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. 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 visible impacts on neutronic characteristics compared with reference full UOX core. The fuel cycle analysis has shown that {sup 233}U mono-recycling with U{sub 3}ThOX fuel could save 13% of natural uranium resource compared with UOX once through fuel cycle, slightly more than that of Plutonium single-recycling with MOX fuel. If {sup 233}U multi-recycling with U{sub 3}ThOX fuel is implemented, more natural uranium resource would be saved. (authors)

  4. 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.

  5. 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.

  6. 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)

  7. Neutronics methods, models, and applications at the Idaho National Engineering Laboratory for the advanced neutron source reactor three-element core design

    SciTech Connect

    Wemple, C.A.; Schnitzler, B.G.; Ryskamp, J.M.

    1995-08-01

    A summary of the methods and models used to perform neutronics analyses on the Advanced Neutron Source reactor three-element core design is presented. The applications of the neutral particle Monte Carlo code MCNP are detailed, as well as the expansion of the static role of MCNP to analysis of fuel cycle depletion calculations. Results to date of these applications are presented also. A summary of the calculations not yet performed is also given to provide a {open_quotes}to-do{close_quotes} list if the project is resurrected.

  8. The Neutronics Design and Analysis of a 200-MW(electric) Simplified Boiling Water Reactor Core

    SciTech Connect

    Tinkler, Daniel R.; Downar, Thomas J.

    2003-06-15

    A 200-MW(electric) simplified boiling water reactor (SBWR) was designed and analyzed under sponsorship of the U.S. Department of Energy Nuclear Energy Research Initiative program. The compact size of a 200-MW(electric) reactor makes it attractive for countries with a less well developed engineering infrastructure, as well as for developed countries seeking to tailor generation capacity more closely to the growth of their electricity demand. The 200-MW(electric) core design reported here is based on the 600-MW(electric) General Electric SBWR core, which was first analyzed in the work performed here in order to qualify the computer codes used in the analysis. Cross sections for the 8 x 8 fuel assembly design were generated with the HELIOS lattice physics code, and core simulation was performed with the U.S. Nuclear Regulatory Commission codes RELAP5/PARCS. In order to predict the critical heat flux, the Hench-Gillis correlation was implemented in the RELAP5 code. An equilibrium cycle was designed for the 200-MW(electric) core, which provided a cycle length of more than 2 yr and satisfied the minimum critical power ratio throughout the core life.

  9. 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.

  10. Neutronics design

    SciTech Connect

    Moir, R.

    1984-10-01

    Initial scoping calculations were done by Lee at LLNL with the TART code and ENDL data to determine the tritium breeding potential of this blanket type. A radially zoned cylindrical nucleonics model was used and is described. Results, local (100% blanket coverage) T and M vs Be zone thickness, are shown. The tritium breeding ratio, T, is seen to vary between 0.5 with no Be to 1.7 with a 60-cm Be zone. Correspondingly, energy multiplication, M, varies between 1.1 and 1.4. The effects of less than 100% blanket coverage on T is shown. For example, if the effective coverage is only 80, a 15-cm Be zone is needed for T = 1.01 compared to 10 cm at full coverage. Higher T can be achieved, of course, by increasing the Be zone thickness. Another possibly attractive use of the excess neutrons generated in Be is for higher M. While this was not the objective here it is clearly possible to include material in the blanket with significantly higher Q's than 4.8 MeV for the Li6(n,t) reaction. Also enriching the Li in Li6 can increase T.

  11. RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design

    SciTech Connect

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

    1996-02-01

    In order to utilize reduced enrichment fuel, the three-element-core design 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. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangers are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components.

  12. Core Design Applications

    Energy Science and Technology Software Center (ESTSC)

    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).

  13. 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 modeling, the amount of heat generated by the fuel is assumed to be transferred totally into the coolant. Therefore, the surface heat flux is applied to the fuel cladding outer surface by considering the depleted fuel composition of each individual fuel rod under a reference core loading condition defined as; 53H at 1MW full power. In order to model the entire PSBR reactor, fine mesh discretization was achieved with 22 millions structured and unstructured computational meshes. The conductive heat transfer inside the fuel rods was ignored in order to decrease the computational mesh requirement. Since the PSBR core operates in the subcooled nucleate boiling region, the CFD simulation of new PSBR design was completed utilizing an Eulerian-Eulerian multiphase flow formulation and RPI wall boiling model. The simulation results showed that the new moderator tank geometry results in secondary flow entering into the core due to decrease in the cross-flow area. Notably, the radial flow improves the local heat transfer conditions by providing radial-mixing in the core. Bubble nucleation occurs on the heated fuel rods but bubbles are collapsing in the subcooled fluid. Furthermore, the bulk fluid properties are not affected by the bubble formation. Yet, subcooled boiling enhances the heat transfer on the fuel rods. Five neutron beam ports are designed for the new reactor. The geometrical configuration, filter and collimator system designs of each neutron beam ports are selected based on the requirements of the experimental facilities. A cold neutron beam port which utilizes cold neutrons from three curved guide tubes is considered. Therefore, there will be seven neutron beams available in the new facility. The neutronic analyses of the new beam port designs were achieved by using MCNP5 code and Burned Coupled Simulation Tool for the PSBR. The MCNP simulation results showed that thermal neutron flux was increased by a factor of minimum 1.23 times and maximum 2.68 times in the new beam port compared to the existing BP4 design. Besides total gamma dose was decreased by a factor of 100 in the new neutron beam facility. Therefore, this study assures both the thermal-hydraulics safety and neutronics performance improvements in the new PSBR design. (Abstract shortened by UMI.)

  14. 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.

  15. 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-lived fission product incineration has not been considered in the studies reported here.)

  16. 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.

  17. 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)

  18. The OECD/NEA/NSC PBMR coupled neutronics/thermal hydraulics transient benchmark: The PBMR-400 core design

    SciTech Connect

    Reitsma, F.; Ivanov, K.; Downar, T.; De Haas, H.; Gougar, H. D.

    2006-07-01

    The Pebble Bed Modular Reactor (PBMR) is a High-Temperature Gas-cooled Reactor (HTGR) concept to be built in South Africa. As part of the verification and validation program the definition and execution of code-to-code benchmark exercises are important. The Nuclear Energy Agency (NEA) of the Organisation for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transient benchmark problem in its program. The OECD benchmark defines steady-state and transients cases, including reactivity insertion transients. It makes use of a common set of cross sections (to eliminate uncertainties between different codes) and includes specific simplifications to the design to limit the need for participants to introduce approximations in their models. In this paper the detailed specification is explained, including the test cases to be calculated and the results required from participants. (authors)

  19. Passive Neutron Dosemeter Design

    NASA Astrophysics Data System (ADS)

    Becerra-Ferrerio, Ana Mara; Vega-Carrillo, Hctor Ren

    2002-08-01

    A passive neutron dosemeter was designed to be used in mixed radiation fields. The design was carried out using Monte Carlo method. The dosemeter model was a 25.4 cmdiameter polyethylene sphere with a thermoluminescent dosemeter, TLD600, located at the sphere center. This model was irradiated with 50 monoenergetic neutron sources with energies from 10-8 to 20 MeV. A 506.71 cm2-area disk was used to model the source term whose center was located at 100 cm from polyethylene sphere's center. The dosemeter response was compared with the responses of SNOOPY, Harwell 95/0075 and PNR-4. With these responses it was calculated the dosemeter responses for 252Cf, 252Cf/D2O and 239PuBe neutron sources. The passive dosemeter relative response has the same shape of SNOOPY, Harwell 95/0075 and PNR-4 dosemeters. Due to the type of thermal neutron detector used in the passive dosemeter the absolute response per unit fluence, is lower than the absolute response of SNOOPY, Harwell 95/0075 and PNR-4 dosemeters. However the passive dosemeter response in function of the average neutron energy of the 252Cf, 252Cf/D2O and 239PuBe neutron energy was more linear.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Neutron beam design, development, and performance for neutron capture therapy

    SciTech Connect

    Harling, O.K.; Bernard, J.A. ); Zamenhof, R.G. )

    1990-01-01

    The report presents topics presented at a workshop on neutron beams and neutron capture therapy. Topics include: neutron beam design; reactor-based neutron beams; accelerator-based neutron beams; and dosimetry and treatment planning. Individual projects are processed separately for the databases. (CBS)

  5. Simplifier cut core inductor design

    NASA Technical Reports Server (NTRS)

    Mclyman, W. T.

    1976-01-01

    Advantages of specifying C cores and cut toroids fabricated from grain oriented silicon steels for use in high frequency power converters and pulse width modulated switching regulators are discussed. A method for rating cores assigns to each core a number which is the product of its window and core cross section area, called 'Area Product A sub p.' A correlation between the A sub p numbers and current density for a given temperature rise was developed. Also, straight line relationships were developed for A sub p and volume, A sub p and surface area, and A sub p and weight. These relationships can be used to simplify and standardize the process of inductor design. They also make it possible to design inductors of small bulk and volume or to optimize efficiency.

  6. 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.

  7. The spin evolution of neutron stars with the superfluid core

    NASA Astrophysics Data System (ADS)

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

    2013-06-01

    We investigate the neutron stars spin evolution (breaking, inclination angle evolution and radiative precession), taking into account the superfluidity of the neutrons in the star core. The neutron star is treated as a two-component system consisting of a `charged' component (including the crust and the core protons, electrons and normal neutrons) and a core superfluid neutron component. The components are supposed to interact through the mutual friction force. We assume that the `charged' component rotates rigidly. The neutron superfluid velocity field is calculated directly from linearized hydrodynamical equations. It is shown that the superfluid core accelerates the evolution of inclination angle and makes all pulsars evolve to either the orthogonal or coaxial state. However, rapid evolution seems to contradict the observation data. Obtained results together with the observations may allow us to examine the superfluid models.

  8. 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.

  9. Heavy water breeder conceptual core design

    SciTech Connect

    Radkowsky, A.

    1981-12-01

    A reference design was developed for a close packed pressurized heavy water cooled and moderated breeder core, operating on the uranium plutonium cycle. The seed-blanket concept was found to be most advantageous. The core is composed of a large number of identical seed-blanket units, each of which is controlled by moving the seed, which is zoned in such a way as to vary the leakage of neutrons between the seed and blanket. The control method is known as Geometry Control. In order to assure a negative void coefficient at all times during life it was necessary to dilute the heavy water with 20% light water. Initial fuel loading was approximately 4700 kg of fissionable plutonium of recycle composition. This leading was sufficient for approximately a two-year lifetime, at the end of which the Fissile Inventory Ratio (FIR) was 1.05. (FIR is the ratio of fissile fuel in the core at any time to that at the beginning of core life). A thermal hydraulic analysis indicated that the core could fit into the pressure vessel of a standard pressurized light reactor and would meet safety requirements in case of a LOCA.

  10. 400 MWth gas cooled ADT core neutronic benchmark

    SciTech Connect

    Chabert, C.; Rimpault, G.; Peneliau, Y.; Tommasi, J.; Plisson-Rieunier, D.; da Cruz, D.F.; Malambu, E.; Rineiski, A.

    2007-07-01

    In France, the 'December 1991 act' has generated much R and D on nuclear waste management. One of the main conclusions is that fast reactors offer the best performances to transmute minor actinides. Two types of fast spectrum reactors can be used: critical or sub-critical ones (in this case, Accelerator Driven systems, ADS). Within this scope, a Gas-cooled ADT core neutronic benchmark has been proposed by CEA for an IAEA Coordinated Research Project (CRP) on 'Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste'. This will help clarifying the future issues associated to the improvement of the core designs with more reliable and accurate tools. The benchmark exercise is based on a 400 MWth gas-cooled ADT with fertile-free fuel (Pu, Np, Am, Cm)O{sub 2} within MgO matrix. (authors)

  11. 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.

  12. Persistent crust-core spin lag in neutron stars

    NASA Astrophysics Data System (ADS)

    Glampedakis, Kostas; Lasky, Paul D.

    2015-06-01

    It is commonly believed that the magnetic field threading a neutron star provides the ultimate mechanism (on top of fluid viscosity) for enforcing long-term corotation between the slowly spun-down solid crust and the liquid core. We show that this argument fails for axisymmetric magnetic fields with closed field lines in the core, the commonly used `twisted torus' field being the most prominent example. The failure of such magnetic fields to enforce global crust-core corotation leads to the development of a persistent spin lag between the core region occupied by the closed field lines and the rest of the crust and core. We discuss the repercussions of this spin lag for the evolution of the magnetic field, suggesting that, in order for a neutron star to settle to a stable state of crust-core corotation, the bulk of the toroidal field component should be deposited into the crust soon after the neutron star's birth.

  13. 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

  14. 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

  15. 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.

  16. Rotating neutron stars with exotic cores: masses, radii, stability

    NASA Astrophysics Data System (ADS)

    Haensel, P.; Bejger, M.; Fortin, M.; Zdunik, L.

    2016-03-01

    A set of theoretical mass-radius relations for rigidly rotating neutron stars with exotic cores, obtained in various theories of dense matter, is reviewed. Two basic observational constraints are used: the largest measured rotation frequency (716Hz) and the maximum measured mass (2 M ⊙ . The present status of measuring the radii of neutron stars is described. The theory of rigidly rotating stars in general relativity is reviewed and limitations of the slow rotation approximation are pointed out. Mass-radius relations for rotating neutron stars with hyperon and quark cores are illustrated using several models. Problems related to the non-uniqueness of the crust-core matching are mentioned. Limits on rigid rotation resulting from the mass-shedding instability and the instability with respect to the axisymmetric perturbations are summarized. The problem of instabilities and of the back-bending phenomenon are discussed in detail. Metastability and instability of a neutron star core in the case of a first-order phase transition, both between pure phases, and into a mixed-phase state, are reviewed. The case of two disjoint families (branches) of rotating neutron stars is discussed and generic features of neutron-star families and of core-quakes triggered by the instabilities are considered.

  17. Design configuration of GCFR core assemblies

    SciTech Connect

    LaBar, M.P.; Lee, G.E.; Meyer, R.J.

    1980-05-01

    The current design configurations of the core assemblies for the gas-cooled fast reactor (GCFR) demonstration plant reactor core conceptual design are described. Primary emphasis is placed upon the design innovations that have been incorporated in the design of the core assemblies since the establishment of the initial design of an upflow GCFR core. A major feature of the design configurations is that they are prototypical of core assemblies for use in commercial plants; a larger number of the same assemblies would be used in a commercial plant.

  18. Design of a Neutron Source for Calibration

    NASA Astrophysics Data System (ADS)

    Carrillo-Nuñez, Aureliano; Vega-Carrillo, Héctor René

    2002-08-01

    The neutron spectra produced by an isotopic neutron source located at the center of moderating media were calculated using Monte Carlo method in the aim to design a neutron source for calibration purposes. To improve the evaluation of the dosimetric quantities, is recomended to calibrate the radiation protection devices with calibrated neutron sources whose neutron spectra being similar to those met in practice. Here, a 239Pu-Be neutron source was inserted in H2O, D2O and polyethylene cylindrical moderators in order to produce neutron spectra that resembles spectra found in workplaces.

  19. 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.

  20. 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)

  1. Nodal weighting factor method for ex-core fast neutron fluence evaluation

    SciTech Connect

    Chiang, R. T.

