisticated neutron dump was designed to reduce the backgrounds to the level of about 10-8 relative to the neutron flux. In this paper, the method to suppress both neutron and gamma backgrounds near a white-spectrum neutron dump is introduced. The optimized geometry structure and materials of the dump are described, and the neutron and gamma space distributions have been calculated by using the FLUKA code for different operation settings which are defined by beam spots of Φ30 mm, Φ60 mm and 90 mm×90 mm, respectively.

  15. Design of a boron neutron capture enhanced fast neutron therapy assembly

    SciTech Connect

    Wang, Zhonglu

    2006-12-01

    The use of boron neutron capture to boost tumor dose in fast neutron therapy has been investigated at several fast neutron therapy centers worldwide. This treatment is termed boron neutron capture enhanced fast neutron therapy (BNCEFNT). It is a combination of boron neutron capture therapy (BNCT) and fast neutron therapy (FNT). It is believed that BNCEFNT may be useful in the treatment of some radioresistant brain tumors, such as glioblastoma multiform (GBM). A boron neutron capture enhanced fast neutron therapy assembly has been designed for the Fermilab Neutron Therapy Facility (NTF). This assembly uses a tungsten filter and collimator near the patient's head, with a graphite reflector surrounding the head to significantly increase the dose due to boron neutron capture reactions. The assembly was designed using Monte Carlo radiation transport code MCNP version 5 for a standard 20x20 cm2 treatment beam. The calculated boron dose enhancement at 5.7-cm depth in a water-filled head phantom in the assembly with a 5x5 cm2 collimation was 21.9% per 100-ppm 10B for a 5.0-cm tungsten filter and 29.8% for a 8.5-cm tungsten filter. The corresponding dose rate for the 5.0-cm and 8.5-cm thick filters were 0.221 and 0.127 Gy/min, respectively; about 48.5% and 27.9% of the dose rate of the standard 10x10 cm2 fast neutron treatment beam. To validate the design calculations, a simplified BNCEFNT assembly was built using four lead bricks to form a 5x5 cm2 collimator. Five 1.0-cm thick 20x20 cm2 tungsten plates were used to obtain different filter thicknesses and graphite bricks/blocks were used to form a reflector. Measurements of the dose enhancement of the simplified assembly in a water-filled head phantom were performed using a pair of tissue-equivalent ion chambers. One of the ion chambers is loaded with 1000-ppm natural boron (184-ppm 10B) to measure dose due to boron neutron capture. The measured

  16. A low resolution model for the chromatin core particle by neutron scattering

    PubMed Central

    Suau, Pedro; Kneale, G.Geoff; Braddock, Gordon W.; Baldwin, John P.; Bradbury, E.Morton

    1977-01-01

    Neutron scattering studies have been applied to chromatin core particles in solution, using the contrast variation technique. On the basis of the contrast dependance of the radius of gyration and the radial distribution function it is shown that the core particle consists of a core containing most of the histone around which is wound the DNA helix,following a path with a mean radius of 4.5 nm,in association with a small proportion of the histones. Separation of the shape from the internal structure, followed by model calculations shows that the overall shape of the particle is that of a flat cylinder with dimensions ca. 11×11×6 nm. Further details of the precise folding of the DNA cannot be deduced from the data, but detailed model calculations support concurrent results from crystallographic studies25. Images PMID:593885

  17. Conceptual design of a polarized 3He neutron spin filter for polarized neutron spectrometer POLANO at J-PARC

    NASA Astrophysics Data System (ADS)

    Ino, T.; Ohoyama, K.; Yokoo, T.; Itoh, S.; Ohkawara, M.; Kira, H.; Hayashida, H.; Sakai, K.; Hiroi, K.; Oku, T.; Kakurai, K.; Chang, L. J.

    2016-04-01

    A 3He neutron spin filter (NSF) has been designed for a new polarized neutron chopper spectrometer called the Polarization Analysis Neutron Spectrometer with Correlation Method (POLANO) at the Materials and Life Science Experimental Facility of the Japan Proton Accelerator Research Complex. It is designed to fit in a limited space on the spectrometer as an initial neutron beam polarizer and is polarized in situ by spin exchange optical pumping. This will be the first generation 3He NSF on POLANO, and a polarized neutron beam up to 100 meV with a diameter of 50 mm will be available for research on magnetism, hydrogen materials, and strongly correlated electron systems.

  18. Neutron core excitations in the N=126 nuclide {sup 210}Po

    SciTech Connect

    Dracoulis, G. D.; Lane, G. J.; Davidson, P. M.; Kibedi, T.; Nieminen, P.; Maier, K. H.; Watanabe, H.; Byrne, A. P.; Wilson, A. N.

    2008-03-15

    Excited states above the 16{sup +} isomer in {sup 210}Po have been identified using time-correlated {gamma}-ray spectroscopy techniques and the {sup 204}Hg({sup 13}C,3n{alpha}){sup 210}Po reaction. States up to {approx}27({Dirac_h}/2{pi}) have been identified, including an isomer at 8074 keV with a mean life of 13(2) ns. Among the new states, a candidate for the 17{sup +} state obtained from maximal coupling of the {pi}[h{sub 9/2}i{sub 13/2}]{sub 11{sup -}} valence proton configuration and the {nu}[p{sub 1/2}{sup -1}i{sub 11/2}]{sub 6{sup -}} neutron core excitation has been identified. This and other results are compared with semiempirical shell-model calculations that predict that single core excitations from the i{sub 13/2} neutron orbital and double core excitations out of the p{sub 1/2} and f{sub 5/2} orbitals, populating the g{sub 9/2},i{sub 11/2}, and j{sub 15/2} orbitals above the N=126 shell, will compete in energy. Good agreement is obtained for the lower states but there are systematic discrepancies at high spins including the absence of states that are calculated to lie low in the spectrum, implying uncertainties for configurations associated either with the i{sub 13/2} neutron hole or double core excitations.

  19. CORE ANALYSIS, DESIGN AND OPTIMIZATION OF A DEEP-BURN PEBBLE BED REACTOR

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-05-01

    Achieving a high burnup in the Deep-Burn pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. To investigate the aforementioned effects a code system using existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

  20. Small Angle Neutron-Scattering Studies of the Core Structure of Intact Neurosecretory Vesicles.

    NASA Astrophysics Data System (ADS)

    Krueger, Susan Takacs

    Small angle neutron scattering (SANS) was used to study the state of the dense cores within intact neurosecretory vesicles. These vesicles transport the neurophysin proteins, along with their associated hormones, oxytocin or vasopressin, from the posterior pituitary gland to the bloodstream, where the entire vesicle contents are released. Knowledge of the vesicle core structure is important in developing an understanding of this release mechanism. Since the core constituents exist in a dense state at concentrations which cannot be reproduced (in solution) in the laboratory, a new method was developed to determine the core structure from SANS experiments performed on intact neurosecretory vesicles. These studies were complemented by biochemical assays performed to determine the role, if any, played by phospholipids in the interactions between the core constituents. H_2O/D_2 O ratio in the solvent can be adjusted, using the method of contrast variation, such that the scattering due to the vesicle membranes is minimized, thus emphasizing the scattering originating from the cores. The applicability of this method for examining the interior of biological vesicles was tested by performing an initial study on human red blood cells, which are similar in structure to other biological vesicles. Changes in intermolecular hemoglobin interactions, occurring when the ionic strength of the solvent was varied or when the cells were deoxygenated, were examined. The results agreed with those expected for dense protein solutions, indicating that the method developed was suitable for the study of hemoglobin within the cells. Similar SANS studies were then performed on intact neurosecretory vesicles. The experimental results were inconsistent with model calculations which assumed that the cores consisted of small, densely-packed particles or large, globular aggregates. Although a unique model could not be determined, the data suggest that the core constituents form long aggregates of

  1. Neutron Imaging of Rapid Water Imbibition in Fractured Sedimentary Rock Cores

    NASA Astrophysics Data System (ADS)

    Cheng, Chu-Lin; Perfect, Edmund; Donnelly, Brendan; Bilheux, Hassina; Tremsin, Anton; McKay, Larry; Distefano, Victoria; Cai, Jianchao; Santodonato, Lou

    2015-03-01

    Advances in nondestructive testing methods, such as neutron, nuclear magnetic resonance, and x-ray imaging, have significantly improved experimental capabilities to visualize fracture flow in various important fossil energy contexts, e.g. enhanced oil recovery and shale gas. We present a theoretical framework for predicting the rapid movement of water into air-filled fractures within a porous medium based on early-time capillary dynamics and spreading over rough fracture surfaces. The theory permits estimation of sorptivity values for the matrix and fracture zone, as well as a dispersion parameter which quantifies the extent of spreading of the wetting front. Dynamic neutron imaging of water imbibition in unsaturated fractured Berea sandstone cores was employed to evaluate the proposed model. The experiments were conducted at the Neutron Imaging Prototype Facility at Oak Ridge National Laboratory. Water uptake into both the matrix and fracture zone exhibited square-root-of-time behavior. Both theory and neutron imaging data indicated that fractures significantly increase imbibition in unsaturated sedimentary rock by capillary action and surface spreading on rough fracture faces. Fractures also increased the dispersion of the wetting front.

  2. Ultracold neutron source at the PULSTAR reactor: Engineering design and cryogenic testing

    NASA Astrophysics Data System (ADS)

    Korobkina, E.; Medlin, G.; Wehring, B.; Hawari, A. I.; Huffman, P. R.; Young, A. R.; Beaumont, B.; Palmquist, G.

    2014-12-01

    Construction is completed and commissioning is in progress for an ultracold neutron (UCN) source at the PULSTAR reactor on the campus of North Carolina State University. The source utilizes two stages of neutron moderation, one in heavy water at room temperature and the other in solid methane at ~ 40 K, followed by a converter stage, solid deuterium at 5 K, that allows a single down scattering of cold neutrons to provide UCN. The UCN source rolls into the thermal column enclosure of the PULSTAR reactor, where neutrons will be delivered from a bare face of the reactor core by streaming through a graphite-lined assembly. The source infrastructure, i.e., graphite-lined assembly, heavy-water system, gas handling system, and helium liquefier cooling system, has been tested and all systems operate as predicted. The research program being considered for the PULSTAR UCN source includes the physics of UCN production, fundamental particle physics, and material surface studies of nanolayers containing hydrogen. In the present paper we report details of the engineering and cryogenic design of the facility as well as results of critical commissioning tests without neutrons.

  3. Pre-conceptual design study of ASTRID core

    SciTech Connect

    Varaine, F.; Marsault, P.; Chenaud, M. S.; Bernardin, B.; Conti, A.; Sciora, P.; Venard, C.; Fontaine, B.; Devictor, N.; Martin, L.; Scholer, A. C.; Verrier, D.

    2012-07-01

    In the framework of the ASTRID project at CEA, core design studies are performed at CEA with the AREVA and EDF support. At the stage of the project, pre-conceptual design studies are conducted in accordance with GEN IV reactors criteria, in particularly for safety improvements. An improved safety for a sodium cooled reactor requires revisiting many aspects of the design and is a rather lengthy process in current design approach. Two types of cores are under evaluation, one classical derived from the SFR V2B and one more challenging called CFV (low void effect core) with a large gain on the sodium void effect. The SFR V2b core have the following specifications: a very low burn-up reactivity swing (due to a small cycle reactivity loss) and a reduced sodium void effect with regard to past designs such as the EFR (around 2$ minus). Its performances are an average burn-up of 100 GWd/t, and an internal conversion ratio equal to one given a very good behavior of this core during a control rod withdrawal transient). The CFV with its specific design offers a negative sodium void worth while maintaining core performances. In accordance of ASTRID needs for demonstration those cores are 1500 MWth power (600 MWe). This paper will focus on the CFV pre-conceptual design of the core and S/A, and the performances in terms of safety will be evaluated on different transient scenario like ULOF, in order to assess its intrinsic behavior compared to a more classical design like V2B core. The gap in term of margin to a severe accident due to a loss of flow initiator underlines the potential capability of this type of core to enhance prevention of severe accident in accordance to safety demonstration. (authors)

  4. Neutronics Design and Fuel Cycle Analysis of a High Conversion BWR with Pu-Th Fuel

    SciTech Connect

    Xu, Yunlin; Downar, T.J.; Takahashi, H.; Rohatgi, U.S.

    2002-07-01

    As part of the U.S. Department of Energy's (DOE) Nuclear Energy Research Initiative (NERI), a 'Generation IV' high conversion Boiling Water Reactor design is being investigated at Purdue University and Brookhaven National Laboratory. One of the primary innovative design features of the core proposed here is the use of Thorium as fertile material. In addition to the advantageous nonproliferation and waste characteristics of thorium fuel cycles, the use of thorium is particularly important in a tight pitch, high conversion lattice in order to insure a negative void coefficient throughout the operating life of the reactor. The principal design objective of a high conversion light water reactor is to substantially increase the conversion ratio (fissile atoms produced per fissile atoms consumed) of the reactor without compromising the safety performance of the plant. Since existing LWRs have a relatively low conversion ratio they require relatively frequent refueling which limits the economic efficiency of the plant. Also, the high volume of spent fuel can pose a burden for waste storage and the accumulation of plutonium in the uranium fuel cycle can become a materials proliferation issue. The development of Fast Breeder Reactors (FBR) as an alternative technology to alleviate some of these concerns has been delayed for various reasons. An intermediate solution has been to examine tight pitch light water reactors which can provide significant improvements in the fuel cycle performance of the existing LWRs by taking advantage of the increased conversion ratios from the harder neutron spectrum in the tight pitch lattice, as well as the by taking advantage of the waste and nonproliferation benefits of the thorium fuel cycle. Several High Conversion BWR designs have been proposed by researchers in Japan and elsewhere during the past several years. One of the more promising HCR designs is the Reduced Moderation Water Reactor (RMWR) proposed by JAERI [1]. Their design was

  5. Core Design Issues of the Supercritcal Water Fast Reactor

    NASA Astrophysics Data System (ADS)

    Mori, Magnus; Rineiski, Andrei; Maschek, Werner; Sinitsa, Valentin

    2006-04-01

    The Super Critical water Fast Reactor is a Generation IV reactor concept, which presents new and challenging design issues. A correct estimation of the void effect for this water-cooled pressurized system is of fundamental importance to assess its theoretical feasibility. Hence, in this work an overview of the void effect analysis is shown together with the resulting core design issues. The effect of the application of different cross section libraries and models on the core design is also treated.

  6. Neutronics analyses in support of the conceptual design of the MAPS NTP reactor

    SciTech Connect

    Raepsaet, X.; Lenain, R.

    1996-03-01

    Within the framework of the French nuclear thermal propulsion program called MAPS (Lenain 1996), several neutronics studies and analyses were performed. The aim was to determine the basic design features of a reactor based on the Pebble Bed Reactor concept (Powell 1985) and needing minimum technological developments. In the concern to further enhance the reactor safety posture and to maintain a minimum engine mass breakdown, a beryllium moderated/reflected reactor using highly enriched UO{sub 2} or UC{sub 2} as fuel has been designed with a mean hydrogen core outlet temperature of 2200 K (theoretical ISP of 859 s). The objective of this paper is to give a detailed neutronics analysis of the MAPS reactor. {copyright} {ital 1996 American Institute of Physics.}

  7. Core and Refueling Design Studies for the Advanced High Temperature Reactor

    SciTech Connect

    Holcomb, David Eugene; Ilas, Dan; Varma, Venugopal Koikal; Cisneros, Anselmo T; Kelly, Ryan P; Gehin, Jess C

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure

  8. Design, synthesis and applications of core-shell, hollow core, and nanorattle multifunctional nanostructures

    NASA Astrophysics Data System (ADS)

    El-Toni, Ahmed Mohamed; Habila, Mohamed A.; Labis, Joselito Puzon; Alothman, Zeid A.; Alhoshan, Mansour; Elzatahry, Ahmed A.; Zhang, Fan

    2016-01-01

    With the evolution of nanoscience and nanotechnology, studies have been focused on manipulating nanoparticle properties through the control of their size, composition, and morphology. As nanomaterial research has progressed, the foremost focus has gradually shifted from synthesis, morphology control, and characterization of properties to the investigation of function and the utility of integrating these materials and chemical sciences with the physical, biological, and medical fields, which therefore necessitates the development of novel materials that are capable of performing multiple tasks and functions. The construction of multifunctional nanomaterials that integrate two or more functions into a single geometry has been achieved through the surface-coating technique, which created a new class of substances designated as core-shell nanoparticles. Core-shell materials have growing and expanding applications due to the multifunctionality that is achieved through the formation of multiple shells as well as the manipulation of core/shell materials. Moreover, core removal from core-shell-based structures offers excellent opportunities to construct multifunctional hollow core architectures that possess huge storage capacities, low densities, and tunable optical properties. Furthermore, the fabrication of nanomaterials that have the combined properties of a core-shell structure with that of a hollow one has resulted in the creation of a new and important class of substances, known as the rattle core-shell nanoparticles, or nanorattles. The design strategies of these new multifunctional nanostructures (core-shell, hollow core, and nanorattle) are discussed in the first part of this review. In the second part, different synthesis and fabrication approaches for multifunctional core-shell, hollow core-shell and rattle core-shell architectures are highlighted. Finally, in the last part of the article, the versatile and diverse applications of these nanoarchitectures in

  9. Design, synthesis and applications of core-shell, hollow core, and nanorattle multifunctional nanostructures.

    PubMed

    El-Toni, Ahmed Mohamed; Habila, Mohamed A; Labis, Joselito Puzon; ALOthman, Zeid A; Alhoshan, Mansour; Elzatahry, Ahmed A; Zhang, Fan

    2016-02-01

    With the evolution of nanoscience and nanotechnology, studies have been focused on manipulating nanoparticle properties through the control of their size, composition, and morphology. As nanomaterial research has progressed, the foremost focus has gradually shifted from synthesis, morphology control, and characterization of properties to the investigation of function and the utility of integrating these materials and chemical sciences with the physical, biological, and medical fields, which therefore necessitates the development of novel materials that are capable of performing multiple tasks and functions. The construction of multifunctional nanomaterials that integrate two or more functions into a single geometry has been achieved through the surface-coating technique, which created a new class of substances designated as core-shell nanoparticles. Core-shell materials have growing and expanding applications due to the multifunctionality that is achieved through the formation of multiple shells as well as the manipulation of core/shell materials. Moreover, core removal from core-shell-based structures offers excellent opportunities to construct multifunctional hollow core architectures that possess huge storage capacities, low densities, and tunable optical properties. Furthermore, the fabrication of nanomaterials that have the combined properties of a core-shell structure with that of a hollow one has resulted in the creation of a new and important class of substances, known as the rattle core-shell nanoparticles, or nanorattles. The design strategies of these new multifunctional nanostructures (core-shell, hollow core, and nanorattle) are discussed in the first part of this review. In the second part, different synthesis and fabrication approaches for multifunctional core-shell, hollow core-shell and rattle core-shell architectures are highlighted. Finally, in the last part of the article, the versatile and diverse applications of these nanoarchitectures in

  10. Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

    SciTech Connect

    Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.; Marshall, Margaret A.

    2014-11-04

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental keff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  11. Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

    DOE PAGES

    Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.; Marshall, Margaret A.

    2014-11-04

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental keff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding asmore » well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

  12. High Flux Isotope Reactor cold neutron source reference design concept

    SciTech Connect

    Selby, D.L.; Lucas, A.T.; Hyman, C.R.

    1998-05-01

    In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

  13. Neutron streaming analysis for shield design of FMIT Facility

    SciTech Connect

    Carter, L.L.

    1980-12-01

    Applications of the Monte Carlo method have been summarized relevant to neutron streaming problems of interest in the shield design for the FMIT Facility. An improved angular biasing method has been implemented to further optimize the calculation of streaming and this method has been applied to calculate streaming within a double bend pipe.

  14. Designing accelerator-based epithermal neutron beams for boron neutron capture therapy

    SciTech Connect

    Bleuel, D.L. |; Donahue, R.J.; Ludewigt, B.A.; Vujic, J.

    1998-09-01

    The {sup 7}Li(p,n){sup 7}Be reaction has been investigated as an accelerator-driven neutron source for proton energies between 2.1 and 2.6 MeV. Epithermal neutron beams shaped by three moderator materials, Al/AlF{sub 3}, {sup 7}LiF, and D{sub 2}O, have been analyzed and their usefulness for boron neutron capture therapy (BNCT) treatments evaluated. Radiation transport through the moderator assembly has been simulated with the Monte Carlo {ital N}-particle code (MCNP). Fluence and dose distributions in a head phantom were calculated using BNCT treatment planning software. Depth-dose distributions and treatment times were studied as a function of proton beam energy and moderator thickness. It was found that an accelerator-based neutron source with Al/AlF{sub 3} or {sup 7}LiF as moderator material can produce depth-dose distributions superior to those calculated for a previously published neutron beam design for the Brookhaven Medical Research Reactor, achieving up to {approximately}50{percent} higher doses near the midline of the brain. For a single beam treatment, a proton beam current of 20 mA, and a {sup 7}LiF moderator, the treatment time was estimated to be about 40 min. The tumor dose deposited at a depth of 8 cm was calculated to be about 21 Gy-Eq. {copyright} {ital 1998 American Association of Physicists in Medicine.}

  15. Overview of the TITAN-2 reversed-field pinch aqueous fusion power core design

    NASA Astrophysics Data System (ADS)

    Wong, C. P. C.; Creedon, R. L.; Grotz, S.; Cheng, E. T.; Sharafat, S.; Cooke, P. I. H.

    1988-03-01

    TITAN-2 is a compact, high power density Reversed-Field Pinch fusion power reactor design based on the aqueous lithium solution fusion power core concept. The selected breeding and structural materials are LiNO3 and 9-C low activation ferritic steel, respectively. TITAN-2 is a viable alternative to the TITAN-1 lithium self-cooled design for the Reversed-Field Pinch reactor to operate at a neutron wall loading of 18 MW/sq m. Submerging the complete fusion power core and the primary loop in a large pool of cool water will minimize the probability of radioactivity release. Since the protection of the large pool integrity is the only requirement for the protection of the public, TITAN-2 is a passive safety assurance design.

  16. Monte Carlo Neutronics and Thermal Hydraulics Analysis of Reactor Cores with Multilevel Grids

    NASA Astrophysics Data System (ADS)

    Bernnat, W.; Mattes, M.; Guilliard, N.; Lapins, J.; Zwermann, W.; Pasichnyk, I.; Velkov, K.

    2014-06-01

    Power reactors are composed of assemblies with fuel pin lattices or other repeated structures with several grid levels, which can be modeled in detail by Monte Carlo neutronics codes such as MCNP6 using corresponding lattice options, even for large cores. Except for fresh cores at beginning of life, there is a varying material distribution due to burnup in the different fuel pins. Additionally, for power states the fuel and moderator temperatures and moderator densities vary according to the power distribution and cooling conditions. Therefore, a coupling of the neutronics code with a thermal hydraulics code is necessary. Depending on the level of detail of the analysis, a very large number of cells with different materials and temperatures must be regarded. The assignment of different material properties to all elements of a multilevel grid is very elaborate and may exceed program limits if the standard input procedure is used. Therefore, an internal assignment is used which overrides uniform input parameters. The temperature dependency of continuous energy cross sections, probability tables for the unresolved resonance region and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. The method is applied with MCNP6 and proven for several full core reactor models. For the coupling of MCNP6 with thermal hydraulics appropriate interfaces were developed for the GRS system code ATHLET for liquid coolant and the IKE thermal hydraulics code ATTICA-3D for gaseous coolant. Examples will be shown for different applications for PWRs with square and hexagonal lattices, fast reactors (SFR) with hexagonal lattices and HTRs with pebble bed and prismatic lattices.

  17. Verification of JUPITER Standard Analysis Method for Upgrading Joyo MK-III Core Design and Management

    NASA Astrophysics Data System (ADS)

    Maeda, Shigetaka; Ito, Chikara; Sekine, Takashi; Aoyama, Takafumi

    In the experimental fast reactor Joyo, loading of irradiation test rigs causes a decrease in excess reactivity because the rigs contain less fissile materials than the driver fuel. In order to carry out duty operation cycles using as many irradiation rigs as possible, it is necessary to upgrade the core performance to increase its excess reactivity and irradiation capacity. Core modification plans have been considered, such as the installation of advanced radial reflectors and reduction of the number of control rods. To implement such core modifications, it is first necessary to improve the prediction accuracy in core design and to optimize safety margins. In the present study, verification of the JUPITER fast reactor standard analysis method was conducted through a comparison between the calculated and the measured Joyo MK-III core characteristics, and it was concluded that the accuracy for a small sodium-cooled fast reactor with a hard neutron spectrum was within 5 % of unity. It was shown that, the performance of the irradiation bed core could be upgraded by the improvement of the prediction accuracy of the core characteristics and optimization of safety margins.

  18. Characterization of core debris/concrete interactions for the Advanced Neutron Source

    SciTech Connect

    Hyman, C.R.; Taleyarkhan, R.P.

    1992-02-01

    This report provides the results of a recent study conducted to explore the molten core/concrete interaction (MCCI) issue for the Advanced Neutron Source (ANS). The need for such a study arises from the potential threats to reactor system integrity posed by MCCI. These threats include direct attack of the concrete basemat of the containment; generation and release of large quantities of gas that can pressurize the containment; the combustion threat of these gases; and the potential generation, release, and transport of radioactive aerosols to the environment.

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

    PubMed

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

    1990-01-01

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

  20. GEANT4 used for neutron beam design of a neutron imaging facility at TRIGA reactor in Morocco

    NASA Astrophysics Data System (ADS)

    Ouardi, A.; Machmach, A.; Alami, R.; Bensitel, A.; Hommada, A.

    2011-09-01

    Neutron imaging has a broad scope of applications and has played a pivotal role in visualizing and quantifying hydrogenous masses in metallic matrices. The field continues to expand into new applications with the installation of new neutron imaging facilities. In this scope, a neutron imaging facility for computed tomography and real-time neutron radiography is currently being developed around 2.0MW TRIGA MARK-II reactor at Maamora Nuclear Research Center in Morocco (Reuscher et al., 1990 [1]; de Menezes et al., 2003 [2]; Deinert et al., 2005 [3]). The neutron imaging facility consists of neutron collimator, real-time neutron imaging system and imaging process systems. In order to reduce the gamma-ray content in the neutron beam, the tangential channel was selected. For power of 250 kW, the corresponding thermal neutron flux measured at the inlet of the tangential channel is around 3×10 11 ncm 2/s. This facility will be based on a conical neutron collimator with two circular diaphragms with diameters of 4 and 2 cm corresponding to L/D-ratio of 165 and 325, respectively. These diaphragms' sizes allow reaching a compromise between good flux and efficient L/D-ratio. Convergent-divergent collimator geometry has been adopted. The beam line consists of a gamma filter, fast neutrons filter, neutron moderator, neutron and gamma shutters, biological shielding around the collimator and several stages of neutron collimator. Monte Carlo calculations by a fully 3D numerical code GEANT4 were used to design the neutron beam line ( http://www.info.cern.ch/asd/geant4/geant4.html[4]). To enhance the neutron thermal beam in terms of quality, several materials, mainly bismuth (Bi) and sapphire (Al 2O 3) were examined as gamma and neutron filters respectively. The GEANT4 simulations showed that the gamma and epithermal and fast neutron could be filtered using the bismuth (Bi) and sapphire (Al 2O 3) filters, respectively. To get a good cadmium ratio, GEANT 4 simulations were used to

  1. De novo design of the hydrophobic cores of proteins.

    PubMed Central

    Desjarlais, J. R.; Handel, T. M.

    1995-01-01

    We have developed and experimentally tested a novel computational approach for the de novo design of hydrophobic cores. A pair of computer programs has been written, the first of which creates a "custom" rotamer library for potential hydrophobic residues, based on the backbone structure of the protein of interest. The second program uses a genetic algorithm to globally optimize for a low energy core sequence and structure, using the custom rotamer library as input. Success of the programs in predicting the sequences of native proteins indicates that they should be effective tools for protein design. Using these programs, we have designed and engineered several variants of the phage 434 cro protein, containing five, seven, or eight sequence changes in the hydrophobic core. As controls, we have produced a variant consisting of a randomly generated core with six sequence changes but equal volume relative to the native core and a variant with a "minimalist" core containing predominantly leucine residues. Two of the designs, including one with eight core sequence changes, have thermal stabilities comparable to the native protein, whereas the third design and the minimalist protein are significantly destabilized. The randomly designed control is completely unfolded under equivalent conditions. These results suggest that rational de novo design of hydrophobic cores is feasible, and stress the importance of specific packing interactions for the stability of proteins. A surprising aspect of the results is that all of the variants display highly cooperative thermal denaturation curves and reasonably dispersed NMR spectra. This suggests that the non-core residues of a protein play a significant role in determining the uniqueness of the folded structure. PMID:8535237

  2. Oak Ridge Spallation Neutron Source (ORSNS) target station design integration

    SciTech Connect

    McManamy, T.; Booth, R.; Cleaves, J.; Gabriel, T.

    1996-06-01

    The conceptual design for a 1- to 3-MW short pulse spallation source with a liquid mercury target has been started recently. The design tools and methods being developed to define requirements, integrate the work, and provide early cost guidance will be presented with a summary of the current target station design status. The initial design point was selected with performance and cost estimate projections by a systems code. This code was developed recently using cost estimates from the Brookhaven Pulsed Spallation Neutron Source study and experience from the Advanced Neutron Source Project`s conceptual design. It will be updated and improved as the design develops. Performance was characterized by a simplified figure of merit based on a ratio of neutron production to costs. A work breakdown structure was developed, with simplified systems diagrams used to define interfaces and system responsibilities. A risk assessment method was used to identify potential problems, to identify required research and development (R&D), and to aid contingency development. Preliminary 3-D models of the target station are being used to develop remote maintenance concepts and to estimate costs.

  3. High-spin isomers in 212Rn in the region of triple neutron core-excitations

    NASA Astrophysics Data System (ADS)

    Dracoulis, G. D.; Lane, G. J.; Byrne, A. P.; Davidson, P. M.; Kibédi, T.; Nieminen, P.; Watanabe, H.; Wilson, A. N.

    2008-04-01

    The level scheme of 212Rn has been extended to spins of ∼ 38 ℏ and excitation energies of about 13 MeV using the 204Hg(13C, 5n)212Rn reaction and γ-ray spectroscopy. Time correlated techniques have been used to obtain sensitivity to weak transitions and channel selectivity. The excitation energy of the 22+ core-excited isomer has been established at 6174 keV. Two isomers with τ = 25 (2) ns and τ = 12 (2) ns are identified at 12211 and 12548 keV, respectively. These are the highest-spin nuclear isomers now known, and are attributed to configurations involving triple neutron core-excitations coupled to the aligned valence protons. Semi-empirical shell-model calculations can account for most states observed, but with significant energy discrepancies for some configurations.

  4. Thermionic in-core heat pipe design and performance

    NASA Astrophysics Data System (ADS)

    Determan, W. R.; Hagelston, G.

    1992-01-01

    The heat pipe cooled thermionic reactor (HPTI) relies on in-core sodium heat pipes to provide a redundant means of cooling the 72 thermionic fuel elements (TFEs) and 36 driver fuel pins which comprise the 40 kWe core assembly. In-core heat pipe cooling was selected for the reactor design to meet the requirements for a system design with the potential to achieve a high survivability level against natural and man-made threats and one that possesses no-mission ending single point failures. A detailed study was performed to determine the potential in-core heat pipe geometries which could be developed for an HPTI concept. Requirements and performance estimates were developed for two in-core heat pipe geometries. Both nominal and faulted operating conditions were evaluated using a two-dimensional thermal model of the core to assess TFE and driver fuel pin temperature profiles. A bow tie in-core heat pipe geometry was selected as the optimum design using a HPTI honeycomb core structure.

  5. De novo design of the hydrophobic core of ubiquitin.

    PubMed Central

    Lazar, G. A.; Desjarlais, J. R.; Handel, T. M.

    1997-01-01

    We have previously reported the development and evaluation of a computational program to assist in the design of hydrophobic cores of proteins. In an effort to investigate the role of core packing in protein structure, we have used this program, referred to as Repacking of Cores (ROC), to design several variants of the protein ubiquitin. Nine ubiquitin variants containing from three to eight hydrophobic core mutations were constructed, purified, and characterized in terms of their stability and their ability to adopt a uniquely folded native-like conformation. In general, designed ubiquitin variants are more stable than control variants in which the hydrophobic core was chosen randomly. However, in contrast to previous results with 434 cro, all designs are destabilized relative to the wild-type (WT) protein. This raises the possibility that beta-sheet structures have more stringent packing requirements than alpha-helical proteins. A more striking observation is that all variants, including random controls, adopt fairly well-defined conformations, regardless of their stability. This result supports conclusions from the cro studies that non-core residues contribute significantly to the conformational uniqueness of these proteins while core packing largely affects protein stability and has less impact on the nature or uniqueness of the fold. Concurrent with the above work, we used stability data on the nine ubiquitin variants to evaluate and improve the predictive ability of our core packing algorithm. Additional versions of the program were generated that differ in potential function parameters and sampling of side chain conformers. Reasonable correlations between experimental and predicted stabilities suggest the program will be useful in future studies to design variants with stabilities closer to that of the native protein. Taken together, the present study provides further clarification of the role of specific packing interactions in protein structure and

  6. Coupled full core neutron transport/CFD simulations of pressurized water reactors

    SciTech Connect

    Kochunas, B.; Stimpson, S.; Collins, B.; Downar, T.; Brewster, R.; Baglietto, E.; Yan, J.

    2012-07-01

    Recently as part of the CASL project, a capability to perform 3D whole-core coupled neutron transport and computational fluid dynamics (CFD) calculations was demonstrated. This work uses the 2D/1D transport code DeCART and the commercial CFD code STAR-CCM+. It builds on previous CASL work demonstrating coupling for smaller spatial domains. The coupling methodology is described along with the problem simulated and results are presented for fresh hot full power conditions. An additional comparison is made to an equivalent model that uses lower order T/H feedback to assess the importance and cost of high fidelity feedback to the neutronics problem. A simulation of a quarter core Combustion Engineering (CE) PWR core was performed with the coupled codes using a Fixed Point Gauss-Seidel iteration technique. The total approximate calculation requirements are nearly 10,000 CPU hours and 1 TB of memory. The problem took 6 coupled iterations to converge. The CFD coupled model and low order T/H feedback model compared well for global solution parameters, with a difference in the critical boron concentration and average outlet temperature of 14 ppm B and 0.94 deg. C, respectively. Differences in the power distribution were more significant with maximum relative differences in the core-wide pin peaking factor (Fq) of 5.37% and average relative differences in flat flux region power of 11.54%. Future work will focus on analyzing problems more relevant to CASL using models with less approximations. (authors)

  7. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1958-04-22

    A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

  8. THE DOUBLE PULSAR: EVIDENCE FOR NEUTRON STAR FORMATION WITHOUT AN IRON CORE-COLLAPSE SUPERNOVA

    SciTech Connect

    Ferdman, R. D.; Kramer, M.; Stappers, B. W.; Lyne, A. G.; Stairs, I. H.; Breton, R. P.; McLaughlin, M. A.; Freire, P. C. C.; Possenti, A.; Kaspi, V. M.; Manchester, R. N.

    2013-04-10

    The double pulsar system PSR J0737-3039A/B is a double neutron star binary, with a 2.4 hr orbital period, which has allowed measurement of relativistic orbital perturbations to high precision. The low mass of the second-formed neutron star, as well as the low system eccentricity and proper motion, point to a different evolutionary scenario compared to most other known double neutron star systems. We describe analysis of the pulse profile shape over 6 years of observations and present the resulting constraints on the system geometry. We find the recycled pulsar in this system, PSR J0737-3039A, to be a near-orthogonal rotator with an average separation between its spin and magnetic axes of 90 Degree-Sign {+-} 11 Degree-Sign {+-} 5 Degree-Sign . Furthermore, we find a mean 95% upper limit on the misalignment between its spin and orbital angular momentum axes of 3. Degree-Sign 2, assuming that the observed emission comes from both magnetic poles. This tight constraint lends credence to the idea that the supernova that formed the second pulsar was relatively symmetric, possibly involving electron capture onto an O-Ne-Mg core.

  9. Advanced Neutron Source: Plant Design Requirements. Revision 4

    SciTech Connect

    Not Available

    1990-07-01

    The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS.

  10. Rare earth elements in core marine sediments of coastal East Malaysia by instrumental neutron activation analysis.

    PubMed

    Ashraf, Ahmadreza; Saion, Elias; Gharibshahi, Elham; Kamari, Halimah Mohamed; Kong, Yap Chee; Hamzah, Mohd Suhaimi; Elias, Md Suhaimi

    2016-01-01

    A study was carried out on the concentration of REEs (Dy, Sm, Eu,Yb, Lu, La and Ce) that are present in the core marine sediments of East Malaysia from three locations at South China Sea and one location each at Sulu Sea and Sulawesi Sea. The sediment samples were collected at a depth of between 49 and 109 m, dried, and crushed to powdery form. The entire core sediments prepared for Instrumental Neutron Activation Analysis (INAA) were weighted approximately 0.0500 g to 0.1000 g for short irradiation and 0.1500 g to 0.2000 g for long irradiation. The samples were irradiated with a thermal neutron flux of 4.0×10(12) cm(-2) s(-1) in a TRIGA Mark II research reactor operated at 750 kW. Blank samples and standard reference materials SL-1 were also irradiated for calibration and quality control purposes. It was found that the concentration of REEs varies in the range from 0.11 to 36.84 mg/kg. The chondrite-normalized REEs for different stations suggest that all the REEs are from similar origins. There was no significant REEs contamination as the enrichment factors normalized for Fe fall in the range of 0.42-2.82. PMID:26405840

  11. Rare earth elements in core marine sediments of coastal East Malaysia by instrumental neutron activation analysis.

    PubMed

    Ashraf, Ahmadreza; Saion, Elias; Gharibshahi, Elham; Kamari, Halimah Mohamed; Kong, Yap Chee; Hamzah, Mohd Suhaimi; Elias, Md Suhaimi

    2016-01-01

    A study was carried out on the concentration of REEs (Dy, Sm, Eu,Yb, Lu, La and Ce) that are present in the core marine sediments of East Malaysia from three locations at South China Sea and one location each at Sulu Sea and Sulawesi Sea. The sediment samples were collected at a depth of between 49 and 109 m, dried, and crushed to powdery form. The entire core sediments prepared for Instrumental Neutron Activation Analysis (INAA) were weighted approximately 0.0500 g to 0.1000 g for short irradiation and 0.1500 g to 0.2000 g for long irradiation. The samples were irradiated with a thermal neutron flux of 4.0×10(12) cm(-2) s(-1) in a TRIGA Mark II research reactor operated at 750 kW. Blank samples and standard reference materials SL-1 were also irradiated for calibration and quality control purposes. It was found that the concentration of REEs varies in the range from 0.11 to 36.84 mg/kg. The chondrite-normalized REEs for different stations suggest that all the REEs are from similar origins. There was no significant REEs contamination as the enrichment factors normalized for Fe fall in the range of 0.42-2.82.

  12. Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

    SciTech Connect

    Sun, K.; Chenu, A.; Mikityuk, K.; Krepel, J.; Chawla, R.

    2012-07-01

    The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout

  13. 5 MW pulsed spallation neutron source, Preconceptual design study

    SciTech Connect

    Not Available

    1994-06-01

    This report describes a self-consistent base line design for a 5 MW Pulsed Spallation Neutron Source (PSNS). It is intended to establish feasibility of design and as a basis for further expanded and detailed studies. It may also serve as a basis for establishing project cost (30% accuracy) in order to intercompare competing designs for a PSNS not only on the basis of technical feasibility and technical merit but also on the basis of projected total cost. The accelerator design considered here is based on the objective of a pulsed neutron source obtained by means of a pulsed proton beam with average beam power of 5 MW, in {approx} 1 {mu}sec pulses, operating at a repetition rate of 60 Hz. Two target stations are incorporated in the basic facility: one for operation at 10 Hz for long-wavelength instruments, and one operating at 50 Hz for instruments utilizing thermal neutrons. The design approach for the proton accelerator is to use a low energy linear accelerator (at 0.6 GeV), operating at 60 Hz, in tandem with two fast cycling booster synchrotrons (at 3.6 GeV), operating at 30 Hz. It is assumed here that considerations of cost and overall system reliability may favor the present design approach over the alternative approach pursued elsewhere, whereby use is made of a high energy linear accelerator in conjunction with a dc accumulation ring. With the knowledge that this alternative design is under active development, it was deliberately decided to favor here the low energy linac-fast cycling booster approach. Clearly, the present design, as developed here, must be carried to the full conceptual design stage in order to facilitate a meaningful technology and cost comparison with alternative designs.

  14. Neutron Skyshine Considerations For The NIF Shielding Design

    SciTech Connect

    Singh, M S; Mecozzi, J M; Tobin, M T

    2004-01-28

    A series of coupled neutron-photon transport Monte-Carlo calculations was performed to estimate the roof shielding required to limit the skyshine dose to less than 1 mrem/y at the site boundary when conducting DT experiments with annual fusion yields up to 1200 MJ (4.2E20 neutrons/y). The NIF shielding design consists of many different components. The basic components include 10-cm-thick Al chamber with 40-cm-thick target chamber gunite shield having multiple penetrations, 1.83-m-thick concrete Target Bay walls, 1.37-m-thick concrete roof, and multiple concrete floors with numerous penetrations. Under this shielding configuration, the skyshine dose at the nearest site-boundary was calculated to be less than 0.2 mrem/y for all possible target illumination configurations. The potential dose at the site boundary would be about one-tenth of the cosmic neutron dose that we measured with bubble neutron detectors on board a commercial roundtrip flight from SF to Rochester. This incremental dose increase is well within the normal fluctuations (noise) of the natural background radiation in the Livermore area. The skyshine dose has no impact on the public. The skyshine dose trends at ground and elevated levels are plotted as a function of distance from 20 m to 1000 m from the center of the target bay. The differential neutron and photon energy flux emerging from the NIF roof and at several locations on the ground is plotted to show how it shifts with distance. The results of this study are compared with the neutron skyshine studies done at high-energy accelerators by R. H. Thomas.

  15. Design of a californium-based epithermal neutron beam for neutron capture therapy.

    PubMed

    Yanch, J C; Kim, J K; Wilson, M J

    1993-08-01

    The potential of the spontaneously fissioning isotope, 252Cf, to provide epithermal neutrons for use in boron neutron capture therapy (BNCT) has been investigated using Monte Carlo simulation. The Monte Carlo code MCNP was used to design an assembly composed of a 26 cm long, 11 cm radius cylindrical D2O moderator followed by a 64 cm long Al filter. Lithium filters are placed between the moderator and the filter and between the Al and the patient. A reflector surrounding the moderator/filter assembly is required in order to maintain adequate therapy flux at the patient position. An ellipsoidal phantom composed of skull- and brain-equivalent material was used to determine the dosimetric effect of this beam. It was found that both advantage depths and advantage ratios compare very favourably with reactor and accelerator epithermal neutron sources. The dose rate obtainable, on the other hand, is 4.1 RBE cGy min-1, based on a very large (1.0 g) source of 252Cf. This dose rate is two to five times lower than those provided by existing reactor beams and can be viewed as a drawback of using 252Cf as a neutron source. Radioisotope sources, however, do offer the advantage of in-hospital installation.

  16. Initial global 2-D shielding analysis for the Advanced Neutron Source core and reflector

    SciTech Connect

    Bucholz, J.A.

    1995-08-01

    This document describes the initial global 2-D shielding analyses for the Advanced Neutron Source (ANS) reactor, the D{sub 2}O reflector, the reflector vessel, and the first 200 mm of light water beyond the reflector vessel. Flux files generated here will later serve as source terms in subsequent shielding analyses. In addition to reporting fluxes and other data at key points of interest, a major objective of this report was to document how these analyses were performed, the phenomena that were included, and checks that were made to verify that these phenomena were properly modeled. In these shielding analyses, the fixed neutron source distribution in the core was based on the `lifetime-averaged` spatial power distribution. Secondary gamma production cross sections in the fuel were modified so as to account intrinsically for delayed fission gammas in the fuel as well as prompt fission gammas. In and near the fuel, this increased the low-energy gamma fluxes by 50 to 250%, but out near the reflector vessel, these same fluxes changed by only a few percent. Sensitivity studies with respect to mesh size were performed, and a new 2-D mesh distribution developed after some problems were discovered with respect to the use of numerous elongated mesh cells in the reflector. All of the shielding analyses were performed sing the ANSL-V 39n/44g coupled library with 25 thermal neutron groups in order to obtain a rigorous representation of the thermal neutron spectrum throughout the reflector. Because of upscatter in the heavy water, convergence was very slow. Ultimately, the fission cross section in the various materials had to be artificially modified in order to solve this fixed source problem as an eigenvalue problem and invoke the Vondy error-mode extrapolation technique which greatly accelerated convergence in the large 2-D RZ DORT analyses. While this was quite effective, 150 outer iterations (over energy) were still required.

  17. Evidence for Gamow-Teller Decay of ^{78}Ni Core from Beta-Delayed Neutron Emission Studies.

    PubMed

    Madurga, M; Paulauskas, S V; Grzywacz, R; Miller, D; Bardayan, D W; Batchelder, J C; Brewer, N T; Cizewski, J A; Fijałkowska, A; Gross, C J; Howard, M E; Ilyushkin, S V; Manning, B; Matoš, M; Mendez, A J; Miernik, K; Padgett, S W; Peters, W A; Rasco, B C; Ratkiewicz, A; Rykaczewski, K P; Stracener, D W; Wang, E H; Wolińska-Cichocka, M; Zganjar, E F

    2016-08-26

    The β-delayed neutron emission of ^{83,84}Ga isotopes was studied using the neutron time-of-flight technique. The measured neutron energy spectra showed emission from states at excitation energies high above the neutron separation energy and previously not observed in the β decay of midmass nuclei. The large decay strength deduced from the observed intense neutron emission is a signature of Gamow-Teller transformation. This observation was interpreted as evidence for allowed β decay to ^{78}Ni core-excited states in ^{83,84}Ge favored by shell effects. We developed shell model calculations in the proton fpg_{9/2} and neutron extended fpg_{9/2}+d_{5/2} valence space using realistic interactions that were used to understand measured β-decay lifetimes. We conclude that enhanced, concentrated β-decay strength for neutron-unbound states may be common for very neutron-rich nuclei. This leads to intense β-delayed high-energy neutron and strong multineutron emission probabilities that in turn affect astrophysical nucleosynthesis models. PMID:27610848

  18. Evidence for Gamow-Teller Decay of 78Ni Core from Beta-Delayed Neutron Emission Studies

    NASA Astrophysics Data System (ADS)

    Madurga, M.; Paulauskas, S. V.; Grzywacz, R.; Miller, D.; Bardayan, D. W.; Batchelder, J. C.; Brewer, N. T.; Cizewski, J. A.; Fijałkowska, A.; Gross, C. J.; Howard, M. E.; Ilyushkin, S. V.; Manning, B.; Matoš, M.; Mendez, A. J.; Miernik, K.; Padgett, S. W.; Peters, W. A.; Rasco, B. C.; Ratkiewicz, A.; Rykaczewski, K. P.; Stracener, D. W.; Wang, E. H.; Wolińska-Cichocka, M.; Zganjar, E. F.

    2016-08-01

    The β -delayed neutron emission of Ga,8483 isotopes was studied using the neutron time-of-flight technique. The measured neutron energy spectra showed emission from states at excitation energies high above the neutron separation energy and previously not observed in the β decay of midmass nuclei. The large decay strength deduced from the observed intense neutron emission is a signature of Gamow-Teller transformation. This observation was interpreted as evidence for allowed β decay to 78Ni core-excited states in Ge,8483 favored by shell effects. We developed shell model calculations in the proton f p g9 /2 and neutron extended f p g9 /2+d5 /2 valence space using realistic interactions that were used to understand measured β -decay lifetimes. We conclude that enhanced, concentrated β -decay strength for neutron-unbound states may be common for very neutron-rich nuclei. This leads to intense β -delayed high-energy neutron and strong multineutron emission probabilities that in turn affect astrophysical nucleosynthesis models.

  19. Neutronic assessment of stringer fuel assembly design for liquid-salt-cooledvery high temperature reactor (LS-VHTR).

    SciTech Connect

    Szakaly, F. J.; Kim, T. K.; Taiwo, T. A.

    2006-09-15

    Neutronic studies of 18-pin and 36-pin stringer fuel assemblies have been performed to ascertain that core design requirements for the Liquid-Salt Cooled Very High Temperature Reactor (LS-VHTR) can be met. Parametric studies were performed to determine core characteristics required to achieve a target core cycle length of 18 months and fuel discharge burnup greater than 100 GWd/t under the constraint that the uranium enrichment be less than 20% in order to support non-proliferation goals. The studies were done using the WIMS9 lattice code and the linear reactivity model to estimate the core reactivity balance, fuel composition, and discharge burnup. The results show that the design goals can be met using a 1-batch fuel management scheme, uranium enrichment of 15% and a fuel packing fraction of 30% or greater for the 36-pin stringer fuel assembly design.

  20. Documented Safety Analysis Addendum for the Neutron Radiography Reactor Facility Core Conversion

    SciTech Connect

    Boyd D. Christensen

    2009-05-01

    The Neutron Radiography Reactor Facility (NRAD) is a Training, Research, Isotope Production, General Atomics (TRIGA) reactor which was installed in the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) in the mid 1970s. The facility provides researchers the capability to examine both irradiated and non-irradiated materials in support of reactor fuel and components programs through non-destructive neutron radiography examination. The facility has been used in the past as one facet of a suite of reactor fuels and component examination facilities available to researchers at the INL and throughout the DOE complex. The facility has also served various commercial research activities in addition to the DOE research and development support. The reactor was initially constructed using Fuel Lifetime Improvement Program (FLIP)- type highly enriched uranium (HEU) fuel obtained from the dismantled Puerto Rico Nuclear Center (PRNC) reactor. In accordance with international non-proliferation agreements, the NRAD core will be converted to a low enriched uranium (LEU) fuel and will continue to utilize the PRNC control rods, control rod drives, startup source, and instrument console as was previously used with the HEU core. The existing NRAD Safety Analysis Report (SAR) was created and maintained in the preferred format of the day, combining sections of both DOE-STD-3009 and Nuclear Regulatory Commission Regulatory Guide 1.70. An addendum was developed to cover the refueling and reactor operation with the LEU core. This addendum follows the existing SAR format combining required formats from both the DOE and NRC. This paper discusses the project to successfully write a compliant and approved addendum to the existing safety basis documents.