    2012-07-01

    The nodal weighting factor method is developed for evaluating ex-core fast neutron flux in a nuclear reactor by utilizing adjoint neutron flux, a fictitious unit detector cross section for neutron energy above 1 or 0.1 MeV, the unit fission source, and relative assembly nodal powers. The method determines each nodal weighting factor for ex-core neutron fast flux evaluation by solving the steady-state adjoint neutron transport equation with a fictitious unit detector cross section for neutron energy above 1 or 0.1 MeV as the adjoint source, by integrating the unit fission source with a typical fission spectrum to the solved adjoint flux over all energies, all angles and given nodal volume, and by dividing it with the sum of all nodal weighting factors, which is a normalization factor. Then, the fast neutron flux can be obtained by summing the various relative nodal powers times the corresponding nodal weighting factors of the adjacent significantly contributed peripheral assembly nodes and times a proper fast neutron attenuation coefficient over an operating period. A generic set of nodal weighting factors can be used to evaluate neutron fluence at the same location for similar core design and fuel cycles, but the set of nodal weighting factors needs to be re-calibrated for a transition-fuel-cycle. This newly developed nodal weighting factor method should be a useful and simplified tool for evaluating fast neutron fluence at selected locations of interest in ex-core components of contemporary nuclear power reactors. (authors)

  2. Core momentum distribution in two-neutron halo nuclei

    NASA Astrophysics Data System (ADS)

    Souza, L. A.; Bellotti, F. F.; Yamashita, M. T.; Frederico, T.; Tomio, Lauro

    2016-06-01

    The core momentum distribution of a weakly-bound neutron-neutron-core exotic nucleus is computed within a renormalized zero-range three-body model, with interactions in the s-wave channel. The halo wave-function in momentum space is obtained by using as inputs the two-body scattering lengths and the two-neutron separation energy. The core momentum densities are computed for 11Li, 14Be 20C and 22C. The model describes the experimental data for 11Li, 14Be and to some extent 20C. The recoil momentum distribution of the 20C from the breakup of 22C nucleus is computed for different two-neutron separation energies, and from the comparison with recent experimental data the two-neutron separation energy is estimated in the range 100 ≲S2n ≲ 400 keV. The recoil momentum distribution depends weakly on the neutron-20C scattering length, while the matter radius is strongly sensitive to it. The expected universality of the momentum distribution width is verified by also considering excited states for the system.

  3. LOCV approach and core-crust transition in neutron stars

    NASA Astrophysics Data System (ADS)

    Bigdeli, M.; Elyasi, S.

    2015-03-01

    In this paper, we have calculated the core-crust transition parameters and the location of inner edge for crust in the neutron stars. We have also investigated the structural properties of neutron stars, such as mass and radius for the core and crust, the moment of inertia, and its crustal fraction. Here we have employed the lowest-order constrained variational approach and used the UV14 + TNI and AV18 potentials to compute the equation of state of nuclear matter. Finally, we have compared our results with those of other techniques.

  4. 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.

  5. 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)

  6. 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 base isolation, and 4) decay heat powered back-up electricity generation. The report provides a preconceptual design of the manipulators, the fuel transfer system, and the salt transfer loops. The mechanical handling of the fuel and how it is accomplished without instrumentation inside the salt is described within the report. All drives for the manipulators reside outside the reactor top flange. The design has also taken into account the transportability of major components and how they will be assembled on site

  7. 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 concept incorporates fully passive responses to all identified design basis or non-very-low frequency beyond design basis accidents as well as multiple levels of radioactive material containment. 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 base isolation, and (4) decay heat powered back-up electricity generation.

  8. Induced gravitational collapse in FeCO Core-Neutron star binaries and Neutron star-Neutron star binary mergers

    NASA Astrophysics Data System (ADS)

    Ruffini, R.; Aimuratov, Y.; Bianco, C. L.; Enderli, M.; Kovacevic, M.; Moradi, R.; Muccino, M.; Penacchioni, A. V.; Pisani, G. B.; Rueda, J. A.; Wang, Y.

    2015-10-01

    We review the recent progress in understanding the nature of gamma-ray bursts (GRBs). The occurrence of GRB is explained by the Induced Gravitational Collapse (IGC) in FeCO Core-Neutron star binaries and Neutron star-Neutron star binary mergers, both processes occur within binary system progenitors. Making use of this most unexpected new paradigm, with the fundamental implications by the neutron star (NS) critical mass, we find that different initial configurations of binary systems lead to different GRB families with specific new physical predictions confirmed by observations.

  9. Shield Design for a Space Based Vapor Core Reactor

    SciTech Connect

    Knight, Travis; Anghaie, Samim

    2002-07-01

    Innovative shielding strategies were sought to reduce the mass of the required shielding for a space based vapor core reactor system with magnetohydrodynamic energy conversion. Gamma-rays directly resultant from fission were found to play no role in the dose rate, while secondary gamma-rays from fission neutron interactions were the dominant contributor to the dose rate. Hydrogen containing materials such as polyethylene were utilized to provide shielding of both radiation from the reactor complex and also solar and galactic cosmic radiation. This shield design was found to contribute 0.125 kg/kWe to the baseline vapor core reactor system specific mass. (authors)

  10. 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 superconducting material," said Peter Shternin of the Ioffe Institute in St Petersburg, Russia, leader of a team with a paper accepted in the journal Monthly Notices of the Royal Astronomical Society. Both teams show that this rapid cooling is explained by the formation of a neutron superfluid in the core of the neutron star within about the last 100 years as seen from Earth. The rapid cooling is expected to continue for a few decades and then it should slow down. "It turns out that Cas A may be a gift from the Universe because we would have to catch a very young neutron star at just the right point in time," said Page's co-author Madappa Prakash, from Ohio University. "Sometimes a little good fortune can go a long way in science." The onset of superfluidity in materials on Earth occurs at extremely low temperatures near absolute zero, but in neutron stars, it can occur at temperatures near a billion degrees Celsius. Until now there was a very large uncertainty in estimates of this critical temperature. This new research constrains the critical temperature to between one half a billion to just under a billion degrees. Cas A will allow researchers to test models of how the strong nuclear force, which binds subatomic particles, behaves in ultradense matter. These results are also important for understanding a range of behavior in neutron stars, including "glitches," neutron star precession and pulsation, magnetar outbursts and the evolution of neutron star magnetic fields. Small sudden changes in the spin rate of rotating neutron stars, called glitches, have previously given evidence for superfluid neutrons in the crust of a neutron star, where densities are much lower than seen in the core of the star. This latest news from Cas A unveils new information about the ultra-dense inner region of the neutron star. "Previously we had no idea how extended superconductivity of protons was in a neutron star," said Shternin's co-author Dmitry Yakovlev, also from the Ioffe Institute. The cooling in the Cas A neutron star was first discovered by co-author Craig Heinke, from the University of Alberta, Canada, and Wynn Ho from the University of Southampton, UK, in 2010. It was the first time that astronomers have measured the rate of cooling of a young neutron star. Page's co-authors were Prakash, James Lattimer (State University of New York at Stony Brook), and Andrew Steiner (Michigan State University.) Shternin's co-authors were Yakovlev, Heinke, Ho, and Daniel Patnaude (Harvard-Smithsonian Center for Astrophysics.) More information, including images and other multimedia, can be found at: http://chandra.harvard.edu and http://chandra.nasa.gov

  11. Hyperons in neutron stars and supernova cores

    NASA Astrophysics Data System (ADS)

    Oertel, Micaela; Gulminelli, Francesca; Providência, Constança; Raduta, Adriana R.

    2016-03-01

    The properties of compact stars and their formation processes depend on many physical ingredients. The composition and the thermodynamics of the involved matter is one of them. We will investigate here uniform strongly interacting matter at densities and temperatures, where potentially other components than free nucleons appear such as hyperons, mesons or even quarks. In this paper we will put the emphasis on two aspects of stellar matter with non-nucleonic degrees of freedom. First, we will study the phase diagram of baryonic matter with strangeness, showing that the onset of hyperons, as that of quark matter, could be related to a very rich phase structure with a large density domain covered by phase coexistence. Second, we will investigate thermal effects on the equation of state (EoS), showing that they favor the appearance of non-nucleonic particles. We will finish by reviewing some recent results on the impact of non-nucleonic degrees freedom in compact star mergers and core-collapse events, where thermal effects cannot be neglected.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. Preliminary design concepts for the advanced neutron source reactor systems

    SciTech Connect

    Peretz, F.J.

    1988-01-01

    This paper describes the initial design work to develop the reactor systems hardware concepts for the advanced neutron source (ANS) reactor. This project has not yet entered the conceptual design phase; thus, design efforts are quite preliminary. This paper presents the collective work of members of the Oak Ridge National Laboratory, Martin Marietta Energy Systems, Inc., Engineering Division, and other participating organizations. The primary purpose of this effort is to show that the ANS reactor concept is realistic from a hardware standpoint and to show that project objectives can be met. It also serves to generate physical models for use in neutronic and thermal-hydraulic core design efforts and defines the constraints and objectives for the design. Finally, this effort will develop the criteria for use in the conceptual design of the reactor.

  18. Neutronic Design Studies for the National Spallation Neutron Source (NSNS)

    SciTech Connect

    Charlton, LA

    2001-08-01

    Neutronics analyses are now in progress to support initial selection of target system design features, materials, geometry, and component sizes for the proposed Spallation Neutron Source (SNS). Calculations have been performed to determine the neutron, proton, heavy ion, and gamma-ray flux spectra as a function of time, energy, and space for the major components of the target station (target, moderators, reflectors, etc.). These analyses were also performed to establish an initial set of performance characteristics for the neutron source. The methodology, reference performance characteristics, and results of initial optimization studies involving moderator poison plate location, target material performance, reflector performance, moderator position and premoderator performance for the target system are presented in this paper.

  19. 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)

  20. 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.

  1. CHINA SPALLATION NEUTRON SOURCE DESIGN.

    SciTech Connect

    WEI,J.

    2007-01-29

    The China Spallation Neutron Source (CSNS) is an accelerator-based high-power project currently in preparation under the direction of the Chinese Academy of Sciences (CAS). The complex is based on an H- linear accelerator, a rapid cycling proton synchrotron accelerating the beam to 1.6 GeV, a solid tungsten target station, and five initial instruments for spallation neutron applications. The facility will operate at 25 Hz repetition rate with a phase-I beam power of about 120 kW. The major challenge is to build a robust and reliable user's facility with upgrade potential at a fractional of ''world standard'' cost.

  2. Simulated magnetic field expulsion in neutron star cores

    NASA Astrophysics Data System (ADS)

    Elfritz, J. G.; Pons, J. A.; Rea, N.; Glampedakis, K.; Viganò, D.

    2016-03-01

    The study of long-term evolution of neutron star (NS) magnetic fields is key to understanding the rich diversity of NS observations, and to unifying their nature despite the different emission mechanisms and observed properties. Such studies in principle permit a deeper understanding of the most important parameters driving their apparent variety, e.g. radio pulsars, magnetars, X-ray dim isolated NSs, gamma-ray pulsars. We describe, for the first time, the results from self-consistent magnetothermal simulations considering not only the effects of the Hall-driven field dissipation in the crust, but also adding a complete set of proposed driving forces in a superconducting core. We emphasize how each of these core-field processes drive magnetic evolution and affect observables, and show that when all forces are considered together in vectorial form, the net expulsion of core magnetic flux is negligible, and will have no observable effect in the crust (consequently in the observed surface emission) on megayear time-scales. Our new simulations suggest that strong magnetic fields in NS cores (and the signatures on the NS surface) will persist long after the crustal magnetic field has evolved and decayed, due to the weak combined effects of dissipation and expulsion in the stellar core.

  3. 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 subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000.

  4. 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.

  5. Contribution of fuel vibrations to ex-core neutron noise during the first and second fuel cycles of the Sequoyah-1 pressurized water reactor

    SciTech Connect

    Sweeney, F.J.; March-Leuba, J.; Smith, C.M.

    1984-01-01

    Noise measurements were performed during the first and second fuel cycles of the Sequoyah-1 pressurized water reactor (PWR) to observe long-term changes in the ex-core neutron signatures. Increases in the ex-core neutron noise amplitude were observed throughout the 0.1- to 50.0-Hz range. In-core noise measurements indicate that fuel assembly vibrations contribute significantly to the ex-core neutron noise at nearly all frequencies in this range, probably due to mechanical or acoustic coupling with other vibrating internal structures. Space-dependent kinetics calculations show that ex-core neutron noise induced by fixed-amplitude fuel assembly vibrations will increase over a fuel cycle because of soluble boron and fuel concentration changes associated with burnup. These reactivity effects can also lead to 180/sup 0/ phase shifts between cross-core detectors. We concluded that it may be difficult to separate the changes in neutron noise due to attenuation (shielding) effects of structural vibrations from changes due to reactivity effects of fuel assembly motion on the basis of neutron noise amplitude or phase information. Amplitudes of core support barrel vibrations inferred from ex-core neutron noise measurements using calculated scale factors are likely to have a high degree of uncertainty, since these scale factors usually do not account for neutron noise generated by fuel assembly vibrations. Modifications in fuel management or design may also lead to altered neutron noise signature behavior over a fuel cycle.

  6. The influence of core superfluidity on the neutron stars long-term rotation evolution

    NASA Astrophysics Data System (ADS)

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

    2013-08-01

    We investigate the evolution of neutron star rotation taking into account the superfluidity of the neutrons in the neutron star core. The neutron star is treated as a two-component system consisting of a charged component (including the crust and the core protons, electrons and normal neutrons) and a core superfluid neutron component. The components are supposed to interact through the mutual friction force. We assume that the charged component rotates rigidly. The neutron superfluid velocity field is calculated directly from linearized hydrodynamical equations. It is shown that the superfluid core accelerates the evolution of inclination angle and makes all pulsars evolve to orthogonal state. But as it is known from observations the rate of the angle evolution is not very high. Therefore, a small size of superfluid cores is more likely. These facts may allow to examine superfluid models.

  7. 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 heating. (authors)

  8. Reactor design of the advanced neutron source

    SciTech Connect

    Ryskamp, J.M. ); Selby, D.L.; Primm, R.T. III )

    1991-03-01

    The ongoing preconceptual and conceptual reactor design of the Advanced Neutron Source (ANS) is explored. The ANS is being designed for materials sciences, isotope production, and fundamental physics research. A reactor design based on previously developed technology can meet the performance requirements set by the user community for a new ANS to serve all fields of neutron science. These requirements include the capability of producing a peak thermal neutron flux over five times higher than that in use at any currently operating steady-state facility. Achievement of these ultrahigh flux levels involves many interesting aspects of reactor design. The reactor characteristics of the current preconceptual reference design are presented. The attainment of this design was reached by following a design strategy that best met the safety and user requirements. The design has evolved over the last 5 yr from two concepts proposed in 1985. The trade-offs and selection of many reactor parameters are described to illustrate how and why the current design was achieved. Further reactor design is planned, leading to an ANS operating by 1999 for use by scientists of many disciplines.

  9. Design and simulation of a neutron facility.

    PubMed

    Studenski, Matthew T; Kearfott, Kimberlee J

    2007-02-01

    State and other regulatory entities require that for any facility housing a particle accelerator the surrounding areas must be restricted to public access unless the dose equivalent rate is less than 0.02 mSv h at 5 cm from any accessible wall surrounding the facility under conditions of maximum radiation output. A Monte Carlo radiation transport simulation code, MCNP5, was used to design a proposed facility to shield two D-T neutron generators and one D-D neutron generator. A number of different designs were simulated, but due to cost and space issues a small concrete cave proved to be the best solution for the shielding problem. With this design, all of the neutron generators could be used and all of the rooms surrounding the neutron facility could be considered unrestricted to public access. To prevent unauthorized access into the restricted area of the neutron facility, light curtains, warning lights, door interlocks, and rope barriers will be built into the facility. PMID:17228186

  10. Advanced Neutron Source: The designer's perspective

    SciTech Connect

    Peretz, F.J.

    1990-01-01

    The Advanced Neutron Source (ANS) is a research facility based on a 350 MW beam reactor, to be brought into service at the Oak Ridge National Laboratory at the end of the century. The primary objective is to provide high-flux neutron beams and guides, with cold, thermal, hot, and ultra-cold neutrons, for research in many fields of science. Secondary objectives include isotopes production, materials irradiation and activation analysis. The design of the ANS is strongly influenced by the historical development of research and power reactor concepts, and of the regulatory infrastructure of the Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC). Current trends in reactor safety also impact the climate for the design of such a reactor.

  11. Design of a Thermal Neutron Beam for a New Neutron Imaging Facility at Tehran Research Reactor

    NASA Astrophysics Data System (ADS)

    Dastjerdi, Mohammad Hossein Choopan; Khalafi, Hossein

    A new neutron imaging facility will be built around the Tehran Research Reactor (TRR). The TRR is an open pool light water moderated5 MW research reactor with six beam tubes. The neutron energy spectrum near the reactor core at the entrance of the beam tube was measured by the foil activation method using the SAND-II code and calculated by the MCNP Monte Carlo code. There was a good similarity between calculated and simulated spectra. The principal component of this facility is its neutron collimator. The collimator is a beam-forming assembly which determines the geometric properties of the beam. In addition, it may contain filters to modify the energy spectrum or to reduce the gamma ray content of the beam. The optimum thickness of filters, the position of the aperture and other details of the neutron collimator were calculated using MCNP Monte Carlo simulations. In this design, the L/D ratio of this facility had the value of 120. The thermal neutron flux at the image plane was about 7.8106 n/cm2.s and n/? ratio about 106 n/cm2.?Sv.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. Scientific Design of the New Neutron Radiography Facility (SANRAD) at SAFARI-1 for South Africa

    NASA Astrophysics Data System (ADS)

    de Beer, F. C.; Gruenauer, F.; Radebe, J. M.; Modise, T.; Schillinger, B.

    The final scientific design for an upgraded neutron radiography/tomography facility at beam port no.2 of the SAFARI-1 nuclear research reactor has been performed through expert advice from Physics Consulting, FRMII in Germany and IPEN, Brazil. A need to upgrade the facility became apparent due to the identification of various deficiencies of the current SANRAD facility during an IAEA-sponsored expert mission of international scientists to Necsa, South Africa. A lack of adequate shielding that results in high neutron background on the beam port floor, a mismatch in the collimator aperture to the core that results in a high gradient in neutron flux on the imaging plane and due to a relative low L/D the quality of the radiographs are poor, are a number of deficiencies to name a few.The new design, based on results of Monte Carlo (MCNP-X) simulations of neutron- and gamma transport from the reactor core and through the new facility, is being outlined. The scientific design philosophy, neutron optics and imaging capabilities that include the utilization of fission neutrons, thermal neutrons, and gamma-rays emerging from the core of SAFARI-1 are discussed.

  17. BEAM-LOSS DRIVEN DESIGN OPTIMIZATION FOR THE SPALLATION NEUTRON SOURCE (SNS) RING.

    SciTech Connect

    WEI,J.; BEEBE-WANG,J.; BLASKIEWICZ,M.; CAMERON,P.; DANBY,G.; GARDNER,C.J.; JACKSON,J.; LEE,Y.Y.; LUDEWIG,H.; MALITSKY,N.; RAPARIA,D.; TSOUPAS,N.; WENG,W.T.; ZHANG,S.Y.

    1999-03-29

    This paper summarizes three-stage design optimization for the Spallation Neutron Source (SNS) ring: linear machine design (lattice, aperture, injection, magnet field errors and misalignment), beam core manipulation (painting, space charge, instabilities, RF requirements), and beam halo consideration (collimation, envelope variation, e-p issues etc.).

  18. Neutron flux measurements in the side-core region of Hunterston B advanced gas-cooled reactor

    SciTech Connect

    Allen, D.A.; Shaw, S.E.; Huggon, A.P.; Steadman, R.J.; Thornton, D.A.; Whiley, G.S.

    2011-07-01

    The core restraints of advanced gas-cooled reactors are important structural components that are required to maintain the geometric integrity of the cores. A review of neutron dosimetry for the sister stations Hunterston B and Hinkley Point B identified that earlier conservative assessments predicted high thermal neutron dose rates to key components of the restraint structure (the restraint rod welds), with the implication that some of them may be predicted to fail during a seismic event. A revised assessment was therefore undertaken [Thornton, D. A., Allen, D. A., Tyrrell, R. J., Meese, T. C., Huggon, A.P., Whiley, G. S., and Mossop, J. R., 'A Dosimetry Assessment for the Core Restraint of an Advanced Gas Cooled Reactor,' Proceedings of the 13. International Symposium on Reactor Dosimetry (ISRD-13, May 2008), World Scientific, River Edge, NJ, 2009, W. Voorbraak, L. Debarberis, and P. D'hondt, Eds., pp. 679-687] using a detailed 3D model and a Monte Carlo radiation transport program, MCBEND. This reassessment resulted in more realistic fast and thermal neutron dose recommendations, the latter in particular being much lower than had been thought previously. It is now desirable to improve confidence in these predictions by providing direct validation of the MCBEND model through the use of neutron flux measurements. This paper describes the programme of work being undertaken to deploy two neutron flux measurement 'stringers' within the side-core region of one of the Hunterston B reactors for the purpose of validating the MCBEND model. The design of the stringers and the determination of the preferred deployment locations have been informed by the use of detailed MCBEND flux calculations. These computational studies represent a rare opportunity to design a flux measurement beforehand, with the clear intention of minimising the anticipated uncertainties and obtaining measurements that are known to be representative of the neutron fields to which the vulnerable steel restraint components are exposed. (authors)

  19. 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.

  20. 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.

  1. 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.

  2. /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.

  3. 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 detecting volume. However, operational limitations guide a) the maximum acceptable size of each detector cell (due to PSD performance and maximum-acceptable per-channel data throughput rates, limited by pulse pile-up and the processing rate of the electronics components of the system) and b) the affordability of a system due to the number of total channels of data to be collected and processed. As a first estimate, it appears that a system comprised of two rows of detectors 5" Ø ? 3" would yield a working prototype system with excellent performance capabilities for assaying Pu-containing items and capable of handling high signal rates likely when measuring items with Pu and other actinides. However, it is still likely that gamma-ray shielding will be needed to reduce the total signal rate in the detectors. As a first step prior to working with these larger-sized detectors, it may be practical to perform scoping studies using small detectors, such as already-on-hand 3" Ø ? 3" detectors.

  4. Interaction of the control system with core nuclear design for fast spectrum space power reactors

    NASA Astrophysics Data System (ADS)

    Lell, R. M.; Hanan, N. A.

    The reactor control system and operating strategy are essential factors in assessing reactor reliability and safety. The control system and its mode of operation also exert major influences on mechanical design of core components and all aspects of nuclear design. This is especially true of reactors for space power applications because of the imposed requirements regarding compactness, minimum mass, and long term operational reliability without external intervention or maintenance. Generic features of the interaction between nuclear design and reactor control system design for fast spectrum space power reactors are outlined. Several basic control concepts were analyzed. These included ex-core control drums, in-core control rods, burnable poisons, dispersed poisons in the core, and movable fuel segments or regions. Cross sections for calculations were generated with MC sup 2 -2, and neutronics calculations were performed with the VIM Monte Carlo code, ONEDANT, and DIF3D.

  5. 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.

  6. 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…

  7. 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

  8. A Design Study of High Breeding Ratio Sodium Cooled Metal Fuel Core without Blanket Fuels

    NASA Astrophysics Data System (ADS)

    Kobayashi, Noboru; Ogawa, Takashi; Ohki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari

    The metal fuel core is superior to the mixed oxide fuel core because of its high breeding ratio and compact core size resulting from hard neutron spectrum and high heavy metal densities. Utilizing these characteristics, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8 , a core height of less than 150 cm, the maximum cladding temperature of 650 °C, and the maximum fuel pin bundle pressure drop of 0.4MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40.

  9. Development and preliminary verification of the 3D core neutronic code: COCO

    SciTech Connect

    Lu, H.; Mo, K.; Li, W.; Bai, N.; Li, J.

    2012-07-01

    As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code, the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)

  10. Overview of core designs and requirements/criteria for core restraint systems

    SciTech Connect

    Sutherland, W.H.

    1984-09-01

    The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System are presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented.

  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. Design trade-off study for a large volume short pulse neutron assembly

    NASA Astrophysics Data System (ADS)

    Griffin, P. J.; Miller, J. D.; Harms, G. A.; Parma, E. J.; Coats, R. L.; Fan, W. C.; Petraglia, J. P.

    There is a continuing need within the radiation effects research community for more intense and larger volume pulsed neutron facilities. To fulfill these requirements a study was performed to examine conceptual designs for a neutron assembly that could produce high-intensity, short pulse neutron environments over a large experimental volume. The desired system characteristics were a cylindrical experimental cavity 60 inches long (152.4 cm) with a 24 inch diameter (60.96 cm), a cavity fluence of phi(sub r) = 8 x 10(exp 14) n/cm(exp 2), and a neutron pulse width of tau = 10-20 (mu)s. Attention was focused on booster assemblies which have been studied since the 1950s at Harwell, General Atomic, and at Sandia National Laboratories. Five conceptual designs were developed and evaluated. Only a two-stage coupled core design with a NpO2 primary core assembly was found to meet the design goals. A program is proposed to refine the design and to construct this nuclear assembly. The proposed three-phase effort represents a conservative approach that will yield large increases in the experiment volume even if the final coupled-core design is not realized.

  13. Sensitivity of ex-core neutron detectors to vibrations of PWR fuel assemblies

    SciTech Connect

    Sweeney, F.J.; Renier, J.P.

    1983-01-01

    The response of an ex-core neutron detector to fuel assembly vibrations in an 1150-MWe Westinghouse pressurized-water reactor (PWR) was determined by performing space-dependent reactor-kinetics calculations. The effect on the detector response of reducing the soluble-boron concentration in the coolant and fuel burnup over the first fuel cycle was also determined. The results of the calculations indicate that the ex-core neutron detectors are sensitive to fuel-assembly vibrations occurring over large spatial regions of the core, and that the detector response increases with fuel burnup and decreasing soluble-boron concentration.

  14. 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.

  15. Core bit design reduces mud invasion, improves ROP

    SciTech Connect

    Clydesdale, G. ); Leseultre, A.; Lamine, E. )

    1994-08-08

    A recently developed core bit reduces fluid invasion in the cut core by minimizing the exposure to the drilling fluid and by increasing the rate of penetration (ROP). A high ROP during coring is one of the major factors in reducing mud filtrate invasion in cores. This new low-invasion polycrystalline diamond compact (PDC) core bit was designed to achieve a higher ROP than conventional PDC core bits without detriment to the cutting structure. The paper describes the bit and its operation, results of lab tests, fluid dynamics, and results of field tests.

  16. 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.

  17. Neutron collimator design of neutron radiography based on the BNCT facility

    NASA Astrophysics Data System (ADS)

    Yang, Xiao-Peng; Yu, Bo-Xiang; Li, Yi-Guo; Peng, Dan; Lu, Jin; Zhang, Gao-Long; Zhao, Hang; Zhang, Ai-Wu; Li, Chun-Yang; Liu, Wan-Jin; Hu, Tao; Lü, Jun-Guang

    2014-02-01

    For the research of CCD neutron radiography, a neutron collimator was designed based on the exit of thermal neutron of the Boron Neutron Capture Therapy (BNCT) reactor. Based on the Geant4 simulations, the preliminary choice of the size of the collimator was determined. The materials were selected according to the literature data. Then, a collimator was constructed and tested on site. The results of experiment and simulation show that the thermal neutron flux at the end of the neutron collimator is greater than 1.0×106 n/cm2/s, the maximum collimation ratio (L/D) is 58, the Cd-ratio(Mn) is 160 and the diameter of collimator end is 10 cm. This neutron collimator is considered to be applicable for neutron radiography.

  18. 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)

  19. An intrinsically safe facility for forefront research and training on nuclear technologies — Core design

    NASA Astrophysics Data System (ADS)

    Viberti, C. M.; Ricco, G.

    2014-04-01

    The core of a subcritical, low-power research reactor in a lead matrix has been designed using the MCNPX code. The main parameters, like geometry, material composition in the fuel assembly and reflector size, have been optimized for a k eff ˜ 0.95 and a thermal power around 200 Kw. A 70 Mev, 1 mA proton beam incident on a beryllium target has been assumed as neutron source and the corresponding thermal power distribution and neutron fluxes in the reactor have been simulated.

  20. Neutronic conceptual design of the ETRR-2 cold-neutron source using the MCNP code

    NASA Astrophysics Data System (ADS)

    Khalil, M. Y.; Shaat, M. K.; Abdelfattah, A. Y.

    2005-04-01

    A conceptual neutronic design of the cold-neutron source (CNS) for the Egyptian second research reactor (ETRR-2) was done using the MCNP code. Parametric analysis to chose the type and geometry of the moderator, and the required CNS dimensions to maximize the cold neutron production was performed. The moderator cell has a spherical annulus structure containing liquid hydrogen. The cold neutron gain and cold neutron brightness are calculated together with the nuclear heat load of the CNS. Analysis of the estimated performance of the CNS has been done regarding the effect of void fraction in the moderator cell together with the ortho: para ratio.

  1. 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.

  2. 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.

  3. Effect of core structure irradiation on the RBMK neutron characteristics

    SciTech Connect

    Balygin, A. A. Krayushkin, A. V.

    2014-12-15

    The effect of changes in the graphite density and fuel channel diameters on the RBMK neutron characteristics is estimated. It is shown that uncertainty of those quantities can lead to a noticeable difference between the calculated and experimental values of the steam coefficient of reactivity and the subcriticality of the reactor.

  4. Effect of core structure irradiation on the RBMK neutron characteristics

    NASA Astrophysics Data System (ADS)

    Balygin, A. A.; Krayushkin, A. V.

    2014-12-01

    The effect of changes in the graphite density and fuel channel diameters on the RBMK neutron characteristics is estimated. It is shown that uncertainty of those quantities can lead to a noticeable difference between the calculated and experimental values of the steam coefficient of reactivity and the subcriticality of the reactor.

  5. 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)

  6. 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)

  7. Design considerations for an air core magnetic actuator

    NASA Technical Reports Server (NTRS)

    Groom, Nelson J.

    1992-01-01

    Equations for the force produced by an air core electromagnet on a permanent magnet core as a function of the coil height, coil inner and outer radii, and core displacement are developed. The magnetization vector of the permanent magnet core is assumed to be aligned with the central axis of the electromagnet and the forces which are produced lie along the same axis. Variations in force due to changes in electromagnet parameters and core displacement are investigated and parameter plots which should be useful for coil design are presented.

  8. Influence of fuel vibration on PWR neutron noise associated with core barrel motion

    SciTech Connect

    Sweeney, F.J.; March-Leuba, J.

    1984-01-01

    Ex-core neutron detector noise has been utilized to monitor core support barrel (CSB) vibrations. In order to observe long-term changes, noise signals at Sequoyah-1 were monitored continuously during the whole first fuel cycle and part of the second cycle. Results suggest that neutron noise measurements performed infrequently may not provide adequate surveillance of the CSB because it may be difficult to separate noise amplitude changes due solely to CSB motion from changes caused by fuel motion and burnup. (DLC)

  9. 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.

  10. TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

    PubMed

    Kurosawa, Masahiko

    2005-01-01

    For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. PMID:16604689

  11. Maximum pulsar mass, equation of state and structure of neutron-star cores

    NASA Astrophysics Data System (ADS)

    Haensel, P.; Zdunik, J. L.

    2016-01-01

    The structure of neutron stars is determined by the equation of state of dense matter in their interiors. Brief review of the equation of state from neutron star surface to its center is presented. Recent discovery of two two-solar-mass pulsars puts interesting constraints on the poorly known equation of state of neutron-star cores for densities greater than normal nuclear matter density. Namely, this equation of state has to be stiff enough to yield maximum allowable mass of neutron stars greater than two solar masses. There are many models of neutron stars cores involving exclusively nucleons that satisfy this constraint. However, for neutron-star models based on recent realistic baryon interaction, and allowing for the presence of hyperons, the hyperon softening of the equation of state yields maximum masses significantly lower than two solar masses. Proposed ways out from this ”hyperon puzzle” are presented. They require a very fine tuning of parameters of dense hadronic matter and quark matter models. Consequences for the mass-radius relation for neutron stars are illustrated. A summary of the present situation and possible perspectives/challenges, as well as possible observational tests, are given.