  1. Bragg optics computer codes for neutron scattering instrument design

    SciTech Connect

    Popovici, M.; Yelon, W.B.; Berliner, R.R.; Stoica, A.D.

    1997-09-01

    Computer codes for neutron crystal spectrometer design, optimization and experiment planning are described. Phase space distributions, linewidths and absolute intensities are calculated by matrix methods in an extension of the Cooper-Nathans resolution function formalism. For modeling the Bragg reflection on bent crystals the lamellar approximation is used. Optimization is done by satisfying conditions of focusing in scattering and in real space, and by numerically maximizing figures of merit. Examples for three-axis and two-axis spectrometers are given.

  2. Neutron transport analysis for nuclear reactor design

    DOEpatents

    Vujic, Jasmina L.

    1993-01-01

    Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values.

  3. Neutron transport analysis for nuclear reactor design

    DOEpatents

    Vujic, J.L.

    1993-11-30

    Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values. 28 figures.

  4. Safety and core design of large liquid-metal cooled fast breeder reactors

    NASA Astrophysics Data System (ADS)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  5. Natural nuclear reactor at Oklo and variation of fundamental constants: Computation of neutronics of a fresh core

    SciTech Connect

    Petrov, Yu. V.; Nazarov, A. I.; Onegin, M. S.; Petrov, V. Yu.; Sakhnovsky, E. G.

    2006-12-15

    Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities, and neutron fluxes were calculated. We also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of {sub 62}{sup 149}Sm and its dependence on the shift of a resonance position E{sub r} (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73{<=}{delta}E{sub r}{<=}62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant {alpha}. We obtain new, more accurate limits of -4x10{sup -17}{<=}{alpha}{center_dot}/{alpha}{<=}3x10{sup -17} yr{sup -1}. Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress.

  6. Natural nuclear reactor at Oklo and variation of fundamental constants: Computation of neutronics of a fresh core

    NASA Astrophysics Data System (ADS)

    Petrov, Yu. V.; Nazarov, A. I.; Onegin, M. S.; Petrov, V. Yu.; Sakhnovsky, E. G.

    2006-12-01

    Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities, and neutron fluxes were calculated. We also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of 62149Sm and its dependence on the shift of a resonance position Er (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73⩽ΔEr⩽62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant α. We obtain new, more accurate limits of -4×10-17⩽α·/α⩽3×10-17yr-1. Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress.

  7. Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

    PubMed

    Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor. PMID:27552124

  8. Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

    PubMed

    Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor.

  9. Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR

    SciTech Connect

    Hanson, A.L.; Diamond, D.

    2011-09-30

    A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Previously, the design of the LEU fuel had been determined in order to provide the users of the NBSR with the same cycle length as exists for the current HEU fueled reactor. The fuel composition at different points within an equilibrium fuel cycle had also been determined. In the present study, neutronics parameters have been calculated for these times in the fuel cycle for both the existing HEU and the proposed LEU equilibrium cores. The results showed differences between the HEU and LEU cores that would not lead to any significant changes in the safety analysis for the converted core. In general the changes were reasonable except that the figure-of-merit for neutrons that can be used by experimentalists shows there will be a 10% reduction in performance. The calculations included kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions.

  10. Design and analysis of a nuclear reactor core for innovative small light water reactors

    NASA Astrophysics Data System (ADS)

    Soldatov, Alexey I.

    In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

  11. Impact of the symmetry energy on nuclear pasta phases and crust-core transition in neutron stars

    NASA Astrophysics Data System (ADS)

    Bao, S. S.; Shen, H.

    2015-01-01

    We study the impact of the symmetry energy on properties of nuclear pasta phases and crust-core transition in neutron stars. We perform a self-consistent Thomas-Fermi calculation employing the relativistic mean-field model. The properties of pasta phases presented in the inner crust of neutron stars are investigated and the crust-core transition is examined. It is found that the slope of the symmetry energy plays an important role in determining the pasta phase structure and the crust-core transition. The correlation between the symmetry energy slope and the crust-core transition density obtained in the Thomas-Fermi approximation is consistent with that predicted by the liquid-drop model.

  12. Design and implementation of a multiaxial loading capability during heating on an engineering neutron diffractometer

    DOE PAGES

    Benafan, O.; Padula, S. A.; Skorpenske, H. D.; An, K.; Vaidyanathan, R.

    2014-10-02

    Here we discuss a gripping capability that was designed, implemented, and tested for in situ neutron diffraction measurements during multiaxial loading and heating on the VULCAN engineering materials diffractometer at the spallation neutron source at Oak Ridge National Laboratory.

  13. Design, construction and characterization of a new neutron beam for neutron radiography at the Tehran Research Reactor

    NASA Astrophysics Data System (ADS)

    Choopan Dastjerdi, M. H.; Khalafi, H.; Kasesaz, Y.; Mirvakili, S. M.; Emami, J.; Ghods, H.; Ezzati, A.

    2016-05-01

    To obtain a thermal neutron beam for neutron radiography applications, a neutron collimator has been designed and implemented at the Tehran Research Reactor (TRR). TRR is a 5 MW open pool light water moderated reactor with seven beam tubes. The neutron collimator is implemented in the E beam tube of the TRR. The design of the neutron collimator was performed using MCNPX Monte Carlo code. In this work, polycrystalline bismuth and graphite have been used as a gamma filter and an illuminator, respectively. The L/D parameter of the facility was chosen in the range of 150-250. The thermal neutron flux at the image plane can be varied from 2.26×106 to 6.5×106 n cm-2 s-1. Characterization of the beam was performed by ASTM standard IQI and foil activation technique to determine the quality of neutron beam. The results show that the obtained neutron beam has a good quality for neutron radiography applications.

  14. Passive neutron design study for 200-L waste drums

    SciTech Connect

    Menlove, H.O.; Beddingfield, D.B.; Pickrell, M.M.

    1997-09-01

    We have developed a passive neutron counter for the measurement of plutonium in 200-L drums of scrap and waste. The counter incorporates high efficiency for the multiplicity counting in addition to the traditional coincidence counting. The {sup 252}Cf add-a-source feature is used to provide an accurate assay over a wide range of waste matrix materials. The room background neutron rate is reduced by using 30 cm of external polyethylene shielding and the cosmic-ray background is reduced by statistical filtering techniques. Monte Carlo Code calculations were used to determine the optimum detector design, including the gas pressure, size, number, and placement of the {sup 3}He tubes in the moderator. Various moderators, including polyethylene, plastics, teflon, and graphite, were evaluated to obtain the maximum efficiency and minimum detectable mass of plutonium.

  15. Core design studies for advanced burner test reactor.

    SciTech Connect

    Yang, W. S.; Kim, T. K.; Hill, R. N.; Nuclear Engineering Division

    2008-01-01

    The U.S. government announced in February 2006 the Global Nuclear Energy Partnership (GNEP) to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. The advanced burner reactor (ABR) based on a fast spectrum is one of the three major technologies to be demonstrated in GNEP. In FY06, a pre-conceptual design study was performed to develop an advanced burner test reactor (ABTR) that supports development of a prototype full-scale ABR, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR were (1) to demonstrate reactor-based transmutation of transuranics (TRU) as part of an advanced fuel cycle, (2) to qualify the TRU-containing fuels and advanced structural materials needed for a full-scale ABR, (3) to support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. Based on these objectives, core design and fuel cycle studies were performed to develop ABTR core designs, which can accommodate the expected changes of the TRU feed and the conversion ratio. Various option and trade-off studies were performed to determine the appropriate power level and conversion ratio. Both ternary metal alloy (U-TRU-10Zr) and mixed oxide (UO{sub 2}-TRUO{sub 2}) fuel forms have been considered with TRU feeds from weapons-grade plutonium (WG-Pu) and TRU recovered from light water reactor spent fuel (LWR-SF). Reactor performances were evaluated in detail including equilibrium cycle core parameters, mass flow, power distribution, kinetic parameters, reactivity feedback coefficient, reactivity control requirements and shutdown margins, and spent fuel characteristics. Trade-off studies on power level suggested that about 250 MWt is a reasonable compromise to allow a low project cost, at the same time providing a reasonable prototypic irradiation environment for demonstrating

  16. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    SciTech Connect

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  17. Development of a neutronics calculation method for designing commercial type Japanese sodium-cooled fast reactor

    SciTech Connect

    Takeda, T.; Shimazu, Y.; Hibi, K.; Fujimura, K.

    2012-07-01

    Under the R and D project to improve the modeling accuracy for the design of fast breeder reactors the authors are developing a neutronics calculation method for designing a large commercial type sodium- cooled fast reactor. The calculation method is established by taking into account the special features of the reactor such as the use of annular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verification and Validation, and Uncertainty Qualification (V and V and UQ) of the calculation method is being performed by using measured data from the prototype FBR Monju. The results of this project will be used in the design and analysis of the commercial type demonstration FBR, known as the Japanese Sodium fast Reactor (JSFR). (authors)

  18. Spin-down Properties of Quark-Core Neutron Stars, Catastrophic Phase Transitions and Gamma-Ray Bursts

    NASA Astrophysics Data System (ADS)

    Ma, Feng; Luo, Shan

    1996-05-01

    We study properties of an isolated quark-core neutron star in its spin-down process. We find that the central density of the star increases with time, and more neutron matter is converted into quark matter via continuous quark-hadron phase transition. As a result, the whole star contracts while its quark core increases in size and mass. This has two observational signatures: first, the fractional moment of inertia of the neutral component of the star (I_n/I) decreases and results in a decrease of the proportional healing parameter (Q) of pulsar glitches; second, due to the contraction of the whole star, the total moment of inertia decreases and results in an increase in the braking index (n) of the pulsar spin-down process. This makes the spin-down behavior of a quark-core neutron star different from that of ``pure'' neutron stars. In comparison, most previous work about rotational properties of neutron and quark stars concentrated on rotation-induced mass ``increase'', which is related to star families and not directly observable. In the extreme case, catastrophic quark-hadron phase transition or pion condensation may happen to neutron stars at a rate of about 10(-6) yr(-1) per galaxy, with an energy release of about 10(52) ergs, and may be a good explanation of gamma-ray bursts (GRBs) at cosmological distances. If so, the detection of gravitational waves (GW) as counterparts of GRBs will be less likely than previously expected. We give an approximate light curve of GW for a catastrophic phase transition in a fast rotating star, and find it to be in sharp contrast to the predictions of neutron star merger models. We also discuss an extremely strong magnetic field that may be formed via the dynamo process during collapse; the effects of this field on electromagnetic and gravitational radiation; and an initial high energy neutrino burst due to the production of strange quarks.

  19. Design of the Mechanical Parts for the Neutron Guide System at HANARO

    SciTech Connect

    Shin, J. W.; Cho, Y. G.; Cho, S. J.; Ryu, J. S.

    2008-03-17

    The research reactor HANARO (High-flux Advanced Neutron Application ReactOr) in Korea will be equipped with a neutron guide system, in order to transport cold neutrons from the neutron source to the neutron scattering instruments in the neutron guide hall near the reactor building. The neutron guide system of HANARO consists of the in-pile plug assembly with in-pile guides, the primary shutter with in-shutter guides, the neutron guides in the guide shielding room with dedicated secondary shutters, and the neutron guides connected to the instruments in the neutron guide hall. Functions of the in-pile plug assembly are to shield the reactor environment from nuclear radiation and to support the neutron guides and maintain them precisely oriented. The primary shutter is a mechanical structure to be installed just after the in-pile plug assembly, which stops neutron flux on demand. This paper describes the design of the in-pile assembly and the primary shutter for the neutron guide system at HANARO. The design of the guide shielding assembly for the primary shutter and the neutron guides is also presented.

  20. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

    SciTech Connect

    Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F.; Lopez, A. Legrand

    2011-03-15

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 x 10{sup 20} n/cm{sup 2}. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

  1. THERMAL: A routine designed to calculate neutron thermal scattering

    SciTech Connect

    Cullen, D.E.

    1995-02-24

    THERMAL is designed to calculate neutron thermal scattering that is isotropic in the center of mass system. At low energy thermal motion will be included. At high energies the target nuclei are assumed to be stationary. The point of transition between low and high energies has been defined to insure a smooth transition. It is assumed that at low energy the elastic cross section is constant in the center of mass system. At high energy the cross section can be of any form. You can use this routine for all energies where the elastic scattering is isotropic in the center of mass system. In most materials this will be a fairly high energy.

  2. A neutronic feasibility study of the AP1000 design loaded with fully ceramic micro-encapsulated fuel

    SciTech Connect

    Liang, C.; Ji, W.

    2013-07-01

    A neutronic feasibility study is performed to evaluate the utilization of fully ceramic microencapsulated (FCM) fuel in the AP1000 reactor design. The widely used Monte Carlo code MCNP is employed to perform the full core analysis at the beginning of cycle (BOC). Both the original AP1000 design and the modified design with the replacement of uranium dioxide fuel pellets with FCM fuel compacts are modeled and simulated for comparison. To retain the original excess reactivity, ranges of fuel particle packing fraction and fuel enrichment in the FCM fuel design are first determined. Within the determined ranges, the reactor control mechanism employed by the original design is directly used in the modified design and the utilization feasibility is evaluated. The worth of control of each type of fuel burnable absorber (discrete/integral fuel burnable absorbers and soluble boron in primary coolant) is calculated for each design and significant differences between the two designs are observed. Those differences are interpreted by the fundamental difference of the fuel form used in each design. Due to the usage of silicon carbide as the matrix material and the fuel particles fuel form in FCM fuel design, neutron slowing down capability is increased in the new design, leading to a much higher thermal spectrum than the original design. This results in different reactivity and fission power density distributions in each design. We conclude that a direct replacement of fuel pellets by the FCM fuel in the AP1000 cannot retain the original optimum reactor core performance. Necessary modifications of the core design should be done and the original control mechanism needs to be re-designed. (authors)

  3. A feasibility design study on a neutron spectrometer for BNCT with liquid moderator.

    PubMed

    Tamaki, S; Sato, F; Murata, I

    2015-12-01

    Neutrons generated by accelerators have various energy spectra. However, only limited methods are available to measure the whole neutron energy spectrum, especially when including the epithermal region that is normally used in BNCT. In the present study, we carried out the design study on a new neutron spectrometer that can measure such a neutron spectrum more accurately, precisely and with higher energy resolution, using an unfolding technique and a liquid moderator.

  4. Simulation of in-core neutron noise measurements for axial void profile reconstruction in boiling water reactors

    SciTech Connect

    Dykin, V.; Pazsit, I.

    2012-07-01

    A possibility to reconstruct the axial void profile from the simulated in-core neutron noise which is caused by density fluctuations in a Boiling Water Reactor (BWR) heated channel is considered. For this purpose, a self-contained model of the two-phase flow regime is constructed which has quantitatively and qualitatively similar properties to those observed in real BWRs. The model is subsequently used to simulate the signals of neutron detectors induced by the corresponding perturbations in the flow density. The bubbles are generated randomly in both space and time using Monte-Carlo techniques. The axial distribution of the bubble production is chosen such that the mean axial void fraction and void velocity follow the actual values of BWRs. The induced neutron noise signals are calculated and then processed by the standard signal analysis methods such as Auto-Power Spectral Density (APSD) and Cross-Power Spectral Density (CPSD). Two methods for axial void and velocity profiles reconstruction are discussed: the first one is based on the change of the break frequency of the neutron auto-power spectrum with axial core elevation, while the second refers to the estimation of transit times of propagating steam fluctuations between different axial detector positions. This paper summarizes the principles of the model and presents a numerical testing of the qualitative applicability to estimate the required parameters for the reconstruction of the void fraction profile from the neutron noise measurements. (authors)

  5. The Design of a Compact Rfq Neutron Generator

    NASA Astrophysics Data System (ADS)

    Hamm, R. W.; Becker, R.

    2014-02-01

    The output and target lifetime of a conventional electrostatic neutron generator are limited by the voltage stand-off capability and the acceleration of molecular species from the ion source. As an alternative, we suggest that the deuterium beam achievable from a compact high intensity ECR source can be injected directly into a compact RFQ to produce a more efficient compact neutron production system. Only the d+ ions are accelerated by the RFQ, which can also produce much higher output energies than electrostatic systems, resulting in a higher neutron output with a longer target lifetime. The direct injection of the beam makes the system more compact than the multielement, electrostatic systems typically used for extraction of the beam and subsequent transport and matching into the RFQ. We have designed and optimized a combined extraction/matching system for a compact high current deuterium ECR ion source injected into a high frequency RFQ structure, allowing a beam of about 12 mA of d+ ions to be injected at a modest ion source voltage of 25 kV. The end wall of the RFQ resonator serves as the ground electrode for the ion source, resembling DPI (direct plasma injection). For this design, we used the features of the code IGUN to take into account the electrostatic field between the ion source and the RFQ end wall, the stray magnetic field of the ECR source, the defocusing space charge of the low energy deuteron beam, and the rf focusing in the fringe field between the RFQ vanes and the RFQ flange.

  6. High-spin states in the semimagic nucleus 89Y and neutron-core excitations in the N =50 isotones

    NASA Astrophysics Data System (ADS)

    Li, Z. Q.; Wang, S. Y.; Niu, C. Y.; Qi, B.; Wang, S.; Sun, D. P.; Liu, C.; Xu, C. J.; Liu, L.; Zhang, P.; Wu, X. G.; Li, G. S.; He, C. Y.; Zheng, Y.; Li, C. B.; Yu, B. B.; Hu, S. P.; Yao, S. H.; Cao, X. P.; Wang, J. L.

    2016-07-01

    The semimagic nucleus 89Y 89 has been investigated using the 82Se(11>B,4 n ) reaction at beam energies of 48 and 52 MeV. More than 24 new transitions have been identified, leading to a considerable extension of the level structures of 89Y. The experimental results are compared with the large-basis shell model calculations. They show that cross-shell neutron excitations play a pivotal role in high-spin level structures of 89Y. The systematic features of neutron-core excitations in the N =50 isotones are also discussed.

  7. Strangeness driven phase transitions in compressed baryonic matter and their relevance for neutron stars and core collapsing supernovae

    SciTech Connect

    Raduta, Ad. R.; Gulminelli, F.; Oertel, M.

    2015-02-24

    We discuss the thermodynamics of compressed baryonic matter with strangeness within non-relativistic mean-field models with effective interactions. The phase diagram of the full baryonic octet under strangeness equilibrium is built and discussed in connection with its relevance for core-collapse supernovae and neutron stars. A simplified framework corresponding to (n, p, Λ)(+e)-mixtures is employed in order to test the sensitivity of the existence of a phase transition on the (poorely constrained) interaction coupling constants and the compatibility between important hyperonic abundances and 2M{sub ⊙} neutron stars.

  8. Design study of a medical proton linac for neutron therapy

    SciTech Connect

    Machida, S.; Raparia, D.

    1988-08-26

    This paper describes a design study which establishes the physical parameters of the low energy beam transport, radiofrequency quadrupole, and linac, using computer programs available at Fermilab. Beam dynamics studies verify that the desired beam parameters can be achieved. The machine described here meets the aforementioned requirements and can be built using existing technology. Also discussed are other technically feasible options which could be attractive to clinicians, though they would complicate the design of the machine and increase construction costs. One of these options would allow the machine to deliver 2.3 MeV protons to produce epithermal neutrons for treating brain tumors. A second option would provide 15 MeV protons for isotope production. 21 refs., 33 figs.

  9. Core compressor exit stage study. 1: Aerodynamic and mechanical design

    NASA Technical Reports Server (NTRS)

    Burdsall, E. A.; Canal, E., Jr.; Lyons, K. A.

    1979-01-01

    The effect of aspect ratio on the performance of core compressor exit stages was demonstrated using two three stage, highly loaded, core compressors. Aspect ratio was identified as having a strong influence on compressors endwall loss. Both compressors simulated the last three stages of an advanced eight stage core compressor and were designed with the same 0.915 hub/tip ratio, 4.30 kg/sec (9.47 1bm/sec) inlet corrected flow, and 167 m/sec (547 ft/sec) corrected mean wheel speed. The first compressor had an aspect ratio of 0.81 and an overall pressure ratio of 1.357 at a design adiabatic efficiency of 88.3% with an average diffusion factor or 0.529. The aspect ratio of the second compressor was 1.22 with an overall pressure ratio of 1.324 at a design adiabatic efficiency of 88.7% with an average diffusion factor of 0.491.

  10. A new paradigm for local-global coupling in whole-core neutron transport.

    SciTech Connect

    Lewis, E.; Smith, M.; Palmiotti, G,; Nuclear Engineering Division; Northwestern Univ.; INL

    2009-01-01

    A new paradigm that increases the efficiency of whole-core neutron transport calculations without lattice homogenization is introduced. Quasi-reflected interface conditions are formulated to partially decouple periodic lattice effects from global flux gradients. The starting point is the finite subelement form of the variational nodal code VARIANT that eliminates fuel-coolant homogenization through the use of heterogeneous nodes. The interface spherical harmonics expansions that couple pin-cell-sized nodes are divided into low-order and high-order terms, and reflected interface conditions are applied to the high-order terms. Combined with an integral transport method within the node, the new approach dramatically reduces both the formation time and the dimensions of the nodal response matrices and leads to sharply reduced memory requirements and computational time. The method is applied to the two-dimensional C5G7 problem, an Organisation for Economic Co-operation and Development/Nuclear Energy Agency pressurized water reactor benchmark containing mixed oxide (MOX) and UO{sub 2} fuel assemblies, as well as to a three-dimensional MOX fuel assembly. Results indicate the new approach results in very little loss of accuracy relative to the corresponding full spherical harmonics expansions while reducing computational times by well over an order of magnitude.

  11. Error Assessment of Homogenized Cross Sections Generation for Whole Core Neutronic Calculation

    NASA Astrophysics Data System (ADS)

    Hursin, Mathieu; Kochunas, Brendan; Downar, Thomas J.

    2007-10-01

    The objective of the work here was to assess the errors introduced by using 2D, few group homogenized cross sections to perform neutronic analysis of BWR problems with significant axial heterogeneities. The 3D method of characteristics code DeCART is used to generate 2-group assembly homogenized cross sections first using a conventional 2D lattice model and then using a full 3D solution of the assembly. A single BWR fuel assembly model based on an advanced BWR lattice design is used with a typical void distribution applied to the fuel channel coolant. This model is validated against an MCNP model. A comparison of the cross sections is performed for the assembly homogenized planar cross sections from the DeCART 3D and DeCART 2D solutions.

  12. Error Assessment of Homogenized Cross Sections Generation for Whole Core Neutronic Calculation

    SciTech Connect

    Hursin, Mathieu; Kochunas, Brendan; Downar, Thomas J.

    2007-10-26

    The objective of the work here was to assess the errors introduced by using 2D, few group homogenized cross sections to perform neutronic analysis of BWR problems with significant axial heterogeneities. The 3D method of characteristics code DeCART is used to generate 2-group assembly homogenized cross sections first using a conventional 2D lattice model and then using a full 3D solution of the assembly. A single BWR fuel assembly model based on an advanced BWR lattice design is used with a typical void distribution applied to the fuel channel coolant. This model is validated against an MCNP model. A comparison of the cross sections is performed for the assembly homogenized planar cross sections from the DeCART 3D and DeCART 2D solutions.

  13. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    SciTech Connect

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  14. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  15. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2011-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  16. Preliminary design study of advanced multistage axial flow core compressors

    NASA Technical Reports Server (NTRS)

    Wisler, D. C.; Koch, C. C.; Smith, L. H., Jr.

    1977-01-01

    A preliminary design study was conducted to identify an advanced core compressor for use in new high-bypass-ratio turbofan engines to be introduced into commercial service in the 1980's. An evaluation of anticipated compressor and related component 1985 state-of-the-art technology was conducted. A parametric screening study covering a large number of compressor designs was conducted to determine the influence of the major compressor design features on efficiency, weight, cost, blade life, aircraft direct operating cost, and fuel usage. The trends observed in the parametric screening study were used to develop three high-efficiency, high-economic-payoff compressor designs. These three compressors were studied in greater detail to better evaluate their aerodynamic and mechanical feasibility.

  17. Validation of the neutron and gamma fields in the JSI TRIGA reactor using in-core fission and ionization chambers.

    PubMed

    Žerovnik, Gašper; Kaiba, Tanja; Radulović, Vladimir; Jazbec, Anže; Rupnik, Sebastjan; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

    2015-02-01

    CEA developed fission chambers and ionization chambers were utilized at the JSI TRIGA reactor to measure neutron and gamma fields. The measured axial fission rate distributions in the reactor core are generally in good agreement with the calculated values using the Monte Carlo model of the reactor thus verifying both the computational model and the fission chambers. In future, multiple absolutely calibrated fission chambers could be used for more accurate online reactor thermal power monitoring. PMID:25479432

  18. Validation of the neutron and gamma fields in the JSI TRIGA reactor using in-core fission and ionization chambers.

    PubMed

    Žerovnik, Gašper; Kaiba, Tanja; Radulović, Vladimir; Jazbec, Anže; Rupnik, Sebastjan; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

    2015-02-01

    CEA developed fission chambers and ionization chambers were utilized at the JSI TRIGA reactor to measure neutron and gamma fields. The measured axial fission rate distributions in the reactor core are generally in good agreement with the calculated values using the Monte Carlo model of the reactor thus verifying both the computational model and the fission chambers. In future, multiple absolutely calibrated fission chambers could be used for more accurate online reactor thermal power monitoring.

  19. THE RF SYSTEM DESIGN FOR THE SPALLATION NEUTRON SOURCE

    SciTech Connect

    D. REES; M. LYNCH; ET AL

    2001-06-01

    Spallation Neutron Source (SNS) accelerator includes a nominally 1000 MeV, 2 mA average current linac consisting of a radio frequency quadrapole (RFQ), drift tube linac (DTL), coupled cavity linac (CCL), a medium and high beta super conducting (SC) linac, and two buncher cavities for beam transport to the ring. Los Alamos is responsible for the RF systems for all sections of the linac. The SNS linac is a pulsed proton linac and the RF system must support a 1 msec beam pulse at up to a 60 Hz repetition rate. The RFQ and DTL utilize seven, 2.5 MW klystrons and operate at 402.5 MHz. The CCL, SC, and buncher cavities operate at 805 MHz. Six, 5 MW klystrons are utilized for the CCL and buncher cavities while eighty-one 550 kW klystrons are used for the SC cavities. All of the RF hardware for the SNS linac is currently in production. This paper will present details of the RF system-level design as well as specific details of the SNS RF equipment. The design parameters will be discussed. One of the design challenges has been achieving a reasonable cost with the very large number of high-power klystrons. The approaches we used to reduce cost and the resulting design compromises will be discussed.

  20. Nuclear design of a vapor core reactor for space nuclear propulsion

    NASA Astrophysics Data System (ADS)

    Dugan, Edward T.; Watanabe, Yoichi; Kuras, Stephen A.; Maya, Isaac; Diaz, Nils J.

    1993-01-01

    Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range of 1000-1200 seconds; a thrust of 75,000 lbs for a hydrogen flow rate of 30 kg/s; average core exit temperatures of 3100 K to 3400 K; and reactor thermal powers of 1400 to 1800 MW. Initial calculations were performed on epithermal NVTRs using ZrC fuel elements. Studies are now directed at thermal NVTRs that use fuel elements made of C-C composite. The large ZrC-moderated reactors resulted in thrust-to-weight ratios of only 1 to 2; the compact C-C composite systems yield thrust-to-weight ratios of 3 to 5.

  1. CHARGED-PARTICLE AND NEUTRON-CAPTURE PROCESSES IN THE HIGH-ENTROPY WIND OF CORE-COLLAPSE SUPERNOVAE

    SciTech Connect

    Farouqi, K.; Truran, J. W.; Kratz, K.-L.; Pfeiffer, B.; Rauscher, T.; Thielemann, F.-K. E-mail: truran@nova.uchicago.ed E-mail: k-l.Kratz@mpic.d E-mail: F-K.Thielemann@unibas.c

    2010-04-01

    The astrophysical site of the r-process is still uncertain, and a full exploration of the systematics of this process in terms of its dependence on nuclear properties from stability to the neutron drip-line within realistic stellar environments has still to be undertaken. Sufficiently high neutron-to-seed ratios can only be obtained either in very neutron-rich low-entropy environments or moderately neutron-rich high-entropy environments, related to neutron star mergers (or jets of neutron star matter) and the high-entropy wind of core-collapse supernova explosions. As chemical evolution models seem to disfavor neutron star mergers, we focus here on high-entropy environments characterized by entropy S, electron abundance Y{sub e} , and expansion velocity V{sub exp}. We investigate the termination point of charged-particle reactions, and we define a maximum entropy S{sub final} for a given V{sub exp} and Y{sub e} , beyond which the seed production of heavy elements fails due to the very small matter density. We then investigate whether an r-process subsequent to the charged-particle freeze-out can in principle be understood on the basis of the classical approach, which assumes a chemical equilibrium between neutron captures and photodisintegrations, possibly followed by a beta-flow equilibrium. In particular, we illustrate how long such a chemical equilibrium approximation holds, how the freeze-out from such conditions affects the abundance pattern, and which role the late capture of neutrons originating from beta-delayed neutron emission can play. Furthermore, we analyze the impact of nuclear properties from different theoretical mass models on the final abundances after these late freeze-out phases and beta-decays back to stability. As only a superposition of astrophysical conditions can provide a good fit to the solar r-abundances, the question remains how such superpositions are attained, resulting in the apparently robust r-process pattern observed in low

  2. Charged-partricle and neutron-capture processes in the high-entropy wind of core-collapse supernovae.

    SciTech Connect

    Farouqi, K.; Kratz, K.-L.; Pfeiffer, B.; Rauscher, T.; Thielemann, F.-K.; Truran, J. W.; Physics; Univ. of Chicago; Joint Inst. for Nuclear Astrophysics; Univ. Mainz; Virtual Inst. for Nuclear Structure and Astrophysics; Max-Planck-Inst. fur Chemie; Univ. of Basel

    2010-04-01

    The astrophysical site of the r-process is still uncertain, and a full exploration of the systematics of this process in terms of its dependence on nuclear properties from stability to the neutron drip-line within realistic stellar environments has still to be undertaken. Sufficiently high neutron-to-seed ratios can only be obtained either in very neutron-rich low-entropy environments or moderately neutron-rich high-entropy environments, related to neutron star mergers (or jets of neutron star matter) and the high-entropy wind of core-collapse supernova explosions. As chemical evolution models seem to disfavor neutron star mergers, we focus here on high-entropy environments characterized by entropy S, electron abundance Y{sub e}, and expansion velocity V{sub exp}. We investigate the termination point of charged-particle reactions, and we define a maximum entropy S{sub final} for a given V{sub exp} and Y{sub e}, beyond which the seed production of heavy elements fails due to the very small matter density. We then investigate whether an r-process subsequent to the charged-particle freeze-out can in principle be understood on the basis of the classical approach, which assumes a chemical equilibrium between neutron captures and photodisintegrations, possibly followed by a {beta}-flow equilibrium. In particular, we illustrate how long such a chemical equilibrium approximation holds, how the freeze-out from such conditions affects the abundance pattern, and which role the late capture of neutrons originating from {beta}-delayed neutron emission can play. Furthermore, we analyze the impact of nuclear properties from different theoretical mass models on the final abundances after these late freeze-out phases and {beta}-decays back to stability. As only a superposition of astrophysical conditions can provide a good fit to the solar r-abundances, the question remains how such superpositions are attained, resulting in the apparently robust r-process pattern observed in low

  3. Charged-particle and neutron-capture processes in the high-entropy wind of core-collapse supernovae.

    SciTech Connect

    Farouqi, K.; Kratz, K.-L.; Pfeiffer, B.; Rauscher, T.; Thielemann, F.-K.; Truran, J.W.; Physics; Univ. of Chicago; Joint Inst. for Nuclear Astrophysics; Univ. Mainz; Virtual Inst. for Nuclear Structure and Astrophysics; Max-Planck-Insti. fur Chemie; Univ. of Basel

    2010-04-01

    The astrophysical site of the r-process is still uncertain, and a full exploration of the systematics of this process in terms of its dependence on nuclear properties from stability to the neutron drip-line within realistic stellar environments has still to be undertaken. Sufficiently high neutron-to-seed ratios can only be obtained either in very neutron-rich low-entropy environments or moderately neutron-rich high-entropy environments, related to neutron star mergers (or jets of neutron star matter) and the high-entropy wind of core-collapse supernova explosions. As chemical evolution models seem to disfavor neutron star mergers, we focus here on high-entropy environments characterized by entropy S, electron abundance Y{sub e}, and expansion velocity V{sub exp}. We investigate the termination point of charged-particle reactions, and we define a maximum entropy S{sub final} for a given V{sub exp} and Y{sub e}, beyond which the seed production of heavy elements fails due to the very small matter density. We then investigate whether an r-process subsequent to the charged-particle freeze-out can in principle be understood on the basis of the classical approach, which assumes a chemical equilibrium between neutron captures and photodisintegrations, possibly followed by a {beta}-flow equilibrium. In particular, we illustrate how long such a chemical equilibrium approximation holds, how the freeze-out from such conditions affects the abundance pattern, and which role the late capture of neutrons originating from {beta}-delayed neutron emission can play. Furthermore, we analyze the impact of nuclear properties from different theoretical mass models on the final abundances after these late freeze-out phases and {beta}-decays back to stability. As only a superposition of astrophysical conditions can provide a good fit to the solar r-abundances, the question remains how such superpositions are attained, resulting in the apparently robust r-process pattern observed in low

  4. Design of neutron beams at the Argonne Continuous Wave Linac (ACWL) for boron neutron capture therapy and neutron radiography

    SciTech Connect

    Zhou, X.L.; McMichael, G.E.

    1994-10-01

    Neutron beams are designed for capture therapy based on p-Li and p-Sc reactions using the Argonne Continuous Wave Linac (ACWL). The p-Li beam will provide a 2.5 {times} 10{sup 9} n/cm{sup 2}s epithermal flux with 7 {times} 10{sup 5} {gamma}/cm{sup 2}s contamination. On a human brain phantom, this beam allows an advantage depth (AD) of 10 cm, an advantage depth dose rate (ADDR) of 78 cGy/min and an advantage ratio (AR) of 3.2. The p-Sc beam offers 5.9 {times} 10{sup 7} n/cm{sup 2}s and a dose performance of AD = 8 cm and AR = 3.5, suggesting the potential of near-threshold (p,n) reactions such as the p-Li reaction at E{sub p} = 1.92 MeV. A thermal radiography beam could also be obtained from ACWL.

  5. Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

    SciTech Connect

    Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi; Zhang, Dingkang

    2013-11-29

    This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.

  6. Neutron micro-beam design simulation by Monte Carlo

    NASA Astrophysics Data System (ADS)

    Pazirandeh, Ali; Taheri, Ali

    2007-09-01

    Over the last two decades neutron micro-beam has increasingly been developing in view of various applications in molecular activation analysis, micro-radiography in space and aviation and in radiation induced bystander effects in bio-cells. In this paper the structure and simulation of a neutron micro-beam is presented. The collimator for micro-beam is made of a polyethylene cylinder with a small hole along the centerline of the cylinder. The hole is filled with very thin needles in triangular or rectangular arrangement. The neutron source was reactor neutrons or a spontaneous Cf-252 neutron source falling on the top side of the collimator. The outgoing thermal and epithermal neutron fluxes were calculated.

  7. Prompt-gamma neutron activation analysis system design: Effects of D-T versus D-D neutron generator source selection

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Prompt-gamma neutron activation (PGNA) analysis is used for the non-invasive measurement of human body composition. Advancements in portable, compact neutron generator design have made those devices attractive as neutron sources. Two distinct generators are available: D-D with 2.5 MeV and D-T with...

  8. Prompt-gamma neutron activation analysis system design: effects of D-T versus D-D neutron generator source selection

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Prompt-gamma neutron activation analysis (PGNAA) is used for the non-invasive measurement of human body composition. Advancements in portable, compact neutron generator design have made those devices attractive as neutron sources. Two distinct generators are available: D-D with 2.5 MeV, and D-T wi...

  9. New approach to the design of core support structures for large LMFBR plants

    SciTech Connect

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-01-01

    The paper describes an innovative design concept for a LMFBR Core Support Structure. A hanging Core Support Structure is described and analyzed. The design offers inherent safety features, constructibility advantages, and potential cost reductions.

  10. Design of the cold neutron triple-axis spectrometer at the China Advanced Research Reactor

    NASA Astrophysics Data System (ADS)

    Cheng, P.; Zhang, Hongxia; Bao, W.; Schneidewind, A.; Link, P.; Grünwald, A. T. D.; Georgii, R.; Hao, L. J.; Liu, Y. T.

    2016-06-01

    The design of the first cold neutron triple-axis spectrometer at the China Advanced Research Reactor is presented. Based on the Monte Carlo simulations using neutron ray-tracing program McStas, the parameters of major neutron optics in this instrument are optimized. The neutron flux at sample position is estimated to be 5.6 ×107 n/cm2/s at neutron incident energy Ei=5 meV when the reactor operates normally at the designed 60 MW power. The performances of several neutron supermirror polarizing devices are compared and their critical parameters are optimized for this spectrometer. The polarization analysis will be realized with a flexible switch from the unpolarized experimental mode.

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

    SciTech Connect

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

    2012-07-01

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

  12. Seismic responses of a pool-type fast reactor with different core support designs

    SciTech Connect

    Wu, Ting-shu; Seidensticker, R.W. )

    1989-01-01

    In designing the core support system for a pool-type fast reactor, there are many issues which must be considered in order to achieve an optimum and balanced design. These issues include safety, reliability, as well as costs. Several design options are possible to support the reactor core. Different core support options yield different frequency ranges and responses. Seismic responses of a large pool-type fast reactor incorporated with different core support designs have been investigated. 4 refs., 3 figs.

  13. Results of a Neutronic Simulation of HTR-Proteus Core 4.2 using PEBBED and other INL Reactor Physics Tools: FY-09 Report

    SciTech Connect

    Hans D. Gougar

    2009-08-01

    The Idaho National Laboratory’s deterministic neutronics analysis codes and methods were applied to the computation of the core multiplication factor of the HTR-Proteus pebble bed reactor critical facility. A combination of unit cell calculations (COMBINE-PEBDAN), 1-D discrete ordinates transport (SCAMP), and nodal diffusion calculations (PEBBED) were employed to yield keff and flux profiles. Preliminary results indicate that these tools, as currently configured and used, do not yield satisfactory estimates of keff. If control rods are not modeled, these methods can deliver much better agreement with experimental core eigenvalues which suggests that development efforts should focus on modeling control rod and other absorber regions. Under some assumptions and in 1D subcore analyses, diffusion theory agrees well with transport. This suggests that developments in specific areas can produce a viable core simulation approach. Some corrections have been identified and can be further developed, specifically: treatment of the upper void region, treatment of inter-pebble streaming, and explicit (multiscale) transport modeling of TRISO fuel particles as a first step in cross section generation. Until corrections are made that yield better agreement with experiment, conclusions from core design and burnup analyses should be regarded as qualitative and not benchmark quality.

  14. Analyzing the effect of geometric factors on designing neutron radiography system.

    PubMed

    Amini, Moharam; Fadaei, Amir Hosein; Gharib, Morteza

    2015-11-01

    Neutron radiography is one of the main applications of research reactors. It is a powerful tool to conduct nondestructive testing of materials. The parameters that affect the quality of a radiographic image must be considered during the design of a neutron radiography system. Hence, this study aims to investigate the effect of geometric factors on the quality of the neutron radiography system. The results show that the performance of the mentioned system can be increased by regulating the geometric factors. PMID:26343340

  15. The interplay between proto{endash}neutron star convection and neutrino transport in core-collapse supernovae

    SciTech Connect

    Mezzacappa, A.; Calder, A.C.; Guidry, M.W.; Strayer, M.R.; Guidry, M.W.; Strayer, M.R.; Umar, A.S. Bruenn, S.W. Blondin, J.M.

    1998-02-01

    We couple two-dimensional hydrodynamics to realistic one-dimensional multigroup flux-limited diffusion neutrino transport to investigate proto{endash}neutron star convection in core-collapse supernovae, and more specifically, the interplay between its development and neutrino transport. Our initial conditions, time-dependent boundary conditions, and neutrino distributions for computing neutrino heating, cooling, and deleptonization rates are obtained from one-dimensional simulations that implement multigroup flux-limited diffusion and one-dimensional hydrodynamics. The development and evolution of proto{endash}neutron star convection are investigated for both 15 and 25M{sub {circle_dot}} models, representative of the two classes of stars with compact and extended iron cores, respectively. For both models, in the absence of neutrino transport, the angle-averaged radial and angular convection velocities in the initial Ledoux unstable region below the shock after bounce achieve their peak values in {approximately}20ms, after which they decrease as the convection in this region dissipates. The dissipation occurs as the gradients are smoothed out by convection. This initial proto{endash}neutron star convection episode seeds additional convectively unstable regions farther out beneath the shock. The additional proto{endash}neutron star convection is driven by successive negative entropy gradients that develop as the shock, in propagating out after core bounce, is successively strengthened and weakened by the oscillating inner core. The convection beneath the shock distorts its sphericity, but on the average the shock radius is not boosted significantly relative to its radius in our corresponding one-dimensional models. In the presence of neutrino transport, proto{endash}neutron star convection velocities are too small relative to bulk inflow velocities to result in any significant convective transport of entropy and leptons. This is evident in our two-dimensional entropy

  16. Double core evolution. 7: The infall of a neutron star through the envelope of its massive star companion

    NASA Technical Reports Server (NTRS)

    Terman, James L.; Taam, Ronald E.; Hernquist, Lars

    1995-01-01

    Binary systems with properties similar to those of high-mass X-ray binaries are evolved through the common envelope phase. Three-dimensional simulations show that the timescale of the infall phase of the neutron star depends upon the evolutionary state of its massive companion. We find that tidal torques more effectively accelerate common envelope evolution for companions in their late core helium-burning stage and that the infall phase is rapid (approximately several initial orbital periods). For less evolved companions the decay of the orbit is longer; however, once the neutron star is deeply embedded within the companion's envelope the timescale for orbital decay decreases rapidly. As the neutron star encounters the high-density region surrounding the helium core of its massive companion, the rate of energy loss from the orbit increases dramatically leading to either partial or nearly total envelope ejection. The outcome of the common envelope phase depends upon the structure of the evolved companion. In particular, it is found that the entire common envelope can be ejected by the interaction of the neutron star with a red supergiant companion in binaries with orbital periods similar to those of long-period Be X-ray binaries. For orbital periods greater than or approximately equal to 0.8-2 yr (for companions of mass 12-24 solar mass) it is likely that a binary will survive the common envelope phase. For these systems, the structure of the progenitor star is characterized by a steep density gradient above the helium core, and the common envelope phase ends with a spin up of the envelope to within 50%-60% of corotation and with a slow mass outflow. The efficiency of mass ejection is found to be approximately 30%-40%. For less evolved companions, there is insufficient energy in the orbit to unbind the common envelope and only a fraction of it is ejected. Since the timescale for orbital decay is always shorter than the mass-loss timescale from the common envelope

  17. NEUTRONIC REACTOR

    DOEpatents

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  18. Design, status and first operations of the spallation neutron source polyphase resonant converter modulator system

    SciTech Connect

    Reass, W. A.; Apgar, S. E.; Baca, D. M.; Doss, James D.; Gonzales, J.; Gribble, R. F.; Hardek, T. W.; Lynch, M. T.; Rees, D. E.; Tallerico, P. J.; Trujillo, P. B.; Anderson, D. E.; Heidenreich, D. A.; Hicks, J. D.; Leontiev, V. N.

    2003-01-01

    The Spallation Neutron Source (SNS) is a new 1.4 MW average power beam, 1 GeV accelerator being built at Oak Ridge National Laboratory. The accelerator requires 15 converter-modulator stations each providing between 9 and 11 MW pulses with up to a 1 .I MW average power. The converter-modulator can be described as a resonant 20 kHz polyphase boost inverter. Each converter modulator derives its buss voltage from a standard substation cast-core transformer. Each substation is followed by an SCR pre-regulator to accommodate voltage changes from no load to full load, in addition to providing a soft-start function. Energy storage is provided by self-clearing metallized hazy polypropylene traction capacitors. These capacitors do not fail short, but clear any internal anomaly. Three 'H-Bridge' IGBT transistor networks are used to generate the polyphase 20 kHz transformer primary drive waveforms. The 20 kHz drive waveforms are time-gated to generate the desired klystron pulse width. Pulse width modulation of the individual 20 lcHz pulses is utilized to provide regulated output waveforms with DSP based adaptive feedforward and feedback techniques. The boost transformer design utilizes nanocrystalline alloy that provides low core loss at design flux levels and switching frequencies. Capacitors are used on the transformer secondary networks to resonate the leakage inductance. The transformers are wound for a specific leakage inductance, not turns ratio. This design technique generates multiple secondary volts per turn as compared to the primary. With the appropriate tuning conditions, switching losses are minimized. The resonant topology has the added benefit of being deQed in a klystron fault condition, with little energy deposited in the arc. This obviates the need of crowbars or other related networks. A review of these design parameters, operational performance, production status, and OWL installation and performance to date will be presented.

  19. Dynamico, an Icosahedral Dynamical Core Designed for Consistency and Versatility

    NASA Astrophysics Data System (ADS)

    Dubos, T.

    2014-12-01

    The design of the icosahedral-hexagonal dynamical core DYNAMICO is presented. DYNAMICO solves the multi-layer rotating shallow-water equations, a compressible variant of the same equivalent to a discretization of the hydrostatic primitive equations (HPE) in a Lagrangian vertical coordinate, and the HPE in a hybrid mass-based vertical coordinate. In line with more general lines of thought known as physics-preserving discretizations and discrete differential geometry, kinematics and dynamics are separated as strictly as possible. This separation means that the transport of mass, scalars and potential temperature uses no information regarding the specific momentum equation being solved. This disregarded information includes the equation of state as well as any metric information, and is used only for certain terms of the momentum budget, written in Hamiltonian, vector-invariant form. The common Hamiltonian structure of the various equations of motion (Tort and Dubos, 2014 ; Dubos and Tort, 2014) is exploited to formulate energy-conserving spatial discretizations in a unified way. Furthermore most of the model code is common to the three sets of equations solved, making it easier to develop and validate each piece of the model separately. This design permits to consider several extensions in the near future, especially to deep-atmosphere, moist and non-hydrostatic equations. Representative academic three-dimensional benchmarks are run and analyzed, showing correctness of the model (Figure : time-zonal statistics from Held and Suarez (1994) simulations). Hopefully preliminary full-physics results will be presented as well. References : T. Dubos and M. Tort, "Equations of atmospheric motion in non-Eulerian vertical coordinates : vector-invariant form and Hamiltonian formulation", accepted by Mon. Wea. Rev. M. Tort and T. Dubos, "Usual approximations to the equations of atmospheric motion : a variational perspective" accepted by J. Atmos. Sci T. Dubos et al., "DYNAMICO

  20. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2013-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has been evaluated as an acceptable benchmark experiment. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  1. Design of a setup for 252Cf neutron source for storage and analysis purpose

    NASA Astrophysics Data System (ADS)

    Hei, Daqian; Zhuang, Haocheng; Jia, Wenbao; Cheng, Can; Jiang, Zhou; Wang, Hongtao; Chen, Da

    2016-11-01

    252Cf is a reliable isotopic neutron source and widely used in the prompt gamma ray neutron activation analysis (PGNAA) technique. A cylindrical barrel made by polymethyl methacrylate contained with the boric acid solution was designed for storage and application of a 5 μg 252Cf neutron source. The size of the setup was optimized with Monte Carlo code. The experiments were performed and the results showed the doses were reduced with the setup and less than the allowable limit. The intensity and collimating radius of the neutron beam could also be adjusted through different collimator.

  2. Proceedings of a workshop on methods for neutron scattering instrumentation design

    SciTech Connect

    Hjelm, R.P.

    1997-09-01

    The future of neutron and x-ray scattering instrument development and international cooperation was the focus of the workshop. The international gathering of about 50 participants representing 15 national facilities, universities and corporations featured oral presentations, posters, discussions and demonstrations. Participants looked at a number of issues concerning neutron scattering instruments and the tools used in instrument design. Objectives included: (1) determining the needs of the neutron scattering community in instrument design computer code and information sharing to aid future instrument development, (2) providing for a means of training scientists in neutron scattering and neutron instrument techniques, and (3) facilitating the involvement of other scientists in determining the characteristics of new instruments that meet future scientific objectives, and (4) fostering international cooperation in meeting these needs. The scope of the meeting included: (1) a review of x-ray scattering instrument design tools, (2) a look at the present status of neutron scattering instrument design tools and models of neutron optical elements, and (3) discussions of the present and future needs of the neutron scattering community. Selected papers were abstracted separately for inclusion to the Energy Science and Technology Database.

  3. Design, construction, and demonstration of a neutron beamline and a neutron imaging facility at a Mark-I TRIGA reactor

    NASA Astrophysics Data System (ADS)

    Craft, Aaron E.

    The fleet of research and training reactors is aging, and no new research reactors are planned in the United States. Thus, there is a need to expand the capabilities of existing reactors to meet users' needs. While many research reactors have beam port facilities, the original design of the United States Geological Survey TRIGA Reactor (GSTR) did not include beam ports. The MInes NEutron Radiography (MINER) facility developed by this thesis and installed at the GSTR provides new capabilities for both researchers and students at the Colorado School of Mines. The facility consists of a number of components, including a neutron beamline and beamstop, an optical table, an experimental enclosure and associated interlocks, a computer control system, a multi-channel plate imaging detector, and the associated electronics. The neutron beam source location, determined through Monte Carlo modeling, provides the best mixture of high neutron flux, high thermal neutron content, and low gamma radiation content. A Monte Carlo n-Particle (MCNP) model of the neutron beam provides researchers with a tool for designing experiments before placing objects in the neutron beam. Experimental multi-foil activation results, compared to calculated multi-foil activation results, verify the model. The MCNP model predicts a neutron beamline flux of 2.2*106 +/- 6.4*105 n/cm2-s based on a source particle rate determined from the foil activation experiments when the reactor is operating at a power of 950 kWt with the beam shutter fully open. The average cadmium ratio of the beamline is 7.4, and the L/D of the neutron beam is approximately 200+/-10. Radiographs of a sensitivity indicator taken using both the digital detector and the transfer foil method provide one demonstration of the radiographic capabilities of the new facility. Calibration fuel pins manufactured using copper and stainless steel surrogate fuel pellets provide additional specimens for demonstration of the new facility and offer a

  4. Coded aperture Fast Neutron Analysis: Latest design advances

    NASA Astrophysics Data System (ADS)

    Accorsi, Roberto; Lanza, Richard C.

    2001-07-01

    Past studies have showed that materials of concern like explosives or narcotics can be identified in bulk from their atomic composition. Fast Neutron Analysis (FNA) is a nuclear method capable of providing this information even when considerable penetration is needed. Unfortunately, the cross sections of the nuclear phenomena and the solid angles involved are typically small, so that it is difficult to obtain high signal-to-noise ratios in short inspection times. CAFNAaims at combining the compound specificity of FNA with the potentially high SNR of Coded Apertures, an imaging method successfully used in far-field 2D applications. The transition to a near-field, 3D and high-energy problem prevents a straightforward application of Coded Apertures and demands a thorough optimization of the system. In this paper, the considerations involved in the design of a practical CAFNA system for contraband inspection, its conclusions, and an estimate of the performance of such a system are presented as the evolution of the ideas presented in previous expositions of the CAFNA concept.

  5. A compact neutron beam generator system designed for prompt gamma nuclear activation analysis.

    PubMed

    Ghassoun, J; Mostacci, D

    2011-08-01

    In this work a compact system was designed for bulk sample analysis using the technique of PGNAA. The system consists of (252)Cf fission neutron source, a moderator/reflector/filter assembly, and a suitable enclosure to delimit the resulting neutron beam. The moderator/reflector/filter arrangement has been optimised to maximise the thermal neutron component useful for samples analysis with a suitably low level of beam contamination. The neutron beam delivered by this compact system is used to irradiate the sample and the prompt gamma rays produced by neutron reactions within the sample elements are detected by appropriate gamma rays detector. Neutron and gamma rays transport calculations have been performed using the Monte Carlo N-Particle transport code (MCNP5).

  6. Magnetic heating properties and neutron activation of tungsten-oxide coated biocompatible FePt core-shell nanoparticles.

    PubMed

    Seemann, K M; Luysberg, M; Révay, Z; Kudejova, P; Sanz, B; Cassinelli, N; Loidl, A; Ilicic, K; Multhoff, G; Schmid, T E

    2015-01-10

    Magnetic nanoparticles are highly desirable for biomedical research and treatment of cancer especially when combined with hyperthermia. The efficacy of nanoparticle-based therapies could be improved by generating radioactive nanoparticles with a convenient decay time and which simultaneously have the capability to be used for locally confined heating. The core-shell morphology of such novel nanoparticles presented in this work involves a polysilico-tungstate molecule of the polyoxometalate family as a precursor coating material, which transforms into an amorphous tungsten oxide coating upon annealing of the FePt core-shell nanoparticles. The content of tungsten atoms in the nanoparticle shell is neutron activated using cold neutrons at the Heinz Maier-Leibnitz (FRMII) neutron facility and thereby transformed into the radioisotope W-187. The sizeable natural abundance of 28% for the W-186 precursor isotope, a radiopharmaceutically advantageous gamma-beta ratio of γβ≈30% and a range of approximately 1mm in biological tissue for the 1.3MeV β-radiation are promising features of the nanoparticles' potential for cancer therapy. Moreover, a high temperature annealing treatment enhances the magnetic moment of nanoparticles in such a way that a magnetic heating effect of several degrees Celsius in liquid suspension - a prerequisite for hyperthermia treatment of cancer - was observed. A rise in temperature of approximately 3°C in aqueous suspension is shown for a moderate nanoparticle concentration of 0.5mg/ml after 15min in an 831kHz high-frequency alternating magnetic field of 250Gauss field strength (25mT). The biocompatibility based on a low cytotoxicity in the non-neutron-activated state in combination with the hydrophilic nature of the tungsten oxide shell makes the coated magnetic FePt nanoparticles ideal candidates for advanced radiopharmaceutical applications.

  7. Near-Core and In-Core Neutron Radiation Monitors for Real Time Neutron Flux Monitoring and Reactor Power Level Measurements

    SciTech Connect

    Douglas S. McGregor; Marvin L. Adams; Igor Carron; Paul Nelson

    2006-06-12

    MPFDs are a new class of detectors that utilize properties from existing radiation detector designs. A majority of these characteristics come from fission chamber designs. These include radiation hardness, gamma-ray background insensitivity, and large signal output.

  8. DESIGN OF A LARGE-AREA FAST NEUTRON DIRECTIONAL DETECTOR.

    SciTech Connect

    VANIER, P.E.

    2006-10-29

    A large-area fast-neutron double-scatter directional detector and spectrometer is being constructed using l-meter-long plastic scintillator paddles with photomultiplier tubes at both ends. The scintillators detect fast neutrons by proton recoil and also gamma rays by Compton scattering. The paddles are arranged in two parallel planes so that neutrons can be distinguished from muons and gamma rays by time of flight between the planes. The signal pulses are digitized with a time resolution of one gigasample per second. The location of an event along each paddle can be determined from the relative amplitudes or timing of the signals at the ends. The angle of deflection of a neutron in the first plane can be estimated from the energy deposited by the recoil proton, combined with the scattered neutron time-of-flight energy. Each scattering angle can be back-projected as a cone, and many intersecting cones define the incident neutron direction from a distant point source. Moreover, the total energy of each neutron can be obtained, allowing some regions of a fission source spectrum to be distinguished from background generated by cosmic rays. Monte Carlo calculations will be compared with measurements.

  9. Designing a minimum-functionality neutron and gamma measurement instrument with a focus on authentication

    SciTech Connect

    Karpius, Peter J; Williams, Richard B

    2009-01-01

    During the design and construction of the Next-Generation Attribute-Measurement System, which included a largely commercial off-the-shelf (COTS), nondestructive assay (NDA) system, we realized that commercial NDA equipment tends to include numerous features that are not required for an attribute-measurement system. Authentication of the hardware, firmware, and software in these instruments is still required, even for those features not used in this application. However, such a process adds to the complexity, cost, and time required for authentication. To avoid these added authenticat ion difficulties, we began to design NDA systems capable of performing neutron multiplicity and gamma-ray spectrometry measurements by using simplified hardware and software that avoids unused features and complexity. This paper discusses one possible approach to this design: A hardware-centric system that attempts to perform signal analysis as much as possible in the hardware. Simpler processors and minimal firmware are used because computational requirements are kept to a bare minimum. By hard-coding the majority of the device's operational parameters, we could cull large sections of flexible, configurable hardware and software found in COTS instruments, thus yielding a functional core that is more straightforward to authenticate.

  10. The 'virtual density' principle of neutronics: Toward rapid computation of reactivity effects in practical core distortion scenarios

    SciTech Connect

    Reed, M.; Smith, K.; Forget, B.

    2013-07-01

    Fast reactor core reactivities are sensitive to geometric distortions arising from three distinct phenomena: (1) irradiation swelling of fuel throughout core lifetime, (2) thermal expansion of fuel during transients, and (3) mechanical oscillations during seismic events. Performing comprehensive reactivity analysis of these distortions requires methods for rapidly computing a multitude of minute reactivity changes. Thus, we introduce the 'virtual density' principle of neutronics as a new perturbation technique to achieve this rapid computation. This new method obviates many of the most challenging aspects of conventional geometric perturbation theory. Essentially, this 'virtual density' principle converts geometric perturbations into equivalent material density perturbations (either isotropic or anisotropic), which are highly accurate and comparatively simple to evaluate. While traditional boundary perturbation theory employs surface integrals, the 'virtual density' principle employs equivalent volume integrals. We introduce and validate this method in three subsequent stages: (1) isotropic 'virtual density', (2) anisotropic 'virtual density' for whole cores, and (3) anisotropic 'virtual density' for interior zones within cores. We numerically demonstrate its accuracy for 2-D core flowering scenarios. (authors)

  11. Preliminary Neutronics Design and Analysis of D2O Cooled High Conversion PWRs

    SciTech Connect

    Hikaru Hiruta; Gilles Youinou

    2012-09-01

    This report presents a neutronics analysis of tight-pitch D2O-cooled PWRs loaded with MOX fuel and focuses essentially on the Pu breeding potential of such reactors as well as on an important safety parameter, the void coefficient, which has to be negative. It is well known that fast reactors have a better neutron economy and are better suited than thermal reactors to breed fissile material from neutron capture in fertile material. Such fast reactors (e.g. sodium-cooled reactors) usually rely on technologies that are very different from those of existing water-cooled reactors and are probably more expensive. This report investigates another possibility to obtain a fast neutron reactor while still relying mostly on a PWR technology by: (1) Tightening the lattice pitch to reduce the water-to-fuel volume ratio compared to that of a standard PWR. Water-to-fuel volume ratios of between 0.45 and 1 have been considered in this study while a value of about 2 is typical of standard PWRs, (2) Using D2O instead of H2O as a coolant. Indeed, because of its different neutron physics properties, the use of D2O hardens the neutron spectrum to an extent impossible with H2O when used in a tight-pitch lattice. The neutron spectra thus obtained are not as fast as those in sodium-cooled reactor but they can still be characterized as fast compared to that of standard PWR neutron spectra. In the phase space investigated in this study we did not find any configurations that would have, at the same time, a positive Pu mass balance (more Pu at the end than at the beginning of the irradiation) and a negative void coefficient. At this stage, the use of radial blankets has only been briefly addressed whereas the impact of axial blankets has been well defined. For example, with a D2O-to-fuel volume ratio of 0.45 and a core driver height of about 60 cm, the fissile Pu mass balance between the fresh fuel and the irradiated fuel (50 GWd/t) would be about -7.5% (i.e. there are 7.5% fewer fissile Pu

  12. Characterization of core debris/concrete interactions for the Advanced Neutron Source. ANS Severe Accident Analysis Program

    SciTech Connect

    Hyman, C.R.; Taleyarkhan, R.P.

    1992-02-01

    This report provides the results of a recent study conducted to explore the molten core/concrete interaction (MCCI) issue for the Advanced Neutron Source (ANS). The need for such a study arises from the potential threats to reactor system integrity posed by MCCI. These threats include direct attack of the concrete basemat of the containment; generation and release of large quantities of gas that can pressurize the containment; the combustion threat of these gases; and the potential generation, release, and transport of radioactive aerosols to the environment.

  13. Preliminary detector design ST862-prototype neutron detector

    SciTech Connect

    Miller, S.D.; Affinito, J.D.; Sisk, D.R.

    1993-12-01

    The detection of fast neutrons has been accomplished with commercially available liquid scintillators in detectors. Liquid scintillators discriminate fast neutrons from gamma radiation by discarding pulses with short decay constants. However, pulse-timing methods require expensive, bulky equipment and a high degree of technical sophistication in the user. Researchers at Pacific Northwest Laboratory have developed a new class of scintillating material, polymerizing crystals of CaF{sub 2}(Eu) and liquid acrylate monomers with matched indexes of refraction. The new detectors avoid the pulse-timing methods of liquid detectors and allow detectors to be large and relatively light. Fast neutrons can be discriminated from gamma radiation solely on the basis of pulse height (i.e., energy deposition). Using these detectors, a hand-held neutron detection instrument is proposed that can operate on battery power for 8 to 12 hours and be easily used in field conditions for surveying vehicles and structures.

  14. Development of 3D full-core ERANOS-2.2/MCNPX-2.7.0 models and neutronic analysis of the BFS-2 zero-power facility

    SciTech Connect

    Girardin, G.; Alonso, M.; Mikityuk, K.

    2012-07-01

    The present paper is addressing the development and validation against experimental data of 3D full-core models of the BFS-2 zero-power fast-reactor using both the deterministic system code ERANOS-2.2 and the stochastic code MCNPX-2.7.0. The model configuration of BFS considered for analysis is the BFS-62-3A benchmark. To extend the - deterministic/stochastic - code-to-code comparison, neutronic parameters, i.e. reactivity, neutron spectrum and reaction rates, were also simulated at the cell level with the Monte Carlo code SERPENT-1.1.7 with two modern data libraries, ENDF-B/VII and JEFF-3.1.1. The BFS-2 critical zero-power facility at the Inst. of Physics and Power Engineering (IPPE) was designed for simulations of the core and shielding of sodium-cooled, fast reactors, for neutron data validation and comparison with experimental results. At the BFS-2 facility, the BFS-62-3A critical benchmark experiment was set-up as a mock-up of the BN-600 reactor, with hybrid MOX fuel and stainless steel reflectors. A UO{sub 2} blanket and a large non-homogeneous stainless-steel reflector surround the core. The lattice is hexagonal of pitch 5.1 cm and metallic dowels are used to keep in central position cylindrical rods made of different types of material (fissile, fertile, blanket, plenum, shielding and absorber). A typical subassembly is formed in piling up various pellets of about 1 cm in height and 4.6 cm in diameter, conferring large heterogeneity in the axial direction. The full-core model development was a complex task due to the large number of subassemblies and the axial subassembly heterogeneity. In ERANOS-2.2, it was necessary to homogenize axially per region the pellets used to form the subassembly. The self-shielded macroscopic cross-sections were calculated using the cell code ECCO in association with JEFF-3.1 and ENDF/B-VI.8 data libraries. The core calculations were performed with broad cross-sections data in 33 neutron energy groups with the solver AVNM in the

  15. Radiation shielding design of BNCT treatment room for D-T neutron source.

    PubMed

    Pouryavi, Mehdi; Farhad Masoudi, S; Rahmani, Faezeh

    2015-05-01

    Recent studies have shown that D-T neutron generator can be used as a proper neutron source for Boron Neutron Capture Therapy (BNCT) of deep-seated brain tumors. In this paper, radiation shielding calculations have been conducted based on the computational method for designing a BNCT treatment room for a recent proposed D-T neutron source. By using the MCNP-4C code, the geometry of the treatment room has been designed and optimized in such a way that the equivalent dose rate out of the treatment room to be less than 0.5μSv/h for uncontrolled areas. The treatment room contains walls, monitoring window, maze and entrance door. According to the radiation protection viewpoint, dose rate results of out of the proposed room showed that using D-T neutron source for BNCT is safe.

  16. IB: A Monte Carlo simulation tool for neutron scattering instrument design under PVM and MPI

    NASA Astrophysics Data System (ADS)

    Zhao, Jinkui

    2011-12-01

    Design of modern neutron scattering instruments relies heavily on Monte Carlo simulation tools for optimization. IB is one such tool written in C++ and implemented under Parallel Virtual Machine and the Message Passing Interface. The program was initially written for the design and optimization of the EQ-SANS instrument at the Spallation Neutron Source. One of its features is the ability to group simple instrument components into more complex ones at the user input level, e.g. grouping neutron mirrors into neutron guides and curved benders. The simulation engine manages the grouped components such that neutrons entering a group are properly operated upon by all components, multiple times if needed, before exiting the group. Thus, only a few basic optical modules are needed at the programming level. For simulations that require higher computer speeds, the program can be compiled and run in parallel modes using either the PVM or the MPI architectures.

  17. ATR LEU Monothlic and Dispersed with 10B Loading Minimization DesignNeutronics Performance Analysis

    SciTech Connect

    G. S. Chang

    2001-10-01

    The Advanced Test Reactor (ATR), currently operating in the United States, is used for material testing at very high neutron fluxes. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting HEU driven reactor cores to low-enriched uranium (LEU) cores. The present work investigates the optimized LEU Monolithic and Dispersed fuel with 10B loading minimization design and evaluates the subsequent neutronics operating effects of these optimized fuel designs. The MCNP ATR 1/8th core model was used to optimize the 235U and minimize the 10B loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The fuel depletion methodology MCWO was used to calculate K eff versus effective full power days (EFPD) in this paper. The MCWO-calculated results for the optimized LEU Monolithic and Dispersed fuel cases demonstrated adequate excess reactivity such that the K-eff versus EFPD plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU Monolithic (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and 235U enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness. The proposed LEU fuel meat varies from 0.203 mm (8.0 mil) to 0.254 mm (10.0 mil) at the inner four fuel plates (1-4) and outer four fuel plates (16-19). In addition, an optimized LEU dispersed (U7Mo) case with all the fuel meat thickness of 0.635 mm (25 mil) was also proposed. Then, for both Monolithic and dispersed cases, a burnable absorber – 10B, was added in the inner and outer plates to reduce the initial excess reactivity, and the higher to average ratio of the inner

  18. Design Core Commonalities: A Study of the College of Design at Iowa State University

    ERIC Educational Resources Information Center

    Venes, Jane

    2015-01-01

    This comprehensive study asks what a group of rather diverse disciplines have in common. It involves a cross-disciplinary examination of an entire college, the College of Design at Iowa State University. This research was intended to provide a sense of direction in developing and assessing possible core content. The reasoning was that material…

  19. Exact versus Taylor-expanded energy density in the study of the neutron star crust-core transition

    NASA Astrophysics Data System (ADS)

    Routray, T. R.; Viñas, X.; Basu, D. N.; Pattnaik, S. P.; Centelles, M.; Robledo, L. B.; Behera, B.

    2016-10-01

    The importance of the fourth and higher order terms in the Taylor series expansion of energy of isospin asymmetric nuclear matter in studies of the neutron star crust-core phase transition is investigated using the finite-range simple effective interaction. Analytic expressions for the evaluation of the second and fourth order derivative terms in the Taylor series expansion for any general finite-range interaction of Yukawa, exponential or Gaussian form have been obtained. The effect of the nuclear matter incompressibility, symmetry energy and slope parameters on the predictions for the crust-core transition density is examined. The crustal moment of inertia is calculated and the prediction for the radius of the Vela pulsar is analyzed using different equations of state.

  20. Development and applications of methodologies for the neutronic design of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR)

    NASA Astrophysics Data System (ADS)

    Fratoni, Massimiliano

    This study investigated the neutronic characteristics of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a novel nuclear reactor concept that combines liquid salt (7LiF-BeF2---flibe) cooling and TRISO coated-particle fuel technology. The use of flibe enables operation at high power density and atmospheric pressure and improves passive decay-heat removal capabilities, but flibe, unlike conventional helium coolant, is not transparent to neutrons. The flibe occupies 40% of the PB-AHTR core volume and absorbs ˜8% of the neutrons, but also acts as an effective neutron moderator. Two novel methodologies were developed for calculating the time dependent and equilibrium core composition: (1) a simplified single pebble model that is relatively fast; (2) a full 3D core model that is accurate and flexible but computationally intensive. A parametric analysis was performed spanning a wide range of fuel kernel diameters and graphite-to-heavy metal atom ratios to determine the attainable burnup and reactivity coefficients. Using 10% enriched uranium ˜130 GWd/tHM burnup was found to be attainable, when the graphite-to-heavy metal atom ratio (C/HM) is in the range of 300 to 400. At this or smaller C/HM ratio all reactivity coefficients examined---coolant temperature, coolant small and full void, fuel temperature, and moderator temperature, were found to be negative. The PB-AHTR performance was compared to that of alternative options for HTRs, including the helium-cooled pebble-bed reactor and prismatic fuel reactors, both gas-cooled and flibe-cooled. The attainable burnup of all designs was found to be similar. The PB-AHTR generates at least 30% more energy per pebble than the He-cooled pebble-bed reactor. Compared to LWRs the PB-AHTR requires 30% less natural uranium and 20% less separative work per unit of electricity generated. For deep burn TRU fuel made from recycled LWR spent fuel, it was found that in a single pass through the core ˜66% of the TRU can be

  1. Verification of the CENTRM Module for Adaptation of the SCALE Code to NGNP Prismatic and PBR Core Designs

    SciTech Connect

    Ganapol, Barry; Maldonado, Ivan

    2014-01-23

    The generation of multigroup cross sections lies at the heart of the very high temperature reactor (VHTR) core design, whether the prismatic (block) or pebble-bed type. The design process, generally performed in three steps, is quite involved and its execution is crucial to proper reactor physics analyses. The primary purpose of this project is to develop the CENTRM cross-section processing module of the SCALE code package for application to prismatic or pebble-bed core designs. The team will include a detailed outline of the entire processing procedure for application of CENTRM in a final report complete with demonstration. In addition, they will conduct a thorough verification of the CENTRM code, which has yet to be performed. The tasks for this project are to: Thoroughly test the panel algorithm for neutron slowing down; Develop the panel algorithm for multi-materials; Establish a multigroup convergence 1D transport acceleration algorithm in the panel formalism; Verify CENTRM in 1D plane geometry; Create and test the corresponding transport/panel algorithm in spherical and cylindrical geometries; and, Apply the verified CENTRM code to current VHTR core design configurations for an infinite lattice, including assessing effectiveness of Dancoff corrections to simulate TRISO particle heterogeneity.

  2. Neutron capture on Pt isotopes in iron meteorites and the Hf-W chronology of core formation in planetesimals

    NASA Astrophysics Data System (ADS)

    Kruijer, Thomas S.; Fischer-Gödde, Mario; Kleine, Thorsten; Sprung, Peter; Leya, Ingo; Wieler, Rainer

    2013-01-01

    The short-lived 182Hf-182W isotope system can provide powerful constraints on the timescales of planetary core formation, but its application to iron meteorites is hampered by neutron capture reactions on W isotopes resulting from exposure to galactic cosmic rays. Here we show that Pt isotopes in magmatic iron meteorites are also affected by capture of (epi)thermal neutrons and that the Pt isotope variations are correlated with variations in 182W/184W. This makes Pt isotopes a sensitive neutron dosimeter for correcting cosmic ray-induced W isotope shifts. The pre-exposure 182W/184W derived from the Pt-W isotope correlations of the IID, IVA and IVB iron meteorites are higher than most previous estimates and are more radiogenic than the initial 182W/184W of Ca-Al-rich inclusions (CAI). The Hf-W model ages for core formation range from +1.6±1.0 million years (Ma; for the IVA irons) to +2.7±1.3 Ma after CAI formation (for the IID irons), indicating that there was a time gap of at least ˜1 Ma between CAI formation and metal segregation in the parent bodies of some iron meteorites. From the Hf-W ages a time limit of <1.5-2 Ma after CAI formation can be inferred for the accretion of the IID, IVA and IVB iron meteorite parent bodies, consistent with earlier conclusions that the accretion of differentiated planetesimals predated that of most chondrite parent bodies.

  3. Computational Benchmark Calculations Relevant to the Neutronic Design of the Spallation Neutron Source (SNS)

    SciTech Connect

    Gallmeier, F.X.; Glasgow, D.C.; Jerde, E.A.; Johnson, J.O.; Yugo, J.J.

    1999-11-14

    The Spallation Neutron Source (SNS) will provide an intense source of low-energy neutrons for experimental use. The low-energy neutrons are produced by the interaction of a high-energy (1.0 GeV) proton beam on a mercury (Hg) target and slowed down in liquid hydrogen or light water moderators. Computer codes and computational techniques are being benchmarked against relevant experimental data to validate and verify the tools being used to predict the performance of the SNS. The LAHET Code System (LCS), which includes LAHET, HTAPE ad HMCNP (a modified version of MCNP version 3b), have been applied to the analysis of experiments that were conducted in the Alternating Gradient Synchrotron (AGS) facility at Brookhaven National Laboratory (BNL). In the AGS experiments, foils of various materials were placed around a mercury-filled stainless steel cylinder, which was bombarded with protons at 1.6 GeV. Neutrons created in the mercury target, activated the foils. Activities of the relevant isotopes were accurately measured and compared with calculated predictions. Measurements at BNL were provided in part by collaborating scientists from JAERI as part of the AGS Spallation Target Experiment (ASTE) collaboration. To date, calculations have shown good agreement with measurements.

  4. Thermal neutron irradiation field design for boron neutron capture therapy of human explanted liver.

    PubMed

    Bortolussi, S; Altieri, S

    2007-12-01

    The selective uptake of boron by tumors compared to that by healthy tissue makes boron neutron capture therapy (BNCT) an extremely advantageous technique for the treatment of tumors that affect a whole vital organ. An example is represented by colon adenocarcinoma metastases invading the liver, often resulting in a fatal outcome, even if surgical resection of the primary tumor is successful. BNCT can be performed by irradiating the explanted organ in a suitable neutron field. In the thermal column of the Triga Mark II reactor at Pavia University, a facility was created for this purpose and used for the irradiation of explanted human livers. The neutron field distribution inside the organ was studied both experimentally and by means of the Monte Carlo N-particle transport code (MCNP). The liver was modeled as a spherical segment in MCNP and a hepatic-equivalent solution was used as an experimental phantom. In the as-built facility, the ratio between maximum and minimum flux values inside the phantom ((phi(max)/phi(min)) was 3.8; this value can be lowered to 2.3 by rotating the liver during the irradiation. In this study, the authors proposed a new facility configuration to achieve a uniform thermal neutron flux distribution in the liver. They showed that a phi(max)/phi(min) ratio of 1.4 could be obtained without the need for organ rotation. Flux distributions and dose volume histograms were reported for different graphite configurations. PMID:18196797

  5. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    NASA Astrophysics Data System (ADS)

    Borodkin, Pavel; Borodkin, Gennady; Khrennikov, Nikolay

    2016-02-01

    Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  6. Boron tracedrug design for neutron dynamic therapeutics for LDL.

    PubMed

    Hori, Hitoshi; Nazumi, Yoshijiro; Uto, Yoshihiro

    2013-01-01

    We describe our solution for removal of the low-density lipoprotein (LDL) depot contained in proteins and lipids as a 'druggable' target for atherosclerotic cardiovascular diseases by neutron dynamic therapy (NDT), which we developed using boron tracedrugs for NDT against bovine serum albumin as a model protein. Thus, we examined, among our developed boron tracedrugs, a boron-containing curcuminoid derivative UTX-51, to destroy freshly isolated human LDL dynamically under irradiated thermal neutron to obtain a decreased intensity of band of LDL treated with UTX-51 and thermal neutron irradiation in their SDS-PAGE and electrophoresis analysis. These results suggest that UTX-51 might be a novel candidate of 'beyond chemical' therapeutic agents for atherosclerotic cardiovascular disease.

  7. Boron tracedrug design for neutron dynamic therapeutics for LDL.

    PubMed

    Hori, Hitoshi; Nazumi, Yoshijiro; Uto, Yoshihiro

    2013-01-01

    We describe our solution for removal of the low-density lipoprotein (LDL) depot contained in proteins and lipids as a 'druggable' target for atherosclerotic cardiovascular diseases by neutron dynamic therapy (NDT), which we developed using boron tracedrugs for NDT against bovine serum albumin as a model protein. Thus, we examined, among our developed boron tracedrugs, a boron-containing curcuminoid derivative UTX-51, to destroy freshly isolated human LDL dynamically under irradiated thermal neutron to obtain a decreased intensity of band of LDL treated with UTX-51 and thermal neutron irradiation in their SDS-PAGE and electrophoresis analysis. These results suggest that UTX-51 might be a novel candidate of 'beyond chemical' therapeutic agents for atherosclerotic cardiovascular disease. PMID:23852519

  8. Design and Rationale for an In Situ Cryogenic Deformation Capability at a Neutron Source

    SciTech Connect

    Livescu, V.; Clausen, B.; Sisneros, T.; Bourke, M.A.M.; Woodruff, T.R.; Vaidyanathan, R.; Notardonato, W.U.

    2004-06-28

    When performed in conjunction with neutron diffraction, in situ loading offers unique insights on microstructural deformation mechanisms. This is by virtue of the penetration and phase sensitivity of neutrons. At Los Alamos National Laboratory room and high temperature (up to 1500 deg. C) polycrystalline constitutive response is modeled using finite element and self-consistent models. The models are compared to neutron diffraction measurements. In doing so the implications of slip and creep to microstructural response have been explored. Recently we have been considering low temperature phenomena. This includes changes in deformation mechanisms such as the increased predilection for twinning over slip. Since this is associated with measurable texture changes as well as microstructural strain effects, it is well suited for study using neutron diffraction. This paper outlines the design and rationale for a cryogenic loading capability that will be used on the Spectrometer for MAterials Research at Temperature and Stress (SMARTS) at the Los Alamos Neutron Science Center (LANSCE)

  9. Thermo-mechanical and neutron lifetime modelling and design of Be pebbles in the neutron multiplier for the LIFE engine

    NASA Astrophysics Data System (ADS)

    DeMange, P.; Marian, J.; Caro, M.; Caro, A.

    2009-11-01

    Concept designs for the laser inertial fusion/fission energy (LIFE) engine include a neutron multiplication blanket containing Be pebbles flowing in a molten salt coolant. These pebbles must be designed to withstand the extreme irradiation and temperature conditions in the blanket to enable a reliable and cost-effective operation of LIFE. In this work, we develop design criteria for spherical Be pebbles on the basis of their thermo-mechanical behaviour under continued neutron exposure. We consider the effects of high fluence and fast fluxes on the elastic, thermal and mechanical properties of nuclear-grade Be. Our results suggest a maximum pebble diameter of 30 mm to avoid tensile failure, coated with an anti-corrosive, high-strength metallic shell to avoid failure by pebble contact. Moreover, we find that the operation temperature must always be kept above 450 °C to enable creep to relax the stresses induced by swelling. Under these circumstances, we estimate the pebble lifetime to be at least 16 months if uncoated, and up to six years when coated. We identify the sources of uncertainty on the properties used and discuss the advantages of new intermetallic beryllides and their use in LIFE's neutron multiplier. To establish Be-pebble lifetimes with improved confidence, reliable experiments to measure irradiation creep must be performed.

  10. Thermo-mechanical and neutron lifetime modeling and design of Be pebbles in the neutron multiplier for the LIFE engine

    SciTech Connect

    DeMange, P; Marian, J; de Caro, M S; Caro, A

    2009-03-16

    Concept designs for the laser-initiated fusion/fission engine (LIFE) include a neutron multiplication blanket containing Be pebbles flowing in a molten salt coolant. These pebbles must be designed to withstand the extreme irradiation and temperature conditions in the blanket to enable a safe and cost-effective operation of LIFE. In this work, we develop design criteria for spherical Be pebbles on the basis of their thermomechanical behavior under continued neutron exposure. We consider the effects of high fluence/fast flux on the elastic, thermal and mechanical properties of nuclear-grade Be. Our results suggest a maximum pebble diameter of 30 mm to avoid tensile failure, coated with an anti-corrosive, high-strength metallic shell to avoid failure by pebble contact. Moreover, we find that the operation temperature must always be kept above 450 C to enable creep to relax the stresses induced by swelling, which we estimate to be at least 16 months if uncoated and up to six years when coated. We identify the sources of uncertainty on the properties used and discuss the advantages of new intermetallic beryllides and their use in LIFE's neutron multiplier. To establish Be-pebble lifetimes with improved confidence, reliable experiments to measure irradiation creep must be performed.

  11. New insights on the spin-up of a neutron star during core collapse

    NASA Astrophysics Data System (ADS)

    Kazeroni, Rémi; Guilet, Jérôme; Foglizzo, Thierry

    2016-02-01

    The spin of a neutron star at birth may be impacted by the asymmetric character of the explosion of its massive progenitor. During the first second after bounce, the spiral mode of the Standing Accretion Shock Instability (SASI) is able to redistribute angular momentum and spin up a neutron star born from a non-rotating progenitor. Our aim is to assess the robustness of this process. We perform 2D numerical simulations of a simplified setup in cylindrical geometry to investigate the timescale over which the dynamics is dominated by a spiral or a sloshing mode. We observe that the spiral mode prevails only if the ratio of the initial shock radius to the neutron star radius exceeds a critical value. In that regime, both the degree of asymmetry and the average expansion of the shock induced by the spiral mode increase monotonously with this ratio, exceeding the values obtained when a sloshing mode is artificially imposed. With a timescale of 2-3 SASI oscillations, the dynamics of SASI takes place fast enough to affect the spin of the neutron star before the explosion. The spin periods deduced from the simulations are compared favourably to analytical estimates evaluated from the measured saturation amplitude of the SASI wave. Despite the simplicity of our setup, numerical simulations revealed unexpected stochastic variations, including a reversal of the direction of rotation of the shock. Our results show that the spin-up of neutron stars by SASI spiral modes is a viable mechanism even though it is not systematic.

  12. Development of Optimized Core Design and Analysis Methods for High Power Density BWRs

    NASA Astrophysics Data System (ADS)

    Shirvan, Koroush

    Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between ~he fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x16 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The plant systems outside the vessel are assumed to be the same as the ABWR-Il design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULA TE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater

  13. Development of Optimized Core Design and Analysis Methods for High Power Density BWRs

    NASA Astrophysics Data System (ADS)

    Shirvan, Koroush

    Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between ~he fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x16 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The plant systems outside the vessel are assumed to be the same as the ABWR-Il design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULA TE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater

  14. Laser Intertial Fusion Energy: Neutronic Design Aspects of a Hybrid Fusion-Fission Nuclear Energy System

    SciTech Connect

    Kramer, Kevin James

    2010-04-08

    This study investigates the neutronics design aspects of a hybrid fusion-fission energy system called the Laser Fusion-Fission Hybrid (LFFH). A LFFH combines current Laser Inertial Confinement fusion technology with that of advanced fission reactor technology to produce a system that eliminates many of the negative aspects of pure fusion or pure fission systems. When examining the LFFH energy mission, a significant portion of the United States and world energy production could be supplied by LFFH plants. The LFFH engine described utilizes a central fusion chamber surrounded by multiple layers of multiplying and moderating media. These layers, or blankets, include coolant plenums, a beryllium (Be) multiplier layer, a fertile fission blanket and a graphite-pebble reflector. Each layer is separated by perforated oxide dispersion strengthened (ODS) ferritic steel walls. The central fusion chamber is surrounded by an ODS ferritic steel first wall. The first wall is coated with 250-500 μm of tungsten to mitigate x-ray damage. The first wall is cooled by Li17Pb83 eutectic, chosen for its neutron multiplication and good heat transfer properties. The Li17Pb83 flows in a jacket around the first wall to an extraction plenum. The main coolant injection plenum is immediately behind the Li17Pb83, separated from the Li17Pb83 by a solid ODS wall. This main system coolant is the molten salt flibe (2LiF-BeF2), chosen for beneficial neutronics and heat transfer properties. The use of flibe enables both fusion fuel production (tritium) and neutron moderation and multiplication for the fission blanket. A Be pebble (1 cm diameter) multiplier layer surrounds the coolant injection plenum and the coolant flows radially through perforated walls across the bed. Outside the Be layer, a fission fuel layer comprised of depleted uranium contained in Tristructural-isotropic (TRISO) fuel particles

  15. A novel design approach for a neutron measurement station for burnt fuel

    NASA Astrophysics Data System (ADS)

    Dietler, Rodolfo; Hursin, Mathieu; Perret, Gregory; Jordan, Kelly; Chawla, Rakesh

    2012-11-01

    The design and characterization of a passive neutron measurement station for highly burnt fuel has been undertaken at the Paul Scherrer Institute (PSI). The measurement station aims at the determination of the total neutron emission rate of full-length light water reactor (LWR) fuel rods, as also the corresponding axial distributions. It is intended that the measurement station be introduced into the hot cells available at PSI to allow measuring the neutron emission of spent fuel rods provided by the Swiss nuclear power plants. In addition, the neutron emission of a large set of burnt fuel samples that have been previously characterized by post-irradiation examination (PIE) will be measured, in order to relate neutron emission to the burnup and isotopic composition of different fuel types. The design of the measurement station is presented in this article. A post-processing algorithm is introduced to improve the spatial resolution of the "measured" axial profile. In order to quantify the accuracy of the reconstructed neutron source distribution, a figure-of-merit (FOM) is defined and adapted to the detection procedure. With the optimized measurement station and procedure, it is estimated that the neutron emission distribution of a highly burnt, full-length fuel rod would be measurable with acceptable accuracy in about 20 min.

  16. Nuclear design of the burst power ultrahigh temperature UF4 vapor core reactor system

    NASA Astrophysics Data System (ADS)

    Kahook, Samer D.; Dugan, Edward T.

    1991-01-01

    Static and dynamic neutronic analyses are being performed, as part of an integrated series of studies, on an innovative burst power UF4 Ultrahigh Temperature Vapor Core Reactor (UTVR)/Disk Magnetohydrodynamic (MHD) generator for space nuclear power applications. This novel reactor concept operates on a direct, closed Rankine cycle in the burst power mode (hundreds of MWe for thousands of seconds). The fuel/working fluid is a mixture of UF4 and metal fluoride. Preliminary calculations indicate high overall system efficiencies (≊20%), small radiator size (≊5 m2/MWe), and high specific power (≊5 kWe/kg). Neutronic analysis has revealed a number of attractive features for this novel reactor concept. These include some unique and very effective inherent negative reactivity control mechanisms such as the vapor-fuel density power coefficient of reactivity, the direct neutronic coupling among the multiple fissioning core regions (the central vapor core and the surrounding boiler columns), and the mass flow coupling feedback between the fissioning cores.

  17. Design and performance of a cryogenic apparatus for magnetically trapping ultracold neutrons

    NASA Astrophysics Data System (ADS)

    Huffman, P. R.; Coakley, K. J.; Doyle, J. M.; Huffer, C. R.; Mumm, H. P.; O'Shaughnessy, C. M.; Schelhammer, K. W.; Seo, P.-N.; Yang, L.

    2014-11-01

    The cryogenic design and performance of an apparatus used to magnetically confine ultracold neutrons (UCN) is presented. The apparatus is part of an effort to measure the beta-decay lifetime of the free neutron and is comprised of a high-current superconducting magnetic trap that surrounds ∼21 l of isotopically pure 4He cooled to approximately 250 mK. A 0.89 nm neutron beam can enter the apparatus from one end of the magnetic trap and a light collection system allows visible light generated within the helium by decays to be transported to detectors at room temperature. Two cryocoolers are incorporated to reduce liquid helium consumption.

  18. Neutronics analysis of an open-cycle high-impulse gas core reactor concept

    NASA Technical Reports Server (NTRS)

    Whitmarsh, C. L., Jr.

    1972-01-01

    A procedure was developed to calculate the critical fuel mass, including the effects of propellant pressure, for coaxial-flow gas-core reactors operating at 196,600 newtons thrust and 4400 seconds specific impulse. Data were generated for a range of cavity diameter, reflector-moderator thickness, and quantity of structural material. Also presented are such core characteristics as upper limits on cavity pressure, spectral hardening in very-high-temperature hydrogen, and reactivity coefficients.

  19. Feasibility study on nuclear core design for soluble boron free small modular reactor

    SciTech Connect

    Rabir, Mohamad Hairie Hah, Chang Joo; Ju, Cho Sung

    2015-04-29

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

  20. Feasibility study on nuclear core design for soluble boron free small modular reactor

    NASA Astrophysics Data System (ADS)

    Rabir, Mohamad Hairie; Hah, Chang Joo; Ju, Cho Sung

    2015-04-01

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

  1. The performance of 3500 MWth homogeneous and heterogeneous metal fueled core designs

    SciTech Connect

    Turski, R.; Yang, Shi-tien

    1987-11-01

    Performance parameters are calculated for a representative 3500 MWth homogeneous and a heterogeneous metal fueled reactor design. The equilibrium cycle neutronic characteristics, safety coefficients, control system requirements, and control rod worths are evaluated. The thermal-hydraulic characteristics for both configurations are also compared. The heavy metal fuel loading requirements and neutronic performance characteristics are also evaluated for the uranium startup option. 14 refs., 14 figs., 20 tabs.

  2. Conceptual design for one megawatt spallation neutron source at Argonne

    SciTech Connect

    Chio, Y.; Bailey, J.; Brown, B.

    1993-12-31

    The feasibility study of a spallation neutron source based on a rapid cycling synchrotron which delivers a proton beam of 2 GeV in energy and 0.5mA time-average current at a 30-Hz repetition rate is presented. The lattice consists of 90-degree phase advanced FODO cells with dispersion-free straight sections, and has a three-fold symmetry. The ring magnet system will be energized by 20-Hz and 60-Hz resonant circuits to decrease the dB/dt during the acceleration cycle. This lowers the peak acceleration voltage requirement to 130kV. The single turn extraction system will be used to extract the beam alternatively to two target stations. The first station will operate at 10Hz for research using long wavelength neutrons, and the second station will use the remaining pulses, collectively, providing 36 neutron beams. The 400-MeV negative-hydrogen-ion injector linac consists of an ion source, rf quadrupole, matching section, 100MeV drift-tube linac, and a 300-Mev coupled-cavity linac.

  3. 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.

  4. Neutron stars: From the inner crust to the core with the (extended) Nambu-Jona-Lasinio model

    NASA Astrophysics Data System (ADS)

    Pais, Helena; Menezes, Débora P.; Providência, Constança

    2016-06-01

    Nucleonic matter is described within an SU(2) extended Nambu-Jona-Lasinio (NJL) model. Several parametrizations with different nuclear matter saturation properties are proposed. At subsaturation, nuclear pasta phases are calculated within two methods: the coexistence-phases approximation and the compressible liquid drop model, with the surface tension coefficient determined using a geometrical approach at zero temperature. A unified equation of state of stellar matter for the inner crust, with the nuclear pasta phases, and the core is calculated. The mass and radius of neutron stars within this framework are obtained for several families of hadronic and hybrid stars. The quark phase of hybrid stars is described within the SU(3) NJL model including a vector term. Stellar macroscopic properties are in accordance with some of the recent results in the literature.

  5. Design of Ultra Small Angle Neutron Scattering (KIST-USANS) at HANARO Cold Neutron Guide, CG4B

    NASA Astrophysics Data System (ADS)

    Kim, Man-Ho

    2013-03-01

    The ultra small angle neutron scattering instrument can measure the lower limit of scattering vector to near Q ~ 2.0x10-5 Å-1 while the upper limit can reach to an intermediate scattering vector Q ~ 10-2 Å-1 of a typical small angle neutron scattering (SANS) depending on the contrast of sample. USANS is useful when measuring objects that are micron to submicron in size while SANS is useful when measuring objects that are micron to nano in size. When both USANS and SANS were used together, we could measure sizes from micron to nano scale, which is useful when studying the hierarchical structures in the wide scale of Q and total cross-section, d Σ/d Ω(Q). Recently, KIST has developed the USANS (so called KIST-USANS) at HANARO cold neutron guide hall of KAERI. We will present the instrument design, performance, future plan, and some examples of measurements that cover approximately 11 orders of magnitude in the d Σ/d Ω(Q) and 4 orders in the Q. This work was partially supported by the KIST (2v02632) and the National Research Foundation of Korea(NRF) grant funded by the Korea government(MEST) (No. 2012M2B2A4030220)

  6. Correlation of the neutron star crust-core properties with the slope of the symmetry energy and the lead skin thickness

    NASA Astrophysics Data System (ADS)

    Pais, H.; Sulaksono, A.; Agrawal, B. K.; Providência, C.

    2016-04-01

    The correlations of the crust-core transition density and pressure in neutron stars with the slope of the symmetry energy and the neutron skin thickness are investigated, using different families of relativistic mean-field parametrizations with constant couplings and nonlinear terms mixing the σ - , ω - , and ρ -meson fields. It is shown that the modification of the density dependence of the symmetry energy, involving the σ or the ω meson, gives rise to different behaviors: the effect of the ω meson may also be reproduced within nonrelativistic phenomenological models, while the effect of the σ meson is essentially relativistic. Depending on the parametrization with σ -ρ or ω -ρ mixing terms, different values of the slope of the symmetry energy at saturation must be considered in order to obtain a neutron matter equation of state compatible with results from chiral effective field theory. This difference leads to different pressures at the crust-core transition density. A linear correlation between the transition density and the symmetry energy slope or the neutron skin thickness of the 208Pb nucleus is obtained, only when the ω meson is used to describe the density dependence of the symmetry energy. A comparison is made between the crust-core transition properties of neutron stars obtained by three different methods, the relativistic random phase approximation (RRPA), the Vlasov equation, and thermodynamical method. It is shown that the RRPA and the Vlasov methods predict similar transition densities for p n e β -equilibrium stellar matter.

  7. Boiling water neutronic reactor incorporating a process inherent safety design

    DOEpatents

    Forsberg, C.W.

    1985-02-19

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  8. Boiling water neutronic reactor incorporating a process inherent safety design

    DOEpatents

    Forsberg, Charles W.

    1987-01-01

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  9. 78 FR 32988 - Core Principles and Other Requirements for Designated Contract Markets; Correction

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-03

    ... and Other Requirements for Designated Contract Markets (77 FR 36612, June 19, 2012). The final rule... COMMISSION 17 CFR Part 38 RIN 3038-AD09 Core Principles and Other Requirements for Designated Contract...: This document corrects the Federal Register release of the final rule regarding Core Principles...

  10. Neutron angular distributions from the core break-up reactions of the {sup 11}Be and {sup 11}Li halo nuclei

    SciTech Connect

    Grevy, S.; Axelsson, L.; Jonson, B.; Nilsson, T.; Nyman, G.; Smedberg, M.; Angelique, J. C.; Orr, N.; Anne, R.; Lewitowicz, M.; Saint-Laurent, M. G.; Guillemaud-Mueller, D.; Mueller, A. C.; Pougheon, F.; Sorlin, O.; Hansen, P. G.; Hornshoj, P.; Riisager, K.

    1998-12-21

    The halo nuclei {sup 11}Be and {sup 11}Li have been studied through core break-up reactions, where the halo neutrons are detected in anti-coincidence with the core of the halo nucleus. In this particular channel, the halo neutrons are not expected to participate in the reaction and should therefore show the same properties as when situated inside the halo nucleus. The widths of the halo neutron momentum distributions have been extracted in coincidence with He fragments, {gamma}=32{+-}4 MeV/c, and Li fragments, {gamma}=42{+-}4 MeV/c for {sup 11}Be and with He fragments, {gamma}=42{+-}6 MeV/c for {sup 11}Li. An experimental value of the shadow effect for {sup 11}Be when breaking up to Li and He fragments has been obtained to be 0.63. A simple theoretical calculation to reproduce this value is given.

  11. Thermally activated post-glitch response of the neutron star inner crust and core. I. Theory

    SciTech Connect

    Link, Bennett

    2014-07-10

    Pinning of superfluid vortices is predicted to prevail throughout much of a neutron star. Based on the idea of Alpar et al., I develop a description of the coupling between the solid and liquid components of a neutron star through thermally activated vortex slippage, and calculate the response to a spin glitch. The treatment begins with a derivation of the vortex velocity from the vorticity equations of motion. The activation energy for vortex slippage is obtained from a detailed study of the mechanics and energetics of vortex motion. I show that the 'linear creep' regime introduced by Alpar et al. and invoked in fits to post-glitch response is not realized for physically reasonable parameters, a conclusion that strongly constrains the physics of a post-glitch response through thermal activation. Moreover, a regime of 'superweak pinning', crucial to the theory of Alpar et al. and its extensions, is probably precluded by thermal fluctuations. The theory given here has a robust conclusion that can be tested by observations: for a glitch in the spin rate of magnitude Δν, pinning introduces a delay in the post-glitch response time. The delay time is t{sub d} = 7(t{sub sd}/10{sup 4} yr)((Δν/ν)/10{sup –6}) d, where t{sub sd} is the spin-down age; t{sub d} is typically weeks for the Vela pulsar and months in older pulsars, and is independent of the details of vortex pinning. Post-glitch response through thermal activation cannot occur more quickly than this timescale. Quicker components of post-glitch response, as have been observed in some pulsars, notably, the Vela pulsar, cannot be due to thermally activated vortex motion but must represent a different process, such as drag on vortices in regions where there is no pinning. I also derive the mutual friction force for a pinned superfluid at finite temperature for use in other studies of neutron star hydrodynamics.

  12. Design and Simulation of a Rotating Aperture & Vacuum System for Neutron Imaging

    SciTech Connect

    Fitsos, P; Hall, J; Rusnak, B; Shen, S

    2006-02-27

    The development of a high-energy (10Mev) neutron imaging system at Lawrence Livermore National Laboratory (LLNL) depends on a precision engineered rotating aperture and vacuum system for generating neutrons that are used for imaging dense objects. This subsystem is part of a larger system which includes a linear accelerator that creates a deuteron beam, a scintillator detector, imaging optics and a high resolution CCD camera. The rotating aperture vacuum system has been successfully simulated and tested. Results show the feasibility of the design and point toward ways to improve the design by minimizing the rotating aperture gap.

  13. An accelerator-based epithermal neutron beam design for BNCT and dosimetric evaluation using a voxel head phantom.

    PubMed

    Lee, Deok-jae; Han, Chi Young; Park, Sung Ho; Kim, Jong Kyung

    2004-01-01

    The beam shaping assembly design has been investigated in order to improve the epithermal neutron beam for accelerator-based boron neutron capture therapy in intensity and quality, and dosimetric evaluation for the beams has been performed using both mathematical and voxel head phantoms with MCNP runs. The neutron source was assumed to be produced from a conventional 2.5 MeV proton accelerator with a thick (7)Li target. The results indicate that it is possible to enhance epithermal neutron flux remarkably as well as to embody a good spectrum shaping to epithermal neutrons only with the proper combination of moderator and reflector. It is also found that a larger number of thermal neutrons can reach deeply into the brain and, therefore, can reduce considerably the treatment time for brain tumours. Consequently, the epithermal neutron beams designed in this study can treat more effectively deep-seated brain tumours.