  12. IRSN working program status on tools for evaluation of SFR cores static neutronics safety parameters

    SciTech Connect

    Ivanov, E.; Tiberi, V.; Ecrabet, F.; Chegrani, Y.; Canuti, E.; Bisogni, D.; Sargeni, A.; Bernard, F.

    2012-07-01

    As technical support of the French Nuclear Safety Authority, IRSN will be in charge of safety assessment of any future project of Sodium Fast Reactor (SFR) that could be built in France. One of the main safety topics will deal with reactivity control. Since the design and safety assessment of the last two SFR plants in France (Phenix and Superphenix, more than thirty years ago), methods, codes and safety objectives have evolved. That is why a working program on core neutronic simulations has been launched in order to be able to evaluate accuracy of future core characteristics computations. The first step consists in getting experienced with the ERANOS well-known deterministic code used in the past for Phenix and Superphenix. Then Monte-Carlo codes have been tested to help in the interpretation of ERANOS results and to define what place this kind of codes can have in a new SFR safety demonstration. This experience is based on open benchmark computations. Different cases are chosen to cover a wide range of configurations. The paper shows, as an example, criticality results obtained with ERANOS, SCALE and MORET, and the first conclusions based on these results. In the future, this work will be extended to other safety parameters such as sodium void and Doppler effects, kinetic parameters or flux distributions. (authors)

  13. Divertor Neutron Flux Monitor: Conceptual Design and Calibration

    NASA Astrophysics Data System (ADS)

    Kaschuck Yu, A.; Batyunin, A. V.; Egorov, O. G.; Krasil'nikov, A. V.; Tsutskih, A. Yu; Frunze, V. V.

    2008-03-01

    The operating conditions of a neutron diagnostic systems which are responsible for measuring of the total neutron yield in the ITER tokamak reactor are analyzed. Based on results of neutronic calculations and analysis of suitable methods for measuring the neutron yield, a concept of a system for neutron flux measurement in the divertor is proposed. The design for the neutron flux monitor located in the divertor cassette of the tokamak is selected in view of the requirements specified for the neutron diagnostic system of the ITER and its operating conditions. Several fission chambers with different sensitivities and radiator materials will be used for measurements. System is capable of neutron fluxes measuring over the entire dynamic range of the ITER neutron yield with an error of δ<10% and a time resolution of 1 ms that are necessary for studying the physical phenomena of ignition and burning plasma. The problems of carrying out of the divertor neutron monitor efficiency calibration with the aim to measure the absolute value of the neutron yield in the ITER tokamak reactor are also discussed.

  14. 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)

  15. Pulsed neutron source cold moderators --- concepts, design and engineering

    SciTech Connect

    Bauer, Guenter S.

    1997-01-01

    Moderator design for pulsed neutron sources is becoming more and more an interface area between source designers and instrument designers. Although there exists a high degree of flexibility, there are also physical and technical limitations. This paper aims at pointing out these limitations and examining ways to extend the current state of moderator technology in order to make the next generation neutron sources even more versatile and flexible tools for science in accordance with the users' requirements. (auth)

  16. 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.

  17. Design of low-energy neutron beams for boron neutron capture synovectomy

    NASA Astrophysics Data System (ADS)

    Yanch, Jacquelyn C.; Shefer, Ruth E.; Binello, E.

    1997-02-01

    A novel application of the 10B(n, (alpha) )7Li nuclear reaction for the treatment of rheumatoid arthritis is under development. this application, called Boron Neutron Capture Synovectomy (BNCS), is briefly described here and the differences between BNCS and Boron Neutron Capture Therapy (BNCT) are discussed in detail. These differences lead to substantially altered demands on neutron beam design for each therapy application. In this paper the considerations for neutron beam design for the treatment of arthritic joints via BNCS are discussed, and comparisons with the design requirements for BNCT are made. This is followed by a description of potential moderator/reflector assemblies that are calculated to produce intense, high- quality neutron beams based on the 7Li(p,n) accelerator- based reactions. Total therapy time and therapeutic ratios are given as a function of both moderator length and boron concentration. Finally, a means of carrying out multi- directional irradiations of arthritic joints is proposed.

  18. 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.

  19. 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

    2016-01-28

    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 catalysis, energy storage, sensing, and biomedicine are presented. PMID:26766598

  20. 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 catalysis, energy storage, sensing, and biomedicine are presented.

  1. 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.

  2. Observational constraints on neutron star crust-core coupling during glitches

    NASA Astrophysics Data System (ADS)

    Newton, W. G.; Berger, S.; Haskell, B.

    2015-12-01

    We demonstrate that observations of glitches in the Vela pulsar can be used to investigate the strength of the crust-core coupling in a neutron star and provide a powerful probe of the internal structure of neutron stars. We assume that glitch recovery is dominated by the torque exerted by the mutual friction-mediated recoupling of superfluid components of the core that were decoupled from the crust during the glitch. Then we use the observations of the recoveries from two recent glitches in the Vela pulsar to infer the fraction of the core that is coupled to the crust during the glitch. We then analyse whether crustal neutrons alone are sufficient to drive glitches in the Vela pulsar, taking into account crustal entrainment. We use two sets of neutron star equations of state (EOSs) which span crust and core consistently and cover a conservative range of the slope of the symmetry energy at saturation density 30 < L < 120 MeV. The two sets differ in the stiffness of the high density EOS. We find that for medium to stiff EOSs, observations imply >70 per cent of the moment of inertia of the core is coupled to the crust during the glitch, though for softer EOSs L ≈ 30 MeV as little as 5 per cent could be coupled. We find that only by extending the region where superfluid vortices are strongly pinned into the core by densities at least 0.016 fm-3 above the crust-core transition density does any EOS reproduce the observed glitch activity.

  3. Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

    SciTech Connect

    Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

    2006-01-01

    The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

  4. Neutronic analysis of the Three Mile Island Unit 2 ex-core detector response

    SciTech Connect

    Malloy, D.J.; Chang, Y.I.

    1981-10-01

    A neutronic analysis has been made with respect to the ex-core neutron detector response during the TMI-2 incident. A series of transport theory calculations quantified the impact upon the detector count rate of various core and downcomer conditions. In particular, various combinations of coolant void content and spatial distributions were investigated to yield the resulting transmission of the photoneutron source to the detector. The impact of a hypothetical distributed source within the downcomer region was also examined in order to simulate the potential effect of the release of neutron producing fission products into the coolant. These results are then offered as potential explanations for the anomalous behavior of the detector during the period of approx. 20 minutes through approx. 3 hours following the reactor scram.

  5. Fuel and Core Design Experiences in Cofrentes NPP

    SciTech Connect

    Garcia-Delgado, L.; Lopez-Carbonell, M.T.; Gomez-Bernal, I.

    2002-07-01

    The electricity market deregulation in Spain is increasing the need for innovations in nuclear power generation, which can be achieved in the fuel area by improving fuel and core designs and by introducing vendors competition. Iberdrola has developed the GIRALDA methodology for design and licensing of Cofrentes reloads, and has introduced mixed cores with fuel from different vendors. The application of GIRALDA is giving satisfactory results, and is showing its capability to adequately reproduce the core behaviour. The nuclear design team is acquiring an invaluable experience and a deep knowledge of the core, very useful to support cycle operation. Continuous improvements are expected for the future in design strategies as well as in the application of new technologies to redesign the methodology processes. (authors)

  6. 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 drop across the core was restricted to no more than 1.5 atm during normal operation to minimize the upward force on the core. Also, the flow velocity in the core was restricted to 3 m/s to minimize erosion of the fuel plates. Section 3.1.1 of this report discusses the design restrictions in more detail.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. Neutronic calculations for the conversion to LEU of a research reactor core

    SciTech Connect

    Varvayanni, M.; Catsaros, N.; Stakakis, E.; Grigoriadis, D.

    2008-07-15

    For a five-year transitional period the Greek Research Reactor (GRR-1) was operating with a mixed core, containing both Low Enrichment (LEU) and High Enrichment (HEU) Uranium MTR- type fuel assemblies. The neutronic study of the GRR-1 conversion to LEU has been performed using a code system comprising the core-analysis code CITATION-LDI2 and the cell-calculation modules XSDRNPM and NITAWL-II of the SCALE code. A conceptual LEU core configuration was defined and analyzed with respect to the three dimensional multi-group neutron fluxes, the power distribution, the control-rod worth and the compliance with pre-defined Operation Limiting Conditions. Perturbation calculations and reactivity feedback computations were also carried out to provide input to a subsequent thermal-hydraulic study. (author)

  12. 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.

  13. Design and Demonstration of a Quasi-monoenergetic Neutron Source

    SciTech Connect

    Joshi, T.; Sangiorgio, Samuele; Mozin, Vladimir V.; Norman, E. B.; Sorensen, Peter F.; Foxe, Michael P.; Bench, G.; Bernstein, A.

    2014-03-05

    The design of a neutron source capable of producing 24 and 70 keV neutron beams with narrow energy spread is presented. The source exploits near-threshold kinematics of the 7Li(p,n)7Be reaction while taking advantage of the interference `notches' found in the scattering cross-sections of iron. The design was implemented and characterized at the Center for Accelerator Mass Spectrometry at Lawrence Livermore National Laboratory. Alternative lters such as vanadium and manganese are also explored and the possibility of studying the response of di*erent materials to low-energy nuclear recoils using the resultant neutron beams is discussed.

  14. 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)

  15. X-ray and Neutron Scattering Study of the Formation of Core-Shell Type Polyoxometalates

    SciTech Connect

    Yin, Panchao; Wu, Bin; Mamontov, Eugene; Daemen, Luke L; Cheng, Yongqiang; Hong, Kunlun; Bonnesen, Peter V; Keum, Jong Kahk; Ramirez-Cuesta, Anibal J

    2016-01-01

    A typical type of core-shell polyoxometalates can be obtained through the Keggin-type polyoxometalate-templated growth of a layer of spherical shell structure of {Mo72Fe30}. Small angle X-ray scattering is used to study the structural features and stability of the core-shell structures in aqueous solutions. Time-resolved small angle X-ray scattering is applied to monitor the synthetic reactions and a three-stage formation mechanism is proposed to describe the synthesis of the core-shell polyoxometalates based on the monitoring results. Quasi-elastic and inelastic neutron scattering are used to probe the dynamics of water molecules in the core-shell structures and two different types of water molecules, the confined and structured water, are observed. These water molecules play an important role in bridging core and shell structures and stabilizing the cluster structures.A typical type of core shell polyoxometalates can be obtained through the Keggin-type polyoxometalate-templated growth of a layer of spherical shell structure of {Mo72Fe30}. Small-angle X-ray scattering is used to study the structural features and stability of the core shell structures in aqueous solutions. Time-resolved small-angle X-ray scattering is applied to monitor the synthetic reactions, and a three-stage formation mechanism is proposed to describe the synthesis of the core shell polyoxometalates based on the monitoring results. New protocols have been developed by fitting the X-ray data with custom physical models, which provide more convincing, objective, and completed data interpretation. Quasi-elastic and inelastic neutron scattering are used to probe the dynamics of water molecules in the core shell structures, and two different types of water molecules, the confined and structured water, are observed. These water molecules play an important role in bridging core and shell structures and stabilizing the cluster structures.

  16. X-ray and Neutron Scattering Study of the Formation of Core-Shell Type Polyoxometalates

    DOE PAGESBeta

    Yin, Panchao; Wu, Bin; Mamontov, Eugene; Daemen, Luke L; Cheng, Yongqiang; Hong, Kunlun; Bonnesen, Peter V; Keum, Jong Kahk; Ramirez-Cuesta, Anibal J

    2016-01-01

    A typical type of core-shell polyoxometalates can be obtained through the Keggin-type polyoxometalate-templated growth of a layer of spherical shell structure of {Mo72Fe30}. Small angle X-ray scattering is used to study the structural features and stability of the core-shell structures in aqueous solutions. Time-resolved small angle X-ray scattering is applied to monitor the synthetic reactions and a three-stage formation mechanism is proposed to describe the synthesis of the core-shell polyoxometalates based on the monitoring results. Quasi-elastic and inelastic neutron scattering are used to probe the dynamics of water molecules in the core-shell structures and two different types of water molecules,morethe confined and structured water, are observed. These water molecules play an important role in bridging core and shell structures and stabilizing the cluster structures.A typical type of core shell polyoxometalates can be obtained through the Keggin-type polyoxometalate-templated growth of a layer of spherical shell structure of {Mo72Fe30}. Small-angle X-ray scattering is used to study the structural features and stability of the core shell structures in aqueous solutions. Time-resolved small-angle X-ray scattering is applied to monitor the synthetic reactions, and a three-stage formation mechanism is proposed to describe the synthesis of the core shell polyoxometalates based on the monitoring results. New protocols have been developed by fitting the X-ray data with custom physical models, which provide more convincing, objective, and completed data interpretation. Quasi-elastic and inelastic neutron scattering are used to probe the dynamics of water molecules in the core shell structures, and two different types of water molecules, the confined and structured water, are observed. These water molecules play an important role in bridging core and shell structures and stabilizing the cluster structures.less

  17. Cooling neutron star in the Cassiopeia A supernova remnant: evidence for superfluidity in the core

    NASA Astrophysics Data System (ADS)

    Shternin, Peter S.; Yakovlev, Dmitry G.; Heinke, Craig O.; Ho, Wynn C. G.; Patnaude, Daniel J.

    2011-03-01

    According to recent results of Ho & Heinke, the Cassiopeia A supernova remnant contains a young (≈330-yr-old) neutron star (NS) which has carbon atmosphere and shows notable decline of the effective surface temperature. We report a new (2010 November) Chandra observation which confirms the previously reported decline rate. The decline is naturally explained if neutrons have recently become superfluid (in triplet state) in the NS core, producing a splash of neutrino emission due to Cooper pair formation (CPF) process that currently accelerates the cooling. This scenario puts stringent constraints on poorly known properties of NS cores: on density dependence of the temperature Tcn(ρ) for the onset of neutron superfluidity [Tcn(ρ) should have a wide peak with maximum ≈ (7-9) × 108 K]; on the reduction factor q of CPF process by collective effects in superfluid matter (q > 0.4) and on the intensity of neutrino emission before the onset of neutron superfluidity (30-100 times weaker than the standard modified Urca process). This is serious evidence for nucleon superfluidity in NS cores that comes from observations of cooling NSs.

  18. 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.

  19. 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.

  20. 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)

  1. A Neutron Streak Camera Designed for ICF Fuel Ion Temperature

    NASA Astrophysics Data System (ADS)

    Chen, Jiabin; Liao, Hua; Chen, Ming

    2007-11-01

    A neutron streak camera was designed for inertial confinement fusion (ICF) fuel ion temperature diagnostic. It is made of a 1 cm thick x8 cm diam piece of 3% benzophenone quenched plastic scintillator with about a 190 ps FWHM and a streak tube (55ps time resolution) with large-area photocathode (φ30 mm) showed no slit. The electron beam from the photocathode is focused into a little spot (φ1mm). Then the spot is scanned directly and multiplied by an internal microchannel plate. This greatly improves the sensitivity of the tube. The neutron streak camera combines the advangtages of scintillation detector (with high neutron detection efficiency) and of streak camera (with fast time response). The whole detection system time resolution is 300ps and can record neutron time of flight signals from ICF implosion target with yields of 10^7 DT neutron per shot.

  2. GFR fuel and core pre-conceptual design studies

    SciTech Connect

    Chauvin, N.; Ravenet, A.; Lorenzo, D.; Pelletier, M.; Escleine, J.M.; Munoz, I.; Bonnerot, J.M.