  14. Core Curriculum Analysis: A Tool for Educational Design

    ERIC Educational Resources Information Center

    Levander, Lena M.; Mikkola, Minna

    2009-01-01

    This paper examines the outcome of a dimensional core curriculum analysis. The analysis process was an integral part of an educational development project, which aimed to compact and clarify the curricula of the degree programmes. The task was also in line with the harmonising of the degree structures as part of the Bologna process within higher…

  15. Teaching to the Common Core by Design, Not Accident

    ERIC Educational Resources Information Center

    Phillips, Vicki; Wong, Carina

    2012-01-01

    The Bill & Melinda Gates Foundation has created tools and supports intended to help teachers adapt to the Common Core State Standards in English language arts and mathematics. The tools seek to find the right balance between encouraging teachers' creativity and giving them enough guidance to ensure quality. They are the product of two years of…

  16. Modified Anchor Shaped Post Core Design for Primary Anterior Teeth

    PubMed Central

    Rajesh, R.; Baroudi, Kusai; Reddy, K. Bala Kasi; Praveen, B. H.; Kumar, V. Sumanth; Amit, S.

    2014-01-01

    Restoring severely damaged primary anterior teeth is challenging to pedodontist. Many materials are tried as a post core but each one of them has its own drawbacks. This a case report describing a technique to restore severely damaged primary anterior teeth with a modified anchor shaped post. This technique is not only simple and inexpensive but also produces better retention. PMID:25379294

  17. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  18. Identifying and using 'core competencies' to help design and assess undergraduate neuroscience curricula.

    PubMed

    Kerchner, Michael; Hardwick, Jean C; Thornton, Janice E

    2012-01-01

    There has been a growing emphasis on the use of core competencies to design and inform curricula. Based on our Faculty for Undergraduate Neuroscience workshop at Pomona we developed a set of neuroscience core competencies. Following the workshop, faculty members were asked to complete an online survey to determine which core competencies are considered most essential and the results are presented. Backward Design principles are then described and we discuss how core competencies, through a backward design process, can be used to design and assess an undergraduate neuroscience curriculum. Oberlin College is used as a case study to describe the use of core competencies to help develop learning objectives, activities, and assessment measures for an undergraduate neuroscience major.

  19. Design and verification of the shielding around the new Neutron Standards Laboratory (LPN) at CIEMAT.

    PubMed

    Méndez-Villafañe, R; Guerrero, J E; Embid, M; Fernández, R; Grandio, R; Pérez-Cejuela, P; Márquez, J L; Alvarez, F; Ortego, P

    2014-10-01

    The construction of the new Neutron Standards Laboratory at CIEMAT (Laboratorio de Patrones Neutrónicos) has been finalised and is ready to provide service. The facility is an ∼8 m×8 m×8 m irradiation vault, following the International Organization for Standardization 8529 recommendations. It relies on several neutron sources: a 5-GBq (5.8× 10(8) s(-1)) (252)Cf source and two (241)Am-Be neutron sources (185 and 11.1 GBq). The irradiation point is located 4 m over the ground level and in the geometrical centre of the room. Each neutron source can be moved remotely from its storage position inside a water pool to the irradiation point. Prior to this, an important task to design the neutron shielding and to choose the most appropriate materials has been developed by the Radiological Security Unit and the Ionizing Radiations Metrology Laboratory. MCNPX was chosen to simulate the irradiation facility. With this information the walls were built with a thickness of 125 cm. Special attention was put on the weak points (main door, air conditioning system, etc.) so that the ambient dose outside the facility was below the regulatory limits. Finally, the Radiation Protection Unit carried out a set of measurements in specific points around the installation with an LB6411 neutron monitor and a Reuter-Stokes high-pressure ion chamber to verify experimentally the results of the simulation.

  20. Optimization of burnable poison design for Pu incineration in fully fertile free PWR core

    SciTech Connect

    Fridman, E.; Shwageraus, E.; Galperin, A.

    2006-07-01

    The design challenges of the fertile-free based fuel (FFF) can be addressed by careful and elaborate use of burnable poisons (BP). Practical fully FFF core design for PWR reactor has been reported in the past [1]. However, the burnable poison option used in the design resulted in significant end of cycle reactivity penalty due to incomplete BP depletion. Consequently, excessive Pu loading were required to maintain the target fuel cycle length, which in turn decreased the Pu burning efficiency. A systematic evaluation of commercially available BP materials in all configurations currently used in PWRs is the main objective of this work. The BP materials considered are Boron, Gd, Er, and Hf. The BP geometries were based on Wet Annular Burnable Absorber (WABA), Integral Fuel Burnable Absorber (IFBA), and Homogeneous poison/fuel mixtures. Several most promising combinations of BP designs were selected for the full core 3D simulation. All major core performance parameters for the analyzed cases are very close to those of a standard PWR with conventional UO{sub 2} fuel including possibility of reactivity control, power peaking factors, and cycle length. The MTC of all FFF cores was found at the full power conditions at all times and very close to that of the UO{sub 2} core. The Doppler coefficient of the FFF cores is also negative but somewhat lower in magnitude compared to UO{sub 2} core. The soluble boron worth of the FFF cores was calculated to be lower than that of the UO{sub 2} core by about a factor of two, which still allows the core reactivity control with acceptable soluble boron concentrations. The main conclusion of this work is that judicial application of burnable poisons for fertile free fuel has a potential to produce a core design with performance characteristics close to those of the reference PWR core with conventional UO{sub 2} fuel. (authors)

  1. Neutronics and thermal design analyses of US solid breeder blanket for ITER

    SciTech Connect

    Gohar, Y.; Billone, M.; Attaya, H. ); Sawan, M. )

    1990-09-01

    The US Solid Breeder Blanket is designed to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Safety, low tritium inventory, reliability, flexibility cost, and minimum R D requirements are the other design criteria. To satisfy these criteria, the produced tritium is recovered continuously during operation and the blanket coolant operates at low pressure. Beryllium multiplier material is used to control the solid-breeder temperature. Neutronics and thermal design analyses were performed in an integrated manner to define the blanket configuration. The reference parameters of ITER including the operating scenarios, the neutron wall loading distribution and the copper stabilizer are included in the design analyses. Several analyses were performed to study the impact of the reactor parameters, blanket dimensions, material characteristics, and heat transfer coefficient at the material interfaces on the blanket performance. The design analyses and the results from the different studies are summarized. 6 refs., 3 figs., 3 tabs.

  2. A neutron pinhole camera for PF-24 source: Conceptual design and optimization

    NASA Astrophysics Data System (ADS)

    Bielecki, J.; Wójcik-Gargula, A.; Wiacek, U.; Scholz, M.; Igielski, A.; Drozdowicz, K.; Woźnicka, U.

    2015-07-01

    A fast-neutron pinhole camera based on small-area (5mm × 5 mm) BCF-12 scintillation detectors with nanosecond time resolution has been designed. The pinhole camera is dedicated to the investigation of the spatial and temporal distributions of DD neutrons from the Plasma Focus (PF-24) source. The geometrical parameters of the camera have been optimized in terms of maximum neutron flux at the imaging plane by means of MCNP calculations. The detection system consists of four closely packed scintillation detectors coupled via long optical fibres to Hamamatsu H3164-10 photomultiplier tubes. The pinhole consists of specially designed 420 mm long copper collimator with an effective aperture of 1.7 mm mounted inside a cylindrical polyethylene tube. The performance of the presented detection system in the mixed (hard X-ray and neutron) radiation field of the PF-24 plasma focus device has been tested. The results of the tests showed that the small-area BCF-12 scintillation detectors can be successfully applied as the detection system of the neutron pinhole camera for the PF-24 device.

  3. Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor

    NASA Astrophysics Data System (ADS)

    Abdullah, Y.; Hamid, N. A.; Mansor, M. A.; Ahmad, M. H. A. R. M.; Yusof, M. R.; Yazid, H.; Mohamed, A. A.

    2013-06-01

    The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

  4. Design and use of a 6 meter neutron small-angle scattering spectrometer at KUR

    NASA Astrophysics Data System (ADS)

    Komura, S.; Takeda, T.; Fujii, H.; Osamura, K.; Mochiki, K.; Hasegawa, K.

    1983-05-01

    A 6 meter neutron small-angle scattering spectrometer has been constructed at the Kyoto University Reactor (KUR) and has been used successfully in various fields of application. The design principles and the characteristics of the spectrometer are described briefly. Some examples of the scattering measurements are presented.

  5. Challenges in the development of high-fidelity LWR core neutronics tools

    SciTech Connect

    Smith, K.; Forget, B.

    2013-07-01

    Modern computing has made possible the solution of extremely large-scale reactor simulations, and the literature has numerous examples of high-resolution methods (often Monte Carlo) applied to full-core reactor problems. However, there are currently no examples in the literature of application of such 'High-Fidelity' or 'First Principles' methods to operating Light Water Reactor (LWR) analysis. This paper seeks to remind code developers, project managers, and analysts of the many important aspects of LWR simulation that must be incorporated to produce truly high-fidelity analysis tools. The authors offer a monetary prize to the first person (or group) that successfully solves a new two-cycle operational PWR depletion benchmark problem using high-fidelity tools and demonstrates acceptable accuracy by comparison with measured operational plant data (open source) provided to the reactor analysis community. (authors)

  6. Improvement of transformer core magnetic properties using the step-lap design

    NASA Astrophysics Data System (ADS)

    Valkovic, Z.; Rezic, A.

    1992-07-01

    Magnetic properties of the step-lap joints have been investigated experimentally on two three-phase three-leg transformer cores. Using the step-lap joint design, a reduction of the total core loss of 2 to 4.4% and of the exciting power of 31 to 37% has been obtained.

  7. Design of the radiation shielding for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

    SciTech Connect

    Du, T. F.; Chen, Z. J.; Peng, X. Y.; Yuan, X.; Zhang, X.; Hu, Z. M.; Cui, Z. Q.; Xie, X. F.; Ge, L. J.; Li, X. Q.; Zhang, G. H.; Chen, J. X.; Fan, T. S.; Gorini, G.; Nocente, M.; Tardocchi, M.; Hu, L. Q.; Zhong, G. Q.; Lin, S. Y.; Wan, B. N.

    2014-11-15

    A radiation shielding has been designed to reduce scattered neutrons and background gamma-rays for the new double-ring Time Of Flight Enhanced Diagnostics (TOFED). The shielding was designed based on simulation with the Monte Carlo code MCNP5. Dedicated model of the EAST tokamak has been developed together with the emission neutron source profile and spectrum; the latter were simulated with the Nubeam and GENESIS codes. Significant reduction of background radiation at the detector can be achieved and this satisfies the requirement of TOFED. The intensities of the scattered and direct neutrons in the line of sight of the TOFED neutron spectrometer at EAST are studied for future data interpretation.

  8. Design of the radiation shielding for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak.

    PubMed

    Du, T F; Chen, Z J; Peng, X Y; Yuan, X; Zhang, X; Gorini, G; Nocente, M; Tardocchi, M; Hu, Z M; Cui, Z Q; Xie, X F; Ge, L J; Hu, L Q; Zhong, G Q; Lin, S Y; Wan, B N; Li, X Q; Zhang, G H; Chen, J X; Fan, T S

    2014-11-01

    A radiation shielding has been designed to reduce scattered neutrons and background gamma-rays for the new double-ring Time Of Flight Enhanced Diagnostics (TOFED). The shielding was designed based on simulation with the Monte Carlo code MCNP5. Dedicated model of the EAST tokamak has been developed together with the emission neutron source profile and spectrum; the latter were simulated with the Nubeam and GENESIS codes. Significant reduction of background radiation at the detector can be achieved and this satisfies the requirement of TOFED. The intensities of the scattered and direct neutrons in the line of sight of the TOFED neutron spectrometer at EAST are studied for future data interpretation.

  9. Design and performance of a large area neutron sensitive anger camera

    NASA Astrophysics Data System (ADS)

    Riedel, R. A.; Donahue, C.; Visscher, T.; Montcalm, C.

    2015-09-01

    We describe the design and performance of a 157 mm×157 mm two dimensional neutron detector. The detector uses the Anger principle to determine the position of neutrons. We have verified FWHM resolution of <1.2 mm with distortion <0.5 mm on over 50 installed Anger Cameras. The performance of the detector is limited by the light yield of the scintillator, and it is estimated that the resolution of the current detector could be doubled with a brighter scintillator. Data collected from small (<1 mm3) single crystal reference samples at the single crystal instrument TOPAZ provide results with low values of the refinement parameter Rw(F).

  10. The Reversed-Field-Pinch (RFP) fusion neutron source: A conceptual design

    NASA Astrophysics Data System (ADS)

    Bathke, C. G.; Krakowski, R. A.; Miller, R. L.; Werley, K. A.

    The conceptual design of an ohmically heated, reversed-field pinch (RFP) operating at approximately 5-MW/m(2) steady-state DT fusion neutron wall loading and approximately 124-MW total fusion power is presented. These results are useful in projecting the development of a cost effective, low input power, approximately 206 MW, source of DT neutrons for large-volume approximately 10 m(3), high-fluence, 3.4 MW yr/m(2), fusion nuclear materials and technology testing.

  11. Design and performance of a large area neutron sensitive anger camera

    DOE PAGES

    Visscher, Theodore; Montcalm, Christopher A.; Donahue, Jr., Cornelius; Riedel, Richard A.

    2015-05-21

    We describe the design and performance of a 157mm x 157mm two dimensional neutron detector. The detector uses the Anger principle to determine the position of neutrons. We have verified FWHM resolution of < 1.2mm with distortion < 0.5mm on over 50 installed Anger Cameras. The performance of the detector is limited by the light yield of the scintillator, and it is estimated that the resolution of the current detector could be doubled with a brighter scintillator. Data collected from small (<1mm3) single crystal reference samples at the single crystal instrument TOPAZ provide results with low Rw(F) values

  12. Design and burn-up analyses of new type holder for silicon neutron transmutation doping.

    PubMed

    Komeda, Masao; Arai, Masaji; Tamai, Kazuo; Kawasaki, Kozo

    2016-07-01

    We have developed a new silicon irradiation holder with a neutron filter to increase the irradiation efficiency. The neutron filter is made of an alloy of aluminum and B4C particles. We fabricated a new holder based on the results of design analyses. This filter has limited use in applications requiring prolonged use due to a decrease in the amount of (10)B in B4C particles. We investigated the influence of (10)B reduction on doping distribution in a silicon ingot by using the Monte Carlo Code MVP. PMID:27131643

  13. Fast Neutron Spectrum Potassium Worth for Space Power Reactor Design Validation

    SciTech Connect

    Bess, John D.; Marshall, Margaret A.; Briggs, J. Blair; Tsiboulia, Anatoli; Rozhikhin, Yevgeniy; Mihalczo, John T.

    2015-03-01

    A variety of critical experiments were constructed of enriched uranium metal (oralloy ) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, one was performed that consisted of uranium metal annuli surrounding a potassium-filled, stainless steel can. The outer diameter of the annuli was approximately 13 inches (33.02 cm) with an inner diameter of 7 inches (17.78 cm). The diameter of the stainless steel can was 7 inches (17.78 cm). The critical height of the configurations was approximately 5.6 inches (14.224 cm). The uranium annulus consisted of multiple stacked rings, each with radial thicknesses of 1 inch (2.54 cm) and varying heights. A companion measurement was performed using empty stainless steel cans; the primary purpose of these experiments was to test the fast neutron cross sections of potassium as it was a candidate for coolant in some early space power reactor designs.The experimental measurements were performed on July 11, 1963, by J. T. Mihalczo and M. S. Wyatt (Ref. 1) with additional information in its corresponding logbook. Unreflected and unmoderated experiments with the same set of highly enriched uranium metal parts were performed at the Oak Ridge Critical Experiments Facility in the 1960s and are evaluated in the International Handbook for Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) with the identifier HEU MET FAST 051. Thin

  14. Design of a north pole Neutron Time-of-Flight (NTOF) system at NIF

    NASA Astrophysics Data System (ADS)

    Caggiano, J. A.; Barbosa, F.; Clancy, T. J.; Eckart, M. J.; Grim, G.; Hartouni, E. P.; Hatarik, R.; Khater, H.; Lee, A.; Sampson, M.; Sayre, D. B.; Yeamans, C.; Yeoman, M.

    2016-05-01

    A north pole NTOF system for neutron spectroscopy is being implemented at the NIF. The design is centered around a fast scintillator with low mass housing fielded 21.6m from target chamber center at θ=18°,ϕ=304°. The line-of-sight (LOS) features a primary port collimator, two secondary collimators in the intervening concrete floors, and a beam dump with a backscatter shield. Because the detector is being fielded on the roof of the NIF building, diagnostic options such as optical and electrical attenuation are remotely controlled, saving setup time and increasing shot rate. The expected performance of the diagnostic is excellent with high sensitivity to both high-energy reaction-in-flight neutrons as well as lower energy down-scattered neutrons.

  15. Design of the neutron imaging pinhole for use at the national ignition facility

    SciTech Connect

    Fatherley, Valerie E; Day, Robert D; Garcia, Felix P; Grim, Gary P; Oertel, John A; Wilde, Carl H; Wilke, Mark D

    2010-01-01

    The Neutron Imaging (NI) diagnostic is designed to be used at the National Ignition Facility (NIF). This instrument will be used to image both primary (14MeV neutrons) and down scattered (6-8MeV neutrons). The pinhole body sits 225mm from the target, while the scintillator and recording systems are located 28m from the target. The diagnostic uses port 90, 315 and the recording system is located in a specifically built room located outside of switchyard I. The location of the pinhole and the recording system combine to give a magnification of 104. The recording of both the primary and downscattered image is done by recording the image from both the front and back side of the scintillator.

  16. A new design of fission detector for prompt fission neutron investigation

    NASA Astrophysics Data System (ADS)

    Zeynalov, Sh.; Zeynalova, O.; Nazarenko, M. A.; Hambsch, F.-J.; Oberstedt, S.

    2012-10-01

    In this work we report recent achievements in design of twin back-to-back ionization chamber (TIC) for fission fragment (FF) mass and kinetic energy spectroscopy. Correlated FF kinetic energies, their masses and the angle of the fission axes in 3D Cartesian coordinates can be determined from analysis of the heights and shapes of the pulses induced by the fission fragments on the anodes of TIC. Anodes of TIC were designed as consisting of isolated strips each having independent electronic circuitry and special multi-channel pulse processing apparatus. Mathematical algorithms were provided along with formulae derived for fission axis angles determination. It was shown how the point of fission fragments origin on the target plane may be determined using the same measured data. The last feature made the TIC a rather powerful tool for prompt fission neutron (PFN) emission investigation in event by event analysis of individual fission reactions from non point fissile source. Position sensitive neutron induced fission detector for neutron imaging applications with both thermal and low energy neutrons was found as another possible implementation of the designed TIC.

  17. High accuracy modeling for advanced nuclear reactor core designs using Monte Carlo based coupled calculations

    NASA Astrophysics Data System (ADS)

    Espel, Federico Puente

    The main objective of this PhD research is to develop a high accuracy modeling tool using a Monte Carlo based coupled system. The presented research comprises the development of models to include the thermal-hydraulic feedback to the Monte Carlo method and speed-up mechanisms to accelerate the Monte Carlo criticality calculation. Presently, deterministic codes based on the diffusion approximation of the Boltzmann transport equation, coupled with channel-based (or sub-channel based) thermal-hydraulic codes, carry out the three-dimensional (3-D) reactor core calculations of the Light Water Reactors (LWRs). These deterministic codes utilize nuclear homogenized data (normally over large spatial zones, consisting of fuel assembly or parts of fuel assembly, and in the best case, over small spatial zones, consisting of pin cell), which is functionalized in terms of thermal-hydraulic feedback parameters (in the form of off-line pre-generated cross-section libraries). High accuracy modeling is required for advanced nuclear reactor core designs that present increased geometry complexity and material heterogeneity. Such high-fidelity methods take advantage of the recent progress in computation technology and coupled neutron transport solutions with thermal-hydraulic feedback models on pin or even on sub-pin level (in terms of spatial scale). The continuous energy Monte Carlo method is well suited for solving such core environments with the detailed representation of the complicated 3-D problem. The major advantages of the Monte Carlo method over the deterministic methods are the continuous energy treatment and the exact 3-D geometry modeling. However, the Monte Carlo method involves vast computational time. The interest in Monte Carlo methods has increased thanks to the improvements of the capabilities of high performance computers. Coupled Monte-Carlo calculations can serve as reference solutions for verifying high-fidelity coupled deterministic neutron transport methods

  18. Large Core Code Evaluation Working Group Benchmark Problem Four: neutronics and burnup analysis of a large heterogeneous fast reactor. Part 1. Analysis of benchmark results. [LMFBR

    SciTech Connect

    Cowan, C.L.; Protsik, R.; Lewellen, J.W.

    1984-01-01

    The Large Core Code Evaluation Working Group Benchmark Problem Four was specified to provide a stringent test of the current methods which are used in the nuclear design and analyses process. The benchmark specifications provided a base for performing detailed burnup calculations over the first two irradiation cycles for a large heterogeneous fast reactor. Particular emphasis was placed on the techniques for modeling the three-dimensional benchmark geometry, and sensitivity studies were carried out to determine the performance parameter sensitivities to changes in the neutronics and burnup specifications. The results of the Benchmark Four calculations indicated that a linked RZ-XY (Hex) two-dimensional representation of the benchmark model geometry can be used to predict mass balance data, power distributions, regionwise fuel exposure data and burnup reactivities with good accuracy when compared with the results of direct three-dimensional computations. Most of the small differences in the results of the benchmark analyses by the different participants were attributed to ambiguities in carrying out the regionwise flux renormalization calculations throughout the burnup step.

  19. PGNAA system preliminary design and measurement of In-Hospital Neutron Irradiator for boron concentration measurement.

    PubMed

    Zhang, Zizhu; Chong, Yizheng; Chen, Xinru; Jin, Congjun; Yang, Lijun; Liu, Tong

    2015-12-01

    A prompt gamma neutron activation analysis (PGNAA) system has been recently developed at the 30-kW research reactor In-Hospital Neutron Irradiator (IHNI) in Beijing. Neutrons from the specially designed thermal neutron beam were used. The thermal flux of this beam is 3.08×10(6) cm(-2) s(-1) at a full reactor power of 30 kW. The PGNAA system consists of an n-type high-purity germanium (HPGe) detector of 40% efficiency, a digital spectrometer, and a shielding part. For both the detector shielding part and the neutron beam shielding part, the inner layer is composed of (6)Li2CO3 powder and the outer layer lead. The boron-10 sensitivity of the PGNAA system is approximately 2.5 cps/ppm. Two calibration curves were produced for the 1-10 ppm and 10-50 ppm samples. The measurement results of the control samples were in accordance with the inductively coupled plasma atomic emission spectroscopy (ICP-AES) results.

  20. Population-based metaheuristic optimization in neutron optics and shielding design

    NASA Astrophysics Data System (ADS)

    DiJulio, D. D.; Björgvinsdóttir, H.; Zendler, C.; Bentley, P. M.

    2016-11-01

    Population-based metaheuristic algorithms are powerful tools in the design of neutron scattering instruments and the use of these types of algorithms for this purpose is becoming more and more commonplace. Today there exists a wide range of algorithms to choose from when designing an instrument and it is not always initially clear which may provide the best performance. Furthermore, due to the nature of these types of algorithms, the final solution found for a specific design scenario cannot always be guaranteed to be the global optimum. Therefore, to explore the potential benefits and differences between the varieties of these algorithms available, when applied to such design scenarios, we have carried out a detailed study of some commonly used algorithms. For this purpose, we have developed a new general optimization software package which combines a number of common metaheuristic algorithms within a single user interface and is designed specifically with neutronic calculations in mind. The algorithms included in the software are implementations of Particle-Swarm Optimization (PSO), Differential Evolution (DE), Artificial Bee Colony (ABC), and a Genetic Algorithm (GA). The software has been used to optimize the design of several problems in neutron optics and shielding, coupled with Monte-Carlo simulations, in order to evaluate the performance of the various algorithms. Generally, the performance of the algorithms depended on the specific scenarios, however it was found that DE provided the best average solutions in all scenarios investigated in this work.

  1. Designing, Leading and Managing the Transition to the Common Core: A Strategy Guidebook for Leaders

    ERIC Educational Resources Information Center

    Brown, Brentt; Vargo, Merrill

    2014-01-01

    The Common Core provides districts an opportunity to renew their focus on teaching and learning. But it also poses a number of design and implementation challenges for school districts. The "Leadership and Design Cycles" described in this guidebook offers an evidenced-based and structured process for leaders to design and implement…

  2. Design and Evaluation of an Enhanced In-Vessel Core Catcher

    SciTech Connect

    Joy L. Rempe

    2004-06-01

    An enhanced in-vessel core catcher is being designed and evaluated as part of a joint United States (U.S.) - Korean International Nuclear Engineering Research Initiative (INERI) investigating methods to insure In-Vessel Retention (IVR) of core materials that may relocate under severe accident conditions in advanced reactors. To reduce cost and simplify manufacture and installation, this new core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. If needed, the core catcher can be manufactured with holes to accommodate lower head penetrations. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary): a base material, which has the capability to support and contain the mass of core materials that may relocate during a severe accident; an oxide coating material on top of the base material, which resists interactions with high-temperature core materials; and an optional coating on the bottom side of the base material to prevent any potential oxidation of the base material during the lifetime of the reactor. This paper summarizes the status of core catcher design and evaluation efforts, including analyses, materials interaction tests, and prototypic testing efforts.

  3. Conceptual Reactor Design Study of Very High Temperature Reactor (VHTR) with Prismatic-Type Core

    NASA Astrophysics Data System (ADS)

    Nakano, Masaaki; Tsuji, Nobumasa; Tazawa, Yujiro

    The preliminary conceptual design study of prismatic-type Very High Temperature Reactor (VHTR) has been performed with 950°C outlet coolant temperature for higher efficient hydrogen and electricity production. First, the core internals that enable higher outlet temperature are considered in the viewpoint of reduction of core bypass flow. Three-dimensional thermal and hydraulic analyses are carried out and show that the 950°C outlet temperature requires approximately 90% fuel flow fraction and it can be achieved with the installation of the seals in bottom blocks, the coolant tubes in the permanent side reflectors and the core restraint devices. Next, the core and fission product (FP) release analyses are performed. The analysis methods that have been developed for the pin-in-block fuel, one type of prismatic VHTR cores, can be applied to multi-hole fuel, another type of the cores, with some adjustments of the analytical models.

  4. Identification of 4-aminoquinoline core for the design of new cholinesterase inhibitors

    PubMed Central

    Chen, Yao; Bian, Yaoyao; Sun, Yuan; Kang, Chen; Yu, Sheng; Fu, Tingming; Li, Wei

    2016-01-01

    Inhibition of acetylcholinesterase (AChE) using small molecules is still one of the most successful therapeutic strategies in the treatment of Alzheimer’s disease (AD). Previously we reported compound T5369186 with a core of quinolone as a new cholinesterase inhibitor. In the present study, in order to identify new cores for the designing of AChE inhibitors, we screened different derivatives of this core with the aim to identify the best core as the starting point for further optimization. Based on the results, we confirmed that only 4-aminoquinoline (compound 04 and 07) had cholinesterase inhibitory effects. Considering the simple structure and high inhibitory potency against AChE, 4-aminoquinoline provides a good starting core for further designing novel multifunctional AChEIs. PMID:27441112

  5. Validation of updated neutronic calculation models proposed for Atucha-II PHWR. Part II: Benchmark comparisons of PUMA core parameters with MCNP5 and improvements due to a simple cell heterogeneity correction

    SciTech Connect

    Grant, C.; Mollerach, R.; Leszczynski, F.; Serra, O.; Marconi, J.; Fink, J.

    2006-07-01

    In 2005 the Argentine Government took the decision to complete the construction of the Atucha-II nuclear power plant, which has been progressing slowly during the last ten years. Atucha-II is a 745 MWe nuclear station moderated and cooled with heavy water, of German (Siemens) design located in Argentina. It has a pressure vessel design with 451 vertical coolant channels and the fuel assemblies (FA) are clusters of 37 natural UO{sub 2} rods with an active length of 530 cm. For the reactor physics area, a revision and update of reactor physics calculation methods and models was recently carried out covering cell, supercell (control rod) and core calculations. This paper presents benchmark comparisons of core parameters of a slightly idealized model of the Atucha-I core obtained with the PUMA reactor code with MCNP5. The Atucha-I core was selected because it is smaller, similar from a neutronic point of view, more symmetric than Atucha-II, and has some experimental data available. To validate the new models benchmark comparisons of k-effective, channel power and axial power distributions obtained with PUMA and MCNP5 have been performed. In addition, a simple cell heterogeneity correction recently introduced in PUMA is presented, which improves significantly the agreement of calculated channel powers with MCNP5. To complete the validation, the calculation of some of the critical configurations of the Atucha-I reactor measured during the experiments performed at first criticality is also presented. (authors)

  6. A new MCNPX PTRAC coincidence capture file capability: a tool for neutron detector design

    SciTech Connect

    Evans, Louise G; Schear, Melissa A; Hendricks, John S; Swinhoe, Martyn T; Tobin, Stephen J; Croft, Stephen

    2011-02-16

    The existing Monte Carlo N-Particle (MCNPX) particle tracking (PTRAC) coincidence capture file allows a full list of neutron capture events to be recorded in any simulated detection medium. The originating event history number (e.g. spontaneous fission events), capture time, location and source particle number are tracked and output to file for post-processing. We have developed a new MCNPX PTRAC coincidence capture file capability to aid detector design studies. New features include the ability to track the nuclides that emitted the detected neutrons as well as induced fission chains in mixed samples before detection (both generation number and nuclide that underwent induced fission). Here, the power of this tool is demonstrated using a detector design developed for the non-destructive assay (NDA) of spent nuclear fuel. Individual capture time distributions have been generated for neutrons originating from Curium-244 source spontaneous fission events and induced fission events in fissile nuclides of interest: namely Plutonium-239, Plutonium-241, and Uranium-235. Through this capability, a full picture for the attribution of neutron capture events in the detector can be simulated.

  7. Design and operation of a passive neutron monitor for assaying the TRU content of solid wastes

    SciTech Connect

    Brodzinski, R.L.; Brown, D.P.; Rieck, H.G. Jr.; Rogers, L.A.

    1984-02-01

    A passive neutron monitor has been designed and built for determining the residual transuranic (TRU) and plutonium content of chopped leached fuel hulls and other solid wastes from spent Fast Flux Test Facility (FFTF) fuel. The system was designed to measure as little as 8 g of plutonium or 88 mg of TRU in a waste package as large as a 208-l drum which could be emitting up to 220,000 R/hr of gamma radiation. For practical purposes, maximum assay times were chosen to be 10,000 sec. The monitor consists of 96 /sup 10/BF/sub 3/ neutron sensitive proportional counting tubes each 5.08 cm in diameter and 183 cm in active length. Tables of neutron emission rates from both spontaneous fission and (..cap alpha..,n) reactions on oxygen are given for all contributing isotopes expected to be present in spent FFTF fuel. Tables of neutron yeilds from isotopic compositions predicted for various exposures and cooling times are also given. Methods of data reduction and sources, magnitude, and control of errors are discussed. Backgrounds and efficiencies have been measured and are reported. A section describing step-by-step operational procedures is included. Guidelines and procedures for quality control and troubleshooting are also given. 13 references, 15 figures, 4 tables.

  8. A new MCNPX PTRAC coincidence capture file capability: a tool for neutron detector design

    SciTech Connect

    Evans, Louise G; Schear, Melissa A; Hendricks, John S; Swinhoe, Martyn T; Tobin, Stephen J; Croft, Stephen

    2010-12-14

    The existing MCNPX{trademark} PTRAC coincidence capture file allows a full list of neutron capture events to be recorded in any simulated detection medium. The originating event history number (e.g. spontaneous fission events), capture time, location and source particle number are tracked and output to file for post-processing. We have developed a new MCNPX PTRAC coincidence capture file capability to aid detector design studies. New features include the ability to track the isotopes that emitted the detected neutrons as well as induced fission chains in mixed samples before detection (both generation number and isotope). Here, the power of this tool is demonstrated using a detector design that has been developed for the non-destructive assay (NDA) of spent nuclear fuel. Individual capture time distributions have been generated for neutrons originating from Curium-244 source spontaneous fission events and induced fission events in fissile isotopes of interest: namely Plutonium-239, Plutonium-241, and Uranium-235. Through this capability, a full picture for the attribution of neutron capture events in the detector can be simulated.

  9. A new NCNPX PTRAC coincidence capture file capability: a tool for neutron detector design

    SciTech Connect

    Evans, Louise G; Schear, Melissa A; Hendricks, John S; Swinhoe, Martyn T; Tobin, Stephen J; Croft, Stephen

    2011-01-13

    The existing Monte Carlo N-Particle (MCNPX) particle tracking (PTRAC) coincidence capture file allows a full list of neutron capture events to be recorded in any simulated detection medium. The originating event history number (e.g. spontaneous fission events), capture time, location and source particle number are tracked and output to file for post-processing. We have developed a new MCNPX PTRAC coincidence capture file capability to aid detector design studies. New features include the ability to track the nuclides that emitted the detected neutrons as well as induced fission chains in mixed samples before detection (both generation number and nuclide that underwent induced fission). Here, the power of this tool is demonstrated using a detector design developed for the non-destructive assay (NDA) of spent nuclear fuel. Individual capture time distributions have been generated for neutrons originating from Curium-244 source spontaneous fission events and induced fission events in fissile nuclides of interest: namely Plutonium-239, Plutonium-241, and Uranium-235. Through this capability, a full picture for the attribution of neutron capture events in the detector can be simulated.

  10. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    SciTech Connect

    Slater, C.O.

    1990-07-01

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

  11. Modified Y-TZP core design improves all-ceramic crown reliability.

    PubMed

    Silva, N R F A; Bonfante, E A; Rafferty, B T; Zavanelli, R A; Rekow, E D; Thompson, V P; Coelho, P G

    2011-01-01

    This study tested the hypothesis that all-ceramic core-veneer system crown reliability is improved by modification of the core design. We modeled a tooth preparation by reducing the height of proximal walls by 1.5 mm and the occlusal surface by 2.0 mm. The CAD-based tooth preparation was replicated and positioned in a dental articulator for core and veneer fabrication. Standard (0.5 mm uniform thickness) and modified (2.5 mm height lingual and proximal cervical areas) core designs were produced, followed by the application of veneer porcelain for a total thickness of 1.5 mm. The crowns were cemented to 30-day-aged composite dies and were either single-load-to-failure or step-stress-accelerated fatigue-tested. Use of level probability plots showed significantly higher reliability for the modified core design group. The fatigue fracture modes were veneer chipping not exposing the core for the standard group, and exposing the veneer core interface for the modified group.

  12. Design progress of cryogenic hydrogen system for China Spallation Neutron Source

    NASA Astrophysics Data System (ADS)

    Wang, G. P.; Zhang, Y.; Xiao, J.; He, C. C.; Ding, M. Y.; Wang, Y. Q.; Li, N.; He, K.

    2014-01-01

    China Spallation Neutron Source (CSNS) is a large proton accelerator research facility with 100 kW beam power. Construction started in October 2011 and is expected to last 6.5 years. The cryogenic hydrogen circulation is cooled by a helium refrigerator with cooling capacity of 2200 W at 20 K and provides supercritical hydrogen to neutron moderating system. Important progresses of CSNS cryogenic system were concluded as follows. Firstly, process design of cryogenic system has been completed including helium refrigerator, hydrogen loop, gas distribution, and safety interlock. Secondly, an accumulator prototype was designed to mitigate pressure fluctuation caused by dynamic heat load from neutron moderation. Performance test of the accumulator has been carried out at room and liquid nitrogen temperature. Results show the accumulator with welding bellows regulates hydrogen pressure well. Parameters of key equipment have been identified. The contract for the helium refrigerator has been signed. Mechanical design of the hydrogen cold box has been completed, and the hydrogen pump, ortho-para hydrogen convertor, helium-hydrogen heat exchanger, hydrogen heater, and cryogenic valves are in procurement. Finally, Hydrogen safety interlock has been finished as well, including the logic of gas distribution, vacuum, hydrogen leakage and ventilation. Generally, design and construction of CSNS cryogenic system is conducted as expected.

  13. Design progress of cryogenic hydrogen system for China Spallation Neutron Source

    SciTech Connect

    Wang, G. P.; Zhang, Y.; Xiao, J.; He, C. C.; Ding, M. Y.; Wang, Y. Q.; Li, N.; He, K.

    2014-01-29

    China Spallation Neutron Source (CSNS) is a large proton accelerator research facility with 100 kW beam power. Construction started in October 2011 and is expected to last 6.5 years. The cryogenic hydrogen circulation is cooled by a helium refrigerator with cooling capacity of 2200 W at 20 K and provides supercritical hydrogen to neutron moderating system. Important progresses of CSNS cryogenic system were concluded as follows. Firstly, process design of cryogenic system has been completed including helium refrigerator, hydrogen loop, gas distribution, and safety interlock. Secondly, an accumulator prototype was designed to mitigate pressure fluctuation caused by dynamic heat load from neutron moderation. Performance test of the accumulator has been carried out at room and liquid nitrogen temperature. Results show the accumulator with welding bellows regulates hydrogen pressure well. Parameters of key equipment have been identified. The contract for the helium refrigerator has been signed. Mechanical design of the hydrogen cold box has been completed, and the hydrogen pump, ortho-para hydrogen convertor, helium-hydrogen heat exchanger, hydrogen heater, and cryogenic valves are in procurement. Finally, Hydrogen safety interlock has been finished as well, including the logic of gas distribution, vacuum, hydrogen leakage and ventilation. Generally, design and construction of CSNS cryogenic system is conducted as expected.

  14. Improvement of advanced nodal method used in 3D core design system

    SciTech Connect

    Rauck, S.; Dall'Osso, A.

    2006-07-01

    This paper deals with AREVA NP progress in the modelling of neutronic phenomena, evaluated through 3D determinist core codes and using 2-group diffusion theory. Our report highlights the advantages of taking into account the assembly environment in the process used for the building of the 2-group collapsed neutronic parameters, such as cross sections or discontinuity factors. The interest of the present method, developed in order to account for the impact of the environment on the above mentioned parameters, resides (i) in the very definition of a global correlation between collapsed neutronic data calculated in an infinite medium and those calculated in a 3D-geometry, and (ii) in the use of a re-homogenization method. Using this approach, computations match better with actual measurements on control rod worth. They also present smaller differences on pin by pin power values compared to the ones computed with another code considered as a reference since it relies on multigroup transport theory. (authors)

  15. Investigation on the reflector/moderator geometry and its effect on the neutron beam design in BNCT.

    PubMed

    Kasesaz, Y; Rahmani, F; Khalafi, H

    2015-12-01

    In order to provide an appropriate neutron beam for Boron Neutron Capture Therapy (BNCT), a special Beam Shaping Assembly (BSA) must be designed based on the neutron source specifications. A typical BSA includes moderator, reflector, collimator, thermal neutron filter, and gamma filter. In common BSA, the reflector is considered as a layer which covers the sides of the moderator materials. In this paper, new reflector/moderator geometries including multi-layer and hexagonal lattice have been suggested and the effect of them has been investigated by MCNP4C Monte Carlo code. It was found that the proposed configurations have a significant effect to improve the thermal to epithermal neutron flux ratio which is an important neutron beam parameter.

  16. Design of a backscatter 14-MeV neutron time-of-flight spectrometer for experiments at ITER

    SciTech Connect

    Dzysiuk, N.; Hellesen, C.; Conroy, S.; Ericsson, G.; Hjalmarsson, A.; Skiba, M.

    2014-08-21

    Neutron energy spectrometry diagnostics play an important role in present-day experiments related to fusion energy research. Measurements and thorough analysis of the neutron emission from the fusion plasma give information on a number of basic fusion performance quantities, on the condition of the neutron source and plasma behavior. Here we discuss the backscatter Time-of-Flight (bTOF) spectrometer concept as a possible instrument for performing high resolution measurements of 14 MeV neutrons. The instrument is based on two sets of scintillators, a first scatterer exposed to a collimated neutron beam and a second detector set placed in the backward direction. The scintillators of the first set are enriched in deuterium to achieve neutron backscattering. The energy resolution and efficiency of a bTOF instrument have been determined for various geometrical configurations. A preliminary design of optimal geometry for the two scintillator sets has been obtained by Monte Carlo simulations based on the MCNPX code.

  17. Design of a high-flux epithermal neutron beam using 235U fission plates at the Brookhaven Medical Research Reactor.

    PubMed

    Liu, H B; Brugger, R M; Rorer, D C; Tichler, P R; Hu, J P

    1994-10-01

    Beams of epithermal neutrons are being used in the development of boron neutron capture therapy for cancer. This report describes a design study in which 235U fission plates and moderators are used to produce an epithermal neutron beam with higher intensity and better quality than the beam currently in use at the Brookhaven Medical Research Reactor (BMRR). Monte Carlo calculations are used to predict the neutron and gamma fluxes and absorbed doses produced by the proposed design. Neutron flux measurements at the present epithermal treatment facility (ETF) were made to verify and compare with the computed results where feasible. The calculations indicate that an epithermal neutron beam produced by a fission-plate converter could have an epithermal neutron intensity of 1.2 x 10(10) n/cm2.s and a fast neutron dose per epithermal neutron of 2.8 x 10(-11) cGy.cm2/nepi plus being forward directed. This beam would be built into the beam shutter of the ETF at the BMRR. The feasibility of remodeling the facility is discussed.

  18. Design, Assembly, and Testing of the Neutron Imaging Lens for the National Ignition Facility

    SciTech Connect

    Malone, Robert M; Fatherley, Valerie E; Frogget, Brent C; Grim, Gary P; Kaufman, Morris I; McGillivray, Kevin D; Oertel, John A; Palagi, Martin J; Skarda, William K; Tibbitts, Aric; Wilde, Carl H; Wilke, Mark D

    2010-09-01

    The National Ignition Facility will begin testing DT fuel capsules yielding greater than 10^13 neutrons during 2010. Neutron imaging is an important diagnostic for understanding capsule behavior. Neutrons are imaged at a scintillator after passing through a pinhole. The pixelated, 160-mm square scintillator is made up of ¼ mm diameter rods 50 mm long. Shielding and distance (28 m) are used to preserve the recording diagnostic hardware. Neutron imaging is light starved. We designed a large nine-element collecting lens to relay as much scintillator light as reasonable onto a 75 mm gated microchannel plate (MCP) intensifier. The image from the intensifier’s phosphor passes through a fiber taper onto a CCD camera for digital storage. Alignment of the pinhole and tilting of the scintillator is performed before the relay lens and MCP can be aligned. Careful tilting of the scintillator is done so that each neutron only passes through one rod (no crosstalk allowed). The 3.2 ns decay time scintillator emits light in the deep blue, requiring special glass materials. The glass within the lens housing weighs 26 lbs, with the largest element being 7.7 inches in diameter. The distance between the scintillator and the MCP is only 27 inches. The scintillator emits light with 0.56 NA and the lens collects light at 0.15 NA. Thus, the MCP collects only 7% of the available light. Baffling the stray light is a major concern in the design of the optics. Glass cost considerations, tolerancing, and alignment of this lens system will be discussed.

  19. Design of a neutron penumbral-aperture microscope with 10-. mu. m resolution

    SciTech Connect

    Ress, D.; Lerche, R.A.; Ellis, R.J.; Lane, S.M.

    1990-05-01

    We are currently designing a 10-{mu}m resolution neutron penumbral-aperture microscope to diagnose high-convergence targets at the Nova laser facility. To achieve such high resolution, the new microscope will require substantial improvements in three areas. First, we have designed thick penumbral apertures with extremely sharp cutoffs over a useful ({approx}100 {mu}m) field of view; fabrication of such apertures appears feasible using gold electroplating techniques. Second, the limited field of view and required close proximity of the aperture to the target (2 cm) necessitates a durable mounting and alignment system with {plus}25 {mu}m accuracy. Finally, a neutron detector containing 160,000 scintillator elements is required; readout and optimization of this large array are outstanding issues. 5 refs., 3 figs.

  20. Design of Real-time Neutron Radiography at China Advanced Research Reactor

    NASA Astrophysics Data System (ADS)

    He, Linfeng; Han, Songbai; Wang, Hongli; Hao, Lijie; Wu, Meimei; Wei, Guohai; Wang, Yu; Liu, Yuntao; Sun, Kai; Chen, Dongfeng

    A real-time detector system for neutron radiography based on CMOS camera has been designed for the thermal neutron imaging facility under construction at China Advanced Research Reactor (CARR). This system is equipped with a new scientific CMOS camera with 5.5 million pixels and speed up to 100 fps at full frame. The readout noise is below 2.4 e/pixel. It is capable of providing images with much higher resolution and sensitivity at high frame rate. With optimized optical design and custom-built lens, the capture of quantitative information may be greatly enhanced. The maximum photon received by detector is calculated to be 2.1 × 103/pixel, while the camera resolution is 0.2 mm at 30 fps according to the expected flux (5 × 107 n/cm2/s) at the sample position.

  1. Conceptual design of thorium-fuelled Mitrailleuse accelerator-driven subcritical reactor using D-Be neutron source

    SciTech Connect

    Kokubo, Y.; Kamei, T.

    2012-07-01

    A distributed accelerator is a charged-particle accelerator that uses a new acceleration method based on repeated electrostatic acceleration. This method offers outstanding benefits not possible with the conventional radio-frequency acceleration method, including: (1) high acceleration efficiency, (2) large acceleration current, and (3) lower failure rate made possible by a fully solid-state acceleration field generation circuit. A 'Mitrailleuse Accelerator' is a product we have conceived to optimize this distributed accelerator technology for use with a high-strength neutron source. We have completed the conceptual design of a Mitrailleuse Accelerator and of a thorium-fuelled subcritical reactor driven by a Mitrailleuse Accelerator. This paper presents the conceptual design details and approach to implementing the subcritical reactor core. We will spend the next year or so on detailed design work, and then will start work on developing a prototype for demonstration. If there are no obstacles in setting up a development organization, we expect to finish verifying the prototype's performance by the third quarter of 2015. (authors)

  2. Increasing Sequence Diversity with Flexible Backbone Protein Design: The Complete Redesign of a Protein Hydrophobic Core

    SciTech Connect

    Murphy, Grant S.; Mills, Jeffrey L.; Miley, Michael J.; Machius, Mischa; Szyperski, Thomas; Kuhlman, Brian

    2015-10-15

    Protein design tests our understanding of protein stability and structure. Successful design methods should allow the exploration of sequence space not found in nature. However, when redesigning naturally occurring protein structures, most fixed backbone design algorithms return amino acid sequences that share strong sequence identity with wild-type sequences, especially in the protein core. This behavior places a restriction on functional space that can be explored and is not consistent with observations from nature, where sequences of low identity have similar structures. Here, we allow backbone flexibility during design to mutate every position in the core (38 residues) of a four-helix bundle protein. Only small perturbations to the backbone, 12 {angstrom}, were needed to entirely mutate the core. The redesigned protein, DRNN, is exceptionally stable (melting point >140C). An NMR and X-ray crystal structure show that the side chains and backbone were accurately modeled (all-atom RMSD = 1.3 {angstrom}).

  3. Shielding analysis and design of the KIPT experimental neutron source facility of Ukraine.

    SciTech Connect

    Zhong, Z.; Gohar, M. Y. A.; Naberezhnev, D.; Duo, J.; Nuclear Engineering Division

    2008-10-31

    Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility based on the use of an electron accelerator driven subcritical (ADS) facility [1]. The facility uses the existing electron accelerators of KIPT in Ukraine. The neutron source of the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and the electron energy in the range of 100 to 200 MeV, [2]. The main functions of the facility are the production of medical isotopes and the support of the Ukraine nuclear power industry. Reactor physics experiments and material performance characterization will also be carried out. The subcritical assembly is driven by neutrons generated by the electron beam interactions with the target material. A fraction of these neutrons has an energy above 50 MeV generated through the photo nuclear interactions. This neutron fraction is very small and it has an insignificant contribution to the subcritical assembly performance. However, these high energy neutrons are difficult to shield and they can be slowed down only through the inelastic scattering with heavy isotopes. Therefore the shielding design of this facility is more challenging relative to fission reactors. To attenuate these high energy neutrons, heavy metals (tungsten, iron, etc.) should be used. To reduce the construction cost, heavy concrete with 4.8 g/cm{sup 3} density is selected as a shielding material. The iron weight fraction in this concrete is about 0.6. The shape and thickness of the heavy concrete shield are defined to reduce the biological dose equivalent outside the shield to an acceptable level during operation. At the same time, special attention was give to reduce the total shield mass to reduce the construction cost. The shield design is configured

  4. Design/Operations review of core sampling trucks and associated equipment

    SciTech Connect

    Shrivastava, H.P.

    1996-03-11

    A systematic review of the design and operations of the core sampling trucks was commissioned by Characterization Equipment Engineering of the Westinghouse Hanford Company in October 1995. The review team reviewed the design documents, specifications, operating procedure, training manuals and safety analysis reports. The review process, findings and corrective actions are summarized in this supporting document.

  5. Energy efficient engine. Core engine bearings, drives and configuration: Detailed design report

    NASA Technical Reports Server (NTRS)

    Broman, C. L.

    1981-01-01

    The detailed design of the forward and aft sumps, the accessory drive system, the lubrication system, and the piping/manifold configuration to be employed in the core engine test of the Energy Efficient Engine is addressed. The design goals for the above components were established based on the requirements of the test cell engine.

  6. Design and performance of horizontal-type neutron reflectometer SOFIA at J-PARC/MLF

    NASA Astrophysics Data System (ADS)

    Yamada, N. L.; Torikai, N.; Mitamura, K.; Sagehashi, H.; Sato, S.; Seto, H.; Sugita, T.; Goko, S.; Furusaka, M.; Oda, T.; Hino, M.; Fujiwara, T.; Takahashi, H.; Takahara, A.