    2007-07-01

    The revision of the GFR core design - plate type - has been undertaken since previous core presented at Global'05. The self-breeding searched for has been achieved with an optimized design ('12/06 E'). The higher core pressure drop was a matter of concern. First of all, the core coolability in natural circulation for pressurized conditions has been studied and preliminary plant transient calculations have been performed. The design and safety criteria are met but no more margin remains. The project is also addressing the feasibility and the design of the fuel S/A. The hexagonal shape together with the principle of closed S/A (wrapper tube) is kept. Ceramic plate type fuel element combines a high enough core power density (minimization of the Pu inventory) and plutonium and minor actinides recycling capabilities. Innovative for many aspects, the fuel element is central to the GFR feasibility. It is supported already by a significant R and D effort also applicable to a pin concept that is considered as the other fuel element of interest. This combination of fuel/core feasibility and performance analysis, safety dispositions and performances analysis will compose the 'GFR preliminary feasibility' which is a project milestone at the end of the year 2007. (authors)

  3. Core design of the upgraded TREAT reactor

    SciTech Connect

    Wade, D.C.; Bhattacharyya, S.K.; Lipinski, W.C.; Stone, C.C.

    1982-01-01

    The upgrading of the TREAT reactor involves the replacement of the central 11 x 11 subzone of the 19 x 19 fuel assembly array by new, Inconel-clad, high-temperature fuel assemblies, and the additions of a new reactor control system, a safety-grade plant protection system, and an enhanced reactor filtration/coolant system. The final design of these modifications will be completed in early 1983. The TREAT facility is scheduled to be shut down for modification in mid-1984, and should resume the safety test program in mid-1985. The upgrading will provide a capability to conduct fast reactor safety tests on clusters of up to 37 prototypic LMFBR pins.

  4. Physics design of a cold neutron source for KIPT neutron source facility.

    SciTech Connect

    Zhong, Z.; Gohar, Y.; Kellogg, R.; Nuclear Engineering Division

    2009-02-17

    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 a neutron source facility. It is based on the use of an electron accelerator driven subcritical (ADS) facility with low enriched uranium fuel, using the existing electron accelerators at KIPT of Ukraine [1]. The neutron source of the subcritical assembly is generated from the interaction of 100-KW electron beam, which has a uniform spatial distribution and the electron energy in the range of 100 to 200 MeV, with a natural uranium target [2]. The main functions of the facility are the production of medical isotopes and the support of the Ukraine nuclear power industry. Neutron beam experiments and material studies are also included. Over the past two-three decades, structures with characteristic lengths of 100 {angstrom} and correspondingly smaller vibrational energies have become increasingly important for both science and technology [3]. The characteristic dimensions of the microstructures can be well matched by neutrons with longer vibrational wavelength and lower energy. In the accelerator-driven subcritical facility, most of the neutrons are generated from fission reactions with energy in the MeV range. They are slowed down to the meV energy range through scattering reactions in the moderator and reflector materials. However, the fraction of neutrons with energies less than 5 meV in a normal moderator spectrum is very low because of up-scattering caused by the thermal motion of moderator or reflector molecules. In order to obtain neutrons with energy less than 5 meV, cryogenically cooled moderators 'cold neutron sources' should be used to slow down the neutrons. These cold moderators shift the neutron energy spectrum down because the thermal motion of moderator molecules as well as the up-scattering is very small, which provides large gains in intensity of low energy neutrons, E < 5 meV. The accelerator driven subcritical facility is designed with a provision to add a cryogenically cooled moderator system. This cold neutron source could provide the neutrons beams with lower energy, which could be utilized in scattering experiment and material structures analysis. This study describes the performed physics analyses to define and characterize the cold neutron source of the KIPT neutron source facility. The cold neutron source is designed to optimize the cold neutron brightness to the experimental instruments outside the radial heavy concrete shield of the facility. Liquid hydrogen or solid methane with 20 K temperature is used as a cold moderator. Monte Carlo computer code MCNPX [4], with ENDF/B-VI nuclear data libraries, is utilized to calculate the cold neutron source performance and estimate the nuclear heat load to the cold moderator. The surface source generation capability of MCNPX code has been used to provide the possibility of analyzing different design configurations and perform design optimization analyses with reasonable computer resources. Several design configurations were analyzed and their performance were characterized and optimized.

  5. Application of ex-vessel neutron dosimetry combined with in-core measurements for correction of neutron source used for RPV fluence calculations

    SciTech Connect

    Borodkin, P.G.; Borodkin, G.I.; Khrennikov, N.N.

    2011-07-01

    This paper deals with calculated and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Neutron activation measurements analyzed in the paper were carried out in an ex-vessel air cavity at different nuclear power plant units with VVER-1000 during different fuel cycles. The time-integrated neutron source distributions used for DORT calculations were prepared via two different approaches based on (a) calculated fuel burnup (standard routine procedure) and (b) in-core measurements by means of self-powered detectors (SPDs) and thermocouples (TCs) (new approach). Considering that fuel burnup distributions in operating VVER may be evaluated now by the use of analytical methods (calculations) only, it is necessary to develop new approaches for the testing and correction of calculated evaluations of a neutron source. The results presented in this paper allow one to consider the reverse task of the alternative estimation of fuel burnup distributions. The proposed approach is based on the adjustment (fitting) of time-integrated neutron source distributions, and thus fuel burnup patterns, in some part of the reactor core, taking into account neutron leakage measurements, neutron-physical calculations, and in-core SPD and TC measurement data. (authors)

  6. China Spallation Neutron Source: Design, R&D, and outlook

    NASA Astrophysics Data System (ADS)

    Wei, Jie; Chen, Hesheng; Chen, Yanwei; Chen, Yuanbo; Chi, Yunlong; Deng, Changdong; Dong, Haiyi; Dong, Lan; Fang, Shouxian; Feng, Ji; Fu, Shinian; He, Lunhua; He, Wei; Heng, Yuekun; Huang, Kaixi; Jia, Xuejun; Kang, Wen; Kong, Xiangcheng; Li, Jian; Liang, Tianjiao; Lin, Guoping; Liu, Zhenan; Ouyang, Huafu; Qin, Qing; Qu, Huamin; Shi, Caitu; Sun, Hong; Tang, Jingyu; Tao, Juzhou; Wang, Chunhong; Wang, Fangwei; Wang, Dingsheng; Wang, Qingbin; Wang, Sheng; Wei, Tao; Xi, Jiwei; Xu, Taoguang; Xu, Zhongxiong; Yin, Wen; Yin, Xuejun; Zhang, Jing; Zhang, Zong; Zhang, Zonghua; Zhou, Min; Zhu, Tao

    2009-02-01

    The China Spallation Neutron Source (CSNS) is an accelerator based multidiscipline user facility planned to be constructed in Dongguan, Guangdong, China. The CSNS complex consists of an negative hydrogen linear accelerator, a rapid cycling proton synchrotron accelerating the beam to 1.6 GeV energy, a solid tungsten target station, and instruments for spallation neutron applications. The facility operates at 25 Hz repetition rate with an initial design beam power of 120 kW and is upgradeable to 500 kW. The primary challenge is to build a robust and reliable user's facility with upgrade potential at a fraction of "world standard" cost. We report the status, design, R&D, and upgrade outlook including applications using spallation neutron, muon, fast neutron, and proton, as well as related programs including medical therapy and accelerator-driven sub-critical reactor (ADS) programs for nuclear waste transmutation.

  7. 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 on a single silicon chip. Individual elements of the array were tested for thermal-neutron detection efficiency and for the n/γ reject ratio. Overall, because of the inadequacies and costs of other neutron detection systems, the MSND is the premier technology for many neutron detection applications.

  8. Design calculations for the ANS (Advanced Neutron Source) cold source

    SciTech Connect

    Lillie, R.A.; Alsmiller, R.G. Jr.

    1988-01-01

    The calculation procedure, based on discrete ordinates transport methods, that is being used to carry out design calculations for the Advanced Neutron Source cold source is described. Calculated results on the gain in cold neutron flux produced by a liquid deuterium cold source are compared with experimental data and with calculated data previously obtained by P. Ageron et al., at the Institute Max von Laue-Paul Langevin in Grenoble, France. Calculated results are also presented that indicated how the flux of cold neutrons vary with cold source parameters. 23 refs., 5 figs., 3 tabs.

  9. A Tight Lattice, Epithermal Core Design for the Integral PWR

    SciTech Connect

    Saccheri, J.G.B.; Todreas, N.E.; Driscoll, M.J.

    2004-07-01

    An 8-year core design for an epithermal, water-cooled reactor has been developed based upon assessments of nuclear reactor physics, thermal-hydraulics and economics. An integral vessel configuration is adopted and self-supporting wire-wrap fuel is employed for the tight lattice of the epithermal core. A streaming path is incorporated in each assembly to ensure a negative void coefficient. A whole-core MCNP simulation of the tight core shows a negative void coefficient for any burnup with positive K{sub EFF}. The VIPRE{sup TM} code has been used to calculate the critical heat flux (CHF) by means of an appropriate wire-wrap CHF correlation, specifically introduced in the source code. Economically, the high fuel enrichment (14% w/o {sup 235}U) and the very long core life (8 ys) lead to high lifetime-levelized unit fuel cycle cost (in mills/kWhre). However, both operation and maintenance and capital-related expenditures strongly benefited from the higher electric output per unit volume, which yielded quite small lifetime-levelized unit capital and operation and maintenance costs for the overall plant. Financing costs are included and an estimate is provided for the total lifetime-levelized unit cost of the epithermal core, which is about 20% lower than that of a more open lattice thermal spectrum core fitting into the same core envelope and with 4-year lifetime. (authors)

  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. Are neutron stars with crystalline color-superconducting cores relevant for the LIGO experiment?

    PubMed

    Haskell, B; Andersson, N; Jones, D I; Samuelsson, L

    2007-12-01

    We estimate the maximal deformation that can be sustained by a rotating neutron star with a crystalline color-superconducting quark core. Our results suggest that current gravitational-wave data from the Laser Interferometer Gravitational-Wave Observatory have already reached the level where a detection would have been possible over a wide range of the poorly constrained QCD parameters. This leads to the nontrivial conclusion that compact objects do not contain maximally strained color crystalline cores drawn from this range of parameter space. We discuss the uncertainties associated with our simple model and how it can be improved in the future. PMID:18233352

  12. Are Neutron Stars with Crystalline Color-Superconducting Cores Relevant for the LIGO Experiment?

    SciTech Connect

    Haskell, B.; Andersson, N.; Jones, D. I.; Samuelsson, L.

    2007-12-07

    We estimate the maximal deformation that can be sustained by a rotating neutron star with a crystalline color-superconducting quark core. Our results suggest that current gravitational-wave data from the Laser Interferometer Gravitational-Wave Observatory have already reached the level where a detection would have been possible over a wide range of the poorly constrained QCD parameters. This leads to the nontrivial conclusion that compact objects do not contain maximally strained color crystalline cores drawn from this range of parameter space. We discuss the uncertainties associated with our simple model and how it can be improved in the future.

  13. Design considerations for neutron activation and neutron source strength monitors for ITER

    SciTech Connect

    Barnes, C.W.; Jassby, D.L.; LeMunyan, G.; Roquemore, A.L.; Walker, C.

    1997-12-31

    The International Thermonuclear Experimental Reactor will require highly accurate measurements of fusion power production in time, space, and energy. Spectrometers in the neutron camera could do it all, but experience has taught us that multiple methods with redundancy and complementary uncertainties are needed. Previously, conceptual designs have been presented for time-integrated neutron activation and time-dependent neutron source strength monitors, both of which will be important parts of the integrated suite of neutron diagnostics for this purpose. The primary goals of the neutron activation system are: to maintain a robust relative measure of fusion energy production with stability and wide dynamic range; to enable an accurate absolute calibration of fusion power using neutronic techniques as successfully demonstrated on JET and TFTR; and to provide a flexible system for materials testing. The greatest difficulty is that the irradiation locations need to be close to plasma with a wide field of view. The routing of the pneumatic system is difficult because of minimum radius of curvature requirements and because of the careful need for containment of the tritium and activated air. The neutron source strength system needs to provide real-time source strength vs. time with {approximately}1 ms resolution and wide dynamic range in a robust and reliable manner with the capability to be absolutely calibrated by in-situ neutron sources as done on TFTR, JT-60U, and JET. In this paper a more detailed look at the expected neutron flux field around ITER is folded into a more complete design of the fission chamber system.

  14. Design of dual-core optical fibers with NEMS functionality.

    PubMed

    Podoliak, Nina; Lian, Zhenggang; Loh, Wei H; Horak, Peter

    2014-01-13

    An optical fiber with nano-electromechanical functionality is presented. The fiber exhibits a suspended dual-core structure that allows for control of the optical properties via nanometer-range mechanical movements. We investigate electrostatic actuation achieved by applying a voltage to specially designed electrodes integrated in the cladding. Numerical and analytical calculations are preformed to optimize the fiber and electrode design. Based on this geometry an all-fiber optical switch is investigated; we find that optical switching of light between the two cores can be achieved in a 10 cm fiber with an operating voltage of 35 V. PMID:24515066

  15. 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.

  16. 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.

  17. Conceptual design of a neutron camera for MAST Upgrade

    SciTech Connect

    Weiszflog, M. Sangaroon, S.; Cecconello, M.; Conroy, S.; Ericsson, G.; Klimek, I.; Keeling, D.; Martin, R.; Turnyanskiy, M.

    2014-11-15

    This paper presents two different conceptual designs of neutron cameras for Mega Ampere Spherical Tokamak (MAST) Upgrade. The first one consists of two horizontal cameras, one equatorial and one vertically down-shifted by 65 cm. The second design, viewing the plasma in a poloidal section, also consists of two cameras, one radial and the other one with a diagonal view. Design parameters for the different cameras were selected on the basis of neutron transport calculations and on a set of target measurement requirements taking into account the predicted neutron emissivities in the different MAST Upgrade operating scenarios. Based on a comparison of the cameras’ profile resolving power, the horizontal cameras are suggested as the best option.

  18. Analysis and Design of ITER 1 MV Core Snubber

    NASA Astrophysics Data System (ADS)

    Wang, Haitian; Li, Ge

    2012-11-01

    The core snubber, as a passive protection device, can suppress arc current and absorb stored energy in stray capacitance during the electrical breakdown in accelerating electrodes of ITER NBI. In order to design the core snubber of ITER, the control parameters of the arc peak current have been firstly analyzed by the Fink-Baker-Owren (FBO) method, which are used for designing the DIIID 100 kV snubber. The B-H curve can be derived from the measured voltage and current waveforms, and the hysteresis loss of the core snubber can be derived using the revised parallelogram method. The core snubber can be a simplified representation as an equivalent parallel resistance and inductance, which has been neglected by the FBO method. A simulation code including the parallel equivalent resistance and inductance has been set up. The simulation and experiments result in dramatically large arc shorting currents due to the parallel inductance effect. The case shows that the core snubber utilizing the FBO method gives more compact design.

  19. Conceptual Neutronic Design of a Lead-Bismuth-Cooled Actinide Burning Reactor

    SciTech Connect

    Hejzlar, Pavel; Driscoll, Michael J.; Kazimi, Mujid S.

    2001-10-15

    A conceptual design of a lead-bismuth-eutectic (LBE)-cooled actinide burner core with innovative streaming fuel assemblies (FAs) is described. The 1800-MW(thermal) core employs metallic, fertile-free fuel where the transuranics (plutonium plus minor actinides) are dispersed in a zirconium matrix. The core contains 157 streaming FAs that enhance neutron streaming by employing gas-filled, sealed streaming tubes at the FA periphery and center. The large reactivity excess at the beginning of life is compensated for by a system of double-entry control rods. The arrangement of top-entry and bottom-entry control rods in a staggered pattern allows the achievement of a very uniform axial power profile and a small reactivity change from control rod driveline expansion. The reactor can operate with an 18- to 24-month cycle length.Safety is provided through negative reactivity coefficients and tight neutronic coupling. The void coefficient is negative for a partially as well as a fully voided core. The effective delayed neutron fraction is 25% less than that of typical oxide-fueled fast reactors, making the requirements on reactor control performance more demanding. The Doppler coefficient is negative with a magnitude appreciably lower than the typical values of oxide fuels in sodium-cooled reactors, but comparable to the values observed in integral fast reactor (IFR) cores with metallic U-Pu-Zr fuels. The fuel thermal expansion coefficient is also negative, having a magnitude approximately equal to the Doppler coefficient. In terms of the transuranic destruction rate per MW(thermal) per effective full-power year, the design is comparable to accelerator-driven systems (ADSs). Long-lived fission products also can be transmuted, albeit at lower incineration efficiency than in ADSs.

  20. 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.

  1. Influence Of Low Boron Core Design On PWR Transient Behavior

    SciTech Connect

    Aleksandrov Papukchiev, Angel; Yubo Liu; Schaefer, Anselm

    2006-07-01

    In conventional pressurized water reactor (PWR) designs, the concentration of boron in primary coolant is limited by the requirement of having a negative moderator density coefficient. As high boron concentrations have significant impact on reactivity feedback properties, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In the framework of an investigation into the feasibility of low boron design, a PWR core configuration based on fuel with higher gadolinium (Gd) content has been developed which permits to reduce the natural boron concentration at begin of cycle (BOC) by approx. 50% compared to current German PWR technology. For the assessment of the potential safety advantages, a Loss-of-Feedwater Anticipated Transient Without Scram (ATWS LOFW) has been simulated with the system code ATHLET for two PWR core designs: a low boron design and a standard core design. The most significant difference in the transient performance of both designs is the total primary fluid mass released through the pressurizer (PRZ) valves. It is reduced by a factor of four for the low boron reactor, indicating its improved density reactivity feedback. (authors)

  2. Experimental determination of residual stress by neutron diffraction in a boiling water reactor core shroud

    SciTech Connect

    Payzant, A.; Spooner, S.; Zhu, Xiaojing; Hubbard, C.R.

    1996-06-01

    Residual strains in a 51 mm (2-inch) thick 304L stainless steel plate have been measured by neutron diffraction and interpreted in terms of residual stress. The plate, measuring (300 mm) in area, was removed from a 6m (20-ft.) diameter unirradiated boiling water reactor core shroud, and included a multiple-pass horizontal weld which joined two of the cylindrical shells which comprise the core shroud. Residual stress mapping was undertaken in the heat affected zone, concentrating on the outside half of the plate thickness. Variations in residual stresses with location appeared consistent with trends expected from finite element calculations, considering that a large fraction of the residual hoop stress was released upon removal of the plate from the core shroud cylinder.

  3. 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 varying cross-sectional diameters. The biochemical experiments not only confirmed the ability of the core constituents to form large aggregates but also established that phospholipids do not play a role in this aggregate formation.

  4. 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.

  5. 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.

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

    NASA Astrophysics Data System (ADS)

    Raepsaet, X.; Lenain, R.; Naury, S.

    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 UO2 or UC2 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.

  8. The Interplay between Proto--Neutron Star Convection and Neutrino Transport in Core-Collapse Supernovae

    NASA Astrophysics Data System (ADS)

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

    1998-01-01

    We couple two-dimensional hydrodynamics to realistic one-dimensional multigroup flux-limited diffusion neutrino transport to investigate proto-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-neutron star convection are investigated for both 15 and 25 M⊙ 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 ~20 ms, 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-neutron star convection episode seeds additional convectively unstable regions farther out beneath the shock. The additional proto-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-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 snapshots, which in this case appear spherically symmetric. The peak angle-averaged radial and angular convection velocities are orders of magnitude smaller than they are in the corresponding ``hydrodynamics-only'' models. A simple analytical model supports our numerical results, indicating that the inclusion of neutrino transport reduces the entropy-driven (lepton-driven) convection growth rates and asymptotic velocities by a factor ~3 (50) at the neutrinosphere and a factor ~250 (1000) at ρ = 1012 g cm-3, for both our 15 and 25 M⊙ models. Moreover, when transport is included, the initial postbounce entropy gradient is smoothed out by neutrino diffusion, whereas the initial lepton gradient is maintained by electron capture and neutrino escape near the neutrinosphere. Despite the maintenance of the lepton gradient, proto-neutron star convection does not develop over the 100 ms duration typical of all our simulations, except in the instance where ``low-test'' intial conditions are used, which are generated by core-collapse and bounce simulations that neglect neutrino-electron scattering and ion-ion screening corrections to neutrino-nucleus elastic scattering. Models favoring the development of proto-neutron star convection either by starting with more favorable, albeit artificial (low-test), initial conditions or by including transport corrections that were ignored in our ``fiducial'' models were considered. Our conclusions nonetheless remained the same. Evidence of proto-neutron star convection in our two-dimensional entropy snapshots was minimal, and, as in our fiducial models, the angle-averaged convective velocities when neutrino transport was included remained orders of magnitude smaller than their counterparts in the corresponding hydrodynamics-only models.

  9. A shielding design for an accelerator-based neutron source for boron neutron capture therapy.

    PubMed

    Hawk, A E; Blue, T E; Woollard, J E

    2004-11-01

    Research in boron neutron capture therapy (BNCT) at The Ohio State University Nuclear Engineering Department has been primarily focused on delivering a high quality neutron field for use in BNCT using an accelerator-based neutron source (ABNS). An ABNS for BNCT is composed of a proton accelerator, a high-energy beam transport system, a (7)Li target, a target heat removal system (HRS), a moderator assembly, and a treatment room. The intent of this paper is to demonstrate the advantages of a shielded moderator assembly design, in terms of material requirements necessary to adequately protect radiation personnel located outside a treatment room for BNCT, over an unshielded moderator assembly design. PMID:15308187

  10. Advanced core design and fuel management for pebble-bed reactors

    NASA Astrophysics Data System (ADS)

    Gougar, Hans David

    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.

  11. 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.

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

    SciTech Connect

    Wang, Zhonglu

    2006-08-01

    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 cm{sup 2} treatment beam. The calculated boron dose enhancement at 5.7-cm depth in a water-filled head phantom in the assembly with a 5x5 cm{sup 2} collimation was 21.9% per 100-ppm {sup 10}B 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 cm{sup 2} fast neutron treatment beam. To validate the design calculations, a simplified BNCEFNT assembly was built using four lead bricks to form a 5x5 cm{sup 2} collimator. Five 1.0-cm thick 20x20 cm{sup 2} 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 {sup 10}B) to measure dose due to boron neutron capture. The measured dose enhancement at 5.0-cm depth in the head phantom for the 5.0-cm thick tungsten filter is (16.6 {+-} 1.8)%, which agrees well with the MCNP simulation of the simplified BNCEFNT assembly, (16.4 {+-} 0.5)%. The error in the calculated dose enhancement only considers the statistical uncertainties. The total dose rate measured at 5.0-cm depth using the non-borated ion chamber is (0.765 {+-} 0.076) Gy/MU, about 61% of the fast neutron standard dose rate (1.255Gy/MU) at 5.0-cm depth for the standard 10x10 cm{sup 2} treatment beam. The increased doses to other organs due to the use of the BNCEFNT assembly were calculated using MCNP5 and a MIRD phantom. The activities of the activation products produced in the BNCEFNT assembly after neutron beam delivery were computed. The photon ambient dose rate due to the radioactive activation products was also estimated.

  13. 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.

  14. 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.

  15. Accelerator-based epithermal neutron beam design for neutron capture therapy.

    PubMed

    Yanch, J C; Zhou, X L; Shefer, R E; Klinkowstein, R E

    1992-01-01

    Recent interest in the production of epithermal neutrons for use in boron neutron capture therapy (BNCT) has promoted an investigation into the feasibility of generating such neutrons with a high current proton accelerator. Energetic protons (2.5 MeV) on a 7Li target produce a spectrum of neutrons with maximum energy of roughly 800 keV. A number of combinations of D2O moderator, lead reflector, 6Li thermal neutron filtration, and D2O/6Li shielding will result in a useful epithermal flux of 1.6 x 10(8) n/s at the patient position. The neutron beam is capable of delivering 3000 RBE-cGy to a tumor at a depth of 7.5 cm in a total treatment time of 60-93 min (depending on RBE values used and based on a 24-cm diameter x 19-cm length D2O moderator). Treatment of deeper tumors with therapeutic advantage would also be possible. Maximum advantage depths (RBE weighted) of 8.2-9.2 (again depending on RBE values and precise moderator configuration) are obtained in a right-circular cylindrical phantom composed of brain-equivalent material with an advantage ratio of 4.7-6.3. A tandem cascade accelerator (TCA), designed and constructed at Science Research Laboratory (SRL) in Somerville MA, can provide the required proton beam parameters for BNCT of deep-seated tumors. An optimized configuration of materials required to shift the accelerator neutron spectrum down to therapeutically useful energies has been designed using Monte Carlo simulation in the Whitaker College Biomedical Imaging and Computation Laboratory at MIT. Actual construction of the moderator/reflector assembly is currently underway. PMID:1324392

  16. 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.

  17. An integrated design of an accelerator-based neutron source for boron neutron capture therapy

    NASA Astrophysics Data System (ADS)

    Dobelbower, Michael Christian

    1997-07-01

    An Accelerator Based Neutron Source (ABNS) for Boron Neutron Capture Therapy (BNCT) was first proposed at The Ohio State University (OSU). Since the conception of the ABNS for BNCT, OSU has designed and optimized a moderator assembly based on in-air and in-phantom parameters. Additionally, the fabrication of the moderator assembly has commenced along with detailed analyses of the target and its heat removal system. In this dissertation, an integrated design of the ABNS is presented. This integrated design includes the high energy beam transport system (HEBT), the target and heat removal system (HRS), and the moderator assembly. In the integration process, a neutronic model of the HRS was developed and incorporated into the moderator assembly model. Additionally, a preliminary design of a HEBT system was developed that is compatible with both the HRS and the facility shielding. This dissertation also includes the completion of the fabrication of the moderator assembly and its experimental verification. The completion of the moderator assembly fabrication included the refabrication of the moderator and delimiter and the fabrication of the 6Li covering on the front of the moderator assembly. The experimental verification included neutron spectrum calculations and measurements in the irradiation port, and 3He detector response calculations and measurements in-phantom downstream of the moderator assembly.

  18. Pebble-bed core design option for VHTRs - Core configuration flexibility and potential applications

    SciTech Connect

    Pritchard, M. L.; Tsvetkov, P. V.

    2006-07-01

    Gas-cooled nuclear reactors have been receiving specific attention for Generation IV possibilities due to desired characteristics such as relatively low cost, short construction period, and inherent safety. Attractive inherent characteristics include an inert, single phase helium coolant, refractory coated fuel with high temperature capability and low fission product release, and graphite moderator with high temperature stability and long response times. The passively safe design has a relatively low power density, annular core, large negative temperature coefficient, and passive decay heat removal system. The objective of the U.S. DOE NERI Project is to assess the possibility, advantages and limitations of VHTRs with fuel loadings containing minor actinides. This paper presents the analysis of pebble-bed core configurations. Whole-core 3D models for pebble-bed design with multi-heterogeneity treatments in SCALE 5.0 are developed to compare computational results with experiments. Obtained results are in agreement with the available HTR-10 data. Actinide fueled VHTR configurations reveal promising performance. With an optimized pebble-bed model, the spectrum shifting abilities become more apparent. Effects of altered moderator to fuel ratio, Dancoff factor, and core and reflector configurations are investigated. This effort is anticipated to contribute to a facilitated development of new fuel cycles in support of future operation of Generation IV nuclear energy systems. (authors)

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

    DOE PAGESBeta

    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

  20. 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.

  1. Surface tension of the core-crust interface of neutron stars with global charge neutrality

    NASA Astrophysics Data System (ADS)

    Rueda, Jorge A.; Ruffini, Remo; Wu, Yuan-Bin; Xue, She-Sheng

    2014-03-01

    It has been shown recently that taking into account strong, weak, electromagnetic, and gravitational interactions, and fulfilling the global charge neutrality of the system, a transition layer will happen between the core and crust of neutron stars, at the nuclear saturation density. We use relativistic mean field theory together with the Thomas-Fermi approximation to study the detailed structure of this transition layer and calculate its surface and Coulomb energy. We find that the surface tension is proportional to a power-law function of the baryon number density in the core bulk region. We also analyze the influence of the electron component and the gravitational field on the structure of the transition layer and the value of the surface tension, to compare and contrast with known phenomenological results in nuclear physics. Based on the above results we study the instability against Bohr-Wheeler surface deformations in the case of neutron stars obeying global charge neutrality. Assuming the core-crust transition at nuclear density ρcore≈2.7×1014 g cm-3, we find that the instability sets the upper limit to the crust density, ρcrustcrit≈1.2×1014 g cm-3. This result implies a nonzero lower limit to the maximum electric field of the core-crust transition surface and makes inaccessible a limit of quasilocal charge neutrality in the limit ρcrust=ρcore. The general framework presented here can be also applied to study the stability of sharp phase transitions in hybrid stars as well as in strange stars, both bare and with outer crust. The results of this work open the way to a more general analysis of the stability of these transition surfaces, accounting for other effects such as gravitational binding, centrifugal repulsion, magnetic field induced by rotating electric field, and therefore magnetic dipole-dipole interactions.

  2. X-ray and Neutron Scattering Study of the Formation of Core-Shell-Type Polyoxometalates.

    PubMed

    Yin, Panchao; Wu, Bin; Mamontov, Eugene; Daemen, Luke L; Cheng, Yongqiang; Li, Tao; Seifert, Soenke; Hong, Kunlun; Bonnesen, Peter V; Keum, Jong Kahk; Ramirez-Cuesta, Anibal J

    2016-03-01

    A typical type of core-shell polyoxometalates can be obtained through the Keggin-type polyoxometalate-templated growth of a layer of spherical shell structure of {Mo72Fe30}. Small-angle X-ray scattering is used to study the structural features and stability of the core-shell structures in aqueous solutions. Time-resolved small-angle X-ray scattering is applied to monitor the synthetic reactions, and a three-stage formation mechanism is proposed to describe the synthesis of the core-shell polyoxometalates based on the monitoring results. New protocols have been developed by fitting the X-ray data with custom physical models, which provide more convincing, objective, and completed data interpretation. Quasi-elastic and inelastic neutron scattering are used to probe the dynamics of water molecules in the core-shell structures, and two different types of water molecules, the confined and structured water, are observed. These water molecules play an important role in bridging core and shell structures and stabilizing the cluster structures. PMID:26848596

  3. Prompt-period measurement of the Annular Core Research Reactor prompt neutron generation time

    SciTech Connect

    Coats, R.L.; Talley, D.G.; Trowbridge, F.R.

    1994-07-01

    The prompt neutron generation time for the Annular Core Research Reactor was experimentally determined using a prompt-period technique. The resultant value of 25.5 {mu}s agreed well with the analytically determined value of 24 {mu}s. The three different methods of reactivity insertion determination yielded {+-}5% agreement in the experimental values of the prompt neutron generation time. Discrepancies observed in reactivity insertion values determined by the three methods used (transient rod position, relative delayed critical control rod positions, and relative transient rod and control rod positions) were investigated to a limited extent. Rod-shadowing and low power fuel/coolant heat-up were addressed as possible causes of the discrepancies.

  4. Designing an Institutional Web-based Core Facility Management System

    PubMed Central

    Tabarini, D.; Clisham, S.; John, D.; Hagen, A.