    2011-11-01

    Neutron reflectometry is a powerful method for investigating the surface and interfacial structures of materials in the spatial range from nanometers to sub-micrometers. At the Japan Proton Accelerator Research Complex (J-PARC), a high-intensity pulsed neutron beam is produced with a proton accelerator at 220kW, which will be upgraded to 1MW in future. Beamline 16 (BL16) at the Materials and Life Science Experimental Facility (MLF) in J-PARC is dedicated to a horizontal-type reflectometer, and in this beamline, neutrons are transported downward at two different angles, 2.2° and 5.7° , relative to the horizontal. In December 2008, we started to accept the neutron beam at BL16 with the old ARISA reflectometer relocated from the KENS facility, KEK, Japan; and we have now replaced it with the brand-new reflectometer SOFIA (SOFt Interface Analyzer). With a high-flux beam and instrumental upgrade, the observable reflectivity of SOFIA reaches around 10-7 within a few hours for specimens on 3" substrates. In this paper, we will present the design and performance of the SOFIA reflectometer, and discuss some preliminary results on the device development for further upgrade.

  7. Magnetic field optimization and design of a superconducting neutron Wollaston prism

    NASA Astrophysics Data System (ADS)

    Li, F.; Parnell, S. R.; Wang, T.; Baxter, D. V.; Pynn, R.

    2016-04-01

    We present finite element simulations of a superconducting magnetic Wollaston prism (WP) for neutron scattering with high encoding efficiency and low Larmor phase aberrations. To achieve this, we develop and quantify the design criteria. The validation of simulation tools used for this work are investigated by using two software packages: RADIA and MagNet©. Based on the optimization criteria, various possible configurations of WP are explored with MagNet, from which the best configuration is chosen for further optimization. To optimize the best configuration, the influence of various physical parameters is investigated, including the dimensions, shapes and arrangements of components of the device. The optimum WP was built and measured at both pulsed and constant wavelength neutron sources. In flipping mode, a neutron spin flipping efficiency of ∼98.5% was measured independent of neutron wavelength and applied current. In a precession mode, measurements showed a highly linear Larmor phase variation along the horizontal direction with low depolarization. Simulations of the device agree well with the experimental measurements. Possible applications of the device are also discussed.

  8. HEIMDAL: A thermal neutron powder diffractometer with high and flexible resolution combined with SANS and neutron imaging - Designed for materials science studies at the European Spallation Source

    NASA Astrophysics Data System (ADS)

    Holm, Sonja L.; Lefmann, Kim; Henry, Paul F.; Bertelsen, Mads; Schefer, Jürg; Christensen, Mogens

    2016-08-01

    HEIMDAL will be a multi length scale neutron scattering instrument for the study of structures covering almost nine orders of magnitude from 0.01 nm to 50 mm. The instrument is accepted for construction at the European Spallation Source (ESS) and features a variable resolution thermal neutron powder diffractometer (TNPD), combined with small angle neutron scattering (SANS) and neutron imaging (NI). The instrument uses a novel combination of a cold and a thermal guide to fulfill the diverse requirements for diffraction and SANS. With an instrument length of 170 m, HEIMDAL will take advantage of the high neutron flux of the long pulse at ESS, whilst maintaining a high q-resolution due to the long flight path. The q-range coverage is up to 20 Å-1 allowing low-resolution PDF analysis. With the addition of SANS, HEIMDAL will be able to cover a uniquely broad length scale within a single instrumental set-up. HEIMDAL will be able to accommodate modern materials research in a broad variety of fields, and the task of the instrument will be to study advanced functional materials in action, as in situ and in operandi at multiple length scales (0.01-100 nm) quasi simultaneously. The instrument combines state-of-the-art neutron scattering techniques (TNPD, SANS, and NI) with the goal of studying real materials, in real time, under real conditions. This article describes the instrument design ideas, calculations and results of simulations and virtual experiments.

  9. Management of Grossly Decayed Mandibular Molar with Different Designs of Split Cast Post and Core

    PubMed Central

    Bansal, Rashmi; Mehrotra, Nakul; Chowdhary, Priyanka; Gurtu, Anuraag

    2016-01-01

    Mandibular molar with extensive loss of tooth structure, especially where no cavity wall is remaining, and insertion of posts in both the roots appear necessary so as to achieve proper retention for the core material. A single unit metal casting with two posts, one in the mesial root and the other in the distal divergent root, is difficult to fabricate due to difference in the path of insertion of the two posts. Multisection post and core or single cast post and core with auxiliary post can be an effective design to manage grossly decayed mandibular molars. PMID:27144038

  10. Sanitation and design of lettuce coring knives for minimizing Escherichia coli O157:H7 contamination.

    PubMed

    Zhou, Bin; Luo, Yaguang; Millner, Patricia; Feng, Hao

    2012-03-01

    This study was undertaken to examine the effect of ultrasound in combination with chlorine on the reduction of Escherichia coli O157:H7 populations on lettuce coring knives. Two new coring devices designed to mitigate pathogen attachment were fabricated and evaluated. The coring rings of the knives were dip inoculated with soil slurry containing 10⁶ E. coli cells and treated with chlorinated water with and without ultrasonication for 30, 60, and 120 s. The rough welding joints on currently used in-field lettuce coring knives provided a site conducive to bacterial attachment and resistant to cell removal during sanitation treatment. The two modified coring knives harbored significantly fewer E. coli cells than did the currently used commercial model, and the efficacy of the disinfection treatment was high (P < 0.05). Ultrasound treatment reduced the E. coli O157:H7 counts to below the detection limit of 1.10 log CFU/cm² at both the coring ring blade and welding joint within 30 s in 1 ppm of chlorinated water. The redesigned coring knives and an ultrasound plus chlorine combination treatment may provide practical options for minimizing the microbial safety hazards of lettuce processed by core-in-field operations.

  11. Design and pilot evaluation of the RAH-66 Comanche Core AFCS

    NASA Technical Reports Server (NTRS)

    Fogler, Donald L., Jr.; Keller, James F.

    1993-01-01

    This paper addresses the design and pilot evaluation of the Core Automatic Flight Control System (AFCS) for the Reconnaissance/Attack Helicopter (RAH-66) Comanche. During the period from November 1991 through February 1992, the RAH-66 Comanche control laws were evaluated through a structured pilot acceptance test using a motion base simulator. Design requirements, descriptions of the control law design, and handling qualities data collected from ADS-33 maneuvers are presented.

  12. Analysis of Stainless Steel Sandwich Panels with a Metal Foam Core for Lightweight Fan Blade Design

    NASA Technical Reports Server (NTRS)

    Min, James B.; Ghosn, Louis J.; Lerch, Bradley A.; Raj, Sai V.; Holland, Frederic A., Jr.; Hebsur, Mohan G.

    2004-01-01

    The quest for cheap, low density and high performance materials in the design of aircraft and rotorcraft engine fan and propeller blades poses immense challenges to the materials and structural design engineers. The present study investigates the use of a sandwich foam fan blade mae up of solid face sheets and a metal foam core. The face sheets and the metal foam core material were an aerospace grade precipitation hardened 17-4 PH stainless steel with high strength and high toughness. The resulting structures possesses a high stiffness while being lighter than a similar solid construction. The material properties of 17-4 PH metal foam are reviewed briefly to describe the characteristics of sandwich structure for a fan blade application. A vibration analysis for natural frequencies and a detailed stress analysis on the 17-4 PH sandwich foam blade design for different combinations of kin thickness and core volume are presented with a comparison to a solid titanium blade.

  13. Core Noise: Implications of Emerging N+3 Designs and Acoustic Technology Needs

    NASA Technical Reports Server (NTRS)

    Hultgren, Lennart S.

    2011-01-01

    This presentation is a summary of the core-noise implications of NASA's primary N+3 aircraft concepts. These concepts are the MIT/P&W D8.5 Double Bubble design, the Boeing/GE SUGAR Volt hybrid gas-turbine/electric engine concept, the NASA N3-X Turboelectric Distributed Propulsion aircraft, and the NASA TBW-XN Truss-Braced Wing concept. The first two are future concepts for the Boeing 737/Airbus A320 US transcontinental mission of 180 passengers and a maximum range of 3000 nm. The last two are future concepts for the Boeing 777 transpacific mission of 350 passengers and a 7500 nm range. Sections of the presentation cover: turbofan design trends on the N+1.5 time frame and the already emerging importance of core noise; the NASA N+3 concepts and associated core-noise challenges; the historical trends for the engine bypass ratio (BPR), overall pressure ratio (OPR), and combustor exit temperature; and brief discussion of a noise research roadmap being developed to address the core-noise challenges identified for the N+3 concepts. The N+3 conceptual aircraft have (i) ultra-high bypass ratios, in the rage of 18 - 30, accomplished by either having a small-size, high-power-density core, an hybrid design which allows for an increased fan size, or by utilizing a turboelectric distributed propulsion design; and (ii) very high OPR in the 50 - 70 range. These trends will elevate the overall importance of turbomachinery core noise. The N+3 conceptual designs specify the need for the development and application of advanced liners and passive and active control strategies to reduce the core noise. Current engineering prediction of core noise uses semi-empirical methods based on older turbofan engines, with (at best) updates for more recent designs. The models have not seen the same level of development and maturity as those for fan and jet noise and are grossly inadequate for the designs considered for the N+3 time frame. An aggressive program for the development of updated noise

  14. A new method based on low background instrumental neutron activation analysis for major, trace and ultra-trace element determination in atmospheric mineral dust from polar ice cores.

    PubMed

    Baccolo, Giovanni; Clemenza, Massimiliano; Delmonte, Barbara; Maffezzoli, Niccolò; Nastasi, Massimiliano; Previtali, Ezio; Prata, Michele; Salvini, Andrea; Maggi, Valter

    2016-05-30

    Dust found in polar ice core samples present extremely low concentrations, in addition the availability of such samples is usually strictly limited. For these reasons the chemical and physical analysis of polar ice cores is an analytical challenge. In this work a new method based on low background instrumental neutron activation analysis (LB-INAA) for the multi-elemental characterization of the insoluble fraction of dust from polar ice cores is presented. Thanks to an accurate selection of the most proper materials and procedures it was possible to reach unprecedented analytical performances, suitable for ice core analyses. The method was applied to Antarctic ice core samples. Five samples of atmospheric dust (μg size) from ice sections of the Antarctic Talos Dome ice core were prepared and analyzed. A set of 37 elements was quantified, spanning from all the major elements (Na, Mg, Al, Si, K, Ca, Ti, Mn and Fe) to trace ones, including 10 (La, Ce, Nd, Sm, Eu, Tb, Ho, Tm, Yb and Lu) of the 14 natural occurring lanthanides. The detection limits are in the range of 10(-13)-10(-6) g, improving previous results of 1-3 orders of magnitude depending on the element; uncertainties lies between 4% and 60%.

  15. A new method based on low background instrumental neutron activation analysis for major, trace and ultra-trace element determination in atmospheric mineral dust from polar ice cores.

    PubMed

    Baccolo, Giovanni; Clemenza, Massimiliano; Delmonte, Barbara; Maffezzoli, Niccolò; Nastasi, Massimiliano; Previtali, Ezio; Prata, Michele; Salvini, Andrea; Maggi, Valter

    2016-05-30

    Dust found in polar ice core samples present extremely low concentrations, in addition the availability of such samples is usually strictly limited. For these reasons the chemical and physical analysis of polar ice cores is an analytical challenge. In this work a new method based on low background instrumental neutron activation analysis (LB-INAA) for the multi-elemental characterization of the insoluble fraction of dust from polar ice cores is presented. Thanks to an accurate selection of the most proper materials and procedures it was possible to reach unprecedented analytical performances, suitable for ice core analyses. The method was applied to Antarctic ice core samples. Five samples of atmospheric dust (μg size) from ice sections of the Antarctic Talos Dome ice core were prepared and analyzed. A set of 37 elements was quantified, spanning from all the major elements (Na, Mg, Al, Si, K, Ca, Ti, Mn and Fe) to trace ones, including 10 (La, Ce, Nd, Sm, Eu, Tb, Ho, Tm, Yb and Lu) of the 14 natural occurring lanthanides. The detection limits are in the range of 10(-13)-10(-6) g, improving previous results of 1-3 orders of magnitude depending on the element; uncertainties lies between 4% and 60%. PMID:27154827

  16. Baseline Design Compliance Matrix for the Rotary Mode Core Sampling System

    SciTech Connect

    LECHELT, J.A.

    2000-10-17

    The purpose of the design compliance matrix (DCM) is to provide a single-source document of all design requirements associated with the fifteen subsystems that make up the rotary mode core sampling (RMCS) system. It is intended to be the baseline requirement document for the RMCS system and to be used in governing all future design and design verification activities associated with it. This document is the DCM for the RMCS system used on Hanford single-shell radioactive waste storage tanks. This includes the Exhauster System, Rotary Mode Core Sample Trucks, Universal Sampling System, Diesel Generator System, Distribution Trailer, X-Ray Cart System, Breathing Air Compressor, Nitrogen Supply Trailer, Casks and Cask Truck, Service Trailer, Core Sampling Riser Equipment, Core Sampling Support Trucks, Foot Clamp, Ramps and Platforms and Purged Camera System. Excluded items are tools such as light plants and light stands. Other items such as the breather inlet filter are covered by a different design baseline. In this case, the inlet breather filter is covered by the Tank Farms Design Compliance Matrix.

  17. Polar-Drive Designs for Optimizing Neutron Yields on the National Ignition Faciltiy

    SciTech Connect

    Cok, A.M.; Craxton, R.S.; McKenty, P.W.

    2008-09-10

    Polar-drive designs are proposed for producing symmetric implosions of thin-shell, DT gas-filled targets leading to high fusion-neutron yields for neutron-diagnostic development. The designs can be implemented as soon as the National Ignition Facility (NIF) [E. M. Campbell and W. J. Hogan, Plasma Phys. Control. Fusion 41, B39 (1999)] is operational as they use indirect-drive phase plates. Two-dimensional simulations using the hydrodynamics code SAGE [R. S. Craxton and R. L. McCrory, J. Appl. Phys. 56, 108 (1984)] have shown that good low-mode uniformity can be obtained by choosing combinations of pointing and defocusing of the beams, including pointing offsets of individual beams within some of the NIF laser-beam quads. The optimizations have been carried out for total laser energies ranging from 350 kJ to 1.5 MJ, enabling the optimum pointing and defocusing parameters to be determined through interpolation for any given laser energy in this range. Neutron yields in the range of 10^15–10^16 are expected.

  18. Polar-drive designs for optimizing neutron yields on the National Ignition Facility

    SciTech Connect

    Cok, A. M.; Craxton, R. S.; McKenty, P. W.

    2008-08-15

    Polar-drive designs are proposed for producing symmetric implosions of thin-shell, DT gas-filled targets leading to high fusion-neutron yields for neutron-diagnostic development. The designs can be implemented as soon as the National Ignition Facility (NIF) [E. M. Campbell and W. J. Hogan, Plasma Phys. Control. Fusion 41, B39 (1999)] is operational as they use indirect-drive phase plates. Two-dimensional simulations using the hydrodynamics code SAGE [R. S. Craxton and R. L. McCrory, J. Appl. Phys. 56, 108 (1984)] have shown that good low-mode uniformity can be obtained by choosing combinations of pointing and defocusing of the beams, including pointing offsets of individual beams within some of the NIF laser-beam quads. The optimizations have been carried out for total laser energies ranging from 350 kJ to 1.5 MJ, enabling the optimum pointing and defocusing parameters to be determined through interpolation for any given laser energy in this range. Neutron yields in the range of 10{sup 15}-10{sup 16} are expected.

  19. Polar-drive designs for optimizing neutron yields on the National Ignition Facility

    NASA Astrophysics Data System (ADS)

    Cok, A. M.; Craxton, R. S.; McKenty, P. W.

    2008-08-01

    Polar-drive designs are proposed for producing symmetric implosions of thin-shell, DT gas-filled targets leading to high fusion-neutron yields for neutron-diagnostic development. The designs can be implemented as soon as the National Ignition Facility (NIF) [E. M. Campbell and W. J. Hogan, Plasma Phys. Control. Fusion 41, B39 (1999)] is operational as they use indirect-drive phase plates. Two-dimensional simulations using the hydrodynamics code SAGE [R. S. Craxton and R. L. McCrory, J. Appl. Phys. 56, 108 (1984)] have shown that good low-mode uniformity can be obtained by choosing combinations of pointing and defocusing of the beams, including pointing offsets of individual beams within some of the NIF laser-beam quads. The optimizations have been carried out for total laser energies ranging from 350kJto1.5MJ, enabling the optimum pointing and defocusing parameters to be determined through interpolation for any given laser energy in this range. Neutron yields in the range of 1015-1016 are expected.

  20. Design and implementation of a multiaxial loading capability during heating on an engineering neutron diffractometer

    SciTech Connect

    Benafan, O.; Padula, S. A.; Skorpenske, H. D.; An, K.; Vaidyanathan, R.

    2014-10-01

    A gripping capability was designed, implemented, and tested for in situ neutron diffraction measurements during multiaxial loading and heating on the VULCAN engineering materials diffractometer at the spallation neutron source at Oak Ridge National Laboratory. The proposed capability allowed for the acquisition of neutron spectra during tension, compression, torsion, and/or complex loading paths at elevated temperatures. The design consisted of age-hardened, Inconel{sup ®} 718 grips with direct attachment to the existing MTS load frame having axial and torsional capacities of 100 kN and 400 N·m, respectively. Internal cooling passages were incorporated into the gripping system for fast cooling rates during high temperature experiments up to ~1000 K. The specimen mounting couplers combined a threaded and hexed end-connection for ease of sample installation/removal without introducing any unwanted loads. Instrumentation of this capability is documented in this work along with various performance parameters. The gripping system was utilized to investigate deformation in NiTi shape memory alloys under various loading/control modes (e.g., isothermal, isobaric, and cyclic), and preliminary results are presented. The measurements facilitated the quantification of the texture, internal strain, and phase fraction evolution in NiTi shape memory alloys under various loading/control modes.

  1. Design and implementation of a multiaxial loading capability during heating on an engineering neutron diffractometer

    NASA Astrophysics Data System (ADS)

    Benafan, O.; Padula, S. A.; Skorpenske, H. D.; An, K.; Vaidyanathan, R.

    2014-10-01

    A gripping capability was designed, implemented, and tested for in situ neutron diffraction measurements during multiaxial loading and heating on the VULCAN engineering materials diffractometer at the spallation neutron source at Oak Ridge National Laboratory. The proposed capability allowed for the acquisition of neutron spectra during tension, compression, torsion, and/or complex loading paths at elevated temperatures. The design consisted of age-hardened, Inconel® 718 grips with direct attachment to the existing MTS load frame having axial and torsional capacities of 100 kN and 400 N.m, respectively. Internal cooling passages were incorporated into the gripping system for fast cooling rates during high temperature experiments up to ˜1000 K. The specimen mounting couplers combined a threaded and hexed end-connection for ease of sample installation/removal without introducing any unwanted loads. Instrumentation of this capability is documented in this work along with various performance parameters. The gripping system was utilized to investigate deformation in NiTi shape memory alloys under various loading/control modes (e.g., isothermal, isobaric, and cyclic), and preliminary results are presented. The measurements facilitated the quantification of the texture, internal strain, and phase fraction evolution in NiTi shape memory alloys under various loading/control modes.

  2. Design and implementation of a multiaxial loading capability during heating on an engineering neutron diffractometer.

    PubMed

    Benafan, O; Padula, S A; Skorpenske, H D; An, K; Vaidyanathan, R

    2014-10-01

    A gripping capability was designed, implemented, and tested for in situ neutron diffraction measurements during multiaxial loading and heating on the VULCAN engineering materials diffractometer at the spallation neutron source at Oak Ridge National Laboratory. The proposed capability allowed for the acquisition of neutron spectra during tension, compression, torsion, and/or complex loading paths at elevated temperatures. The design consisted of age-hardened, Inconel(®) 718 grips with direct attachment to the existing MTS load frame having axial and torsional capacities of 100 kN and 400 N·m, respectively. Internal cooling passages were incorporated into the gripping system for fast cooling rates during high temperature experiments up to ∼1000 K. The specimen mounting couplers combined a threaded and hexed end-connection for ease of sample installation/removal without introducing any unwanted loads. Instrumentation of this capability is documented in this work along with various performance parameters. The gripping system was utilized to investigate deformation in NiTi shape memory alloys under various loading/control modes (e.g., isothermal, isobaric, and cyclic), and preliminary results are presented. The measurements facilitated the quantification of the texture, internal strain, and phase fraction evolution in NiTi shape memory alloys under various loading/control modes.

  3. Neutron and Synchrotron Radiation Studies for Designer Materials, Sustainable Energy and Healthy Lives

    NASA Astrophysics Data System (ADS)

    Gibson, J. Murray

    2009-05-01

    Probably the most prolific use of large accelerators today is in the creation of bright beams of x-ray photons or neutrons. The number of scientific users of such sources in the US alone is approaching 10,000. I will describe the some of the major applications of synchrotron and neutron radiation and their impact on society. If you have AIDS, need a better IPOD or a more efficient car, or want to clean up a superfund site, you are benefitting from these accelerators. The design of new materials is becoming more and more dependent on structural information from these sources. I will identify the trends in applications which are demanding new sources with greater capabilities.

  4. Design and performance of a large area neutron sensitive anger camera

    SciTech Connect

    Visscher, Theodore; Montcalm, Christopher A.; Donahue, Jr., Cornelius; Riedel, Richard A.

    2015-05-21

    We describe the design and performance of a 157mm x 157mm two dimensional neutron detector. The detector uses the Anger principle to determine the position of neutrons. We have verified FWHM resolution of < 1.2mm with distortion < 0.5mm on over 50 installed Anger Cameras. The performance of the detector is limited by the light yield of the scintillator, and it is estimated that the resolution of the current detector could be doubled with a brighter scintillator. Data collected from small (<1mm3) single crystal reference samples at the single crystal instrument TOPAZ provide results with low Rw(F) values

  5. Design and analysis of large-core single-mode windmill single crystal sapphire optical fiber

    NASA Astrophysics Data System (ADS)

    Cheng, Yujie; Hill, Cary; Liu, Bo; Yu, Zhihao; Xuan, Haifeng; Homa, Daniel; Wang, Anbo; Pickrell, Gary

    2016-06-01

    We present a large-core single-mode "windmill" single crystal sapphire optical fiber (SCSF) design, which exhibits single-mode operation by stripping off the higher-order modes (HOMs) while maintaining the fundamental mode. The "windmill" SCSF design was analyzed using the finite element analysis method, in which all the HOMs are leaky. The numerical simulation results show single-mode operation in the spectral range from 0.4 to 2 μm in the windmill SCSF, with an effective core diameter as large as 14 μm. Such fiber is expected to improve the performance of many of the current sapphire fiber optic sensor structures.

  6. Core-shell designed scaffolds for drug delivery and tissue engineering.

    PubMed

    Perez, Roman A; Kim, Hae-Won

    2015-07-01

    Scaffolds that secure and deliver therapeutic ingredients like signaling molecules and stem cells hold great promise for drug delivery and tissue engineering. Employing a core-shell design for scaffolds provides a promising solution. Some unique methods, such as co-concentric nozzle extrusion, microfluidics generation, and chemical confinement reactions, have been successful in producing core-shelled nano/microfibers and nano/microspheres. Signaling molecules and drugs, spatially allocated to the core and/or shell part, can be delivered in a controllable and sequential manner for optimal therapeutic effects. Stem cells can be loaded within the core part on-demand, safely protected from the environments, which ultimately affords ex vivo culture and in vivo tissue engineering. The encapsulated cells experience three-dimensional tissue-mimic microenvironments in which therapeutic molecules are secreted to the surrounding tissues through the semi-permeable shell. Tuning the material properties of the core and shell, changing the geometrical parameters, and shaping them into proper forms significantly influence the release behaviors of biomolecules and the fate of the cells. This topical issue highlights the immense usefulness of core-shell designs for the therapeutic actions of scaffolds in the delivery of signaling molecules and stem cells for tissue regeneration and disease treatment.

  7. Design and Performance of South Ukraine Nuclear Power Plant Mixed Cores

    SciTech Connect

    Abdullayev, A. M.; Baydulin, V.; Zhukov, A. I.; Latorre, Richard

    2011-09-24

    In 2010, 42 Westinghouse fuel assemblies (WFAs) were loaded into the core of South Ukraine Nuclear Power Plant (SUNPP) Unit 3 after four successful cycles with 6 Westinghouse Lead Test Assemblies. The scope of safety substantiating documents required for the regulatory approval of this mixed core was extended considerably, particularly with development and implementation of new methodologies and 3-D kinetic codes. Additional verification for all employed codes was also performed. Despite the inherent hydraulic non-uniformity of a mixed core, it was possible to demonstrate that all design and operating restrictions for three different types of fuel (TVS-M, TVSA and WFA) loaded in the core were conservatively met. This paper provides the main results from the first year of operation of the core loaded with 42 WFAs, the predicted parameters for the transition and equilibrium cycles with WFAs, comparisons of predicted versus measured core parameters, as well as the acceptable margin evaluation results for reactivity accidents using the 3-D kinetic codes. To date WFA design parameters have been confirmed by operation experience.

  8. The design and performance of IceCube DeepCore

    NASA Astrophysics Data System (ADS)

    Abbasi, R.; Abdou, Y.; Abu-Zayyad, T.; Ackermann, M.; Adams, J.; Aguilar, J. A.; Ahlers, M.; Allen, M. M.; Altmann, D.; Andeen, K.; Auffenberg, J.; Bai, X.; Baker, M.; Barwick, S. W.; Bay, R.; Bazo Alba, J. L.; Beattie, K.; Beatty, J. J.; Bechet, S.; Becker, J. K.; Becker, K.-H.; Benabderrahmane, M. L.; BenZvi, S.; Berdermann, J.; Berghaus, P.; Berley, D.; Bernardini, E.; Bertrand, D.; Besson, D. Z.; Bindig, D.; Bissok, M.; Blaufuss, E.; Blumenthal, J.; Boersma, D. J.; Bohm, C.; Bose, D.; Böser, S.; Botner, O.; Brown, A. M.; Buitink, S.; Caballero-Mora, K. S.; Carson, M.; Chirkin, D.; Christy, B.; Clevermann, F.; Cohen, S.; Colnard, C.; Cowen, D. F.; Cruz Silva, A. H.; D'Agostino, M. V.; Danninger, M.; Daughhetee, J.; Davis, J. C.; De Clercq, C.; Degner, T.; Demirörs, L.; Descamps, F.; Desiati, P.; de Vries-Uiterweerd, G.; DeYoung, T.; Díaz-Vélez, J. C.; Dierckxsens, M.; Dreyer, J.; Dumm, J. P.; Dunkman, M.; Eisch, J.; Ellsworth, R. W.; Engdegård, O.; Euler, S.; Evenson, P. A.; Fadiran, O.; Fazely, A. R.; Fedynitch, A.; Feintzeig, J.; Feusels, T.; Filimonov, K.; Finley, C.; Fischer-Wasels, T.; Fox, B. D.; Franckowiak, A.; Franke, R.; Gaisser, T. K.; Gallagher, J.; Gerhardt, L.; Gladstone, L.; Glüsenkamp, T.; Goldschmidt, A.; Goodman, J. A.; Góra, D.; Grant, D.; Griesel, T.; Groß, A.; Grullon, S.; Gurtner, M.; Ha, C.; Haj Ismail, A.; Hallgren, A.; Halzen, F.; Han, K.; Hanson, K.; Heinen, D.; Helbing, K.; Hellauer, R.; Hickford, S.; Hill, G. C.; Hoffman, K. D.; Hoffmann, B.; Homeier, A.; Hoshina, K.; Huelsnitz, W.; Hülß, J.-P.; Hulth, P. O.; Hultqvist, K.; Hussain, S.; Ishihara, A.; Jacobi, E.; Jacobsen, J.; Japaridze, G. S.; Johansson, H.; Kampert, K.-H.; Kappes, A.; Karg, T.; Karle, A.; Kenny, P.; Kiryluk, J.; Kislat, F.; Klein, S. R.; Köhne, J.-H.; Kohnen, G.; Kolanoski, H.; Köpke, L.; Koskinen, D. J.; Kowalski, M.; Kowarik, T.; Krasberg, M.; Kroll, G.; Kurahashi, N.; Kuwabara, T.; Labare, M.; Laihem, K.; Landsman, H.; Larson, M. J.; Lauer, R.; Lünemann, J.; Madsen, J.; Marotta, A.; Maruyama, R.; Mase, K.; Matis, H. S.; Meagher, K.; Merck, M.; Mészáros, P.; Meures, T.; Miarecki, S.; Middell, E.; Milke, N.; Miller, J.; Montaruli, T.; Morse, R.; Movit, S. M.; Nahnhauer, R.; Nam, J. W.; Naumann, U.; Nygren, D. R.; Odrowski, S.; Olivas, A.; Olivo, M.; O'Murchadha, A.; Panknin, S.; Paul, L.; Pérez de los Heros, C.; Petrovic, J.; Piegsa, A.; Pieloth, D.; Porrata, R.; Posselt, J.; Price, P. B.; Przybylski, G. T.; Rawlins, K.; Redl, P.; Resconi, E.; Rhode, W.; Ribordy, M.; Richman, M.; Rodrigues, J. P.; Rothmaier, F.; Rott, C.; Ruhe, T.; Rutledge, D.; Ruzybayev, B.; Ryckbosch, D.; Sander, H.-G.; Santander, M.; Sarkar, S.; Schatto, K.; Schmidt, T.; Schönwald, A.; Schukraft, A.; Schultes, A.; Schulz, O.; Schunck, M.; Seckel, D.; Semburg, B.; Seo, S. H.; Sestayo, Y.; Seunarine, S.; Silvestri, A.; Spiczak, G. M.; Spiering, C.; Stamatikos, M.; Stanev, T.; Stezelberger, T.; Stokstad, R. G.; Stößl, A.; Strahler, E. A.; Ström, R.; Stüer, M.; Sullivan, G. W.; Swillens, Q.; Taavola, H.; Taboada, I.; Tamburro, A.; Tepe, A.; Ter-Antonyan, S.; Tilav, S.; Toale, P. A.; Toscano, S.; Tosi, D.; van Eijndhoven, N.; Vandenbroucke, J.; Van Overloop, A.; van Santen, J.; Vehring, M.; Voge, M.; Walck, C.; Waldenmaier, T.; Wallraff, M.; Walter, M.; Weaver, Ch.; Wendt, C.; Westerhoff, S.; Whitehorn, N.; Wiebe, K.; Wiebusch, C. H.; Williams, D. R.; Wischnewski, R.; Wissing, H.; Wolf, M.; Wood, T. R.; Woschnagg, K.; Xu, C.; Xu, D. L.; Xu, X. W.; Yanez, J. P.; Yodh, G.; Yoshida, S.; Zarzhitsky, P.; Zoll, M.

    2012-05-01

    The IceCube neutrino observatory in operation at the South Pole, Antarctica, comprises three distinct components: a large buried array for ultrahigh energy neutrino detection, a surface air shower array, and a new buried component called DeepCore. DeepCore was designed to lower the IceCube neutrino energy threshold by over an order of magnitude, to energies as low as about 10 GeV. DeepCore is situated primarily 2100 m below the surface of the icecap at the South Pole, at the bottom center of the existing IceCube array, and began taking physics data in May 2010. Its location takes advantage of the exceptionally clear ice at those depths and allows it to use the surrounding IceCube detector as a highly efficient active veto against the principal background of downward-going muons produced in cosmic-ray air showers. DeepCore has a module density roughly five times higher than that of the standard IceCube array, and uses photomultiplier tubes with a new photocathode featuring a quantum efficiency about 35% higher than standard IceCube PMTs. Taken together, these features of DeepCore will increase IceCube's sensitivity to neutrinos from WIMP dark matter annihilations, atmospheric neutrino oscillations, galactic supernova neutrinos, and point sources of neutrinos in the northern and southern skies. In this paper we describe the design and initial performance of DeepCore.

  9. The Design and Performance of IceCube DeepCore

    NASA Technical Reports Server (NTRS)

    Stamatikos, M.

    2012-01-01

    The IceCube neutrino observatory in operation at the South Pole, Antarctica, comprises three distinct components: a large buried array for ultrahigh energy neutrino detection, a surface air shower array, and a new buried component called DeepCore. DeepCore was designed to lower the IceCube neutrino energy threshold by over an order of magnitude, to energies as low as about 10 GeV. DeepCore is situated primarily 2100 m below the surface of the icecap at the South Pole, at the bottom center of the existing IceCube array, and began taking pbysics data in May 2010. Its location takes advantage of the exceptionally clear ice at those depths and allows it to use the surrounding IceCube detector as a highly efficient active veto against the principal background of downward-going muons produced in cosmic-ray air showers. DeepCore has a module density roughly five times higher than that of the standard IceCube array, and uses photomultiplier tubes with a new photocathode featuring a quantum efficiency about 35% higher than standard IceCube PMTs. Taken together, these features of DeepCore will increase IceCube's sensitivity to neutrinos from WIMP dark matter annihilations, atmospheric neutrino oscillations, galactic supernova neutrinos, and point sources of neutrinos in the northern and southern skies. In this paper we describe the design and initial performance of DeepCore.

  10. THERMAL NEUTRONIC REACTOR

    DOEpatents

    Spinrad, B.I.

    1960-01-12

    A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

  11. Design of the Testing Set-up for a Nuclear Fuel Rod by Neutron Radiography at CARR

    NASA Astrophysics Data System (ADS)

    Wei, Guohai; Han, Songbai; Wang, Hongli; Hao, Lijie; Wu, Meimei; He, Linfeng; Wang, Yu; Liu, Yuntao; Sun, Kai; Chen, Dongfeng

    In this paper, an experimental set-up dedicated to non-destructively test a 15cm-long Pressurized Water Reactor (PWR) nuclear fuel rod by neutron radiography (NR) is described. It consists of three parts: transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo Simulation by the MCNP code. The material for the shell of the transport container was chosen to be lead with the thickness of 13 cm. Also, the mechanical devices were designed to control fuel rod movement inside the container. The imaging block was designed as the exposure platform, with three openings for the neutron beam, neutron converter foil, and specimen. Development and application of this experimental set-up will help gain much experience for investigating the actual irradiated fuel rod by neutron radiography at CARR in the future.

  12. Enhancement of REBUS-3/DIF3D for whole-core neutronic analysis of prismatic very high temperature reactor (VHTR).

    SciTech Connect

    Lee, C. H.; Zhong, Z.; Taiwo, T.A.; Yang, W.S.; Khalil, H.S.; Smith, M.A.; Nuclear Engineering Division

    2006-10-13

    Enhancements have been made to the REBUS-3/DIF3D code suite to facilitate its use for the design and analysis of prismatic Very High Temperature Reactors (VHTRs). A new cross section structure, using table-lookup, has been incorporated to account for cross section changes with burnup and fuel and moderator temperatures. For representing these cross section dependencies, three new modules have been developed using FORTRAN 90/95 object-oriented data structures and implemented within the REBUS-3 code system. These modules provide a cross section storage procedure, construct microscopic cross section data for all isotopes, and contain a single block of banded scattering data for efficient data management. Fission products other than I, Xe, Pm, and Sm, can be merged into a single lumped fission product to save storage space, memory, and computing time without sacrificing the REBUS-3 solution accuracy. A simple thermal-hydraulic (thermal-fluid) feedback model has been developed for prismatic VHTR cores and implemented in REBUS-3 for temperature feedback calculations. Axial conduction was neglected in the formulation because of its small magnitude compared to radial (planar) conduction. With the simple model, the average fuel and graphite temperatures are accurately estimated compared to reference STAR-CD results. The feedback module is currently operational for the non-equilibrium fuel cycle analysis option of REBUS-3. Future work should include the extension of this capability to the equilibrium cycle option of the code and additional verification of the feedback module. For the simulation of control rods in VHTR cores, macroscopic cross section deviations (deltas) have been defined to account for the effect of control rod insertion. The REBUS-3 code has been modified to use the appropriately revised cross sections when control rods are inserted in a calculation node. In order to represent asymmetric core blocks (e.g., fuel blocks or reflector blocks containing

  13. Designing with advanced composites; Report on the European Core Conference, 1st, Zurich, Switzerland, Oct. 20, 21, 1988, Conference Papers

    SciTech Connect

    Not Available

    1988-01-01

    The present conference discusses the development history of sandwich panel construction, production methods and quality assurance for Nomex sandwich panel core papers, the manufacture of honeycomb cores, state-of-the-art design methods for honeycomb-core panels, the Airbus A320 airliner's CFRP rudder structure, and the design tradeoffs encountered in honeycomb-core structures' design. Also discussed are sandwich-construction aircraft cabin interiors meeting new FAA regulations, the use of Nomex honeycomb cores in composite structures, a low-cost manufacturing technique for sandwich structures, and the Starship sandwich panel-incorporating airframe primary structure.

  14. Modeling and design of a reload PWR core for a 48-month fuel cycle

    SciTech Connect

    McMahon, M.V.; Driscoll, M.J.; Todreas, N.E.

    1997-05-01

    The objective of this research was to use state-of-the-art nuclear and fuel performance packages to evaluate the feasibility and costs of a 48 calendar month core in existing pressurized water reactor (PWR) designs, considering the full range of practical design and economic considerations. The driving force behind this research is the desire to make nuclear power more economically competitive with fossil fuel options by expanding the scope for achievement of higher capacity factors. Using CASMO/SIMULATE, a core design with fuel enriched to 7{sup w}/{sub o} U{sup 235} for a single batch loaded, 48-month fuel cycle has been developed. This core achieves an ultra-long cycle length without exceeding current fuel burnup limits. The design uses two different types of burnable poisons. Gadolinium in the form of gadolinium oxide (Gd{sub 2}O{sub 3}) mixed with the UO{sub 2} of selected pins is sued to hold down initial reactivity and to control flux peaking throughout the life of the core. A zirconium di-boride (ZrB{sub 2}) integral fuel burnable absorber (IFBA) coating on the Gd{sub 2}O{sub 3}-UO{sub 2} fuel pellets is added to reduce the critical soluble boron concentration in the reactor coolant to within acceptable limits. Fuel performance issues of concern to this design are also outlined and areas which will require further research are highlighted.

  15. Preliminary Neutronics Design Studies for a Molten Salt Blanket LIFE Engine

    SciTech Connect

    Powers, J

    2008-10-23

    The Laser Inertial Confinement Fusion Fission Energy (LIFE) Program being developed at Lawrence Livermore National Laboratory (LLNL) aims to design a hybrid fission-fusion subcritical nuclear engine that uses a laser-driven Inertial Confinement Fusion (ICF) system to drive a subcritical fission blanket. This combined fusion-fission hybrid system could be used for generating electricity, material transmutation or incineration, or other applications. LIFE does not require enriched fuel since it is a sub-critical system and LIFE can sustain power operation beyond the burnup levels at which typical fission reactors need to be refueled. In light of these factors, numerous options have been suggested and are being investigated. Options being investigated include fueling LIFE engines with spent nuclear fuel to aid in disposal/incineration of commercial spent nuclear fuel or using depleted uranium or thorium fueled options to enhance proliferation resistance and utilize non-fissile materials [1]. LIFE engine blanket designs using a molten salt fuel system represent one area of investigation. Possible applications of a LIFE engine with a molten salt blanket include uses as a spent nuclear fuel burner, fissile fuel breeding platform, and providing a backup alternative to other LIFE engine blanket designs using TRISO fuel particles in case the TRISO particles are found to be unable to withstand the irradiation they will be subjected to. These molten salts consist of a mixture of LiF with UF{sub 4} or ThF{sub 4} or some combination thereof. Future systems could look at using PuF{sub 3} or PuF{sub 4} as well, though no work on such system with initial plutonium loadings has been performed for studies documented in this report. The purpose of this report is to document preliminary neutronics design studies performed to support the development of a molten salt blanket LIFE engine option, as part of the LIFE Program being performed at Lawrence Livermore National laboratory

  16. A conceptual design of a beam-shaping assembly for boron neutron capture therapy based on deuterium-tritium neutron generators.

    PubMed

    Martín, Guido; Abrahantes, Arian

    2004-05-01

    A conceptual design of a beam-shaping assembly for boron neutron capture therapy using deuterium-tritium accelerator based neutrons source is developed. Calculations based on a simple geometry model for the radiation transport are initially performed to estimate the assembly materials and their linear dimensions. Afterward, the assembly geometry is produced, optimized and verified. In order to perform these calculations the general-purpose MCNP code is used. Irradiation time and therapeutic gain are utilized as beam assessment parameters. Metallic uranium and manganese are successfully tested for fast-to-epithermal neutron moderation. In the present beam-shaping assembly proposal, the therapeutic gain is improved by 23% and the accelerator current required for a fixed irradiation period is reduced by six times compared to previous proposals based on the same D-T reaction.

  17. A multi-group Monte Carlo core analysis method and its application in SCWR design

    SciTech Connect

    Zhang, P.; Wang, K.; Yu, G.

    2012-07-01

    Complex geometry and spectrum have been the characteristics of many newly developed nuclear energy systems, so the suitability and precision of the traditional deterministic codes are doubtable while being applied to simulate these systems. On the contrary, the Monte Carlo method has the inherent advantages of dealing with complex geometry and spectrum. The main disadvantage of Monte Carlo method is that it takes long time to get reliable results, so the efficiency is too low for the ordinary core designs. A new Monte Carlo core analysis scheme is developed, aimed to increase the calculation efficiency. It is finished in two steps: Firstly, the assembly level simulation is performed by continuous energy Monte Carlo method, which is suitable for any geometry and spectrum configuration, and the assembly multi-group constants are tallied at the same time; Secondly, the core level calculation is performed by multi-group Monte Carlo method, using the assembly group constants generated in the first step. Compared with the heterogeneous Monte Carlo calculations of the whole core, this two-step scheme is more efficient, and the precision is acceptable for the preliminary analysis of novel nuclear systems. Using this core analysis scheme, a SCWR core was designed based on a new SCWR assembly design. The core output is about 1,100 MWe, and a cycle length of about 550 EFPDs can be achieved with 3-batch refueling pattern. The average and maximum discharge burn-up are about 53.5 and 60.9 MWD/kgU respectively. (authors)

  18. Use of Solid Hydride Fuel for Improved long-Life LWR Core Designs

    SciTech Connect

    Greenspan, E

    2006-04-30

    The primary objective of this project was to assess the feasibility of improving the performance of PWR and BWR cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the bus-bar cost of electricity (COE). Additional performance measures considered are safety, fuel bundle design simplicity – in particular for BWR’s, and plutonium incineration capability. It was found that hydride fuel can safely operate in PWR’s and BWR’s without restricting the linear heat generation rate of these reactors relative to that attainable with oxide fuel. A couple of promising applications of hydride fuel in PWR’s and BWR’s were identified: (1) Eliminating dedicated water moderator volumes in BWR cores thus enabling to significantly increase the cooled fuel rods surface area as well as the coolant flow cross section area in a given volume fuel bundle while significantly reducing the heterogeneity of BWR fuel bundles thus achieving flatter pin-by-pin power distribution. The net result is a possibility to significantly increase the core power density – on the order of 30% and, possibly, more, while greatly simplifying the fuel bundle design. Implementation of the above modifications is, though, not straightforward; it requires a design of completely different control system that could probably be implemented only in newly designed plants. It also requires increasing the coolant pressure drop across the core. (2) Recycling plutonium in PWR’s more effectively than is possible with oxide fuel by virtue of a couple of unique features of hydride fuel – reduced inventory of U-238 and increased inventory of hydrogen. As a result, the hydride fuelled core achieves nearly double the average discharge burnup and the fraction of the loaded Pu it incinerates in one pass is double that of the MOX fuel. The fissile fraction of the Pu in the discharged hydride fuel is only ~2/3 that of the MOX fuel and the

  19. Design of a Neutron Temporal Diagnostic for measuring DD or DT burn histories at the NIF

    NASA Astrophysics Data System (ADS)

    Lahmann, B.; Frenje, J. A.; Sio, H.; Petrasso, R. D.; Bradley, D. K.; Le Pape, S.; MacKinnon, A. J.; Isumi, N.; Macphee, A.; Zayas, C.; Spears, B. K.; Hermann, H.; Hilsabeck, T. J.; Kilkenny, J. D.

    2015-11-01

    The DD or DT burn history in Inertial Confinement Fusion (ICF) implosions provides essential information about implosion performance and helps to constrain numerical modeling. The capability of measuring this burn history is thus important for the NIF in its pursuit of ignition. Currently, the Gamma Reaction History (GRH) diagnostic is the only system capable of measuring the burn history for DT implosions with yields greater than ~ 1e14. To complement GRH, a new NIF Neutron Temporal Diagnostic (NTD) is being designed for measuring the DD or DT burn history with yields greater than ~ 1e10. A traditional scintillator-based design and a pulse-dilation-based design are being considered. Using MCNPX simulations, both designs have been optimized, validated and contrasted for various types of implosions at the NIF. This work was supported in part by the U.S. DOE, LLNL and LLE.

  20. Soft error rate simulation and initial design considerations of neutron intercepting silicon chip (NISC)

    NASA Astrophysics Data System (ADS)

    Celik, Cihangir

    -scale technologies. Prevention of SEEs has been studied and applied in the semiconductor industry by including radiation protection precautions in the system architecture or by using corrective algorithms in the system operation. Decreasing 10B content (20%of natural boron) in the natural boron of Borophosphosilicate glass (BPSG) layers that are conventionally used in the fabrication of semiconductor devices was one of the major radiation protection approaches for the system architecture. Neutron interaction in the BPSG layer was the origin of the SEEs because of the 10B (n,alpha) 7Li reaction products. Both of the particles produced have the capability of ionization in the silicon substrate region, whose thickness is comparable to the ranges of these particles. Using the soft error phenomenon in exactly the opposite manner of the semiconductor industry can provide a new neutron detection system based on the SERs in the semiconductor memories. By investigating the soft error mechanisms in the available semiconductor memories and enhancing the soft error occurrences in these devices, one can convert all memory using intelligent systems into portable, power efficient, directiondependent neutron detectors. The Neutron Intercepting Silicon Chip (NISC) project aims to achieve this goal by introducing 10B-enriched BPSG layers to the semiconductor memory architectures. This research addresses the development of a simulation tool, the NISC Soft Error Analysis Tool (NISCSAT), for soft error modeling and analysis in the semiconductor memories to provide basic design considerations for the NISC. NISCSAT performs particle transport and calculates the soft error probabilities, or SER, depending on energy depositions of the particles in a given memory node model of the NISC. Soft error measurements were performed with commercially available, off-the-shelf semiconductor memories and microprocessors to observe soft error variations with the neutron flux and memory supply voltage. Measurement

  1. Reliability Design for Neutron Induced Single-Event Burnout of IGBT

    NASA Astrophysics Data System (ADS)

    Shoji, Tomoyuki; Nishida, Shuichi; Ohnishi, Toyokazu; Fujikawa, Touma; Nose, Noboru; Hamada, Kimimori; Ishiko, Masayasu

    Single-event burnout (SEB) caused by cosmic ray neutrons leads to catastrophic failures in insulated gate bipolar transistors (IGBTs). It was found experimentally that the incident neutron induced SEB failure rate increases as a function of the applied collector voltage. Moreover, the failure rate increased sharply with an increase in the applied collector voltage when the voltage exceeded a certain threshold value (SEB cutoff voltage). In this paper, transient device simulation results indicate that impact ionization at the n-drift/n+ buffer boundary is a crucially important factor in the turning-on of the parasitic pnp transistor, and eventually latch-up of the parasitic thyristor causes SEB. In addition, the device parameter dependency of the SEB cutoff voltage was analytically derived from the latch-up condition of the parasitic thyristor. As a result, it was confirmed that reducing the current gain of the parasitic transistor, such as by increasing the n-drift region thickness d was effective in increasing the SEB cutoff voltage. Furthermore, `white' neutron-irradiation experiments demonstrated that suppressing the inherent parasitic thyristor action leads to an improvement of the SEB cutoff voltage. It was confirmed that current gain optimization of the parasitic transistor is a crucial factor for establishing highly reliable design against chance failures.

  2. ASPUN: design for an Argonne super-intense pulsed neutron source

    SciTech Connect

    Khoe, T.K.; Kustom, R.L.

    1983-01-01

    Argonne pioneered the pulsed spallation neutron source with the ZING-P and IPNS-I concepts. IPNS-I is now a reliable and actively used source for pulsed spallation neutrons. The accelerator is a 500-MeV, 8 to 9 ..mu..a, 30-Hz rapid-cycling proton synchrotron. Other proton spallation sources are now in operation or in construction. These include KENS-I at the National Laboratory for High Energy Physics in Japan, the WNR/PSR at Los Alamos National Laboratory in the USA, and the SNS at the Rutherford Appleton Laboratory in England. Newer and bolder concepts are being developed for more-intense pulsed spallation neutron sources. These include SNQ at the KFA Laboratory in Juelich, Germany, ASTOR at the Swiss Institute for Nuclear Physics in Switzerland, and ASPUN, the Argonne concept. ASPUN is based on the Fixed-Field Alternating Gradient concept. The design goal is to provide a time-averaged beam of 3.5 ma at 1100 MeV on a spallation target in intense bursts, 100 to 200 nanoseconds long, at a repetition rate of no more than 60 to 85 Hz.

  3. Light source design using Kagome-lattice hollow core photonic crystal fibers

    NASA Astrophysics Data System (ADS)

    Hossain, Md. Anwar; Namihira, Yoshinori

    2014-09-01

    Supercontinuum (SC) light source is designed using high pressure Xe-filled hollow core Kagome-lattice photonic crystal fiber. Using finite element method with perfectly matched layer, SC spectra in normal chromatic dispersion region have been generated using picosecond optical pulses from relatively less expensive laser sources.