    2011-01-01

    The authors and their four institutions collaborated to (i) identify the key challenges to core facility management; (ii) identify the requirements for an effective core facility management system; (iii) design, test and deploy such a system. Through a series of interviews with all participants in the core work flow (customers, core staff, administrators), the team identified a number of key challenges, including: (i) difficulty for researchers in identifying available services; (ii) inconsistent processes for requesting services; (iii) inadequate controls for approving service requests; (iv) inefficient processes for tracking and communicating about project processes; (v) time-consuming billing practices; (vi) incomplete revenue capture; (vii) manual reporting processes. The team identified the following requirements for a system to address these challenges: (i) ability to support a broad range of core business practices such as complex quote generation and project management; calendaring/equipment reservation management; sample tracking; complex forms; and import of usage data from hardware; (ii) ability to offer services for both internal and external customers, including flexible pricing and off-site access; (iii) ability to interact with institutional financial systems (e.g. SAP, PeopleSoft, Lawson, SunGard Banner) and identify management systems (e.g. Microsoft Active Directory, LDAP, and other SAML 2.0-compliant services). The team developed and deployed this system across the collaborative partners, as well as other major research institutions.

  5. CHINA SPALLATION NEUTRON SOURCE ACCELERATORS: DESIGN, RESEARCH, AND DEVELOPMENT.

    SciTech Connect

    WEI, J.; FU, S.; FANG, S.

    2006-06-26

    The China Spallation Neutron Source (CSNS) is a newly approved high-power accelerator project based on a H{sup -} linear accelerator and a rapid cycling synchrotron. During the past year, several major revisions were made on the design including the type of the front end, the linac frequency, the transport layout, the ring lattice, and the type of ring components. Here, we discuss the rationale of design revisions, status of the R&D efforts, and upgrade considerations.

  6. 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.

  7. Neutronic design of a Liquid Salt-cooled Pebble Bed Reactor (LSPBR)

    SciTech Connect

    De Zwaan, S. J.; Boer, B.; Lathouwers, D.; Kloosterman, J. L.

    2006-07-01

    A renewed interest has been raised for liquid salt cooled nuclear reactors. The excellent heat transfer properties of liquid salt coolants provide several benefits, like lower fuel temperatures, higher coolant outlet temperatures, increased core power density and better decay heat removal. In order to benefit from the online refueling capability of a pebble bed reactor, the Liquid Salt Pebble Bed Reactor (LSPBR) is proposed. This is a high temperature pebble-bed reactor with a fuel design similar to existing HTRs, but using a liquid salt as a coolant. In this paper, the selection criteria for the liquid salt coolant are described. Based on its neutronic properties, LiF-BeF{sub 2} (FLIBE) was selected for the LSPBR. Two designs of the LSPBR were considered: a cylindrical core and an annular core with a graphite inner reflector. Coupled neutronic-thermal hydraulic calculations were performed to obtain the steady state power distribution and the corresponding fuel temperatures. Finally, calculations were performed to investigate the decay heat removal capability in a protected loss-of-forced cooling accident. The maximum allowable power that can be produced with the LSPBR is hereby determined. (authors)

  8. Neutron source design for the Los Alamos APT system

    SciTech Connect

    Cappiello, M.W. )

    1993-01-01

    The Los Alamos National Laboratory accelerator production of tritium (APT) design incorporates a high-energy (1-GeV), high-current (200-mA) proton accelerator, a tungsten neutron source, and a moderating blanket that contains [sup 3]He for the production of tritium. To reduce the power density in the neutron source to acceptable levels, the incident proton beam is expanded to a 0.5- x 1.0-m [open quotes]beam spot.[close quotes] The beam passes through a double-wall vacuum interface window that is constructed of Inconel and is cooled with D[sub 2]O. The protons then enter a 1.3-m-diam x 3.5-m-long vessel also made of Inconel that contains [sup 3]He at moderate pressure (0.66 MPa) and 104 tungsten rod bundles that are distributed along the length of the vessel. The protons impinge on the tungsten and produce high-energy neutrons through the process of spallation. Behind the Inconel vessel is a proton backstop made of zirconium and lead that provides an additional source of neutrons and fully stops the proton beam. The Inconel vessel is surrounded by an annulus of aluminum-clad lead that is cooled with D[sub 2]O. Neutron multiplication occurs in the lead through additional spallation and (n,xn) reactions. Surrounding the lead is a thin annulus of [sup 3]He contained in aluminum tubing, a D[sub 2]O moderator, another annulus of [sup 3]He, and a D[sub 2]O reflector. Neutrons are moderated to near thermal energies in the D[sub 2]O. Tritium is produced through neutron capture in [sup 3]He that is circulated and processed continually. Thermal neutrons that are scattered back into the neutron source are preferentially captured in the [sup 3]He, thereby reducing the parasitic capture in the tungsten. The entire assembly of the neutron source vessel and surrounding blanket of lead, [sup 3]He, and D[sub 2]O are contained in a steel vessel.

  9. Neutronic design studies for an unattended, low power reactor

    SciTech Connect

    Palmer, R.G.; Durkee, J.W. Jr.

    1986-01-01

    The Los Alamos National Laboratory is involved in the design and demonstrations of a small, long-lived nuclear heat and electric power source for potential applications at remote sites where alternate fossil energy systems would not be cost effective. This paper describes the neutronic design analysis that was performed to arrive at two conceptual designs, one using thermoelectric conversion, the other using an organic Rankine cycle. To meet the design objectives and constraints a number of scoping and optimization studies were carried out. The results of calculations of control worths, temperature coefficients of reactivity and fuel depletion effects are reported.

  10. 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.

  11. Space neutron spectrometer design with SSPM-based instrumentation

    NASA Astrophysics Data System (ADS)

    Stapels, Christopher J.; Johnson, Erik B.; Chen, Xiao J.; Prettyman, Thomas H.; Benton, Eric R.; Christian, James F.

    2011-10-01

    The compact, robust nature of the CMOS solid-state photomultiplier (SSPM) allows the creation of small, low-power scintillation-based radiation measurement devices. Monitoring space radiation including solar protons and secondary neutrons generated from high-energy protons impinging on spacecraft is required to determine the dose to astronauts. Small size and highly integrated design are desired to minimize consumption of payload resources. RMD is developing prototype radiation measurement and personal dosimeter devices using emerging scintillation materials coupled to CMOS SSPM's for multiple applications. Spectroscopic measurements of high-energy protons and gamma-rays using tissue-equivalent, inorganic scintillators coupled to SSPM devices demonstrate the ability of an SSPM device to monitor the dose from proton and heavy ion particles, providing real time feedback to astronauts. Measurement of the dose from secondary neutrons introduces additional challenges due to the need to discriminate neutrons from other particle types and to accurately determine their energy deposition. We present strategies for measuring neutron signatures and assessing neutron dose including simulations of relevant environments and detector materials.

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

    PubMed

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

    1998-09-01

    The 7Li(p,n)7Be 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/AlF3, 7LiF, and D2O, 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 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/AlF3 or 7LiF 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% higher doses near the midline of the brain. For a single beam treatment, a proton beam current of 20 mA, and a 7LiF 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. PMID:9775379

  13. 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.

  14. 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.

  15. 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.

  16. 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)

  17. 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 is prevented. As a next step, the classical sodium plenum is replaced by a fission gas plenum (with lower sodium fraction), thus improving flow stability. Stable boiling at a steady power level is achieved in this final configuration. (authors)

  18. Pre-conceptual design and preliminary neutronic analysis of the proposed National Spallation Neutron Source (NSNS)

    SciTech Connect

    Johnson, J.O.; Barnes, J.M.; Charlton, L.A.

    1997-03-01

    The Department of Energy (DOE) has initiated a pre-conceptual design study for the National Spallation Neutron Source (NSNS) and given preliminary approval for the proposed facility to be built at Oak Ridge National Laboratory (ORNL). The pre-conceptual design of the NSNS initially consists of an accelerator system capable of delivering a 1 to 2 GeV proton beam with 1 MW of beam power in an approximate 0.5 {micro}s pulse at a 60 Hz frequency onto a single target station. The NSNS will be upgradable to a significantly higher power level with two target stations (a 60 Hz station and a 10 Hz station). There are many possible layouts and designs for the NSNS target stations. This paper gives a brief overview of the proposed NSNS with respect to the target station, as well as the general philosophy adopted for the neutronic design of the NSNS target stations. A reference design is presented, and some preliminary neutronic results for the NSNS are briefly discussed.

  19. Hyper-heuristic applied to nuclear reactor core design

    NASA Astrophysics Data System (ADS)

    Domingos, R. P.; Platt, G. M.

    2013-02-01

    The design of nuclear reactors gives rises to a series of optimization problems because of the need for high efficiency, availability and maintenance of security levels. Gradient-based techniques and linear programming have been applied, as well as genetic algorithms and particle swarm optimization. The nonlinearity, multimodality and lack of knowledge about the problem domain makes de choice of suitable meta-heuristic models particularly challenging. In this work we solve the optimization problem of a nuclear reactor core design through the application of an optimal sequence of meta-heuritics created automatically. This combinatorial optimization model is known as hyper-heuristic.

  20. 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.

  1. System design description for GCFR-core flow test loop

    SciTech Connect

    Huntley, W.R.; Grindell, A.G.

    1980-12-01

    The Core Flow Test Loop is a high-pressure, high-temperature, out-of-reactor helium circulation system that is being constructed to permit detailed study of the thermomechanical and thermal performance at prototypic steady-state and transient operating conditions of simulated segments of core assemblies for a GCFR Demonstration Plant, as designed by General Atomic Company. It will also permit the expermental verification of predictive analytical models of the GCFR core assemblies needed to reduce operational and safety uncertainties of the GCFR. Full-sized blanket assemblies and segments of fuel rod and control rod fuel assemblies will be simulated with test bundles of electrically powered fuel rod or blanket rod simulators. The loop will provide the steady-state and margin test requirements of bundle power and heat removal, and of helium coolant flow rate, pressure, and temperature for test bundles having up to 91 rods; these requirements set the maximum power, coolant helium flow, and thermal requirements for the loop. However, the size of the test vessel that contains the test bundles will be determined by the bundles that simulate a full-sized GCFR blanket assembly. The loop will also provide for power and coolant transients to simulate transient operation of GCFR core assemblies, including the capability for rapid helium depressurization to simulate the depressurization class of GCFR accidents. In addition, the loop can be used as an out-of-reactor test bed for characterizing in-reactor test bundle configurations.

  2. Conceptual design of a camera system for neutron imaging in low fusion power tokamaks

    NASA Astrophysics Data System (ADS)

    Xie, X.; Yuan, X.; Zhang, X.; Nocente, M.; Chen, Z.; Peng, X.; Cui, Z.; Du, T.; Hu, Z.; Li, T.; Fan, T.; Chen, J.; Li, X.; Zhang, G.; Yuan, G.; Yang, J.; Yang, Q.

    2016-02-01

    The basic principles for designing a camera system for neutron imaging in low fusion power tokamaks are illustrated for the case of the HL-2A tokamak device. HL-2A has an approximately circular cross section, with total neutron yields of about 1012 n/s under 1 MW neutral beam injection (NBI) heating. The accuracy in determining the width of the neutron emission profile and the plasma vertical position are chosen as relevant parameters for design optimization. Typical neutron emission profiles and neutron energy spectra are calculated by Monte Carlo method. A reference design is assumed, for which the direct and scattered neutron fluences are assessed and the neutron count profile of the neutron camera is obtained. Three other designs are presented for comparison. The reference design is found to have the best performance for assessing the width of peaked to broadened neutron emission profiles. It also performs well for the assessment of the vertical position.

  3. 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.

  4. 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.

  5. 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.

  6. 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

  7. 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.

  8. Isoscalar-vector interaction and hybrid quark core in massive neutron stars

    NASA Astrophysics Data System (ADS)

    Shao, G. Y.; Colonna, M.; Di Toro, M.; Liu, Y. X.; Liu, B.

    2013-05-01

    The hadron-quark phase transition in the core of massive neutron stars is studied with a newly constructed two-phase model. For nuclear matter, a nonlinear Walecka type model with general nucleon-meson and meson-meson couplings, recently calibrated by Steiner, Hemper and Fischer, is taken. For quark matter, a modified Polyakov-NambuJona-Lasinio model, which gives consistent results with lattice QCD data, is used. Most importantly, we introduce an isoscalar-vector interaction in the description of quark matter, and we study its influence on the hadron-quark phase transition in the interior of massive neutron stars. With the constraints of neutron star observations, our calculation shows that the isoscalar-vector interaction between quarks is indispensable if massive hybrids star exist in the universe, and its strength determines the onset density of quark matter, as well as the mass-radius relations of hybrid stars. Furthermore, as a connection with heavy-ion-collision experiments we give some discussions about the strength of isoscalar-vector interaction and its effect on the signals of hadron-quark phase transition in heavy-ion collisions, in the energy range of the NICA at JINR-Dubna and FAIR at GSI-Darmstadt facilities.

  9. Development of an inconel self powered neutron detector for in-core reactor monitoring

    NASA Astrophysics Data System (ADS)

    Alex, M.; Ghodgaonkar, M. D.

    2007-04-01

    The paper describes the development and testing of an Inconel600 (2 mm diameter×21 cm long) self-powered neutron detector for in-core neutron monitoring. The detector has 3.5 mm overall diameter and 22 cm length and is integrally coupled to a 12 m long mineral insulated cable. The performance of the detector was compared with cobalt and platinum detectors of similar dimensions. Gamma sensitivity measurements performed at the 60Co irradiation facility in 14 MR/h gamma field showed values of -4.4×10 -18 A/R/h/cm (-9.3×10 -24 A/ γ/cm 2-s/cm), -5.2×10 -18 A/R/h/cm (-1.133×10 -23 A/ γ/cm 2-s/cm) and 34×10 -18 A/R/h/cm (7.14×10 -23 A/ γ/cm 2-s/cm) for the Inconel, Co and Pt detectors, respectively. The detectors together with a miniature gamma ion chamber and fission chamber were tested in the in-core Apsara Swimming Pool type reactor. The ion chambers were used to estimate the neutron and gamma fields. With an effective neutron cross-section of 4b, the Inconel detector has a total sensitivity of 6×10 -23 A/nv/cm while the corresponding sensitivities for the platinum and cobalt detectors were 1.69×10 -22 and 2.64×10 -22 A/nv/cm. The linearity of the detector responses at power levels ranging from 100 to 200 kW was within ±5%. The response of the detectors to reactor scram showed that the prompt response of the Inconel detector was 0.95 while it was 0.7 and 0.95 for the platinum and cobalt self-powered detectors, respectively. The detector was also installed in the horizontal flux unit of 540 MW Pressurised Heavy Water Reactor (PHWR). The neutron flux at the detector location was calculated by Triveni code. The detector response was measured from 0.02% to 0.07% of full power and showed good correlation between power level and detector signals. Long-term tests and the dynamic response of the detector to shut down in PHWR are in progress.

  10. Design of MR brake featuring tapered inner magnetic core

    NASA Astrophysics Data System (ADS)

    Sohn, Jung Woo; Oh, Jong-Soek; Choi, Seung-Bok

    2015-04-01

    In this work, a new type of MR brake featuring tapered inner magnetic core is proposed and its braking performance is numerically evaluated. In order to achieve high braking torque with restricted size and weight of MR brake system, tapered inner magnetic core is designed and expands the area that the magnetic flux is passing by MR fluid-filled gap. The mathematical braking torque model of the proposed MR brake is derived based on the field-dependent Bingham rheological model of MR fluid. Finite element analysis is carried out to identify electromagnetic characteristics of the conventional and the proposed MR brake configuration. To demonstrate the superiority of the proposed MR brake, the braking torque of the proposed MR brake is numerically evaluated and compared with that of conventional MR brake model.

  11. Toward a final design for the Birmingham boron neutron capture therapy neutron beam.

    PubMed

    Allen, D A; Beynon, T D; Green, S; James, N D

    1999-01-01

    This paper is concerned with the proposed Birmingham accelerator-based epithermal neutron beam for boron neutron capture therapy (BNCT). Details of the final moderator design, such as beam delimiter, shield, and beam exit surface shape are considered. Monte Carlo radiation transport simulations with a head and body phantom have shown that a simple flat moderator beam exit surface is preferable to the previously envisioned spherical design. Dose rates to individual body organs during treatment have been calculated using a standard MIRD phantom. We have shown that a simple polyethylene shield, doped with natural lithium, is sufficient to provide adequate protection to the rest of the body during head irradiations. The effect upon the head phantom dose distributions of the use of such a shield to delimit the therapy beam has been evaluated. PMID:9949401

  12. 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.