  4. Narrative Plus: Designing and Implementing the Common Core State Standards with the Gift Essay

    ERIC Educational Resources Information Center

    Chandler-Olcott, Kelly; Zeleznik, John

    2013-01-01

    The authors of this article describe their inquiry into implementation of the writing-focused Common Core State Standards in a co-taught English 9 class in an urban school. They describe instructional moves designed to increase student success with an assignment called the Gift Essay, with particular focus on planning and other organizational…

  5. Spring design for use in the core of a nuclear reactor

    DOEpatents

    Willard, Jr., H. James

    1993-01-01

    A spring design particularly suitable for use in the core of a nuclear reactor includes one surface having a first material oriented in a longitudinal direction, and another surface having a second material oriented in a transverse direction. The respective surfaces exhibit different amounts of irraditation induced strain.

  6. Laser inertial fusion-based energy: Neutronic design aspects of a hybrid fusion-fission nuclear energy system

    NASA Astrophysics Data System (ADS)

    Kramer, Kevin James

    This study investigates the neutronics design aspects of a hybrid fusion-fission energy system called the Laser Fusion-Fission Hybrid (LFFH). A LFFH combines current Laser Inertial Confinement fusion technology with that of advanced fission reactor technology to produce a system that eliminates many of the negative aspects of pure fusion or pure fission systems. When examining the LFFH energy mission, a significant portion of the United States and world energy production could be supplied by LFFH plants. The LFFH engine described utilizes a central fusion chamber surrounded by multiple layers of multiplying and moderating media. These layers, or blankets, include coolant plenums, a beryllium (Be) multiplier layer, a fertile fission blanket and a graphite-pebble reflector. Each layer is separated by perforated oxide dispersion strengthened (ODS) ferritic steel walls. The central fusion chamber is surrounded by an ODS ferritic steel first wall. The first wall is coated with 250-500 mum of tungsten to mitigate x-ray damage. The first wall is cooled by Li17Pb83 eutectic, chosen for its neutron multiplication and good heat transfer properties. The Li17Pb 83 flows in a jacket around the first wall to an extraction plenum. The main coolant injection plenum is immediately behind the Li17Pb83, separated from the Li17Pb83 by a solid ODS wall. This main system coolant is the molten salt flibe (2LiF-BeF2), chosen for beneficial neutronics and heat transfer properties. The use of flibe enables both fusion fuel production (tritium) and neutron moderation and multiplication for the fission blanket. A Be pebble (1 cm diameter) multiplier layer surrounds the coolant injection plenum and the coolant flows radially through perforated walls across the bed. Outside the Be layer, a fission fuel layer comprised of depleted uranium contained in Tristructural-isotropic (TRISO) fuel particles having a packing fraction of 20% in 2 cm diameter fuel pebbles. The fission blanket is cooled by

  7. Fusion-power-core design of a Compact Reversed-Field Pinch Reactor (CRFPR)

    NASA Astrophysics Data System (ADS)

    Copenhaver, C.; Schnurr, N. M.; Krakowski, R. A.; Hagenson, R. L.; Mynard, R. C.; Cappiello, C.; Lujan, R. E.; Davidson, J. W.; Chaffee, A. D.; Battat, M. E.

    A conceptual design of a fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, coils) based on a Reversed-Field Pinch (RFP) has been completed. After a brief statement of rationale and description of the reactor configuraton, the FPC integration is described in terms of power balance, thermal-hydraulics, and mechanical design. The engineering versatility, promise, and problems of this high-power-density approach to fusion are addressed.

  8. Methodology for worker neutron exposure evaluation in the PDCF facility design.

    PubMed

    Scherpelz, R I; Traub, R J; Pryor, K H

    2004-01-01

    A project headed by Washington Group International is meant to design the Pit Disassembly and Conversion Facility (PDCF) to convert the plutonium pits from excessed nuclear weapons into plutonium oxide for ultimate disposition. Battelle staff are performing the shielding calculations that will determine appropriate shielding so that the facility workers will not exceed target exposure levels. The target exposure levels for workers in the facility are 5 mSv y(-1) for the whole body and 100 mSv y(-1) for the extremity, which presents a significant challenge to the designers of a facility that will process tons of radioactive material. The design effort depended on shielding calculations to determine appropriate thickness and composition for glove box walls, and concrete wall thicknesses for storage vaults. Pacific Northwest National Laboratory (PNNL) staff used ORIGEN-S and SOURCES to generate gamma and neutron source terms, and Monte Carlo (computer code for) neutron photon (transport) (MCNP-4C) to calculate the radiation transport in the facility. The shielding calculations were performed by a team of four scientists, so it was necessary to develop a consistent methodology. There was also a requirement for the study to be cost-effective, so efficient methods of evaluation were required. The calculations were subject to rigorous scrutiny by internal and external reviewers, so acceptability was a major feature of the methodology. Some of the issues addressed in the development of the methodology included selecting appropriate dose factors, developing a method for handling extremity doses, adopting an efficient method for evaluating effective dose equivalent in a non-uniform radiation field, modelling the reinforcing steel in concrete, and modularising the geometry descriptions for efficiency. The relative importance of the neutron dose equivalent compared with the gamma dose equivalent varied substantially depending on the specific shielding conditions and lessons

  9. METHODOLOGY FOR WORKER NEUTRON EXPOSURE EVALUATION IN THE PDCF FACILITY DESIGN

    SciTech Connect

    Scherpelz, Robert I.; Traub, Richard J.; Pryor, Kathryn H.

    2004-08-01

    A project headed by Washington Group International is meant to design the Pit Disassembly and Conversion Facility (PDCF) to convert the plutonium pits from excessed nuclear weapons into plutonium oxide for ultimate disposition. Battelle staff are performing the shielding calculations that will determine appropriate shielding so that the facility workers will not exceed target exposure levels. The target exposure levels for workers in the facility are 5 mSv y?1 for the whole body and 100 mSv y?1 for the extremity, which presents a significant challenge to the designers of a facility that will process tons of radioactive material. The design effort depended on shielding calculations to determine appropriate thickness and composition for glove box walls, and concrete wall thicknesses for storage vaults. Pacific Northwest National Laboratory (PNNL) staff used ORIGEN-S and SOURCES to generate gamma and neutron source terms, and Monte Carlo (computer code for) neutron photon (transport) (MCNP-4C) to calculate the radiation transport in the facility. The shielding calculations were performed by a team of four scientists, so it was necessary to develop a consistent methodology. There was also a requirement for the study to be cost-effective, so efficient methods of evaluation were required. The calculations were subject to rigorous scrutiny by internal and external reviewers, so acceptability was a major feature of the methodology. Some of the issues addressed in the development of the methodology included selecting appropriate dose factors, developing a method for handling extremity doses, adopting an efficient method for evaluating effective dose equivalent in a non-uniform radiation field, modeling the reinforcing steel in concrete, and modularizing the geometry descriptions for efficiency. The relative importance of the neutron dose equivalent compared with the gamma dose equivalent varied substantially depending on the specific shielding conditions and lessons were

  10. Methodology for worker neutron exposure evaluation in the PDCF facility design.

    PubMed

    Scherpelz, R I; Traub, R J; Pryor, K H

    2004-01-01

    A project headed by Washington Group International is meant to design the Pit Disassembly and Conversion Facility (PDCF) to convert the plutonium pits from excessed nuclear weapons into plutonium oxide for ultimate disposition. Battelle staff are performing the shielding calculations that will determine appropriate shielding so that the facility workers will not exceed target exposure levels. The target exposure levels for workers in the facility are 5 mSv y(-1) for the whole body and 100 mSv y(-1) for the extremity, which presents a significant challenge to the designers of a facility that will process tons of radioactive material. The design effort depended on shielding calculations to determine appropriate thickness and composition for glove box walls, and concrete wall thicknesses for storage vaults. Pacific Northwest National Laboratory (PNNL) staff used ORIGEN-S and SOURCES to generate gamma and neutron source terms, and Monte Carlo (computer code for) neutron photon (transport) (MCNP-4C) to calculate the radiation transport in the facility. The shielding calculations were performed by a team of four scientists, so it was necessary to develop a consistent methodology. There was also a requirement for the study to be cost-effective, so efficient methods of evaluation were required. The calculations were subject to rigorous scrutiny by internal and external reviewers, so acceptability was a major feature of the methodology. Some of the issues addressed in the development of the methodology included selecting appropriate dose factors, developing a method for handling extremity doses, adopting an efficient method for evaluating effective dose equivalent in a non-uniform radiation field, modelling the reinforcing steel in concrete, and modularising the geometry descriptions for efficiency. The relative importance of the neutron dose equivalent compared with the gamma dose equivalent varied substantially depending on the specific shielding conditions and lessons

  11. The Design and Construction of a Cold Neutron Source for Use in the Cornell University Triga Reactor

    NASA Astrophysics Data System (ADS)

    Young, Lydia Jane

    A cold neutron source has been designed and constructed for insertion into the 6"-radial beam port of the Cornell University TRIGA reactor for use with a neutron guide tube system. The main differences between this cold source and other existing sources are the use of heat conduction as the method of cooling and the use of mesitylene (1,3,5 -trimethylbenzene; melting point, 228(DEGREES)K; boiling point, 437(DEGREES)K) as the moderating material. This thesis describes the design and construction details of the cold neutron source, discusses its safety aspects, and presents its cryogenic performance curves and also the results of a test of its neutron moderating ability. A closed-cycle helium gas refrigerator, located outside the reactor shielding, cools the 500 cm('3) moderator chamber and its surrounding heat shield by heat conduction through two meters of copper and rod tubing. Moderator temperatures of 23 (+OR-) 3(DEGREES)K have been achieved. Mesitylene, a hydrocarbon, is an effective cold moderator because even at low temperatures the weakly hindered rotational motions of its methyl groups enable the absorption of small amounts of energy ((LESSTHEQ) 0.005 eV) from neutrons. The use of mesitylene simplifies the cold source design because it is a liquid at room temperature and thus, the usual design safeguards required for sources using gaseous moderators are not necessary. Moreover, the flammability of mesitylene is much smaller than that of hydrogen and methane, which are the commonly used cold moderators. A method of transferring and handling the mesitylene, a carcinogen, was devised to ensure minimal contact with this substance. To test the neutron moderating ability of the cold neutron source, an out-of-reactor neutron transmission experiment was performed with the moderator chamber first at room temperature and then at about 23(DEGREES)K. The results indicate that the neutron energy spectrum is strongly shifted to lower energies when the chamber is cold

  12. Thermal-hydraulic criteria for the APT tungsten neutron source design

    SciTech Connect

    Pasamehmetoglu, K.

    1998-03-01

    This report presents the thermal-hydraulic design criteria (THDC) developed for the tungsten neutron source (TNS). The THDC are developed for the normal operations, operational transients, and design-basis accidents. The requirements of the safety analyses are incorporated into the design criteria, consistent with the integrated safety management and the safety-by-design philosophy implemented throughout the APT design process. The phenomenology limiting the thermal-hydraulic design and the confidence level requirements for each limit are discussed. The overall philosophy of the uncertainty analyses and the confidence level requirements also are presented. Different sets of criteria are developed for normal operations, operational transients, anticipated accidents, unlikely accidents, extremely unlikely accidents, and accidents during TNS replacement. In general, the philosophy is to use the strictest criteria for the high-frequency events. The criteria is relaxed as the event frequencies become smaller. The THDC must be considered as a guide for the design philosophy and not as a hard limit. When achievable, design margins greater than those required by the THDC must be used. However, if a specific event sequence cannot meet the THDC, expensive design changes are not necessary if the single event sequence results in sufficient margin to safety criteria and does not challenge the plant availability or investment protection considerations.

  13. Design and performance of a new high accuracy combined small sample neutron/gamma detector

    SciTech Connect

    Menlove, H.; Davidson, D.; Verplancke, J.; Vermeulen, P.; Wagner, H.G.; Wellum, R.; Brandelise, B.; Mayer, K.

    1993-08-01

    This paper describes the design of an optimized combined neutron and gamma detector installed around a measurement well protruding from the floor of a glove box. The objective of this design was to achieve an overall accuracy for the plutonium element concentration in gram-sized samples of plutonium oxide powder approaching the {approximately}0.1--0.2% accuracies routinely achieved by inspectors` chemical analysis. The efficiency of the clam-shell neutron detector was increased and the flat response zone extended in axial and radial directions. The sample holder introduced from within the glove box was designed to form the upper reflector, while two graphite half-shells fitted around the thin neck of the high-resolution LEGE detector replaced the lower plug. The Institute for Reference Materials and Measurements (IRMM) in Geel prepared special plutonium oxide test samples whose plutonium concentration was determined to better than 0.05%. During a three week initial performance test in July 1992 at ITU Karlsruhe and in long term tests, it was established that the target accuracy can be achieved provided sufficient care is taken to assure the reproducibility of sample bottling and sample positioning. The paper presents and discusses the results of all test measurements.

  14. Design and performance of a new high accuracy combined small sample neutron/gamma detector

    SciTech Connect

    Menlove, H.; Davidson, D.; Verplancke, J.; Vermeulen, P.; Wagner, H.G.; Wellum, R.; Brandelise, B.; Mayer, K.

    1993-12-31

    This paper describes the design of an optimized combined neutron and gamma detector installed around a measurement well protruding from the floor of a glove box. The objective of this design was to achieve an overall accuracy for the plutonium element concentration in gram-sized samples of plutonium oxide powder approaching the {approximately}0.1--0.2% accuracies routinely achieved by inspectors` chemical analysis. The efficiency of the clam-shell neutron detector was increased and the flat response zone extended in axial and radial directions. The sample holder introduced from within the glove box was designed to form the upper reflector, while two graphite half-shells fitted around the thin neck of the high-resolution LEGe detector replaced the lower plug. The Institute for Reference Materials and Measurements (IRMM) in Geel prepared special plutonium oxide test samples whose plutonium concentration was determined to better than 0.05%. During a three week initial performance test in July 1992 at ITU Karlsruhe and in long term tests, it was established that the target accuracy can be achieved provided sufficient care is taken to assure the reproducibility of sample bottling and sample positioning. The paper presents and discusses the results of all test measurements.

  15. Systematic approach for designing zero-DGD coupled multi-core optical fibers.

    PubMed

    Parto, Midya; Eftekhar, Mohammad Amin; Miri, Mohammad-Ali; Amezcua-Correa, Rodrigo; Li, Guifang; Christodoulides, Demetrios N

    2016-05-01

    An analytical method is presented for designing N-coupled multi-core fibers with zero differential group delay. This approach effectively reduces the problem to a system of N-1 algebraic equations involving the associated coupling coefficients and propagation constants, as obtained from coupled mode theory. Once the parameters of one of the cores are specified, the roots of the resulting N-1 equations can be used to determine the characteristics of the remaining waveguide elements. Using this technique, a number of pertinent geometrical configurations are investigated to minimize intermodal dispersion.

  16. High-Order Homogenization Method in Diffusion Theory for Reactor Core Simulation and Design Calculation

    SciTech Connect

    Farzad Rahnema

    2003-09-30

    Most modern nodal methods in use by the reactor vendors and utilities are based on the generalized equivalence theory (GET) that uses homogenized cross sections and flux discontinuity factors. These homogenized parameters, referred to as infinite medium parameters, are precomputed by performing single bundle fine-mesh calculations with zero current boundary conditions. It is known that for configurations in which the node-to-node leakage (e.g., surface current-to-flux ratio) is large the use of the infinite medium parameters could lead to large errors in the nodal solution. This would be the case for highly heterogeneous core configurations, typical of modern reactor core designs.

  17. Klystron Modulator Design for the Los Alamos Neutron Science Center Accelerator

    SciTech Connect

    Reass, William A.; Baca, David M.; Partridge, Edward R.; Rees, Daniel E.

    2012-06-22

    This paper will describe the design of the 44 modulator systems that will be installed to upgrade the Los Alamos Neutron Science Center (LANSCE) accelerator RF system. The klystrons can operate up to 86 kV with a nominal 32 Amp beam current with a 120 Hz repetition rate and 15% duty cycle. The klystrons are a mod-anode design. The modulator is designed with analog feedback control to ensure the klystron beam current is flat-top regulated. To achieve fast switching while maintaining linear feedback control, a grid-clamp, totem-pole modulator configuration is used with an 'on' deck and an 'off' deck. The on and off deck modulators are of identical design and utilize a cascode connected planar triode, cathode driven with a high speed MOSFET. The derived feedback is connected to the planar triode grid to enable the flat-top control. Although modern design approaches suggest solid state designs may be considered, the planar triode (Eimac Y-847B) is very cost effective, is easy to integrate with the existing hardware, and provides a simplified linear feedback control mechanism. The design is very compact and fault tolerant. This paper will review the complete electrical design, operational performance, and system characterization as applied to the LANSCE installation.

  18. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    SciTech Connect

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  19. Hollow Core Bragg Waveguide Design and Fabrication for Enhanced Raman Spectroscopy

    NASA Astrophysics Data System (ADS)

    Ramanan, Janahan

    Raman spectroscopy is a widely used technique to unambiguously ascertain the chemical composition of a sample. The caveat with this technique is its extremely weak optical cross-section, making it difficult to measure Raman signal with standard optical setups. In this thesis, a novel hollow core Bragg Reflection Waveguide was designed to simultaneously increase the generation and collection of Raman scattered photons. A robust fabrication process of this waveguide was developed employing flip-chip bonding methods to securely seal the hollow core channel. The waveguide air-core propagation loss was experimentally measured to be 0.17 dB/cm, and the Raman sensitivity limit was measured to be 3 mmol/L for glycerol solution. The waveguide was also shown to enhance Raman modes of standard household aerosols that could not be seen with other devices.

  20. MCNP5 and GEANT4 comparisons for preliminary Fast Neutron Pencil Beam design at the University of Utah TRIGA system

    NASA Astrophysics Data System (ADS)

    Adjei, Christian Amevi

    The main objective of this thesis is twofold. The starting objective was to develop a model for meaningful benchmarking of different versions of GEANT4 against an experimental set-up and MCNP5 pertaining to photon transport and interactions. The following objective was to develop a preliminary design of a Fast Neutron Pencil Beam (FNPB) Facility to be applicable for the University of Utah research reactor (UUTR) using MCNP5 and GEANT4. The three various GEANT4 code versions, GEANT4.9.4, GEANT4.9.3, and GEANT4.9.2, were compared to MCNP5 and the experimental measurements of gamma attenuation in air. The average gamma dose rate was measured in the laboratory experiment at various distances from a shielded cesium source using a Ludlum model 19 portable NaI detector. As it was expected, the gamma dose rate decreased with distance. All three GEANT4 code versions agreed well with both the experimental data and the MCNP5 simulation. Additionally, a simple GEANT4 and MCNP5 model was developed to compare the code agreements for neutron interactions in various materials. Preliminary FNPB design was developed using MCNP5; a semi-accurate model was developed using GEANT4 (because GEANT4 does not support the reactor physics modeling, the reactor was represented as a surface neutron source, thus a semi-accurate model). Based on the MCNP5 model, the fast neutron flux in a sample holder of the FNPB is obtained to be 6.52×107 n/cm2s, which is one order of magnitude lower than gigantic fast neutron pencil beam facilities existing elsewhere. The MCNP5 model-based neutron spectrum indicates that the maximum expected fast neutron flux is at a neutron energy of ~1 MeV. In addition, the MCNP5 model provided information on gamma flux to be expected in this preliminary FNPB design; specifically, in the sample holder, the gamma flux is to be expected to be around 108 γ/cm 2s, delivering a gamma dose of 4.54×103 rem/hr. This value is one to two orders of magnitudes below the gamma

  1. Transmission and signal loss in mask designs for a dual neutron and gamma imager applied to mobile standoff detection

    NASA Astrophysics Data System (ADS)

    Ayaz-Maierhafer, Birsen; Hayward, Jason P.; Ziock, Klaus P.; Blackston, Matthew A.; Fabris, Lorenzo

    2013-06-01

    In order to design a next-generation, dual neutron and gamma imager for mobile standoff detection which uses coded aperture imaging as its primary detection modality, the following design parameters have been investigated for gamma and neutron radiation incident upon a hybrid, coded mask: (1) transmission through mask elements for various mask materials and thicknesses; and (2) signal attenuation in the mask versus angle of incidence. Each of these parameters directly affects detection significance, as quantified by the signal-to-noise ratio. The hybrid mask consists of two or three layers: organic material for fast neutron attenuation and scattering, Cd for slow neutron absorption (if applied), and one of three of the following photon or photon and slow neutron attenuating materials—Linotype alloy, CLYC, or CZT. In the MCNP model, a line source of gamma rays (100-2500 keV), fast neutrons (1000-10,000 keV) or thermal neutrons was positioned above the hybrid mask. The radiation penetrating the mask was simply tallied at the surface of an ideal detector, which was located below the surface of the last mask layer. The transmission was calculated as the ratio of the particles transmitted through the fixed aperture to the particles passing through the closed mask. In order to determine the performance of the mask considering relative motion between the source and detector, simulations were used to calculate the signal attenuation for incident radiation angles of 0-50°. The results showed that a hybrid mask can be designed to sufficiently reduce both transmission through the mask and signal loss at large angles of incidence, considering both gamma ray and fast neutron radiations. With properly selected material thicknesses, the signal loss of a hybrid mask, which is necessarily thicker than the mask required for either single mode imaging, is not a setback to the system's detection significance.

  2. NEUTRONIC REACTOR

    DOEpatents

    Wade, E.J.

    1958-09-16

    This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

  3. Evaluating secondary neutron doses of a refined shielded design for a medical cyclotron using the TLD approach

    NASA Astrophysics Data System (ADS)

    Lin, Jye-Bin; Tseng, Hsien-Chun; Liu, Wen-Shan; Lin, Ding-Bang; Hsieh, Teng-San; Chen, Chien-Yi

    2013-11-01

    An increasing number of cyclotrons at medical centers in Taiwan have been installed to generate radiopharmaceutical products. An operating cyclotron generates immense amounts of secondary neutrons from reactions such the 18O(p, n)18F, used in the production of FDG. This intense radiation can be hazardous to public health, particularly to medical personnel. To increase the yield of 18F-FDG from 4200 GBq in 2005 to 48,600 GBq in 2011, Chung Shan Medical University Hospital (CSMUH) has prolonged irradiation time without changing the target or target current to meet requirements regarding the production 18F. The CSMUH has redesigned the CTI Radioisotope Delivery System shield. The lack of data for a possible secondary neutron doses has increased due to newly designed cyclotron rooms. This work aims to evaluate secondary neutron doses at a CTI cyclotron center using a thermoluminescent dosimeter (TLD-600). Two-dimensional neutron doses were mapped and indicated that neutron doses were high as neutrons leaked through self-shielded blocks and through the L-shaped concrete shield in vault rooms. These neutron doses varied markedly among locations close to the H218O target. The Monte Carlo simulation and minimum detectable dose are also discussed and demonstrated the reliability of using the TLD-600 approach. Findings can be adopted by medical centers to identify radioactive hot spots and develop radiation protection.

  4. Advanced Neutron Source (ANS) Project progress report

    SciTech Connect

    McBee, M.R.; Chance, C.M. ); Selby, D.L.; Harrington, R.M.; Peretz, F.J. )

    1990-04-01

    This report discusses the following topics on the advanced neutron source: quality assurance (QA) program; reactor core development; fuel element specification; corrosion loop tests and analyses; thermal-hydraulic loop tests; reactor control concepts; critical and subcritical experiments; material data, structural tests, and analysis; cold source development; beam tube, guide, and instrument development; hot source development; neutron transport and shielding; I C research and development; facility concepts; design; and safety.

  5. Homogeneous immunoconjugates for boron neutron-capture therapy: design, synthesis, and preliminary characterization.

    PubMed

    Guan, L; Wims, L A; Kane, R R; Smuckler, M B; Morrison, S L; Hawthorne, M F

    1998-10-27

    The application of immunoprotein-based targeting strategies to the boron neutron-capture therapy of cancer poses an exceptional challenge, because viable boron neutron-capture therapy by this method will require the efficient delivery of 10(3) boron-10 atoms by each antigen-binding protein. Our recent investigations in this area have been focused on the development of efficient methods for the assembly of homogeneous immunoprotein conjugates containing the requisite boron load. In this regard, engineered immunoproteins fitted with unique, exposed cysteine residues provide attractive vehicles for site-specific modification. Additionally, homogeneous oligomeric boron-rich phosphodiesters (oligophosphates) have been identified as promising conjugation reagents. The coupling of two such boron-rich oligophosphates to sulfhydryls introduced to the CH2 domain of a chimeric IgG3 has been demonstrated. The resulting boron-rich immunoconjugates are formed efficiently, are readily purified, and have promising in vitro and in vivo characteristics. Encouragingly, these studies showed subtle differences in the properties of the conjugates derived from the two oligophosphate molecules studied, providing a basis for the application of rational design to future work. Such subtle details would not have been as readily discernible in heterogeneous conjugates, thus validating the rigorous experimental design employed here.

  6. Homogeneous immunoconjugates for boron neutron-capture therapy: Design, synthesis, and preliminary characterization

    PubMed Central

    Guan, Lufeng; Wims, Letitia A.; Kane, Robert R.; Smuckler, Mark B.; Morrison, Sherie L.; Hawthorne, M. Frederick

    1998-01-01

    The application of immunoprotein-based targeting strategies to the boron neutron-capture therapy of cancer poses an exceptional challenge, because viable boron neutron-capture therapy by this method will require the efficient delivery of 103 boron-10 atoms by each antigen-binding protein. Our recent investigations in this area have been focused on the development of efficient methods for the assembly of homogeneous immunoprotein conjugates containing the requisite boron load. In this regard, engineered immunoproteins fitted with unique, exposed cysteine residues provide attractive vehicles for site-specific modification. Additionally, homogeneous oligomeric boron-rich phosphodiesters (oligophosphates) have been identified as promising conjugation reagents. The coupling of two such boron-rich oligophosphates to sulfhydryls introduced to the CH2 domain of a chimeric IgG3 has been demonstrated. The resulting boron-rich immunoconjugates are formed efficiently, are readily purified, and have promising in vitro and in vivo characteristics. Encouragingly, these studies showed subtle differences in the properties of the conjugates derived from the two oligophosphate molecules studied, providing a basis for the application of rational design to future work. Such subtle details would not have been as readily discernible in heterogeneous conjugates, thus validating the rigorous experimental design employed here. PMID:9789066

  7. Design and analysis of large-core multitrench channel waveguide for high-power applications.

    PubMed

    Saini, Than Singh; Kumar, Ajeet; Sinha, Ravindra Kumar

    2015-07-01

    We present a multitrench leaky channel waveguide design that supports effective single-mode operation even with large-core size. The proposed waveguide structure has a uniform rectangular core and a geometrically shaped trench-assisted cladding in such a way that all the confined modes become leaky. The effective single-mode operation is achieved by choosing the appropriate geometrical parameters that introduce very large leakage losses for higher-order modes with very low leakage losses for fundamental mode. A power law profile for cladding geometry is considered to explore the effect of trenches on the effective single-mode operation. The finite element method is used to calculate the leakage losses of the modes. Our numerical results show that the waveguide ensures extended single-mode operation in the wavelength range of 1.25-2.0 μm with the rectangular core area as large as 100  μm2. Such a large-core-area waveguide structure efficiently suppresses unwanted nonlinear optical effects. A proposed channel waveguide structure with a large core size is suitable for high-power delivery devices such as high-power waveguide lasers and amplifiers.

  8. Insert Design and Manufacturing for Foam-Core Composite Sandwich Structures

    NASA Astrophysics Data System (ADS)

    Lares, Alan

    Sandwich structures have been used in the aerospace industry for many years. The high strength to weight ratios that are possible with sandwich constructions makes them desirable for airframe applications. While sandwich structures are effective at handling distributed loads such as aerodynamic forces, they are prone to damage from concentrated loads at joints or due to impact. This is due to the relatively thin face-sheets and soft core materials typically found in sandwich structures. Carleton University's Uninhabited Aerial Vehicle (UAV) Project Team has designed and manufactured a UAV (GeoSury II Prototype) which features an all composite sandwich structure fuselage structure. The purpose of the aircraft is to conduct geomagnetic surveys. The GeoSury II Prototype serves as the test bed for many areas of research in advancing UAV technologies. Those areas of research include: low cost composite materials manufacturing, geomagnetic data acquisition, obstacle detection, autonomous operations and magnetic signature control. In this thesis work a methodology for designing and manufacturing inserts for foam-core sandwich structures was developed. The results of this research work enables a designer wishing to design a foam-core sandwich airframe structure, a means of quickly manufacturing optimized inserts for the safe introduction of discrete loads into the airframe. The previous GeoSury II Prototype insert designs (v.1 & v.2) were performance tested to establish a benchmark with which to compare future insert designs. Several designs and materials were considered for the new v.3 inserts. A plug and sleeve design was selected, due to its ability to effectively transfer the required loads to the sandwich structure. The insert material was chosen to be epoxy, reinforced with chopped carbon fibre. This material was chosen for its combination of strength, low mass and also compatibility with the face-sheet material. The v.3 insert assembly is 60% lighter than the

  9. Performance of truss panels with kagome cores and design of a high authority shape morphing structure

    NASA Astrophysics Data System (ADS)

    Wang, Ju

    This dissertation includes two parts: First, the performance of a light weight truss panels with Kagome cores; Second, design of a high authority morphing structure for hinging and twisting. The performance characteristics of a truss core sandwich panel design based on the 3D Kagome are measured and compared with earlier numerical simulations and the consistency is demonstrated. Panels are fabricated by investment casting and tested in compression, shear and 3-point bending. The isotropic nature of this core design is confirmed. The superior performance relative to truss designs based on the tetrahedron is demonstrated and attributed to the greater resistance to plastic buckling at the equivalent core density. The failed samples are examined in the scanning electron microscope and imperfections are identified to have caused the premature failures. A concept for a high authority shape morphing plate is described. The design incorporates an active Kagome back-plane capable of changing the shape of a solid face by transmitting loads through a tetrahedral core. The two shape deformations to be achieved and demonstrated consist of hinging and twisting. The design is performed by a combination of analytic estimation and numerical simulation, guided by previous assessments of the Kagome configuration. The objective is to ascertain designs that provide the maximum edge displacement subjected to specified passive load. An optimization is used to ascertain the largest displacement achievable within the force capacity of the actuators. These displacements have been demonstrated and shown to correspond with values predicted by numerical simulation. Non-linear effects, such as face wrinkling, are probed by using a finite element method and the fidelity of the results assessed through comparison with measurements. The numerical results are used to validate a dimensional analysis of trends in the actuation resistance of the structure with geometry, as well as the passive load

  10. Neutron cross-sections above 20 MeV for design and modeling of accelerator driven systems

    NASA Astrophysics Data System (ADS)

    Blomgren, J.

    2007-02-01

    One of the outstanding new developments in the field of partitioning and transmutation (P{&}T) concerns accelerator-driven systems (ADS) which consist of a combination of a high-power, high-energy accelerator, a spallation target for neutron production and a sub-critical reactor core. The development of the commercial critical reactors of today motivated a large effort on nuclear data up to about 20 MeV, and presently several million data points can be found in various data libraries. At higher energies, data are scarce or even non-existent. With the development of nuclear techniques based on neutrons at higher energies, nowadays there is a need also for higher-energy nuclear data. To provide alternative to this lack of data, a wide program on neutron-induced data related to ADS for P{&}T is running at the 20-180 MeV neutron beam facility at `The Svedberg Laboratory' (TSL), Uppsala. The programme encompasses studies of elastic scattering, inelastic neutron production, i.e., (n, xn') reactions, light-ion production, fission and production of heavy residues. Recent results are presented and future program of development is outlined.

  11. NEUTRONIC REACTOR

    DOEpatents

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  12. The design of a high-efficiency neutron counter for waste drums to provide optimized sensitivity for plutonium assay

    SciTech Connect

    Menlove, H.O.; Beddingfield, D.H.; Pickrell, M.M.

    1997-11-01

    An advanced passive neutron counter has been designed to improve the accuracy and sensitivity for the nondestructive assay of plutonium in scrap and waste containers. The High-Efficiency Neutron Counter (HENC) was developed under a Cooperative Research Development Agreement between the Los Alamos National Laboratory and Canberra Industries. The primary goal of the development was to produce a passive assay system for 200-L drums that has detectability limits and multiplicity counting features that are superior to previous systems. A detectability limit figure of merit (FOM) was defined that included the detector efficiency, the neutron die-away time, and the detector`s active volume and density that determine the cosmic-ray background. Monte Carlo neutron calculations were performed to determine the parameters to provide an optimum FOM. The system includes the {sup 252}Cf {open_quotes}add-a-source{close_quotes} feature to improve the accuracy as well as statistical filters to reduce the cosmic-ray spallation neutron background. The final decision gave an efficiency of 32% for plutonium with a detector {sup 3}He tube volume that is significantly smaller than for previous high-efficiency systems for 200-L drums. Because of the high efficiency of the HENC, we have incorporated neutron multiplicity counting for matrix corrections for those cases where the plutonium is localized in nonuniform hydrogenous materials. The paper describes the design and performance testing of the advanced system. 5 refs., 8 figs., 3 tabs.

  13. GENERALISATION OF RADIATOR DESIGN TECHNIQUES FOR PERSONAL NEUTRON DOSEMETERS BY UNFOLDING METHOD.

    PubMed

    Oda, K; Nakayama, T; Umetani, K; Kajihara, M; Yamauchi, T

    2016-09-01

    A novel technique for designing a radiator suitable for personal neutron dosemeter based on plastic track detector was discussed. A multi-layer structure has been proposed in the previous report, where the thicknesses of plural polyethylene (PE) layers and insensitive ones were determined by iterative calculations of double integral. In order to arrange this procedure and make it more systematic, unfolding calculation has been employed to estimate an ideal radiator containing an arbitrary hydrogen concentration. In the second step, realistic materials replaced it with consideration of minimisation of the layer number and commercial availability. A radiator consisting of three layers of PE, Upilex and Kapton sheets was finally designed, for which a deviation in the energy dependence between 0.1 and 20 MeV could be controlled within 18 %. An applicability of fluorescent nuclear track detector element has also been discussed.

  14. GENERALISATION OF RADIATOR DESIGN TECHNIQUES FOR PERSONAL NEUTRON DOSEMETERS BY UNFOLDING METHOD.

    PubMed

    Oda, K; Nakayama, T; Umetani, K; Kajihara, M; Yamauchi, T

    2016-09-01

    A novel technique for designing a radiator suitable for personal neutron dosemeter based on plastic track detector was discussed. A multi-layer structure has been proposed in the previous report, where the thicknesses of plural polyethylene (PE) layers and insensitive ones were determined by iterative calculations of double integral. In order to arrange this procedure and make it more systematic, unfolding calculation has been employed to estimate an ideal radiator containing an arbitrary hydrogen concentration. In the second step, realistic materials replaced it with consideration of minimisation of the layer number and commercial availability. A radiator consisting of three layers of PE, Upilex and Kapton sheets was finally designed, for which a deviation in the energy dependence between 0.1 and 20 MeV could be controlled within 18 %. An applicability of fluorescent nuclear track detector element has also been discussed. PMID:26378225

  15. Design, construction, and characterization of a facility for neutron capture gamma ray analysis of sulfur in coal using californium-252

    SciTech Connect

    Layfield, J.R.

    1980-03-01

    A study of neutron capture gamma ray analysis of sulfur in coal using californium-252 as a neutron source is reported. Both internal and external target geometries are investigated. The facility designed for and used in this study is described. The external target geometry is found to be inappropriate because of the low thermal neutron flux at the sample location, which must be outside the biological shielding. The internal target geometry is found to have a sufficient thermal neutron flux, but an excessive gamma ray background. A water filled plastic facility, rather than the paraffin filled steel one used in this study, is suggested as a means of increasing flexibility and decreasing the beackground in the internal target geometry.

  16. Design and development of an in-line sputtering system and process development of thin film multilayer neutron supermirrors.

    PubMed

    Biswas, A; Sampathkumar, R; Kumar, Ajaya; Bhattacharyya, D; Sahoo, N K; Lagoo, K D; Veerapur, R D; Padmanabhan, M; Puri, R K; Bhattacharya, Debarati; Singh, Surendra; Basu, S

    2014-12-01

    Neutron supermirrors and supermirror polarizers are thin film multilayer based devices which are used for reflecting and polarizing neutrons in various neutron based experiments. In the present communication, the in-house development of a 9 m long in-line dc sputtering system has been described which is suitable for deposition of neutron supermirrors on large size (1500 mm × 150 mm) substrates and in large numbers. The optimisation process of deposition of Co and Ti thin film, Co/Ti periodic multilayers, and a-periodic supermirrors have also been described. The system has been used to deposit thin film multilayer supermirror polarizers which show high reflectivity up to a reasonably large critical wavevector transfer of ∼0.06 Å(-1) (corresponding to m = 2.5, i.e., 2.5 times critical wavevector transfer of natural Ni). The computer code for designing these supermirrors has also been developed in-house. PMID:25554268

  17. Design and development of an in-line sputtering system and process development of thin film multilayer neutron supermirrors

    SciTech Connect

    Biswas, A.; Sampathkumar, R.; Kumar, Ajaya; Bhattacharyya, D.; Sahoo, N. K.; Lagoo, K. D.; Veerapur, R. D.; Padmanabhan, M.; Puri, R. K.; Bhattacharya, Debarati; Singh, Surendra; Basu, S.

    2014-12-15

    Neutron supermirrors and supermirror polarizers are thin film multilayer based devices which are used for reflecting and polarizing neutrons in various neutron based experiments. In the present communication, the in-house development of a 9 m long in-line dc sputtering system has been described which is suitable for deposition of neutron supermirrors on large size (1500 mm × 150 mm) substrates and in large numbers. The optimisation process of deposition of Co and Ti thin film, Co/Ti periodic multilayers, and a-periodic supermirrors have also been described. The system has been used to deposit thin film multilayer supermirror polarizers which show high reflectivity up to a reasonably large critical wavevector transfer of ∼0.06 Å{sup −1} (corresponding to m = 2.5, i.e., 2.5 times critical wavevector transfer of natural Ni). The computer code for designing these supermirrors has also been developed in-house.

  18. Design and optimization of 3-mode×12-core dual-ring structured few-mode multi-core fiber

    NASA Astrophysics Data System (ADS)

    Tu, Jiajing; Long, Keping; Saitoh, Kunimasa

    2016-12-01

    We adopt dual-ring structure (DRS) for the core arrangement of 3-mode (LP01, LP11a and LP11b)×12-core few-mode multi-core fiber (FM-MCF) and then introduce the design method for this DRS-FM-MCF. After investigating the characteristics such as differential mode delay (DMD), inter-core crosstalk (XT), threshold value of bending radius (Rpk), relative core multiplicity factor (RCMF) and cable cutoff wavelength (λcc), we present an optimized scheme for this DRS-FM-MCF. For the optimized DRS-FM-MCF, | DMD | is ≤ 100 ps / km over C+L band, the maximum XT at wavelength (λ) of 1625 nm achieves -33 dB/100 km, maximum Rpk is 11.03 cm, RCMF (LP01, LP11a and LP11b) reaches 25.49 and maximum λcc is ≤ 1530 nm. Compared with one-ring structure (ORS), DRS has much more space to enlarge core pitch (Λ) so that lower XT can be achieved. Furthermore, DRS has less confinement degree on mode than square-lattice structure (SLS) if Λ and cladding diameter (Dcl) are set at similar values. It means that it is easier for DRS to make sure λcc would not be larger than the lower limit of C+L bands. In this paper, DRS is proved as a suitable core arrangement for 3-mode×12-core FM-MCF.

  19. Design and analysis of large-core single-mode windmill single crystal sapphire optical fiber

    DOE PAGES

    Cheng, Yujie; Hill, Cary; Liu, Bo; Yu, Zhihao; Xuan, Haifeng; Homa, Daniel; Wang, Anbo; Pickrell, Gary

    2016-06-01

    We present a large-core single-mode “windmill” single crystal sapphire optical fiber (SCSF) design, which exhibits single-mode operation by stripping off the higher-order modes (HOMs) while maintaining the fundamental mode. The “windmill” SCSF design was analyzed using the finite element analysis method, in which all the HOMs are leaky. The numerical simulation results show single-mode operation in the spectral range from 0.4 to 2 μm in the windmill SCSF, with an effective core diameter as large as 14 μm. Such fiber is expected to improve the performance of many of the current sapphire fiber optic sensor structures.

  20. Preliminary design report for SCDAP/RELAP5 lower core plate model

    SciTech Connect

    Coryell, E.W.; Griffin, F.P.

    1998-07-01

    The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. Under primary sponsorship of the US Nuclear Regulatory Commission (NRC), Idaho National Engineering and Environmental Laboratory (INEEL) is responsible for overall maintenance of this code and for improvements for pressurized water reactor (PWR) applications. Since 1991, Oak Ridge National Laboratory (ORNL) has been improving SCDAP/RELAP5 for boiling water reactor (BWR) applications. The RELAP5 portion of the code performs the thermal-hydraulic calculations for both normal and severe accident conditions. The structures within the reactor vessel and coolant system can be represented with either RELAP5 heat structures or SCDAP/RELAP5 severe accident structures. The RELAP5 heat structures are limited to normal operating conditions (i.e., no structural oxidation, melting, or relocation), while the SCDAP portion of the code is capable of representing structural degradation and core damage progression that can occur under severe accident conditions. DCDAP/RELAP5 currently assumes that molten material which leaves the core region falls into the lower vessel head without interaction with structural materials. The objective of this design report is to describe the modifications required for SCDAP/RELAP5 to treat the thermal response of the structures in the core plate region as molten material relocates downward from the core, through the core plate region, and into the lower plenum. This has been a joint task between INEEL and ORNL, with INEEL focusing on PWR-specific design, and ORNL focusing upon the BWR-specific aspects. Chapter 2 describes the structures in the core plate region that must be represented by the proposed model. Chapter 3 presents the available information about the damage progression that is anticipated to occur in the core plate region during a severe accident, including typical SCDAP/RELAP5 simulation results. Chapter 4 provides a

  1. The design, construction and performance of a variable collimator for epithermal neutron capture therapy beams.

    PubMed

    Riley, K J; Binns, P J; Ali, S J; Harling, O K

    2004-05-21

    A patient collimator for the fission converter based epithermal neutron beam (FCB) at the Massachusetts Institute of Technology Research Reactor (MITR-II) was built for clinical trials of boron neutron capture therapy (BNCT). A design was optimized by Monte Carlo simulations of the entire beam line and incorporates a modular construction for easy modifications in the future. The device was formed in-house by casting a mixture of lead spheres (7.6 mm diameter) in epoxy resin loaded with either 140 mg cm(-3) of boron carbide or 210 mg cm(-3) of lithium fluoride (95% enriched in 6Li). The cone shaped collimator allows easy field placement anywhere on the patient and is equipped with a laser indicator of central axis, beam's eye view optics and circular apertures of 80, 100, 120 and 160 mm diameter. Beam profiles and the collateral dose in a half-body phantom were measured for the 160 mm field using fission counters, activation foils as well as tissue equivalent (A-150) and graphite walled ionization chambers. Leakage radiation through the collimator contributes less than 10% to the total collateral dose up to 0.15 m beyond the edge of the aperture and becomes relatively more prominent with lateral displacement. The measured whole body dose equivalent of 24 +/- 2 mSv per Gy of therapeutic dose is comparable to doses received during conventional therapy and is due principally (60-80%) to thermal neutron capture reactions with boron. These findings, together with the dose distributions for the primary beam, demonstrate the suitability of this patient collimator for BNCT.

  2. Energy Efficient Engine integrated core/low spool design and performance report

    NASA Technical Reports Server (NTRS)

    Stearns, E. Marshall

    1985-01-01

    The Energy Efficient Engine (E3) is a NASA program to create fuel saving technology for future transport aircraft engines. The E3 technology advancements were demonstrated to operate reliably and achieve goal performance in tests of the Integrated Core/Low Spool vehicle. The first build of this undeveloped technology research engine set a record for low fuel consumption. Its design and detailed test results are herein presented.

  3. Design of the polar neutron-imaging aperture for use at the National Ignition Facility

    NASA Astrophysics Data System (ADS)

    Fatherley, V. E.; Barker, D. A.; Fittinghoff, D. N.; Hibbard, R. L.; Martinez, J. I.; Merrill, F. E.; Oertel, J. A.; Schmidt, D. W.; Volegov, P. L.; Wilde, C. H.

    2016-11-01

    The installation of a neutron imaging diagnostic with a polar view at the National Ignition Facility (NIF) required design of a new aperture, an extended pinhole array (PHA). This PHA is different from the pinhole array for the existing equatorial system due to significant changes in the alignment and recording systems. The complex set of component requirements, as well as significant space constraints in its intended location, makes the design of this aperture challenging. In addition, lessons learned from development of prior apertures mandate careful aperture metrology prior to first use. This paper discusses the PHA requirements, constraints, and the final design. The PHA design is complex due to size constraints, machining precision, assembly tolerances, and design requirements. When fully assembled, the aperture is a 15 mm × 15 mm × 200 mm tungsten and gold assembly. The PHA body is made from 2 layers of tungsten and 11 layers of gold. The gold layers include 4 layers containing penumbral openings, 4 layers containing pinholes and 3 spacer layers. In total, there are 64 individual, triangular pinholes with a field of view (FOV) of 200 μm and 6 penumbral apertures. Each pinhole is pointed to a slightly different location in the target plane, making the effective FOV of this PHA a 700 μm square in the target plane. The large FOV of the PHA reduces the alignment requirements both for the PHA and the target, allowing for alignment with a laser tracking system at NIF.

  4. Lunar in-core thermionic nuclear reactor power system conceptual design

    NASA Technical Reports Server (NTRS)

    Mason, Lee S.; Schmitz, Paul C.; Gallup, Donald R.

    1991-01-01

    This paper presents a conceptual design of a lunar in-core thermionic reactor power system. The concept consists of a thermionic reactor located in a lunar excavation with surface mounted waste heat radiators. The system was integrated with a proposed lunar base concept representative of recent NASA Space Exploration Initiative studies. The reference mission is a permanently-inhabited lunar base requiring a 550 kWe, 7 year life central power station. Performance parameters and assumptions were based on the Thermionic Fuel Element (TFE) Verification Program. Five design cases were analyzed ranging from conservative to advanced. The cases were selected to provide sensitivity effects on the achievement of TFE program goals.

  5. Design and construction of a thermal neutron beam for BNCT at Tehran Research Reactor.

    PubMed

    Kasesaz, Yaser; Khalafi, Hossein; Rahmani, Faezeh; Ezzati, Arsalan; Keyvani, Mehdi; Hossnirokh, Ashkan; Shamami, Mehrdad Azizi; Amini, Sepideh

    2014-12-01

    An irradiation facility has been designed and constructed at Tehran Research Reactor (TRR) for the treatment of shallow tumors using Boron Neutron Capture Therapy (BNCT). TRR has a thermal column which is about 3m in length with a wide square cross section of 1.2×1.2m(2). This facility is filled with removable graphite blocks. The aim of this work is to perform the necessary modifications in the thermal column structure to meet thermal BNCT beam criteria recommended by International Atomic Energy Agency. The main modifications consist of rearranging graphite blocks and reducing the gamma dose rate at the beam exit. Activation foils and TLD700 dosimeter have been used to measure in-air characteristics of the neutron beam. According to the measurements, a thermal flux is 5.6×10(8) (ncm(-2)s(-1)), a cadmium ratio is 186 for gold foils and a gamma dose rate is 0.57Gy h(-1).

  6. Design and expected performance of a fast neutron attenuation probe for light element density measurements

    NASA Astrophysics Data System (ADS)

    Sweany, M.; Marleau, P.

    2016-10-01

    We present the design and expected performance of a proof-of-concept 32 channel material identification system. Our system is based on the energy-dependent attenuation of fast neutrons for four elements: hydrogen, carbon, nitrogen and oxygen. We describe a new approach to obtaining a broad range of neutron energies to probe a sample, as well as our technique for reconstructing the molar densities within a sample. The system's performance as a function of time-of-flight energy resolution is explored using a Geant4-based Monte Carlo. Our results indicate that, with the expected detector response of our system, we will be able to determine the molar density of all four elements to within a 20-30% accuracy in a two hour scan time. In many cases this error is systematically low, thus the ratio between elements is more accurate. This degree of accuracy is enough to distinguish, for example, a sample of water from a sample of pure hydrogen peroxide: the ratio of oxygen to hydrogen is reconstructed to within 8±0.5% of the true value. Finally, with future algorithm development that accounts for backgrounds caused by scattering within the sample itself, the accuracy of molar densities, not ratios, may improve to the 5-10% level for a two hour scan time.

  7. Design and expected performance of a fast neutron attenuation probe for light element density measurements

    DOE PAGES

    Sweany, M.; Marleau, P.

    2016-07-08

    In this paper, we present the design and expected performance of a proof-of-concept 32 channel material identification system. Our system is based on the energy-dependent attenuation of fast neutrons for four elements: hydrogen, carbon, nitrogen and oxygen. We describe a new approach to obtaining a broad range of neutron energies to probe a sample, as well as our technique for reconstructing the molar densities within a sample. The system's performance as a function of time-of-flight energy resolution is explored using a Geant4-based Monte Carlo. Our results indicate that, with the expected detector response of our system, we will be ablemore » to determine the molar density of all four elements to within a 20–30% accuracy in a two hour scan time. In many cases this error is systematically low, thus the ratio between elements is more accurate. This degree of accuracy is enough to distinguish, for example, a sample of water from a sample of pure hydrogen peroxide: the ratio of oxygen to hydrogen is reconstructed to within 8±0.5% of the true value. Lastly, with future algorithm development that accounts for backgrounds caused by scattering within the sample itself, the accuracy of molar densities, not ratios, may improve to the 5–10% level for a two hour scan time.« less

  8. Design, implementation, and testing of a cryogenic loading capability on an engineering neutron diffractometer

    SciTech Connect

    Woodruff, T. R.; Krishnan, V. B.; Vaidyanathan, R.; Clausen, B.; Sisneros, T.; Livescu, V.; Brown, D. W.; Bourke, M. A. M.

    2010-06-15

    A novel capability was designed, implemented, and tested for in situ neutron diffraction measurements during loading at cryogenic temperatures on the spectrometer for materials research at temperature and stress at Los Alamos National Laboratory. This capability allowed for the application of dynamic compressive forces of up to 250 kN on standard samples controlled at temperatures between 300 and 90 K. The approach comprised of cooling thermally isolated compression platens that in turn conductively cooled the sample in an aluminum vacuum chamber which was nominally transparent to the incident and diffracted neutrons. The cooling/heat rate and final temperature were controlled by regulating the flow of liquid nitrogen in channels inside the platens that were connected through bellows to the mechanical actuator of the load frame and by heaters placed on the platens. Various performance parameters of this system are reported here. The system was used to investigate deformation in Ni-Ti-Fe shape memory alloys at cryogenic temperatures and preliminary results are presented.

  9. Design, implementation, and testing of a cryogenic loading capability on an engineering neutron diffractometer.

    PubMed

    Woodruff, T R; Krishnan, V B; Clausen, B; Sisneros, T; Livescu, V; Brown, D W; Bourke, M A M; Vaidyanathan, R

    2010-06-01

    A novel capability was designed, implemented, and tested for in situ neutron diffraction measurements during loading at cryogenic temperatures on the spectrometer for materials research at temperature and stress at Los Alamos National Laboratory. This capability allowed for the application of dynamic compressive forces of up to 250 kN on standard samples controlled at temperatures between 300 and 90 K. The approach comprised of cooling thermally isolated compression platens that in turn conductively cooled the sample in an aluminum vacuum chamber which was nominally transparent to the incident and diffracted neutrons. The cooling/heat rate and final temperature were controlled by regulating the flow of liquid nitrogen in channels inside the platens that were connected through bellows to the mechanical actuator of the load frame and by heaters placed on the platens. Various performance parameters of this system are reported here. The system was used to investigate deformation in Ni-Ti-Fe shape memory alloys at cryogenic temperatures and preliminary results are presented.

  10. Design, implementation, and testing of a cryogenic loading capability on an engineering neutron diffractometer.

    PubMed

    Woodruff, T R; Krishnan, V B; Clausen, B; Sisneros, T; Livescu, V; Brown, D W; Bourke, M A M; Vaidyanathan, R

    2010-06-01

    A novel capability was designed, implemented, and tested for in situ neutron diffraction measurements during loading at cryogenic temperatures on the spectrometer for materials research at temperature and stress at Los Alamos National Laboratory. This capability allowed for the application of dynamic compressive forces of up to 250 kN on standard samples controlled at temperatures between 300 and 90 K. The approach comprised of cooling thermally isolated compression platens that in turn conductively cooled the sample in an aluminum vacuum chamber which was nominally transparent to the incident and diffracted neutrons. The cooling/heat rate and final temperature were controlled by regulating the flow of liquid nitrogen in channels inside the platens that were connected through bellows to the mechanical actuator of the load frame and by heaters placed on the platens. Various performance parameters of this system are reported here. The system was used to investigate deformation in Ni-Ti-Fe shape memory alloys at cryogenic temperatures and preliminary results are presented. PMID:20590248

  11. Neutron and Gamma Fluxes and dpa Rates for HFIR Vessel Beltline Region (Present and Upgrade Designs)

    SciTech Connect

    Blakeman, E.D.

    2001-01-11

    The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) is currently undergoing an upgrading program, a part of which is to increase the diameters of two of the four radiation beam tubes (HB-2 and HB-4). This change will cause increased neutron and gamma radiation dose rates at and near locations where the tubes penetrate the vessel wall. Consequently, the rate of radiation damage to the reactor vessel wall at those locations will also increase. This report summarizes calculations of the neutron and gamma flux (particles/cm{sup 2}/s) and the dpa rate (displacements/atom/s) in iron at critical locations in the vessel wall. The calculated dpa rate values have been recently incorporated into statistical damage evaluation codes used in the assessment of radiation induced embrittlement. Calculations were performed using models based on the discrete ordinates methodology and utilizing ORNL two-dimensional and three-dimensional discrete ordinates codes. Models for present and proposed beam tube designs are shown and their results are compared. Results show that for HB-2, the dpa rate in the vessel wall where the tube penetrates the vessel will be increased by {approximately}10 by the proposed enlargement. For HB-4, a smaller increase of {approximately}2.6 is calculated.

  12. Design of a Brassica rapa core collection for association mapping studies.

    PubMed

    Zhao, Jianjun; Artemyeva, Anna; Del Carpio, Dunia Pino; Basnet, Ram Kumar; Zhang, Ningwen; Gao, Jie; Li, Fei; Bucher, Johan; Wang, Xiaowu; Visser, Richard G F; Bonnema, Guusje

    2010-11-01

    A Brassica rapa collection of 239 accessions, based on two core collections representing different morphotypes from different geographical origins, is presented and its use for association mapping is illustrated for flowering time. We analyzed phenotypic variation of leaf and seed pod traits, plant architecture, and flowering time using data collected from three field experiments and evaluated the genetic diversity with a set of SSR markers. The Wageningen University and Research Centre (WUR) and the Vavilov Research Institute of Plant Industry (VIR) core collections had similar representations of most morphotypes, as illustrated by the phenotypic and genetic variation within these groups. The analysis of population structure revealed five subgroups in the collection, whereas previous studies of the WUR core collection indicated four subgroups; the fifth group identified consisted mainly of oil accessions from the VIR core collection, winter oils from Pakistan, and a number of other types. A very small group of summer oils is described, that is not related to other oil accessions. A candidate gene approach was chosen for association mapping of flowering time with a BrFLC1 biallelic CAPS marker and a BrFLC2 multiallelic SSR marker. The two markers were significantly associated with flowering time, but their effects were confined to certain morphotypes and (or) alleles. Based on these results, we discuss the optimal design for an association mapping population and the need to fix the heterogeneous accessions to facilitate phenotyping and genotyping.

  13. Design and operation of the core topography data acquisition system for TMI-2

    SciTech Connect

    Beller, L S; Brown, H L

    1984-05-01

    Development of effective procedures for recovery from the 1979 accident at the Three Mile Island 2 nuclear station requires a detailed and quantitative description of the postaccident configuration of the core. This report describes the techniques, equipment, and procedures used for making precise ultrasonic, sonar-like measurements of the cavity left in the upper core region as a result of the accident and details the primary results of the measurements. The system developed for the measurements uses computer techniques for the command and control of remote mechanical and electronic equipment, and for data acquisition and reduction. The system was designed, fabricated, and tested; procedures developed; and personnel trained in 4-1/2 months. The primary results are detailed topographic maps of the cavity. A variety of visual aids was developed to supplement the maps and aid in interpreting companion videotape surveys. The measurements reveal a cavity of 9.3 m/sup 3/, approximately 26% of the total core volume. The cavity occupies most of the full diameter of the core to an average depth of about 1.5 m and approaches 2 m in places.

  14. Design of a Brassica rapa core collection for association mapping studies.

    PubMed

    Zhao, Jianjun; Artemyeva, Anna; Del Carpio, Dunia Pino; Basnet, Ram Kumar; Zhang, Ningwen; Gao, Jie; Li, Fei; Bucher, Johan; Wang, Xiaowu; Visser, Richard G F; Bonnema, Guusje

    2010-11-01

    A Brassica rapa collection of 239 accessions, based on two core collections representing different morphotypes from different geographical origins, is presented and its use for association mapping is illustrated for flowering time. We analyzed phenotypic variation of leaf and seed pod traits, plant architecture, and flowering time using data collected from three field experiments and evaluated the genetic diversity with a set of SSR markers. The Wageningen University and Research Centre (WUR) and the Vavilov Research Institute of Plant Industry (VIR) core collections had similar representations of most morphotypes, as illustrated by the phenotypic and genetic variation within these groups. The analysis of population structure revealed five subgroups in the collection, whereas previous studies of the WUR core collection indicated four subgroups; the fifth group identified consisted mainly of oil accessions from the VIR core collection, winter oils from Pakistan, and a number of other types. A very small group of summer oils is described, that is not related to other oil accessions. A candidate gene approach was chosen for association mapping of flowering time with a BrFLC1 biallelic CAPS marker and a BrFLC2 multiallelic SSR marker. The two markers were significantly associated with flowering time, but their effects were confined to certain morphotypes and (or) alleles. Based on these results, we discuss the optimal design for an association mapping population and the need to fix the heterogeneous accessions to facilitate phenotyping and genotyping. PMID:21076504

  15. Design and performance of a pulse transformer based on Fe-based nanocrystalline core

    NASA Astrophysics Data System (ADS)

    Yi, Liu; Xibo, Feng; Lin, Fuchang

    2011-08-01

    A dry-type pulse transformer based on Fe-based nanocrystalline core with a load of 0.88 nF, output voltage of more than 65 kV, and winding ratio of 46 is designed and constructed. The dynamic characteristics of Fe-based nanocrystalline core under the impulse with the pulse width of several microseconds were studied. The pulse width and incremental flux density have an important effect on the pulse permeability, so the pulse permeability is measured under a certain pulse width and incremental flux density. The minimal volume of the toroidal pulse transformer core is determined by the coupling coefficient, the capacitors of the resonant charging circuit, incremental flux density, and pulse permeability. The factors of the charging time, ratio, and energy transmission efficiency in the resonant charging circuit based on magnetic core-type pulse transformer are analyzed. Experimental results of the pulse transformer are in good agreement with the theoretical calculation. When the primary capacitor is 3.17 μF and charge voltage is 1.8 kV, a voltage across the secondary capacitor of 0.88 nF with peak value of 68.5 kV, rise time (10%-90%) of 1.80 μs is obtained.

  16. Design and performance of a pulse transformer based on Fe-based nanocrystalline core.

    PubMed

    Yi, Liu; Xibo, Feng; Lin, Fuchang

    2011-08-01

    A dry-type pulse transformer based on Fe-based nanocrystalline core with a load of 0.88 nF, output voltage of more than 65 kV, and winding ratio of 46 is designed and constructed. The dynamic characteristics of Fe-based nanocrystalline core under the impulse with the pulse width of several microseconds were studied. The pulse width and incremental flux density have an important effect on the pulse permeability, so the pulse permeability is measured under a certain pulse width and incremental flux density. The minimal volume of the toroidal pulse transformer core is determined by the coupling coefficient, the capacitors of the resonant charging circuit, incremental flux density, and pulse permeability. The factors of the charging time, ratio, and energy transmission efficiency in the resonant charging circuit based on magnetic core-type pulse transformer are analyzed. Experimental results of the pulse transformer are in good agreement with the theoretical calculation. When the primary capacitor is 3.17 μF and charge voltage is 1.8 kV, a voltage across the secondary capacitor of 0.88 nF with peak value of 68.5 kV, rise time (10%-90%) of 1.80 μs is obtained. PMID:21895262

  17. Cost-Optimal Design of a 3-Phase Core Type Transformer by Gradient Search Technique

    NASA Astrophysics Data System (ADS)

    Basak, R.; Das, A.; Sensarma, A. K.; Sanyal, A. N.

    2014-04-01

    3-phase core type transformers are extensively used as power and distribution transformers in power system and their cost is a sizable proportion of the total system cost. Therefore they should be designed cost-optimally. The design methodology for reaching cost-optimality has been discussed in details by authors like Ramamoorty. It has also been discussed in brief in some of the text-books of electrical design. The paper gives a method for optimizing design, in presence of constraints specified by the customer and the regulatory authorities, through gradient search technique. The starting point has been chosen within the allowable parameter space the steepest decent path has been followed for convergence. The step length has been judiciously chosen and the program has been maneuvered to avoid local minimal points. The method appears to be best as its convergence is quickest amongst different optimizing techniques.

  18. First principles design of a core bioenergetic transmembrane electron-transfer protein.

    PubMed

    Goparaju, Geetha; Fry, Bryan A; Chobot, Sarah E; Wiedman, Gregory; Moser, Christopher C; Dutton, P Leslie; Discher, Bohdana M

    2016-05-01

    Here we describe the design, Escherichia coli expression and characterization of a simplified, adaptable and functionally transparent single chain 4-α-helix transmembrane protein frame that binds multiple heme and light activatable porphyrins. Such man-made cofactor-binding oxidoreductases, designed from first principles with minimal reference to natural protein sequences, are known as maquettes. This design is an adaptable frame aiming to uncover core engineering principles governing bioenergetic transmembrane electron-transfer function and recapitulate protein archetypes proposed to represent the origins of photosynthesis. This article is part of a Special Issue entitled Biodesign for Bioenergetics--the design and engineering of electronic transfer cofactors, proteins and protein networks, edited by Ronald L. Koder and J.L. Ross Anderson.

  19. First principles design of a core bioenergetic transmembrane electron-transfer protein.

    PubMed

    Goparaju, Geetha; Fry, Bryan A; Chobot, Sarah E; Wiedman, Gregory; Moser, Christopher C; Dutton, P Leslie; Discher, Bohdana M

    2016-05-01

    Here we describe the design, Escherichia coli expression and characterization of a simplified, adaptable and functionally transparent single chain 4-α-helix transmembrane protein frame that binds multiple heme and light activatable porphyrins. Such man-made cofactor-binding oxidoreductases, designed from first principles with minimal reference to natural protein sequences, are known as maquettes. This design is an adaptable frame aiming to uncover core engineering principles governing bioenergetic transmembrane electron-transfer function and recapitulate protein archetypes proposed to represent the origins of photosynthesis. This article is part of a Special Issue entitled Biodesign for Bioenergetics--the design and engineering of electronic transfer cofactors, proteins and protein networks, edited by Ronald L. Koder and J.L. Ross Anderson. PMID:26672896

  20. Design of bus-on-chip core for micro-satellite avionics

    NASA Astrophysics Data System (ADS)

    Liu, Youjun; You, Zheng; Li, Bin; Zhang, Xiangqi; Meng, Ziyang

    2007-11-01

    This paper discusses a layout of bus-on-chip core referring to SoC thinking which is composed of six sections based on a physical chip of FPGA: multi-Processor cache coherence unit, external bus control module, TT&C module, Ethernet Mac interface, EDAC/DMA module, and AMBA bridges. Multi-processor cache coherence unit, as a key part of the bus core, is used to serve the rapid parallel computing by means of the breakthrough of write/read speed of EMS memory and enhances the reliability of OBC with the service of supporting the hot standby of redundancy and the reconfiguration of fault-tolerance. External bus control module is made to support the PnP of external components applying varieties of buses, which is designed by means of soft-core in order to adapt the variation of macro-design and improve the flexibility of external application. TT&C module is the interface of subsystems of telemetry, telecommand and communication, which involves the protocols of HDLC. Ethernet Mac interface based on TCP/IP acts as the access of ISL for formation flying, constellation, etc. EDAC/DMA module mainly manages the data exchange between AMBA bus and RAM, and assigns DMA for the payloads.

  1. Design and Demonstration of a Material-Plasma Exposure Target Station for Neutron Irradiated Samples

    SciTech Connect

    Rapp, Juergen; Aaron, A. M.; Bell, Gary L.; Burgess, Thomas W.; Ellis, Ronald James; Giuliano, D.; Howard, R.; Kiggans, James O.; Lessard, Timothy L.; Ohriner, Evan Keith; Perkins, Dale E.; Varma, Venugopal Koikal

    2015-10-20

    Fusion energy is the most promising energy source for the future, and one of the most important problems to be solved progressing to a commercial fusion reactor is the identification of plasma-facing materials compatible with the extreme conditions in the fusion reactor environment. The development of plasma–material interaction (PMI) science and the technology of plasma-facing components are key elements in the development of the next step fusion device in the United States, the so-called Fusion Nuclear Science Facility (FNSF). All of these PMI issues and the uncertain impact of the 14-MeV neutron irradiation have been identified in numerous expert panel reports to the fusion community. The 2007 Greenwald report classifies reactor plasma-facing materials (PFCs) and materials as the only Tier 1 issues, requiring a “. . . major extrapolation from the current state of knowledge, need for qualitative improvements and substantial development for both the short and long term.” The Greenwald report goes on to list 19 gaps in understanding and performance related to the plasma–material interface for the technology facilities needed for DEMO-oriented R&D and DEMO itself. Of the 15 major gaps, six (G7, G9, G10, G12, G13) can possibly be addressed with ORNL’s proposal of an advanced Material Plasma Exposure eXperiment. Establishing this mid-scale plasma materials test facility at ORNL is a key element in ORNL’s strategy to secure a leadership role for decades of fusion R&D. That is to say, our end goal is to bring the “signature facility” FNSF home to ORNL. This project is related to the pre-conceptual design of an innovative target station for a future Material–Plasma Exposure eXperiment (MPEX). The target station will be designed to expose candidate fusion reactor plasma-facing materials and components (PFMs and PFCs) to conditions anticipated in fusion reactors, where PFCs will be exposed to dense high-temperature hydrogen plasmas providing steady

  2. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    SciTech Connect

    Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

    2004-10-06

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are

  3. A new designed dual-guided ring-core fiber for OAM mode transmission

    NASA Astrophysics Data System (ADS)

    Zhu, Min; Zhang, Wenbo; Xi, Lixia; Tang, Xianfeng; Zhang, Xiaoguang

    2015-10-01

    We propose and analyze a design of multi-OAM-modes ring-core fiber with two guided modes regions which possess relatively large effective index separations required for the vector modes transmission. The proposed fiber can support up to 28 information states bearing OAM spanning 8 OAM orders in the ring region, and two degenerate fundamental polarization modes in the core region across the whole C bands. Fiber features such as dispersion, differential mode delay, effective mode area, power isolation between two guided regions and modal birefringence have been analyzed in this paper. For the reason that the proposed fiber possess the relatively good features, it has potential applications in the next generation fiber communication systems either in the quantum domain or in the classical domain.

  4. Dosimetric comparison of four new design {sup 103}Pd brachytherapy sources: Optimal design using silver and copper rod cores

    SciTech Connect

    Hosseini, S. Hamed; Sadeghi, Mahdi; Ataeinia, Vahideh

    2009-07-15

    Four new brachytherapy sources, IRA1-{sup 103}Pd, IRA2-{sup 103}Pd, IRA3-{sup 103}Pd, and IRA4-{sup 103}Pd, have been developed at Agricultural, Medical, and Industrial Research School and are designed for permanent implant application. With the goal of determining an optimal design for a {sup 103}Pd source, this article compares the dosimetric properties of these sources with reference to the authors' earlier IRA-{sup 103}Pd source. The four new sources differ in end cap configuration and thickness and in the core material, silver or copper, that carries the adsorbed {sup 103}Pd. Dosimetric data derived from the authors' Monte Carlo simulation results are reported in accordance with the updated AAPM Task Group No. 43 report (TG-43U1). For each source, the authors obtained detailed results for the dose rate constant {Lambda}, the radial dose function g(r), the anisotropy function F(r,{theta}), and the anisotropy factor {phi}{sub an}(r). In this study, the optimal source IRA3-{sup 103}Pd provides the most isotropic dose distribution in water with the dose rate constant of 0.678({+-}0.1%) cGy h{sup -1} U{sup -1}. The IRA3-{sup 103}Pd design has a silver rod core combined with thin-wall, concave end caps. Finally, the authors compared the results for their optimal source with published results for those of other source manufacturers.

  5. Design of a 100 J Dense Plasma Focus Z-pinch Device as a Portable Neutron Source

    NASA Astrophysics Data System (ADS)

    Jiang, Sheng; Higginson, Drew; Link, Anthony; Liu, Jason; Schmidt, Andrea

    2015-11-01

    The dense plasma focus (DPF) Z-pinch devices are capable of accelerating ions to high energies through MV/mm-scale electric fields. When deuterium is used as the filling gas, neutrons are generated through beam-target fusion when fast D beams collide with the bulk plasma. The neutron yield on a DPF scales favorably with current, and could be used as portable sources for active interrogation. Past DPF experiments have been optimized empirically. Here we use the particle-in-cell (PIC) code LSP to optimize a portable DPF for high neutron yield prior to building it. In this work, we are designing a DPF device with about 100 J of energy which can generate 106 - 107 neutrons. The simulations are run in the fluid mode for the rundown phase and are switched to kinetic to capture the anomalous resistivity and beam acceleration process during the pinch. A scan of driver parameters, anode geometries and gas pressures are studied to maximize the neutron yield. The optimized design is currently under construction. This work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344 and supported by the Laboratory Directed Research and Development Program (15-ERD-034) at LLNL.

  6. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    SciTech Connect

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

  7. The design of an intense accelerator-based epithermal neutron beam prototype for BNCT using near-threshold reactions

    NASA Astrophysics Data System (ADS)

    Lee, Charles L.

    Near-threshold boron neutron capture therapy (BNCT) uses proton energies only tens of rev above the (pan) reaction threshold in lithium in order to reduce the moderation requirements of the neutron source. The goals of this research were to prove the feasibility of this near-threshold concept for BNCT applications, using both calculation and experiment, and design a compact neutron source prototype from these results. This required a multidisciplinary development of methods for calculation of neutron yields, head phantom dosimetry, and accelerator target heat removal. First, a method was developed to accurately calculate thick target neutron yields for both near-threshold and higher energy proton beams, in lithium metal as well as lithium compounds. After these yields were experimentally verified, they were used as neutron sources for Monte Carlo (MCNP) simulations of neutron and photon transport in head phantoms. The theoretical and experimental determination of heat removal from a target backing with multiple fins, as well as numerical calculations of heat deposition profiles based on proton energy loss in target and backing materials, demonstrated that lithium integrity can be maintained for proton beam currents up to 2.5 mA. The final design uses a proton beam energy of 1.95 MeV and has a centerline epithermal neutron flux of 2.2 × 108 n/cm2- sec/mA, an advantage depth of 5.7 cm, an advantage ratio of 4.3, and an advantage depth dose rate of 6.7 RBE- cGy/min/mA, corresponding to an irradiation time of 38 minutes with a 5 mA beam. Moderator, reflector, and shielding weigh substantially less than other accelerator BNCT designs based on higher proton energies, e.g. 2.5 MeV. The near-threshold concept is useful as a portable neutron source for hospital settings, with applications ranging from glioblastomas to melanomas and synovectomy. (Copies available exclusively from MIT Libraries, Rm. 14- 0551, Cambridge, MA 02139-4307. Ph. 617-253-5668; Fax 617-253-1690.)

  8. Radiation transport analyses in support of the SNS Target Station Neutron Beam Line Shutters Title I Design

    SciTech Connect

    Miller, T.M.; Pevey, R.E.; Lillie, R.A.; Johnson, J.O.

    2000-12-01

    A detailed radiation transport analysis of the Spallation Neutron Source (SNS) shutters is important for the construction of the SNS because of its impact on conventional facility design, normal operation of the facility, and maintenance operations. Thus far the analysis of the SNS shutter travel gaps has been completed. This analysis was performed using coupled Monte Carlo and multi-dimensional discrete ordinates calculations.

  9. Core compressor exit stage study. Volume 1: Blading design. [turbofan engines

    NASA Technical Reports Server (NTRS)

    Wisler, D. C.

    1977-01-01

    A baseline compressor test stage was designed as well as a candidate rotor and two candidate stators that have the potential of reducing endwall losses relative to the baseline stage. These test stages are typical of those required in the rear stages of advanced, highly-loaded core compressors. The baseline Stage A is a low-speed model of Stage 7 of the 10 stage AMAC compressor. Candidate Rotor B uses a type of meanline in the tip region that unloads the leading edge and loads the trailing edge relative to the baseline Rotor A design. Candidate Stator B embodies twist gradients in the endwall region. Candidate Stator C embodies airfoil sections near the endwalls that have reduced trailing edge loading relative to Stator A. Tests will be conducted using four identical stages of blading so that the designs described will operate in a true multistage environment.

  10. The ARIES-RS power core -- Recent development in Li/V designs

    SciTech Connect

    Sze, D.K.; Billone, M.C.; Hua, T.Q.

    1997-04-01

    The ARIES-RS fusion power plant design study is based on reversed-shear (RS) physics with a Li/V (lithium breeder and vanadium structure) blanket. The reversed-shear discharge has been documented in many large tokamak experiments. The plasma in the RS mode has a high beta, low current, and low current drive requirements. Therefore, it is an attractive physics regime for a fusion power plant. The blanket system based on a Li/V has high temperature operating capability, good tritium breeding, excellent high heat flux removal capability, long structural life time, low activation, low after heat and good safety characteristics. For these reasons, the ARIES-RS reactor study selected Li/V as the reference blanket. The combination of attractive physics and attractive blanket engineering is expected to result in a superior power plant design. This paper summarizes the power core design of the ARIES-RS power plant study.

  11. Solves the Multigroup Neutron Diffusion Equation

    1995-06-23

    GNOMER is a program which solves the multigroup neutron diffusion equation in 1D, 2D and 3D cartesian geometry. The program is designed to calculate the global core power distributions (with thermohydraulic feedbacks), as well as power distribution and homogenized cross sections over a fuel assembly.

  12. Core physics analysis of 100% MOX Core in IRIS

    SciTech Connect

    Franceschini, F.; Petrovic, B.

    2006-07-01

    International Reactor Innovative and Secure (IRIS) is an advanced small-to-medium-size (1000 MWt) Pressurized Water Reactor (PWR), targeting deployment around 2015. Its reference core design is based on the current Westinghouse UO{sub 2} fuel with less than 5% {sup 235}U, and the analysis has been previously completed confirming good performance. The full MOX fuel core is currently under evaluation as one of the alternatives for the second wave of IRIS reactors. A full 3-D neutronic analysis has been performed to examine main core performance parameters, such as critical boron concentration, peaking factors, discharge burnup, etc. The enhanced moderation of the IRIS fuel lattice facilitates MOX core design, and all the obtained results are within the requirements, confirming viability of this option from the reactor physics standpoint. (authors)

  13. ACT-CCREC Core Research Program: Study Questions and Design. ACT Working Paper Series. WP-2015-01

    ERIC Educational Resources Information Center

    Cruce, Ty M.

    2015-01-01

    This report provides a non-technical overview of the guiding research questions and research design for the ACT-led core research program conducted on behalf of the GEAR UP College and Career Readiness Evaluation Consortium (CCREC). The core research program is a longitudinal study of the effectiveness of 14 GEAR UP state grants on the academic…

  14. Common Core State Standards for Mathematics. Appendix A: Designing High School Mathematics Courses Based on the Common Core State Standards

    ERIC Educational Resources Information Center

    Common Core State Standards Initiative, 2011

    2011-01-01

    The Common Core State Standards (CCSS) for Mathematics are organized by grade level in Grades K-8. At the high school level, the standards are organized by conceptual category (number and quantity, algebra, functions, geometry, modeling and probability and statistics), showing the body of knowledge students should learn in each category to be…

  15. Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

    NASA Technical Reports Server (NTRS)

    Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim

    2007-01-01

    A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

  16. Efficient Design and Analysis of Lightweight Reinforced Core Sandwich and PRSEUS Structures

    NASA Technical Reports Server (NTRS)

    Bednarcyk, Brett A.; Yarrington, Phillip W.; Lucking, Ryan C.; Collier, Craig S.; Ainsworth, James J.; Toubia, Elias A.

    2012-01-01

    Design, analysis, and sizing methods for two novel structural panel concepts have been developed and incorporated into the HyperSizer Structural Sizing Software. Reinforced Core Sandwich (RCS) panels consist of a foam core with reinforcing composite webs connecting composite facesheets. Boeing s Pultruded Rod Stitched Efficient Unitized Structure (PRSEUS) panels use a pultruded unidirectional composite rod to provide axial stiffness along with integrated transverse frames and stitching. Both of these structural concepts are ovencured and have shown great promise applications in lightweight structures, but have suffered from the lack of efficient sizing capabilities similar to those that exist for honeycomb sandwich, foam sandwich, hat stiffened, and other, more traditional concepts. Now, with accurate design methods for RCS and PRSEUS panels available in HyperSizer, these concepts can be traded and used in designs as is done with the more traditional structural concepts. The methods developed to enable sizing of RCS and PRSEUS are outlined, as are results showing the validity and utility of the methods. Applications include several large NASA heavy lift launch vehicle structures.

  17. Space power reactor in-core thermionic multicell evolutionary (S-prime) design

    NASA Astrophysics Data System (ADS)

    Determan, William R.; Van Hagan, Tom H.

    1993-01-01

    A 5- to 40-kWe moderated in-core thermionic space nuclear power system (TI-SNPS) concept was developed to address the TI-SNPS program requirements. The 40-kWe baseline design uses multicell Thermionic Fuel Elements (TFEs) in a zirconium hydride moderated reactor to achieve a specific mass of 18.2 We/kg and a net end-of-mission (EOM) efficiency of 8.2%. The reactor is cooled with a single NaK-78 pumped loop, which rejects the heat through a 24 m2 heat pipe space radiator.

  18. Space power reactor in-core thermionic multicell evolutionary (S-prime) design

    SciTech Connect

    Determan, W.R. ); Van Hagan, T.H. )

    1993-01-20

    A 5- to 40-kWe moderated in-core thermionic space nuclear power system (TI-SNPS) concept was developed to address the TI-SNPS program requirements. The 40-kWe baseline design uses multicell Thermionic Fuel Elements (TFEs) in a zirconium hydride moderated reactor to achieve a specific mass of 18.2 We/kg and a net end-of-mission (EOM) efficiency of 8.2%. The reactor is cooled with a single NaK-78 pumped loop, which rejects the heat through a 24 m[sup 2] heat pipe space radiator.

  19. Fuel performance models for high-temperature gas-cooled reactor core design

    SciTech Connect

    Stansfield, O.M.; Simon, W.A.; Baxter, A.M.

    1983-09-01

    Mechanistic fuel performance models are used in high-temperature gas-cooled reactor core design and licensing to predict failure and fission product release. Fuel particles manufactured with defective or missing SiC, IPyC, or fuel dispersion in the buffer fail at a level of less than 5 x 10/sup -4/ fraction. These failed particles primarily release metallic fission products because the OPyC remains intact on 90% of the particles and retains gaseous isotopes. The predicted failure of particles using performance models appears to be conservative relative to operating reactor experience.

  20. A single aromatic core mutation converts a designed "primitive" protein from halophile to mesophile folding.

    PubMed

    Longo, Liam M; Tenorio, Connie A; Kumru, Ozan S; Middaugh, C Russell; Blaber, Michael

    2015-01-01

    The halophile environment has a number of compelling aspects with regard to the origin of structured polypeptides (i.e., proteogenesis) and, instead of a curious niche that living systems adapted into, the halophile environment is emerging as a candidate "cradle" for proteogenesis. In this viewpoint, a subsequent halophile-to-mesophile transition was a key step in early evolution. Several lines of evidence indicate that aromatic amino acids were a late addition to the codon table and not part of the original "prebiotic" set comprising the earliest polypeptides. We test the hypothesis that the availability of aromatic amino acids could facilitate a halophile-to-mesophile transition by hydrophobic core-packing enhancement. The effects of aromatic amino acid substitutions were evaluated in the core of a "primitive" designed protein enriched for the 10 prebiotic amino acids (A,D,E,G,I,L,P,S,T,V)-having an exclusively prebiotic core and requiring halophilic conditions for folding. The results indicate that a single aromatic amino acid substitution is capable of eliminating the requirement of halophile conditions for folding of a "primitive" polypeptide. Thus, the availability of aromatic amino acids could have facilitated a critical halophile-to-mesophile protein folding adaptation-identifying a selective advantage for the incorporation of aromatic amino acids into the codon table.

  1. LOW LOSS DESIGN OF THE LINAC AND ACCUMULATOR RING FOR THE SPALLATION NEUTRON SOURCE.

    SciTech Connect

    RAPARIA,D.

    2003-02-03

    The Spallation Neutron Source (SNS) is a second generation pulsed neutron source and is presently in the fourth year of a seven-year construction cycle at Oak Ridge National Laboratory. A collaboration of six national laboratories (ANL, BNL, LANL, LBNL, ORNL, TJNAF) is responsible for the design and construction of the various subsystems. The operation of the facility will begin in 2006 and deliver a 1.0 GeV, 1.4 MW proton beam with pulse length of 650 nanosecond at a repetition rate of 60 Hz, on a liquid mercury target. It consists of an RF volume H{sup -} source of 50 mA peak current at 6% duty; an all electrostatic Low-Energy Beam Transport (LEBT) which also serves as a first stage beam chopper with {+-} 25 ns rise/fall time; a 402.5 MHz, 4-vane Radio-Frequency Quadrupole (RFQ) for acceleration up to 2.5 MeV; a Medium Energy Beam Transport (MEBT) housing a second stage chopper (<{+-} 10ns rise/fall), an adjustable beam halo scraper, and diagnostics devices; a 6-tank Drift Tube Linac (DTL) with permanent magnet quadrupoles up to 87 MeV; an 805 MHz, 4-module, Side Coupled Cavity Linac (CCL) up to 186 MeV; an 805 MHz, superconducting RF (SRF) linac with eleven medium beta ({beta} = 0.61) cryo-modules and twelve high beta ({beta} = 0.81) cryo-modules accelerating the beam to the full energy; a High Energy Beam transport (HEBT) for diagnostics, transverse and longitudinal collimation, energy correction, painting and matching; an accumulator ring compressing the 1 GeV, 1 ms pulse to 650 ns for delivery onto the target through a Ring to Target Beam Transport (RTBT) with transverse collimators.

  2. The neutron imaging diagnostic at NIF (invited).

    PubMed

    Merrill, F E; Bower, D; Buckles, R; Clark, D D; Danly, C R; Drury, O B; Dzenitis, J M; Fatherley, V E; Fittinghoff, D N; Gallegos, R; Grim, G P; Guler, N; Loomis, E N; Lutz, S; Malone, R M; Martinson, D D; Mares, D; Morley, D J; Morgan, G L; Oertel, J A; Tregillis, I L; Volegov, P L; Weiss, P B; Wilde, C H; Wilson, D C

    2012-10-01

    A neutron imaging diagnostic has recently been commissioned at the National Ignition Facility (NIF). This new system is an important diagnostic tool for inertial fusion studies at the NIF for measuring the size and shape of the burning DT plasma during the ignition stage of Inertial Confinement Fusion (ICF) implosions. The imaging technique utilizes a pinhole neutron aperture, placed between the neutron source and a neutron detector. The detection system measures the two dimensional distribution of neutrons passing through the pinhole. This diagnostic has been designed to collect two images at two times. The long flight path for this diagnostic, 28 m, results in a chromatic separation of the neutrons, allowing the independently timed images to measure the source distribution for two neutron energies. Typically the first image measures the distribution of the 14 MeV neutrons and the second image of the 6-12 MeV neutrons. The combination of these two images has provided data on the size and shape of the burning plasma within the compressed capsule, as well as a measure of the quantity and spatial distribution of the cold fuel surrounding this core.

  3. Design and development of a 3He replacement safeguards neutron counter based on 10B-lined proportional detector technology

    SciTech Connect

    Henzlova, Daniela; Evans, Louise; Menlove, Howard O.; Swinhoe, Martyn T.; Rael, Carlos D.; Martinez, Isaac P.; Marlow, Johnna B.

    2012-07-16

    This presentation represents an overview of the experimental evaluation of a boron-lined proportional technology performed within an NA-241 sponsored project on testing of boron-lined proportional counters for the purpose of replacement of {sup 3}He technologies. The presented boron-lined technology will be utilized in a design of a full scale safeguards neutron coincidence counter. The design considerations and the Monte Carlo performance predictions for the counter are also presented.

  4. Optimum design of a moderator system based on dose calculation for an accelerator driven Boron Neutron Capture Therapy.

    PubMed

    Inoue, R; Hiraga, F; Kiyanagi, Y

    2014-06-01

    An accelerator based BNCT has been desired because of its therapeutic convenience. However, optimal design of a neutron moderator system is still one of the issues. Therefore, detailed studies on materials consisting of the moderator system are necessary to obtain the optimal condition. In this study, the epithermal neutron flux and the RBE dose have been calculated as the indicators to look for optimal materials for the filter and the moderator. As a result, it was found that a combination of MgF2 moderator with Fe filter gave best performance, and the moderator system gave a dose ratio greater than 3 and an epithermal neutron flux over 1.0×10(9)cm(-2)s(-1).

  5. Pulsed neutron generators based on the sealed chambers of plasma focus design with D and DT fillings

    NASA Astrophysics Data System (ADS)

    Yurkov, D. I.; Dulatov, A. K.; Lemeshko, B. D.; Golikov, A. V.; Andreev, D. A.; Mikhailov, Yu V.; Prokuratov, I. A.; Selifanov, A. N.

    2015-11-01

    Development of neutron generators using plasma focus (PF) chambers is being conducted in the All-Russia Scientific Research Institute of Automatics (VNIIA) during more than 25 years. PF is a source of soft and hard x-rays and neutrons 2.5 MeV (D) or 14 MeV (DT). Pulses of x-rays and neutrons have a duration of about several tens of nanoseconds, which defines the scope of such generators—the study of ultrafast processes. VNIIA has developed a series of pulse neutron generators covering the range of outputs 107-1012 n/pulse with resources on the order of 103-104 switches, depending on purposes. Generators have weights in the range of 30-700 kg, which allows referring them to the class of transportable generators. Generators include sealed PF chambers, whose manufacture was mastered by VNIIA vacuum tube production plant. A number of optimized PF chambers, designed for use in generators with a certain yield of neutrons has been developed. The use of gas generator based on gas absorber of hydrogen isotopes, enabled to increase the self-life and resource of PF chambers. Currently, the PF chambers withstand up to 1000 switches and have the safety of not less than 5 years. Using a generator with a gas heater, significantly increased security of PF chambers, because deuterium-tritium mixture is released only during work, other times it is in a bound state in the working element of the gas generator.

  6. Fusion-neutron-yield, activation measurements at the Z accelerator: Design, analysis, and sensitivity

    SciTech Connect

    Hahn, K. D. Ruiz, C. L.; Fehl, D. L.; Chandler, G. A.; Knapp, P. F.; Smelser, R. M.; Torres, J. A.; Cooper, G. W.; Nelson, A. J.; Leeper, R. J.

    2014-04-15

    We present a general methodology to determine the diagnostic sensitivity that is directly applicable to neutron-activation diagnostics fielded on a wide variety of neutron-producing experiments, which include inertial-confinement fusion (ICF), dense plasma focus, and ion beam-driven concepts. This approach includes a combination of several effects: (1) non-isotropic neutron emission; (2) the 1/r{sup 2} decrease in neutron fluence in the activation material; (3) the spatially distributed neutron scattering, attenuation, and energy losses due to the fielding environment and activation material itself; and (4) temporally varying neutron emission. As an example, we describe the copper-activation diagnostic used to measure secondary deuterium-tritium fusion-neutron yields on ICF experiments conducted on the pulsed-power Z Accelerator at Sandia National Laboratories. Using this methodology along with results from absolute calibrations and Monte Carlo simulations, we find that for the diagnostic configuration on Z, the diagnostic sensitivity is 0.037% ± 17% counts/neutron per cm{sup 2} and is ∼ 40% less sensitive than it would be in an ideal geometry due to neutron attenuation, scattering, and energy-loss effects.

  7. AN Core Analysis

    NASA Astrophysics Data System (ADS)

    Barbarino, Andrea; Tomatis, Daniele

    2014-06-01

    Several alternative approximations of neutron transport have been proposed in years to move around the known limitations imposed by neutron diffusion in the modeling of nuclear cores. However, only a few complied with the industrial requirements of fast numerical computation, concentrating more on physical accuracy. In this work, the AN transport methodology is discussed with particular interest in core performance calculations. The implementation of the methodology in full core codes is discussed with particular attention to numerical issues and to the integration within the entire simulation process. Finally, first results from core studies in AN transport are analyzed in detail and compared to standard results of neutron diffusion.

  8. Simulation of the full-core pin-model by JMCT Monte Carlo neutron-photon transport code

    SciTech Connect

    Li, D.; Li, G.; Zhang, B.; Shu, L.; Shangguan, D.; Ma, Y.; Hu, Z.

    2013-07-01

    Since the large numbers of cells over a million, the tallies over a hundred million and the particle histories over ten billion, the simulation of the full-core pin-by-pin model has become a real challenge for the computers and the computational methods. On the other hand, the basic memory of the model has exceeded the limit of a single CPU, so the spatial domain and data decomposition must be considered. JMCT (J Monte Carlo Transport code) has successful fulfilled the simulation of the full-core pin-by-pin model by the domain decomposition and the nested parallel computation. The k{sub eff} and flux of each cell are obtained. (authors)

  9. Dynamic neutronic and stability analysis of a burst mode, single cavity gas core reactor Brayton cycle space power system

    NASA Astrophysics Data System (ADS)

    Dugan, Edward T.; Kutikkad, Kiratadas

    The conceptual, burst-mode gaseous-core reactor (GCR) space nuclear power system presently subjected to reactor-dynamics and system stability studies operates on a closed Brayton cycle, via disk MHD generator for energy conversion. While the gaseous fuel density power coefficient of reactivity is found to be capable of rapidly stabilizing the GCR system, the power of this feedback renders standard external reactivity insertions inadequate for significant power-level changes during normal operation.

  10. Impact of a Core Ferrule Design on Fracture Resistance of Teeth Restored with Cast Post and Core.

    PubMed

    Shamseddine, Loubna; Chaaban, Farid

    2016-01-01

    Objectives. To investigate the influence of a contra bevel on the fracture resistance of teeth restored with cast post and core. Materials and Methods. Sixty plastic analogues of an upper incisor were endodontically treated and prepared with 6° internal taper and 2 mm of ferrule in order to receive a cast post and core. The prepared samples were divided into two groups (n = 30); the first group serves as control while the second group was prepared with an external 30° bevel on the buccal and lingual walls. All samples crowned were exposed to a compressive load at 130° to their long axis until fractures occurred. Fracture resistance loads were recorded and failure modes were also observed. Mann-Whitney test was carried out to compare the two groups. Results. Mean failure loads for the groups were, respectively, 1038.69 N (SD ±243.52 N) and 1078.89 N (SD ±352.21 N). Statistically, there was no significant difference between the two groups (P = 0.7675 > 0.05). Conclusion. In the presence of a ferrule and a crown in the anterior teeth, adding a secondary ferrule to the cast post and core will not increase the resistance to fracture. PMID:27419202

  11. Impact of a Core Ferrule Design on Fracture Resistance of Teeth Restored with Cast Post and Core

    PubMed Central

    Chaaban, Farid

    2016-01-01

    Objectives. To investigate the influence of a contra bevel on the fracture resistance of teeth restored with cast post and core. Materials and Methods. Sixty plastic analogues of an upper incisor were endodontically treated and prepared with 6° internal taper and 2 mm of ferrule in order to receive a cast post and core. The prepared samples were divided into two groups (n = 30); the first group serves as control while the second group was prepared with an external 30° bevel on the buccal and lingual walls. All samples crowned were exposed to a compressive load at 130° to their long axis until fractures occurred. Fracture resistance loads were recorded and failure modes were also observed. Mann-Whitney test was carried out to compare the two groups. Results. Mean failure loads for the groups were, respectively, 1038.69 N (SD ±243.52 N) and 1078.89 N (SD ±352.21 N). Statistically, there was no significant difference between the two groups (P = 0.7675 > 0.05). Conclusion. In the presence of a ferrule and a crown in the anterior teeth, adding a secondary ferrule to the cast post and core will not increase the resistance to fracture. PMID:27419202

  12. Fission reactor neutron sources for neutron capture therapy--a critical review.

    PubMed

    Harling, Otto K; Riley, Kent J

    2003-01-01

    The status of fission reactor-based neutron beams for neutron capture therapy (NCT) is reviewed critically. Epithermal neutron beams, which are favored for treatment of deep-seated tumors, have been constructed or are under construction at a number of reactors worldwide. Some of the most recently constructed epithermal neutron beams approach the theoretical optimum for beam purity. Of these higher quality beams, at least one is suitable for use in high through-put routine therapy. It is concluded that reactor-based epithermal neutron beams with near optimum characteristics are currently available and more can be constructed at existing reactors. Suitable reactors include relatively low power reactors using the core directly as a source of neutrons or a fission converter if core neutrons are difficult to access. Thermal neutron beams for NCT studies with small animals or for shallow tumor treatments, with near optimum properties have been available at reactors for many years. Additional high quality thermal beams can also be constructed at existing reactors or at new, small reactors. Furthermore, it should be possible to design and construct new low power reactors specifically for NCT, which meet all requirements for routine therapy and which are based on proven and highly safe reactor technology.

  13. Nuclear Engineering Computer Models for In-Core Fuel Management Analysis.

    1992-06-12

    Version 00 VPI-NECM is a nuclear engineering computer system of modules for in-core fuel management analysis. The system consists of 6 independent programs designed to calculate: (1) FARCON - neutron slowing down and epithermal group constants, (2) SLOCON - thermal neutron spectrum and group constants, (3) DISFAC - slow neutron disadvantage factors, (4) ODOG - solution of a one group neutron diffusion equation, (5) ODMUG - three group criticality problem, (6) FUELBURN - fuel burnupmore » in slow neutron fission reactors.« less

  14. Lead-Cooled Fast Reactor (LFR) Design: Safety, Neutronics, Thermal Hydraulics, Structural Mechanics, Fuel, Core, and Plant Design

    SciTech Connect

    Smith, C

    2010-02-22

    The idea of developing fast spectrum reactors with molten lead (or lead alloy) as a coolant is not a new one. Although initially considered in the West in the 1950s, such technology was not pursued to completion because of anticipated difficulties associated with the corrosive nature of these coolant materials. However, in the Soviet Union, such technology was actively pursued during the same time frame (1950s through the 1980s) for the specialized role of submarine propulsion. More recently, there has been a renewal of interest in the West for such technology, both for critical systems as well as for Accelerator Driven Subcritical (ADS) systems. Meanwhile, interest in the former Soviet Union, primarily Russia, has remained strong and has expanded well beyond the original limited mission of submarine propulsion. This section reviews the past and current status of LFR development.

  15. Composite Cores

    NASA Technical Reports Server (NTRS)

    1990-01-01

    Spang & Company's new configuration of converter transformer cores is a composite of gapped and ungapped cores assembled together in concentric relationship. The net effect of the composite design is to combine the protection from saturation offered by the gapped core with the lower magnetizing requirement of the ungapped core. The uncut core functions under normal operating conditions and the cut core takes over during abnormal operation to prevent power surges and their potentially destructive effect on transistors. Principal customers are aerospace and defense manufacturers. Cores also have applicability in commercial products where precise power regulation is required, as in the power supplies for large mainframe computers.

  16. Small Launch Vehicle Design Approaches: Clustered Cores Compared with Multi-Stage Inline Concepts

    NASA Technical Reports Server (NTRS)

    Waters, Eric D.; Beers, Benjamin; Esther, Elizabeth; Philips, Alan; Threet, Grady E., Jr.

    2013-01-01

    In an effort to better define small launch vehicle design options two approaches were investigated from the small launch vehicle trade space. The primary focus was to evaluate a clustered common core design against a purpose built inline vehicle. Both designs focused on liquid oxygen (LOX) and rocket propellant grade kerosene (RP-1) stages with the terminal stage later evaluated as a LOX/methane (CH4) stage. A series of performance optimization runs were done in order to minimize gross liftoff weight (GLOW) including alternative thrust levels, delivery altitude for payload, vehicle length to diameter ratio, alternative engine feed systems, re-evaluation of mass growth allowances, passive versus active guidance systems, and rail and tower launch methods. Additionally manufacturability, cost, and operations also play a large role in the benefits and detriments for each design. Presented here is the Advanced Concepts Office's Earth to Orbit Launch Team methodology and high level discussion of the performance trades and trends of both small launch vehicle solutions along with design philosophies that shaped both concepts. Without putting forth a decree stating one approach is better than the other; this discussion is meant to educate the community at large and let the reader determine which architecture is truly the most economical; since each path has such a unique set of limitations and potential payoffs.

  17. Neutron Star Compared to Manhattan

    NASA Video Gallery

    A pulsar is a neutron star, the crushed core of a star that has exploded. Neutron stars crush half a million times more mass than Earth into a sphere no larger than Manhattan, as animated in this s...

  18. Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

    NASA Astrophysics Data System (ADS)

    Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

    2016-01-01

    Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

  19. Conceptual design of the early implementation of the NEutron Detector Array (NEDA) with AGATA

    NASA Astrophysics Data System (ADS)

    Hüyük, Tayfun; Di Nitto, Antonio; Jaworski, Grzegorz; Gadea, Andrés; Javier Valiente-Dobón, José; Nyberg, Johan; Palacz, Marcin; Söderström, Pär-Anders; Jose Aliaga-Varea, Ramon; de Angelis, Giacomo; Ataç, Ayşe; Collado, Javier; Domingo-Pardo, Cesar; Egea, Francisco Javier; Erduran, Nizamettin; Ertürk, Sefa; de France, Gilles; Gadea, Rafael; González, Vicente; Herrero-Bosch, Vicente; Kaşkaş, Ayşe; Modamio, Victor; Moszynski, Marek; Sanchis, Enrique; Triossi, Andrea; Wadsworth, Robert

    2016-03-01

    The NEutron Detector Array (NEDA) project aims at the construction of a new high-efficiency compact neutron detector array to be coupled with large γ-ray arrays such as AGATA. The application of NEDA ranges from its use as selective neutron multiplicity filter for fusion-evaporation reaction to a large solid angle neutron tagging device. In the present work, possible configurations for the NEDA coupled with the Neutron Wall for the early implementation with AGATA has been simulated, using Monte Carlo techniques, in order to evaluate their performance figures. The goal of this early NEDA implementation is to improve, with respect to previous instruments, efficiency and capability to select multiplicity for fusion-evaporation reaction channels in which 1, 2 or 3 neutrons are emitted. Each NEDA detector unit has the shape of a regular hexagonal prism with a volume of about 3.23l and it is filled with the EJ301 liquid scintillator, that presents good neutron- γ discrimination properties. The simulations have been performed using a fusion-evaporation event generator that has been validated with a set of experimental data obtained in the 58Ni + 56Fe reaction measured with the Neutron Wall detector array.

  20. Moderator design studies for a new neutron reference source based on the D-T fusion reaction

    NASA Astrophysics Data System (ADS)

    Mozhayev, Andrey V.; Piper, Roman K.; Rathbone, Bruce A.; McDonald, Joseph C.

    2016-06-01

    The radioactive isotope Californium-252 (252Cf) is relied upon internationally as a neutron calibration source for ionizing radiation dosimetry because of its high specific activity. The source may be placed within a heavy-water (D2O) moderating sphere to produce a softened spectrum representative of neutron fields common to commercial nuclear power plant environments, among others. Due to termination of the U.S. Department of Energy loan/lease program in 2012, the expense of obtaining 252Cf sources has undergone a significant increase, rendering high output sources largely unattainable. On the other hand, the use of neutron generators in research and industry applications has increased dramatically in recent years. Neutron generators based on deuteriumtritium (D-T) fusion reaction provide high neutron fluence rates and, therefore, could possibly be used as a replacement for 252Cf. To be viable, the 14 MeV D-T output spectrum must be significantly moderated to approximate common workplace environments. This paper presents the results of an effort to select appropriate moderating materials and design a configuration to reshape the primary neutron field toward a spectrum approaching that from a nuclear power plant workplace. A series of Monte-Carlo (MCNP) simulations of single layer high- and low-Z materials are used to identify initial candidate moderators. Candidates are refined through a similar series of simulations involving combinations of 2-5 different materials. The simulated energy distribution using these candidate moderators are rated in comparison to a target spectrum. Other properties, such as fluence preservation and/or enhancement, prompt gamma production and other characteristics are also considered.

  1. Design and initial 1D radiography tests of the FANTOM mobile fast-neutron radiography and tomography system

    NASA Astrophysics Data System (ADS)

    Andersson, P.; Valldor-Blücher, J.; Andersson Sundén, E.; Sjöstrand, H.; Jacobsson-Svärd, S.

    2014-08-01

    The FANTOM system is a tabletop sized fast-neutron radiography and tomography system newly developed at the Applied Nuclear Physics Division of Uppsala University. The main purpose of the system is to provide time-averaged steam-and-water distribution measurement capability inside the metallic structures of two-phase test loops for light water reactor thermal-hydraulic studies using a portable fusion neutron generator. The FANTOM system provides a set of 1D neutron transmission data, which may be inserted into tomographic reconstruction algorithms to achieve a 2D mapping of the steam-and-water distribution. In this paper, the selected design of FANTOM is described and motivated. The detector concept is based on plastic scintillator elements, separated for spatial resolution. Analysis of pulse heights on an event-to-event basis is used for energy discrimination. Although the concept allows for close stacking of a large number of detector elements, this demonstrator is equipped with only three elements in the detector and one additional element for monitoring the yield from the neutron generator. The first measured projections on test objects of known configurations are presented. These were collected using a Sodern Genie 16 neutron generator with an isotropic yield of about 1E8 neutrons per second, and allowed for characterization of the instrument's capabilities. At an energy threshold of 10 MeV, the detector offered a count rate of about 500 cps per detector element. The performance in terms of spatial resolution was validated by fitting a Gaussian Line Spread Function to the experimental data, a procedure that revealed a spatial unsharpness in good agreement with the predicted FWHM of 0.5 mm.

  2. Optimal design at inner core of the shaped pyramidal truss structure

    SciTech Connect

    Lee, Sung-Uk; Yang, Dong-Yol

    2013-12-16

    Sandwich material is a type of composite material with lightweight, high strength, good dynamic properties and high bending stiffness-to-weight ratio. This can be found well such structures in the nature (for example, internal structure of bones, plants, etc.). New trend which prefers eco-friendly products and energy efficiency is emerging in industries recently. Demand for materials with high strength and light weight is also increasing. In line with these trends, researches about manufacturing methods of sandwich material have been actively conducted. In this study, a sandwich structure named as “Shaped Pyramidal Truss Structure” is proposed to improve mechanical strength and to apply a manufacturing process suitable for massive production. The new sandwich structure was designed to enhance compressive strength by changing the cross-sectional shape at the central portion of the core. As the next step, optimization of the shape was required. Optimization technique used here was the SZGA(Successive Zooming Genetic Algorithm), which is one of GA(Genetic Algorithm) methods gradually reducing the area of design variable. The objective function was defined as moment of inertia of the cross-sectional shape of the strut. The control points of cubic Bezier curve, which was assumed to be the shape of the cross section, were used as design variables. By using FEM simulation, it was found that the structure exhibited superior mechanical properties compared to the simple design of the prior art.

  3. THERMAL: A routine designed to calculate neutron thermal scattering. Revision 1

    SciTech Connect

    Cullen, D.E.

    1995-09-19

    THERMAL is designed to calculate neutron thermal scattering that is elastic and isotropic in the center of mass system. At low energy thermal motion will be included. At high energies the target nuclei are assumed to be stationary. The point of transition between low and high energies has been defined to insure a smooth transition. It is assumed that at low energy the elastic cross section is constant in the relative system. At high energy the cross section can be of any form. You can use this routine for all energies where the elastic scattering is isotropic in the center of mass system. In most materials this will be a fairly high energy, e.g., the keV energy range. The THERMAL method is simple, clean, easy to understand, and most important very efficient; on a SUN SPARC-10 workstation, at low energies with thermal scattering it can do almost 6 million scatters a minute and at high energy over 13 million. Warning: This version of THERMAL completely supersedes the original version described in the same report number, dated February 24, 1995. The method used in the original code is incorrect, as explained in this report.

  4. Form and structural response calculations for NIF neutron exposure sample case assembly design

    SciTech Connect

    DiPeso, G.; Serduke, F.; Pillenger, L.

    1996-12-31

    We describe the calculations used to design an aluminum foam protection layer for a stainless steel neutron exposure sample case. The layer protects the case from impulsive loads generated by a 20 MJ NIF capsule 10 cm from the sample case assembly. Impulse only from ablating x-rays and hohlraum plasma debris is considered. One dimensional CALE foam response calculations and analytic estimates are used to show that 1 cm of aluminum 6101-T6 foam 10 % of solid density is sufficient to attenuate the incoming peak pressure without complete melting on crush-up. Two dimensional DYNA calculations show that a 304 stainless steel spherical shell sample case with an inner radius of 1 cm and a wall thickness of 2 mm encased in 1 cm of foam does not yield to the pressure that is transmitted through the foam by a 220 Pa-sec (2.2 ktap), 2 GPa (20 kbar) load due to recoil of x- ray ablation. An unprotected spherical shell case subjected to a gentler load with peak pressure reduced to 0.2 GPa (2 kbar) not only yields but its effective plastic strain exceeds the failure point of 0.4 in 304 stainless steel after 160 microseconds. Doubling the impulse for the protected case to approximately account for debris loading results in very localized yield and an effective plastic strain that does not exceed 0.014. (U)

  5. America's Next Great Ship: Space Launch System Core Stage Transitioning from Design to Manufacturing

    NASA Technical Reports Server (NTRS)

    Birkenstock, Benjamin; Kauer, Roy

    2014-01-01

    The Space Launch System (SLS) Program is essential to achieving the Nation's and NASA's goal of human exploration and scientific investigation of the solar system. As a multi-element program with emphasis on safety, affordability, and sustainability, SLS is becoming America's next great ship of exploration. The SLS Core Stage includes avionics, main propulsion system, pressure vessels, thrust vector control, and structures. Boeing manufactures and assembles the SLS core stage at the Michoud Assembly Facility (MAF) in New Orleans, LA, a historical production center for Saturn V and Space Shuttle programs. As the transition from design to manufacturing progresses, the importance of a well-executed manufacturing, assembly, and operation (MA&O) plan is crucial to meeting performance objectives. Boeing employs classic techniques such as critical path analysis and facility requirements definition as well as innovative approaches such as Constraint Based Scheduling (CBS) and Cirtical Chain Project Management (CCPM) theory to provide a comprehensive suite of project management tools to manage the health of the baseline plan on both a macro (overall project) and micro level (factory areas). These tools coordinate data from multiple business systems and provide a robust network to support Material & Capacity Requirements Planning (MRP/CRP) and priorities. Coupled with these tools and a highly skilled workforce, Boeing is orchestrating the parallel buildup of five major sub assemblies throughout the factory. Boeing and NASA are transforming MAF to host state of the art processes, equipment and tooling, the most prominent of which is the Vertical Assembly Center (VAC), the largest weld tool in the world. In concert, a global supply chain is delivering a range of structural elements and component parts necessary to enable an on-time delivery of the integrated Core Stage. SLS is on plan to launch humanity into the next phase of space exploration.

  6. Melt spreading code assessment, modifications, and application to the EPR core catcher design.

    SciTech Connect

    Farmer, M. T .; Nuclear Engineering Division

    2009-03-30

    The Evolutionary Power Reactor (EPR) is under consideration by various utilities in the United States to provide base load electrical production, and as a result the design is undergoing a certification review by the U.S. Nuclear Regulatory Commission (NRC). The severe accident design philosophy for this reactor is based upon the fact that the projected power rating results in a narrow margin for in-vessel melt retention by external cooling of the reactor vessel. As a result, the design addresses ex-vessel core melt stabilization using a mitigation strategy that includes: (1) an external core melt retention system to temporarily hold core melt released from the vessel; (2) a layer of 'sacrificial' material that is admixed with the melt while in the core melt retention system; (3) a melt plug in the lower part of the retention system that, when failed, provides a pathway for the mixture to spread to a large core spreading chamber; and finally, (4) cooling and stabilization of the spread melt by controlled top and bottom flooding. The overall concept is illustrated in Figure 1.1. The melt spreading process relies heavily on inertial flow of a low-viscosity admixed melt to a segmented spreading chamber, and assumes that the melt mass will be distributed to a uniform height in the chamber. The spreading phenomenon thus needs to be modeled properly in order to adequately assess the EPR design. The MELTSPREAD code, developed at Argonne National Laboratory, can model segmented, and both uniform and nonuniform spreading. The NRC is thus utilizing MELTSPREAD to evaluate melt spreading in the EPR design. MELTSPREAD was originally developed to support resolution of the Mark I containment shell vulnerability issue. Following closure of this issue, development of MELTSPREAD ceased in the early 1990's, at which time the melt spreading database upon which the code had been validated was rather limited. In particular, the database that was utilized for initial validation consisted

  7. Scintillation fiber array detector for measurement of neutron beam profile

    NASA Astrophysics Data System (ADS)

    Kim, Chong; Hong, Byungsik; Jo, Mihee; Lee, Kyong Sei; Sim, Kwang-Souk

    2009-10-01

    We built and tested a detector to measure the profile of fast-neutron beams delivered by the MC50 cyclotron at the Korea Institute of Radiological and Medical Science (KIRAMS). The core component of the detector is a 2×46 array of scintillation fibers. The light output of the scintillation fibers is transformed into a current signal by a 46-channel silicon photodiode and digitized by a current-mode signal processor. This scanning device was designed to cover a neutron beam area of 30×32 cm2. The detector was tested in a neutron beam delivered by the MC50 cyclotron at KIRAMS. We demonstrate that the detector can successfully measure the neutron beam profile at various beam currents from 10 to 20 μA. The proposed neutron beam profile detector will be useful, for example, in radiotherapy applications with neutron intensities above 107 Hz/cm2.

  8. Chemical and Colloidal Stability of Carboxylated Core-Shell Magnetite Nanoparticles Designed for Biomedical Applications

    PubMed Central

    Szekeres, Márta; Tóth, Ildikó Y.; Illés, Erzsébet; Hajdú, Angéla; Zupkó, István; Farkas, Katalin; Oszlánczi, Gábor; Tiszlavicz, László; Tombácz, Etelka

    2013-01-01

    Despite the large efforts to prepare super paramagnetic iron oxide nanoparticles (MNPs) for biomedical applications, the number of FDA or EMA approved formulations is few. It is not known commonly that the approved formulations in many instances have already been withdrawn or discontinued by the producers; at present, hardly any approved formulations are produced and marketed. Literature survey reveals that there is a lack for a commonly accepted physicochemical practice in designing and qualifying formulations before they enter in vitro and in vivo biological testing. Such a standard procedure would exclude inadequate formulations from clinical trials thus improving their outcome. Here we present a straightforward route to assess eligibility of carboxylated MNPs for biomedical tests applied for a series of our core-shell products, i.e., citric acid, gallic acid, poly(acrylic acid) and poly(acrylic acid-co-maleic acid) coated MNPs. The discussion is based on physicochemical studies (carboxylate adsorption/desorption, FTIR-ATR, iron dissolution, zeta potential, particle size, coagulation kinetics and magnetization measurements) and involves in vitro and in vivo tests. Our procedure can serve as an example to construct adequate physico-chemical selection strategies for preparation of other types of core-shell nanoparticles as well. PMID:23857054

  9. Novel design of dual-core microstructured fiber with enhanced longitudinal strain sensitivity

    NASA Astrophysics Data System (ADS)

    Szostkiewicz, Lukasz; Tenderenda, T.; Napierala, M.; Szymański, M.; Murawski, M.; Mergo, P.; Lesiak, P.; Marc, P.; Jaroszewicz, L. R.; Nasilowski, T.

    2014-05-01

    Constantly refined technology of manufacturing increasingly complex photonic crystal fibers (PCF) leads to new optical fiber sensor concepts. The ways of enhancing the influence of external factors (such as hydrostatic pressure, temperature, acceleration) on the fiber propagating conditions are commonly investigated in literature. On the other hand longitudinal strain analysis, due to the calculation difficulties caused by the three dimensional computation, are somehow neglected. In this paper we show results of such a 3D numerical simulation and report methods of tuning the fiber strain sensitivity by changing the fiber microstructure and core doping level. Furthermore our approach allows to control whether the modes' effective refractive index is increasing or decreasing with strain, with the possibility of achieving zero strain sensitivity with specific fiber geometries. The presented numerical analysis is compared with experimental results of the fabricated fibers characterization. Basing on the aforementioned methodology we propose a novel dual-core fiber design with significantly increased sensitivity to longitudinal strain for optical fiber sensor applications. Furthermore the reported fiber satisfies all conditions necessary for commercial applications like good mode matching with standard single-mode fiber, low confinement loss and ease of manufacturing with the stack-and-draw technique. Such fiber may serve as an integrated Mach-Zehnder interferometer when highly coherent source is used. With the optimization of single mode transmission to 850 nm, we propose a VCSEL source to be used in order to achieve a low-cost, reliable and compact strain sensing transducer.

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

    SciTech Connect

    Burns, T.D. Jr.

    1995-05-01

    The non-surgical brain cancer treatment modality, Boron Neutron Capture Therapy (BNCT), requires the use of an epithermal neutron beam. This purpose of this thesis was to design an epithermal neutron beam at the University of Virginia Research Reactor (UVAR) suitable for BNCT applications. A suitable epithermal neutron beam for BNCT must have minimal fast neutron and gamma radiation contamination, and yet retain an appreciable intensity. The low power of the UVAR core makes reaching a balance between beam quality and intensity a very challenging design endeavor. The MCNP monte carlo neutron transport code was used to develop an equivalent core radiation source, and to perform the subsequent neutron transport calculations necessary for beam model analysis and development. The code accuracy was validated by benchmarking output against experimental criticality measurements. An epithermal beam was designed for the UVAR, with performance characteristics comparable to beams at facilities with cores of higher power. The epithermal neutron intensity of this beam is 2.2 {times} 10{sup 8} n/cm{sup 2} {center_dot} s. The fast neutron and gamma radiation KERMA factors are 10 {times} 10{sup {minus}11}cGy{center_dot}cm{sup 2}/n{sub epi} and 20 {times} 10{sup {minus}11} cGy{center_dot}cm{sup 2}/n{sub epi}, respectively, and the current-to-flux ratio is 0.85. This thesis has shown that the UVAR has the capability to provide BNCT treatments, however the performance characteristics of the final beam of this study were limited by the low core power.

  11. Nuclear Data Needs for the Neutronic Design of MYRRHA Fast Spectrum Research Reactor

    NASA Astrophysics Data System (ADS)

    Stankovskiy, A.; Malambu, E.; Van den Eynde, G.; Díez, C. J.

    2014-04-01

    A global sensitivity analysis of effective neutron multiplication factor to the change of nuclear data library has been performed. It revealed that the test version of JEFF-3.2 neutron-induced evaluated data library produces closer results to ENDF/B-VII.1 than JEFF-3.1.2 does. The analysis of contributions of individual evaluations into keff sensitivity resulted in the priority list of nuclides, uncertainties on cross sections and fission neutron multiplicities of which have to be improved by setting up dedicated differential and integral experiments.

  12. Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report

    SciTech Connect

    G. S. Chang; M. A. Lillo; R. G. Ambrosek

    2008-06-01

    The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of the

  13. Simulated Performance of the Integrated Passive Neutron Albedo Reactivity and Self-Interrogation Neutron Resonance Densitometry Detector Designed for Spent Fuel Measurement at the Fugen Reactor in Japan

    SciTech Connect

    Ulrich, Timothy J. II; Lafleur, Adrienne M.; Menlove, Howard O.; Swinhoe, Martyn T.; Tobin, Stephen J.; Seya, Michio; Bolind, Alan M.

    2012-07-16

    An integrated nondestructive assay instrument, which combined the Passive Neutron Albedo Reactivity (PNAR) and the Self-Interrogation Neutron Resonance Densitometry (SINRD) techniques, is the research focus for a collaborative effort between Los Alamos National Laboratory (LANL) and the Japanese Atomic Energy Agency as part of the Next Generation Safeguard Initiative. We will quantify the anticipated performance of this experimental system in two physical environments: (1) At LANL we will measure fresh Low Enriched Uranium (LEU) assemblies for which the average enrichment can be varied from 0.2% to 3.2% and for which Gd laced rods will be included. (2) At Fugen we will measure spent Mixed Oxide (MOX-B) and LEU spent fuel assemblies from the heavy water moderated Fugen reactor. The MOX-B assemblies will vary in burnup from {approx}3 GWd/tHM to {approx}20 GWd/tHM while the LEU assemblies ({approx}1.9% initial enrichment) will vary from {approx}2 GWd/tHM to {approx}7 GWd/tHM. The estimated count rates will be calculated using MCNPX. These preliminary results will help the finalization of the hardware design and also serve a guide for the experiment. The hardware of the detector is expected to be fabricated in 2012 with measurements expected to take place in 2012 and 2013. This work is supported by the Next Generation Safeguards Initiative, Office of Nuclear Safeguards and Security, National Nuclear Security Administration.

  14. Neutron chopper development at LANSCE

    SciTech Connect

    Nutter, M.; Lewis, L.; Tepper, S.; Silver, R.N.; Heffner, R.H.

    1985-01-01

    Progress is reported on neutron chopper systems for the Los Alamos Neutron Scattering Center pulsed spallation neutron source. This includes the development of 600+ Hz active magnetic bearing neutron chopper and a high speed control system designed to operate with the Proton Storage Ring to phase the chopper to the neutron source. 5 refs., 3 figs.

  15. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2006

    SciTech Connect

    Primm, R. T.; Ellis, R. J.; Gehin, J. C.; Clarno, K. T.; Williams, K. A.; Moses, D. L.

    2006-11-01

    Neutronics and thermal-hydraulics studies show that, for equivalent operating power [85 MW(t)], a low-enriched uranium (LEU) fuel cycle based on uranium-10 wt % molybdenum (U-10Mo) metal foil with radially, “continuously graded” fuel meat thickness results in a 15% reduction in peak thermal flux in the beryllium reflector of the High Flux Isotope Reactor (HFIR) as compared to the current highly enriched uranium (HEU) cycle. The uranium-235 content of the LEU core is almost twice the amount of the HEU core when the length of the fuel cycle is kept the same for both fuels. Because the uranium-238 content of an LEU core is a factor of 4 greater than the uranium-235 content, the LEU HFIR core would weigh 30% more than the HEU core. A minimum U-10Mo foil thickness of 84 μm is required to compensate for power peaking in the LEU core although this value could be increased significantly without much penalty. The maximum U-10Mo foil thickness is 457μm. Annual plutonium production from fueling the HFIR with LEU is predicted to be 2 kg. For dispersion fuels, the operating power for HFIR would be reduced considerably below 85 MW due to thermal considerations and due to the requirement of a 26-d fuel cycle. If an acceptable fuel can be developed, it is estimated that $140 M would be required to implement the conversion of the HFIR site at Oak Ridge National Laboratory from an HEU fuel cycle to an LEU fuel cycle. To complete the conversion by fiscal year 2014 would require that all fuel development and qualification be completed by the end of fiscal year 2009. Technological development areas that could increase the operating power of HFIR are identified as areas for study in the future.

  16. Developing standardized connection analysis techniques for slim hole core rod designs

    SciTech Connect

    Fehr, G.; Bailey, E.I.

    1994-12-31

    Slim hole core rod design remains essentially in the proprietary domain. API standardization provides the ability to perform engineering analyses and dimensional inspections through the use of documents, ie: Specifications, Bulletins, and Recommended Practices. In order to provide similar engineering capability for non-API slim hole connections, this paper develops the initial phase of what may evolve into an engineering tool to provide at least an indication of relative serviceability between two connection styles for a given application. The starting point for this process will look at bending strength ratios and connection strength calculations. Since empirical data are yet needed to verify the approaches proposed in this paper, it is recognized that the alternatives presented here are only a first step to developing useful rules of thumb which may lead to later standardization.

  17. The IPE Database: providing information on plant design, core damage frequency and containment performance

    SciTech Connect

    Lehner, J.R.; Lin, C.C.; Pratt, W.T.; Su, T.; Danziger, L.

    1996-08-01

    A database, called the IPE Database has been developed that stores data obtained from the Individual Plant Examinations (IPEs) which licensees of nuclear power plants have conducted in response to the Nuclear Regulatory Commission`s (NRC) Generic Letter GL88-20. The IPE Database is a collection of linked files which store information about plant design, core damage frequency (CDF), and containment performance in a uniform, structured way. The information contained in the various files is based on data contained in the IPE submittals. The information extracted from the submittals and entered into the IPE Database can be manipulated so that queries regarding individual or groups of plants can be answered using the IPE Database.

  18. PVDF core-free actuator for Braille displays: design, fabrication process, and testing

    NASA Astrophysics Data System (ADS)

    Levard, Thomas; Diglio, Paul J.; Lu, Sheng-Guo; Gorny, Lee J.; Rahn, Christopher D.; Zhang, Q. M.

    2011-04-01

    Refreshable Braille displays require many, small diameter actuators to move the pins. The electrostrictive P(VDF-TrFECFE) terpolymer can provide the high strain and actuation force under modest electric fields that are required of this application. In this paper, we develop core-free tubular actuators and integrate them into a 3 × 2 Braille cell. The films are solution cast, stretched to 6 μm thick, electroded, laminated into a bilayer, rolled into a 2 mm diameter tube, bonded, and provided with top and bottom contacts. Experimental testing of 17 actuators demonstrates significant strains (up to 4%). A novel Braille cell is designed and fabricated using six of these actuators.

  19. Core design study of a supercritical light water reactor with double row fuel rods

    SciTech Connect

    Zhao, C.; Wu, H.; Cao, L.; Zheng, Y.; Yang, J.; Zhang, Y.

    2012-07-01

    An equilibrium core for supercritical light water reactor has been designed. A novel type of fuel assembly with dual rows of fuel rods between water rods is chosen and optimized to get more uniform assembly power distributions. Stainless steel is used for fuel rod cladding and structural material. Honeycomb structure filled with thermal isolation is introduced to reduce the usage of stainless steel and to keep moderator temperature below the pseudo critical temperature. Water flow scheme with ascending coolant flow in inner regions is carried out to achieve high outlet temperature. In order to enhance coolant outlet temperature, the radial power distributions needs to be as flat as possible through operation cycle. Fuel loading pattern and control rod pattern are optimized to flatten power distribution at inner regions. Axial fuel enrichment is divided into three parts to control axial power peak, which affects maximum cladding surface temperature. (authors)

  20. Design Review Report for formal review of safety class features of exhauster system for rotary mode core sampling

    SciTech Connect

    JANICEK, G.P.

    2000-06-08

    Report documenting Formal Design Review conducted on portable exhausters used to support rotary mode core sampling of Hanford underground radioactive waste tanks with focus on Safety Class design features and control requirements for flammable gas environment operation and air discharge permitting compliance.

  1. Phononic band gap design in honeycomb lattice with combinations of auxetic and conventional core

    NASA Astrophysics Data System (ADS)

    Mukherjee, Sushovan; Scarpa, Fabrizio; Gopalakrishnan, S.

    2016-05-01

    We present a novel design of a honeycomb lattice geometry that uses a seamless combination of conventional and auxetic cores, i.e. elements showing positive and negative Poisson’s ratio. The design is aimed at tuning and improving the band structure of periodic cellular structures. The proposed cellular configurations show a significantly wide band gap at much lower frequencies compared to their pure counterparts, while still retaining their major dynamic features. Different topologies involving both auxetic inclusions in a conventional lattice and conversely hexagonal cellular inclusions in auxetic butterfly lattices are presented. For all these cases the impact of the varying degree of auxeticity on the band structure is evaluated. The proposed cellular designs may offer significant advantages in tuning high-frequency bandgap behaviour, which is relevant to phononics applications. The configurations shown in this paper may be made iso-volumetric and iso-weight to a given regular hexagonal topology, making possible to adapt the hybrid lattices to existing sandwich structures with fixed dimensions and weights. This work also features a comparative study of the wave speeds corresponding to different configurations vis-a vis those of a regular honeycomb to highlight the superior behaviour of the combined hybrid lattice.

  2. IMPACT OF NEUTRON AND GAMMA RADIATION ON THE DESIGN OF DIAGNOSTICS AND OTHER TARGET-BAY SYSTEMS

    SciTech Connect

    Eder, D C; Song, P M; Latkowski, J F; Reyes, S; O'Brien, D W; Lee, F D; Young, B K; Koch, J A; Watts, P W; Kimbrough, J R; Ng, E W; Landen, O L; MacGowan, B J

    2005-09-01

    The design of a wide range of components in and near the target bay of the National Ignition Facility (NIF) must allow for significant radiation from neutrons and gammas. Detailed 3D Monte Carlo simulations are critical to determine neutron and gamma fluxes for all target-bay components to allow optimization of location and auxiliary shielding. Demonstration of ignition poses unique challenges because of the large range ({approx}3 orders of magnitude) in the yield for any given attempt at ignition. Some diagnostics will provide data independent of yield, while others will provide data for lower yields and only survive high yields with little or no damage. In addition, for a given yield there is a more than 10 orders of magnitude range in neutron and gamma fluxes depending on location in the facility. For example, sensitive components in the diagnostic mezzanines and switchyards require auxiliary shielding for high-yield shots even though they are greater than 17 meters from target chamber center (TCC) and shielded by the 2 m-thick target-bay wall. In contrast, there are components 0.2 to 2 m from TCC with little or no shielding. For these components, particular attention is being made to use low-activation material because of the extremely high neutron loading levels. Many of the components closest to target center are designed to be single use to reduce worker dose from having to refurbish highly activated components. The cryogenic target positioner is an example where activation and ease of component replacement is an important part of the design. We are developing a design process for all target-bay systems that will assure reliable operation for the full range of planned yields.

  3. Replacing a 252Cf source with a neutron generator in a shuffler - a conceptual design performed with MCNPX

    SciTech Connect

    Schear, Melissa A; Tobin, Stephen J

    2009-01-01

    The {sup 252}Cf shuffler has been widely used in nuclear safeguards and radioactive waste management to assay fissile isotopes, such as {sup 235}U or {sup 239}Pu, present in a variety of samples, ranging from small cans of uranium waste to metal samples weighing several kilograms. Like other non-destructive assay instruments, the shuffler uses an interrogating neutron source to induce fissions in the sample. Although shufflers with {sup 252}Cf sources have been reliably used for several decades, replacing this isotopic source with a neutron generator presents some distinct advantages. Neutron generators can be run in a continuous or pulsed mode, and may be turned off, eliminating the need for shielding and a shuffling mechanism in the shuffler. There is also essentially no dose to personnel during installation, and no reliance on the availability of {sup 252}Cf. Despite these advantages, the more energetic neutrons emitted from the neutron generator (141 MeV for D-T generators) present some challenges for certain material types. For example when the enrichment of a uranium sample is unknown, the fission of {sup 238}U is generally undesirable. Since measuring uranium is one of the main uses of a shuffler, reducing the delayed neutron contribution from {sup 238}U is desirable. Hence, the shuffler hardware must be modified to accommodate a moderator configuration near the source to tailor the interrogating spectrum in a manner which promotes sub-threshold fissions (below 1 MeV) but avoids the over-moderation of the interrogating neutrons so as to avoid self-shielding. In this study, where there are many material and geometry combinations, the Monte Carlo N-Particle eXtended (MCNPX) transport code was used to model, design, and optimize the moderator configuration within the shuffler geometry. The code is then used to evaluate and compare the assay performances of both the modified shuffler and the current {sup 252}Cf shuffler designs for different test samples. The

  4. Mechanical Design and Analysis of a 200 MHz, Bolt-together RFQ forthe Accelerator Driven Neutron Source

    SciTech Connect

    Virostek, Steve; Hoff, Matt; Li, Derun; Staples, John; Wells,Russell

    2007-06-20

    A high-yield neutron source to screen sea-land cargocontainers for shielded Special Nuclear Materials (SNM) has been designedat LBNL [1,2]. The Accelerator-Driven Neutron Source (ADNS) uses theD(d,n)3He reaction to create a forward directed neutron beam. Keycomponents are a high-current radio-frequency quadrupole (RFQ)accelerator and a high-power target capable of producing a neutron fluxof>107 n/(cm2 cdot s) at a distance of 2.5 m. The mechanical designand analysis of the four-module, bolt-together RFQ will be presentedhere. Operating at 200 MHz, the 5.1 m long RFQ will accelerate a 40 mAdeuteron beam to 6 MeV. At a 5 percent duty factor, the time-average d+beam current on target is 1.5 mA. Each of the 1.27 m long RFQ moduleswill consist of four solid OFHC copper vanes. A specially designed 3-DO-ring will provide vacuum sealing between both the vanes and themodules. RF connections are made with canted coil spring contacts. Aseries of 60 water-cooled pi-mode rods provides quadrupole modestabilization. A set of 80 evenly spaced fixed slug tuners is used forfinal frequency adjustment and local field perturbationcorrection.

  5. Vacuum seals design and testing for a linear accelerator of the National Spallation Neutron Source

    SciTech Connect

    Z. Chen; C. Gautier; F. Hemez; N. K. Bultman

    2000-02-01

    Vacuum seals are very important to ensure that the Spallation Neutron Source (SNS) Linac has an optimum vacuum system. The vacuum joints between flanges must have reliable seals to minimize the leak rate and meet vacuum and electrical requirements. In addition, it is desirable to simplify the installation and thereby also simplify the maintenance required. This report summarizes an investigation of the metal vacuum seals that include the metal C-seal, Energized Spring seal, Helcoflex Copper Delta seal, Aluminum Delta seal, delta seal with limiting ring, and the prototype of the copper diamond seals. The report also contains the material certifications, design, finite element analysis, and testing for all of these seals. It is a valuable reference for any vacuum system design. To evaluate the suitability of several types of metal seals for use in the SNS Linac and to determine the torque applied on the bolts, a series of vacuum leak rate tests on the metal seals have been completed at Los Alamos Laboratory. A copper plated flange, using the same type of delta seal that was used for testing with the stainless steel flange, has also been studied and tested. A vacuum seal is desired that requires significantly less loading than a standard ConFlat flange with a copper gasket for the coupling cavity assembly. To save the intersegment space the authors use thinner flanges in the design. The leak rate of the thin ConFlat flange with a copper gasket is a baseline for the vacuum test on all seals and thin flanges. A finite element analysis of a long coupling cavity flange with a copper delta seal has been performed in order to confirm the design of the long coupling cavity flange and the welded area of a cavity body with the flange. This analysis is also necessary to predict a potential deformation of the cavity under the combined force of atmospheric pressure and the seating load of the seal. Modeling of this assembly has been achieved using both HKS/Abaqus and COSMOS

  6. Calculation of neutron and gamma ray energy spectra for fusion reactor shield design: comparison with experiment

    SciTech Connect

    Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.; Chapman, G.T.

    1980-08-01

    Integral experiments that measure the transport of approx. 14 MeV D-T neutrons through laminated slabs of proposed fusion reactor shield materials have been carried out. Measured and calculated neutron and gamma ray energy spectra are compared as a function of the thickness and composition of stainless steel type 304, borated polyethylene, and Hevimet (a tungsten alloy), and as a function of detector position behind these materials. The measured data were obtained using a NE-213 liquid scintillator using pulse-shape discrimination methods to resolve neutron and gamma ray pulse height data and spectral unfolding methods to convert these data to energy spectra. The calculated data were obtained using two-dimensional discrete ordinates radiation transport methods in a complex calculational network that takes into account the energy-angle dependence of the D-T neutrons and the nonphysical anomalies of the S/sub n/ method.

  7. Design Challenges and Performance of Nested Neutron Mirrors for Microfocusing on SNAP

    SciTech Connect

    Ice, G.E.; Takacs, P.; Pang, J.W.L.; Tulk, C.; Molaison, J.; Choi, J-Y.; Vaughn, C.; Lytle, L.; Andersen, K.H.; Bigault, T.; Khounsary, A.

    2009-09-16

    Kirkpatrick-Baez (KB) neutron supermirrors can efficiently focus polychromatic neutron beams to micrometre dimensions. The ultimate size is determined mainly by the perfection of the mirrors and by the size of the beam needed to have sufficient experimental signal. Nested or Montel KB mirrors can collect {approx}2.6 times more beam than standard sequential KB optics, but require good figure perfection at the edge of one mirror. This paper describes the characterization of the figure errors over the important reflective portions of the two mirrors needed for a Montel focusing pair. The measurements are placed in context with theoretical predictions and are used to predict mirror focusing performance. Strategies to improve on the focusing of this class of optics are suggested and early results from these mirrors installed on the Spallation Neutrons at Pressure (SNAP) Beamline 3 at the Spallation Neutron Source (SNS) at Oak Ridge are presented.

  8. Design challenges and performance of nested neutron mirrors for microfocusing on SNAP

    SciTech Connect

    Ice, Gene E; Pang, Judy; Tulk, Christopher A; Molaison, Jamie J; Choi, JY; Vaughn, Cody; Lytle, Lauren; Takacs, P. Z.; Anderson, Ken; Bigault, T; Khounsary, Ali

    2009-01-01

    Kirkpatrick-Baez (KB) neutron supermirrors can efficiently focus polychromatic neutron beams to micrometre dimensions. The ultimate size is determined mainly by the perfection of the mirrors and by the size of the beam needed to have sufficient experimental signal. Nested or Montel KB mirrors can collect ~2.6 times more beam than standard sequential KB optics, but require good figure perfection at the edge of one mirror. This paper describes the characterization of the figure errors over the important reflective portions of the two mirrors needed for a Montel focusing pair. The measurements are placed in context with theoretical predictions and are used to predict mirror focusing performance. Strategies to improve on the focusing of this class of optics are suggested and early results from these mirrors installed on the Spallation Neutrons at Pressure (SNAP) Beamline 3 at the Spallation Neutron Source (SNS) at Oak Ridge are presented.

  9. Parameter Design and Optimal Control of an Open Core Flywheel Energy Storage System

    NASA Technical Reports Server (NTRS)

    Pang, D.; Anand, D. K.; Kirk, J. A.

    1996-01-01

    In low earth orbit (LEO) satellite applications spacecraft power is provided by photovoltaic cells and batteries. To overcome battery shortcomings the University of Maryland, working in cooperation with NASA/GSFC and NASA/LeRC, has developed a magnetically suspended flywheel for energy storage applications. The system is referred to as an Open Core Composite Flywheel (OCCF) energy storage system. Successful application of flywheel energy storage requires integration of several technologies, viz. bearings, rotor design, motor/generator, power conditioning, and system control. In this paper we present a parameter design method which has been developed for analyzing the linear SISO model of the magnetic bearing controller for the OCCF. The objective of this continued research is to principally analyze the magnetic bearing system for nonlinear effects in order to increase the region of stability, as determined by high speed and large air gap control. This is achieved by four tasks: (1) physical modeling, design, prototyping, and testing of an improved magnetically suspended flywheel energy storage system, (2) identification of problems that limit performance and their corresponding solutions, (3) development of a design methodology for magnetic bearings, and (4) design of an optimal controller for future high speed applications. Both nonlinear SISO and MIMO models of the magnetic system were built to study limit cycle oscillations and power amplifier saturation phenomenon observed in experiments. The nonlinear models include the inductance of EM coils, the power amplifier saturation, and the physical limitation of the flywheel movement as discussed earlier. The control program EASY5 is used to study the nonlinear SISO and MIMO models. Our results have shown that the characteristics and frequency responses of the magnetic bearing system obtained from modeling are comparable to those obtained experimentally. Although magnetic saturation is shown in the bearings, there

  10. LTA Physics Design: Description of All MOX Pin LTA Design

    SciTech Connect

    Pavlovichev, A.M.

    2001-09-28

    In this document issued according to Work Release 02. P. 99-lb the results of neutronics studies of <<100%Pu>> MOX LTA design are presented. The parametric studies of infinite MOX-UOX grids, MOX-UOX core fragments and of VVER-1000 core with 3 MOX LTAs are performed. The neutronics parameters of MOX fueled core have been performed for the chosen design MOX LTA using the Russian 3D code BIPR-7A and 2D code PERMAK-A with the constants prepared by the cell spectrum code TVS-M.

  11. The ITER Radial Neutron Camera Detection System

    SciTech Connect

    Marocco, D.; Belli, F.; Esposito, B.; Petrizzi, L.; Riva, M.; Bonheure, G.; Kaschuck, Y.

    2008-03-12

    A multichannel neutron detection system (Radial Neutron Camera, RNC) will be installed on the ITER equatorial port plug 1 for total neutron source strength, neutron emissivity/ion temperature profiles and n{sub t}/n{sub d} ratio measurements [1]. The system is composed by two fan shaped collimating structures: an ex-vessel structure, looking at the plasma core, containing tree sets of 12 collimators (each set lying on a different toroidal plane), and an in-vessel structure, containing 9 collimators, for plasma edge coverage. The RNC detecting system will work in a harsh environment (neutron fiux up to 10{sup 8}-10{sup 9} n/cm{sup 2} s, magnetic field >0.5 T or in-vessel detectors), should provide both counting and spectrometric information and should be flexible enough to cover the high neutron flux dynamic range expected during the different ITER operation phases. ENEA has been involved in several activities related to RNC design and optimization [2,3]. In the present paper the up-to-date design and the neutron emissivity reconstruction capabilities of the RNC will be described. Different options for detectors suitable for spectrometry and counting (e.g. scintillators and diamonds) focusing on the implications in terms of overall RNC performance will be discussed. The increase of the RNC capabilities offered by the use of new digital data acquisition systems will be also addressed.

  12. Technical bases of the second generation SARIS core model (Task Number: 90-008-0)

    SciTech Connect

    Gregory, M.V.

    1991-11-01

    A methodology has been developed to rigorously derive the constants in the Savannah River Simulator (SARIS) core model from detailed, charge-design, diffusion theory solutions. This methodology is intended to replace the ill-defined, ad hoc iterative process used in the past to generate these constants. Along the development path, three shortcomings of the current core model were identified and corrected. The updated core model with revised constants is termed the second generation core model. In addition, changes in the decay heat and delayed neutron precursor models are also recommended, all in the interest of improving simulator neutronics fidelity.

  13. MIC-SVM: Designing A Highly Efficient Support Vector Machine For Advanced Modern Multi-Core and Many-Core Architectures

    SciTech Connect

    You, Yang; Song, Shuaiwen; Fu, Haohuan; Marquez, Andres; Mehri Dehanavi, Maryam; Barker, Kevin J.; Cameron, Kirk; Randles, Amanda; Yang, Guangwen

    2014-08-16

    Support Vector Machine (SVM) has been widely used in data-mining and Big Data applications as modern commercial databases start to attach an increasing importance to the analytic capabilities. In recent years, SVM was adapted to the field of High Performance Computing for power/performance prediction, auto-tuning, and runtime scheduling. However, even at the risk of losing prediction accuracy due to insufficient runtime information, researchers can only afford to apply offline model training to avoid significant runtime training overhead. To address the challenges above, we designed and implemented MICSVM, a highly efficient parallel SVM for x86 based multi-core and many core architectures, such as the Intel Ivy Bridge CPUs and Intel Xeon Phi coprocessor (MIC).

  14. Neutron and Gamma-ray Measurements

    SciTech Connect

    Krasilnikov, Anatoly V.; Sasao, Mamiko; Kaschuck, Yuri A.; Kiptily, Vasily G.; Popovichev, Sergey V.; Nishitani, Takeo; Bertalot, Luciano

    2008-03-12

    Due to high neutron and gamma-ray yields and large size plasmas many future fusion reactor plasma parameters such as fusion power, fusion power density, ion temperature, fuel mixture, fast ion energy and spatial distributions can be well measured by various fusion product diagnostics. Neutron diagnostics provide information on fusion reaction rate, which indicates how close is the plasma to the ultimate goal of nuclear fusion and fusion power distribution in the plasma core, which is crucial for optimization of plasma breakeven and burn. Depending on the plasma conditions neutron and gamma-ray diagnostics can provide important information, namely about dynamics of fast ion energy and spatial distributions during neutral beam injection, ion cyclotron heating and generated by fast ions MHD instabilities. The influence of the fast particle population on the 2-D neutron source profile was clearly demonstrated in JET experiments. 2-D neutron and gamma-ray source measurements could be important for driven plasma heating profile optimization in fusion reactors. To meat the measurement requirements in ITER the planned set of neutron and gamma ray diagnostics includes radial and vertical neutron and gamma cameras, neutron flux monitors, neutron activation systems and neutron spectrometers. The necessity of using massive radiation shielding strongly influences the diagnostic designs in fusion reactor, determines angular fields of view of neutron and gamma-ray cameras and spectrometers and gives rise to unavoidable difficulties in the absolute calibration. The development, testing in existing tokomaks and a possible engineering integration of neuron and gamma-ray diagnostic systems into ITER are presented.

  15. Neutron and Gamma-ray Measurements

    NASA Astrophysics Data System (ADS)

    Krasilnikov, Anatoly V.; Sasao, Mamiko; Kaschuck, Yuri A.; Kiptily, Vasily G.; Nishitani, Takeo; Popovichev, Sergey V.; Bertalot, Luciano

    2008-03-01

    Due to high neutron and gamma-ray yields and large size plasmas many future fusion reactor plasma parameters such as fusion power, fusion power density, ion temperature, fuel mixture, fast ion energy and spatial distributions can be well measured by various fusion product diagnostics. Neutron diagnostics provide information on fusion reaction rate, which indicates how close is the plasma to the ultimate goal of nuclear fusion and fusion power distribution in the plasma core, which is crucial for optimization of plasma breakeven and burn. Depending on the plasma conditions neutron and gamma-ray diagnostics can provide important information, namely about dynamics of fast ion energy and spatial distributions during neutral beam injection, ion cyclotron heating and generated by fast ions MHD instabilities. The influence of the fast particle population on the 2-D neutron source profile was clearly demonstrated in JET experiments. 2-D neutron and gamma-ray source measurements could be important for driven plasma heating profile optimization in fusion reactors. To meat the measurement requirements in ITER the planned set of neutron and gamma ray diagnostics includes radial and vertical neutron and gamma cameras, neutron flux monitors, neutron activation systems and neutron spectrometers. The necessity of using massive radiation shielding strongly influences the diagnostic designs in fusion reactor, determines angular fields of view of neutron and gamma-ray cameras and spectrometers and gives rise to unavoidable difficulties in the absolute calibration. The development, testing in existing tokomaks and a possible engineering integration of neuron and gamma-ray diagnostic systems into ITER are presented.

  16. Small-angle x-ray and neutron scattering studies of the volume phase transition in thermosensitive core-shell colloids

    NASA Astrophysics Data System (ADS)

    Seelenmeyer, S.; Deike, I.; Rosenfeldt, S.; Norhausen, Ch.; Dingenouts, N.; Ballauff, M.; Narayanan, T.; Lindner, P.

    2001-06-01

    The volume transition in thermosensitive colloidal core-shell particles is investigated by small-angle x-ray scattering (SAXS), small-angle Neutron scattering (SANS), and dynamic light scattering (DLS). The latex particles are dispersed in water and consist of a solid poly(styrene) core with a diameter of 100 nm. The thermosensitive shell is made up of poly(N-isopropylacrylamide) (PNIPA) chains crosslinked by 2.5 mol % N,N'-methylenbisacrylamide (BIS). Water is a good solvent for PNIPA at room temperature but becomes a poor solvent above 32 °C. The PNIPA network of the shell undergoes a volume transition at this temperature. As a result the diameter of the particle shrinks. The scattering intensities of the particles measured by SAXS and SANS as a function of temperature may be decomposed into a part deriving from the overall structure and a part originating from the fluctuations within the network. The analysis of the overall structure leads to the volume fraction of the swollen network at different temperatures. SANS in conjunction with contrast variation demonstrates that the network is confined in a well-defined shell. SAXS and SANS data therefore allow the phase diagram of the network in the shell of the particles to be derived, i.e., the average volume fraction of the network in the shell can be determined as a function of temperature. DLS corroborates this result but demonstrates that there is a small fraction of chains exceeding the outer radius derived from SAXS and SANS. The static intensity caused by the fluctuations of the network becomes the leading contribution at high scattering angles. SAXS data show that this part can be described by a Lorentzian both below and above the volume transition. The analysis demonstrates that critical fluctuations of the network around the transition temperature are fully suppressed. This finding is explained by the strong steric constraint of the network by its confinement within a shell of colloidal dimension. The

  17. Design of a magnetic shielding system for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

    SciTech Connect

    Cui, Z. Q.; Chen, Z. J.; Xie, X. F.; Peng, X. Y.; Hu, Z. M.; Du, T. F.; Ge, L. J.; Zhang, X.; Yuan, X.; Fan, T. S.; Chen, J. X.; Li, X. Q. E-mail: guohuizhang@pku.edu.cn; Zhang, G. H. E-mail: guohuizhang@pku.edu.cn; Xia, Z. W.; Hu, L. Q.; Zhong, G. Q.; Lin, S. Y.; Wan, B. N.

    2014-11-15

    The novel neutron spectrometer TOFED (Time of Flight Enhanced Diagnostics), comprising 90 individual photomultiplier tubes coupled with 85 plastic scintillation detectors through light guides, has been constructed and installed at Experimental Advanced Superconducting Tokamak. A dedicated magnetic shielding system has been constructed for TOFED, and is designed to guarantee the normal operation of photomultiplier tubes in the stray magnetic field leaking from the tokamak device. Experimental measurements and numerical simulations carried out employing the finite element method are combined to optimize the design of the magnetic shielding system. The system allows detectors to work properly in an external magnetic field of 200 G.

  18. Design and commissioning of a high magnetic field muon spin relaxation spectrometer at the ISIS pulsed neutron and muon source.

    PubMed

    Lord, J S; McKenzie, I; Baker, P J; Blundell, S J; Cottrell, S P; Giblin, S R; Good, J; Hillier, A D; Holsman, B H; King, P J C; Lancaster, T; Mitchell, R; Nightingale, J B; Owczarkowski, M; Poli, S; Pratt, F L; Rhodes, N J; Scheuermann, R; Salman, Z

    2011-07-01

    The high magnetic field (HiFi) muon instrument at the ISIS pulsed neutron and muon source is a state-of-the-art spectrometer designed to provide applied magnetic fields up to 5 T for muon studies of condensed matter and molecular systems. The spectrometer is optimised for time-differential muon spin relaxation studies at a pulsed muon source. We describe the challenges involved in its design and construction, detailing, in particular, the magnet and detector performance. Commissioning experiments have been conducted and the results are presented to demonstrate the scientific capabilities of the new instrument.

  19. Design of a magnetic shielding system for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak.

    PubMed

    Cui, Z Q; Chen, Z J; Xie, X F; Peng, X Y; Hu, Z M; Du, T F; Ge, L J; Zhang, X; Yuan, X; Xia, Z W; Hu, L Q; Zhong, G Q; Lin, S Y; Wan, B N; Fan, T S; Chen, J X; Li, X Q; Zhang, G H

    2014-11-01

    The novel neutron spectrometer TOFED (Time of Flight Enhanced Diagnostics), comprising 90 individual photomultiplier tubes coupled with 85 plastic scintillation detectors through light guides, has been constructed and installed at Experimental Advanced Superconducting Tokamak. A dedicated magnetic shielding system has been constructed for TOFED, and is designed to guarantee the normal operation of photomultiplier tubes in the stray magnetic field leaking from the tokamak device. Experimental measurements and numerical simulations carried out employing the finite element method are combined to optimize the design of the magnetic shielding system. The system allows detectors to work properly in an external magnetic field of 200 G.

  20. Design and commissioning of a high magnetic field muon spin relaxation spectrometer at the ISIS pulsed neutron and muon source

    SciTech Connect

    Lord, J. S.; McKenzie, I.; Baker, P. J.; Cottrell, S. P.; Giblin, S. R.; Hillier, A. D.; Holsman, B. H.; King, P. J. C.; Nightingale, J. B.; Pratt, F. L.; Rhodes, N. J.; Blundell, S. J.; Lancaster, T.; Good, J.; Mitchell, R.; Owczarkowski, M.; Poli, S.; Scheuermann, R.; Salman, Z.

    2011-07-15

    The high magnetic field (HiFi) muon instrument at the ISIS pulsed neutron and muon source is a state-of-the-art spectrometer designed to provide applied magnetic fields up to 5 T for muon studies of condensed matter and molecular systems. The spectrometer is optimised for time-differential muon spin relaxation studies at a pulsed muon source. We describe the challenges involved in its design and construction, detailing, in particular, the magnet and detector performance. Commissioning experiments have been conducted and the results are presented to demonstrate the scientific capabilities of the new instrument.