  13. 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.

  14. 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.

  15. Core damage frequency (reactor design) perspectives based on IPE results

    SciTech Connect

    Camp, A.L.; Dingman, S.E.; Forester, J.A.

    1996-12-31

    This paper provides perspectives gained from reviewing 75 Individual Plant Examination (IPE) submittals covering 108 nuclear power plant units. Variability both within and among reactor types is examined to provide perspectives regarding plant-specific design and operational features, and C, modeling assumptions that play a significant role in the estimates of core damage frequencies in the IPEs. Human actions found to be important in boiling water reactors (BWRs) and in pressurized water reactors (PWRs) are presented and the events most frequently found important are discussed.

  16. Benchmark Evaluation of the Neutron Radiography (NRAD) Reactor Upgraded LEU-Fueled Core

    SciTech Connect

    John D. Bess

    2001-09-01

    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. The final upgraded core configuration with 64 fuel elements has been completed. Evaluated benchmark measurement data include criticality, control-rod worth measurements, shutdown margin, and excess reactivity. Dominant uncertainties in keff include the manganese content and impurities contained within the stainless steel cladding of the fuel and the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 nuclear data are approximately 1.4% greater than the benchmark model eigenvalue, supporting contemporary research regarding errors in the cross section data necessary to simulate TRIGA-type reactors. Uncertainties in reactivity effects measurements are estimated to be ~10% with calculations in agreement with benchmark experiment values within 2s. The completed benchmark evaluation de-tails are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Experiments (IRPhEP Handbook). Evaluation of the NRAD LEU cores containing 56, 60, and 62 fuel elements have also been completed, including analysis of their respective reactivity effects measurements; they are also available in the IRPhEP Handbook but will not be included in this summary paper.

  17. Dynamic behavior of homogeneous and heterogeneous LMFBR core-design concepts

    SciTech Connect

    Chang, Y.I.; Henryson, H. II; Orechwa, Y.; Su, S.F.; Greenman, G.; Blomquist, R.

    1981-01-01

    The emphasis is placed on obtaining an understanding of the inherent difference between homogeneous and heterogeneous core configurations regarding neutronic characteristics related to the dynamic behavior. The space-time neutronic and thermal-hydraulic behavior was analyzed in detail for various core configurations by using the FX2-TH, a two-dimensional kinetics code with thermal-hydraulic feedback. In addition, the relationship between the flux tilt and the fundamental-to-first harmonic eigenvalue separation, and the sodium void reactivity in heterogeneous cores were also sutdied.

  18. SPALLATION NEUTRON SOURCE RING-DESIGN AND CONSTRUCTION SUMMARY.

    SciTech Connect

    WEI,J.

    2005-05-16

    After six years, the delivery of components for the Spallation Neutron Source (SNS) accumulator ring (AR) and the transport lines was completed in Spring 2005. Designed to deliver 1.5 MW beam power (1.5 x 10{sup 14} protons of 1 GeV kinetic energy at a repetition rate of 60 Hz), stringent measures were implemented in the fabrication, test, and assembly to ensure the quality of the accelerator systems. This paper summarizes the design, R&D, and construction of the ring and transport systems.

  19. Burning of two-flavor quark matter into strange matter in neutron stars and in supernova cores

    SciTech Connect

    Anand, J.D.; Goyal, A.; Gupta, V.K.; Singh, S.

    1997-06-01

    Assuming a first-order phase transition from nuclear to quark matter in neutron stars and in supernova cores, we have studied the phase transition from two-flavor quark matter to strange matter. This transition has bearing on the cooling of neutron stars and may lead to observable signals in the form of a second neutrino burst. In the case of transition occurring in a supernova core, it has the effect of raising the core temperature and the energy of the shock wave and thus affecting the evolution of the core. In this study we have systematically taken into account the effect of strong interactions perturbatively to order {alpha}{sub c} and the effect of finite temperature and strange quark mass. {copyright} {ital 1997} {ital The American Astronomical Society}

  20. 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.

  1. Core transform design for high-efficiency video coding

    NASA Astrophysics Data System (ADS)

    Dong, Jie; Ye, Yan

    2012-10-01

    High Efficiency Video Coding (HEVC) is the next generation video coding standard currently being developed by the Joint Collaborative Team on Video Coding (JCT-VC). It employs various coding unit sizes 2K×2K, where K is a positive integer with the typical values from 3 to 6; it also uses larger transform sizes up to 32×32. This raises the interest in seeking high performance higher order integer transforms with low computation requirements. This paper presents approaches to designing order-N (N=4, 8, 16, 32) integer transforms, by which the derived integer transforms have special symmetry structures to ensure the matrix factorization. The proposed set of high order integer transforms with well selected elements demonstrates excellent coding performance, compared with the core transform design in HEVC.

  2. 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.

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

    DOE PAGESBeta

    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.

  4. ORR core re-configuration measurements to increase the fast neutron flux in the Magnetic Fusion Energy (MFE) experiments

    NASA Astrophysics Data System (ADS)

    Hobbs, R. W.; Stinnett, R. M.; Sims, T. M.

    1985-06-01

    The relative increases obtainable in the fast neutron flux in the Magnetic Fusion Energy (MFE) experiment positions were studied by reconfiguring the current ORR core. The percentage increase possible in the current displacement per atom (dpa) rate was examined. The principle methods to increase the fast flux, consisted of reducing the current core size (number of fuel elements), to increase the core average power density and arrangement of the fuel elements in the reduced-size core to tilt the core power distribution towards the MFE positions were investigated. It is concluded that fast fluxes in the E-3 core position can be increased by approximately 15 to 20% over current values and in E-5 by approximately 45 to 55%.

  5. 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, an icosahedral hydrostatic dynamical core designed for consistency and versatility", in preparation.

  6. 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)

  7. Design of a multi-element TEPC for neutron monitoring.

    PubMed

    Waker, A J; Aslam; Lori, J

    2011-02-01

    Tissue-equivalent proportional counters (TEPCs) have long been considered suitable candidate instruments for more accurate neutron monitors in nuclear power plants. It has also been recognised that the production of truly light-weight devices based on TEPCs requires further effort directed towards increasing their sensitivity and decreasing their physical size. This paper deals with the construction of a multi-element TEPC (METEPC) designed to have the sensitivity of a 12.7-cm (5-in.) diameter spherical TEPC, but with approximately one-tenth of its physical size. Construction of the METEPC is achieved by machining 61 elongated cylindrical cavities in a single block of A150 TE plastic. Comparative measurements carried out in neutron fields with mean energies ranging from 34 to 354 keV demonstrate that the METEPC constructed does match the sensitivity of a 5-in. spherical TEPC and that microdosimetric lineal energy spectra measured with both detectors have the same features and show the same changes with neutron radiation quality. PMID:21186210

  8. 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.

  9. On the charge to mass ratio of neutron cores and heavy nuclei

    SciTech Connect

    Patricelli, B.; Rotondo, M.; Ruffini, R.

    2008-01-03

    We determine theoretically the relation between the total number of protons N{sub p} and the mass number A (the charge to mass ratio) of nuclei and neutron cores with the model recently proposed by Ruffini et al. (2007) and we compare it with other N{sub p} versus A relations: the empirical one, related to the Periodic Table, and the semi-empirical relation, obtained by minimizing the Weizsaecker mass formula. We find that there is a very good agreement between all the relations for values of A typical of nuclei, with differences of the order of per cent. Our relation and the semi-empirical one are in agreement up to A{approx}10{sup 4}; for higher values, we find that the two relations differ. We interprete the different behaviour of our theoretical relation as a result of the penetration of electrons (initially confined in an external shell) inside the core, that becomes more and more important by increasing A; these effects are not taken into account in the semi-empirical mass-formula.

  10. 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…

  11. 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

  12. Quadrupole Transition Strength in the 7Ni4 Nucleus and Core Polarization Effects in the Neutron-Rich Ni Isotopes

    NASA Astrophysics Data System (ADS)

    Marchi, T.; de Angelis, G.; Valiente-Dobn, J. J.; Bader, V. M.; Baugher, T.; Bazin, D.; Berryman, J.; Bonaccorso, A.; Clark, R.; Coraggio, L.; Crawford, H. L.; Doncel, M.; Farnea, E.; Gade, A.; Gadea, A.; Gargano, A.; Glasmacher, T.; Gottardo, A.; Gramegna, F.; Itaco, N.; John, P. R.; Kumar, R.; Lenzi, S. M.; Lunardi, S.; McDaniel, S.; Michelagnoli, C.; Mengoni, D.; Modamio, V.; Napoli, D. R.; Quintana, B.; Ratkiewicz, A.; Recchia, F.; Sahin, E.; Stroberg, R.; Weisshaar, D.; Wimmer, K.; Winkler, R.

    2014-10-01

    The reduced transition probability B (E 2 ;0+?2+) has been measured for the neutron-rich nucleus 7Ni4 in an intermediate energy Coulomb excitation experiment performed at the National Superconducting Cyclotron Laboratory at Michigan State University. The obtained B (E 2 ;0+?2+)=64 2-226+216 e2 fm4 value defines a trend which is unexpectedly small if referred to 7Ni0 and to a previous indirect determination of the transition strength in 7Ni4 . This indicates a reduced polarization of the Z =28 core by the valence neutrons. Calculations in the p f g d model space reproduce well the experimental result indicating that the B (E 2 ) strength predominantly corresponds to neutron excitations. The ratio of the neutron and proton multipole matrix elements supports such an interpretation.

  13. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2008-03-01

    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.

  15. 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.

  16. 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.

  17. A comparative neutronic feasibility study for a hydrogen, deuterium and helium cold neutron sources situated in the center of a nuclear reactor core

    NASA Astrophysics Data System (ADS)

    Chatila, Malek

    A tool was developed to calculate the average cold neutron flux that could be generated for a spherically shaped cold neutron source situated in the center of a nuclear reactor core. The tool also estimates the subsequent nuclear heating of the cold source. The results were compared for three different cold source mediums; hydrogen, deuterium and helium. The tool utilizes the consistent energy dependent P1 equations to generate the fast neutron energy spectrum, the Grueling-Goertzel equations to generate the slow spectrum and the Proton Gas Model to generate the cold energy spectrum. These spectrums are then used to collapse the group constants into three energy groups. The cold flux that can be generated in different mediums is then calculated by utilizing the three energy group constants in SN-6, 2 regions calculations.

  18. Tokamak Fusion Core Experiment: design studies based on superconducting and hybrid toroidal field coils. Design overview

    SciTech Connect

    Flanagan, C.A.

    1984-10-01

    This document is a design overview that describes the scoping studies and preconceptual design effort performed in FY 1983 on the Tokamak Fusion Core Experiment (TFCX) class of device. These studies focussed on devices with all-superconducting toroidal field (TF) coils and on devices with superconducting TF coils supplemented with copper TF coil inserts located in the bore of the TF coils in the shield region. Each class of device is designed to satisfy the mission of ignition and long pulse equilibrium burn. Typical design parameters are: major radius = 3.75 m, minor radius = 1.0 m, field on axis = 4.5 T, plasma current = 7.0 MA. These designs relay on lower hybrid (LHRH) current rampup and heating to ignition using ion cyclotron range of frequency (ICRF). A pumped limiter has been assumed for impurity control. The present document is a design overview; a more detailed design description is contained in a companion document.

  19. Design study of a medical linac for neutron therapy

    NASA Astrophysics Data System (ADS)

    Raparia, D.; Machida, S.

    1989-04-01

    A 66 MeV proton linac for neutron therapy has been studied and conceptual designs for low energy beam transport (LEBT), a radio-frequency quadrupole (RFQ) and a drift tube linac (DTL) have been achieved. The machine is compact and simple enough to be operated in hospitals. The LEBT consists of two Einzel lenses and is 22 cm long. The 425 MHz RFQ is designed for 50 mA peak current at 60 Hz and is 1 m long. The 18-m-long DTL has the same frequency as the RFQ and uses permanent quadrupole magnets. Beam dynamics are very important for this compact and high-brightness machine. This paper emphasizes our investigation of the beam dynamics.

  20. 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.

  1. Westinghouse AP600 core design diversification for compliance with utility requirements

    SciTech Connect

    Carlson, W.R.; Hoerner, J.A.

    1994-12-31

    The utility requirements document (URD) has necessitated diversification in the reactor system designs of advanced LWRs, especially in the area of core design. The AP600 core design incorporates both proven and innovative design features to insure compliance with these utility functional requirements. A fuel rod diagram is included in the paper.

  2. Core design studies for a 1000 MW{sub th} advanced burner reactor.

    SciTech Connect

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

    2009-04-01

    This paper describes the core design and performance characteristics of 1000 MW{sub th} Advanced Burner Reactor (ABR) core concepts with a wide range of TRU conversion ratio. Using ternary metal alloy and mixed oxide fuels, reference core designs of a medium TRU conversion ratio of {approx}0.7 were developed by trade-off between burnup reactivity loss and TRU conversion ratio. Based on these reference core concepts, TRU burner cores with low and high TRU conversion ratios were developed by changing the intra-assembly design parameters and core configurations. Reactor performance characteristics were evaluated in detail, including equilibrium cycle core performances, reactivity feedback coefficients, and shutdown margins. The results showed that by employing different assembly designs, a wide range of TRU conversion ratios from {approx}0.2 to break-even can be achieved within the same core without introducing significant performance and safety penalties.

  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. PMID:26297075

  4. Facing Challenges for Monte Carlo Analysis of Full PWR Cores : Towards Optimal Detail Level for Coupled Neutronics and Proper Diffusion Data for Nodal Kinetics

    NASA Astrophysics Data System (ADS)

    Nuttin, A.; Capellan, N.; David, S.; Doligez, X.; El Mhari, C.; Méplan, O.

    2014-06-01

    Safety analysis of innovative reactor designs requires three dimensional modeling to ensure a sufficiently realistic description, starting from steady state. Actual Monte Carlo (MC) neutron transport codes are suitable candidates to simulate large complex geometries, with eventual innovative fuel. But if local values such as power densities over small regions are needed, reliable results get more difficult to obtain within an acceptable computation time. In this scope, NEA has proposed a performance test of full PWR core calculations based on Monte Carlo neutron transport, which we have used to define an optimal detail level for convergence of steady state coupled neutronics. Coupling between MCNP for neutronics and the subchannel code COBRA for thermal-hydraulics has been performed using the C++ tool MURE, developed for about ten years at LPSC and IPNO. In parallel with this study and within the same MURE framework, a simplified code of nodal kinetics based on two-group and few-point diffusion equations has been developed and validated on a typical CANDU LOCA. Methods for the computation of necessary diffusion data have been defined and applied to NU (Nat. U) and Th fuel CANDU after assembly evolutions by MURE. Simplicity of CANDU LOCA model has made possible a comparison of these two fuel behaviours during such a transient.

  5. 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.

  6. Designing a new type of neutron detector for neutron and gamma-ray discrimination via GEANT4.

    PubMed

    Shan, Qing; Chu, Shengnan; Ling, Yongsheng; Cai, Pingkun; Jia, Wenbao

    2016-04-01

    Design of a new type of neutron detector, consisting of a fast neutron converter, plastic scintillator, and Cherenkov detector, to discriminate 14-MeV fast neutrons and gamma rays in a pulsed n-γ mixed field and monitor their neutron fluxes is reported in this study. Both neutrons and gamma rays can produce fluorescence in the scintillator when they are incident on the detector. However, only the secondary charged particles of the gamma rays can produce Cherenkov light in the Cherenkov detector. The neutron and gamma-ray fluxes can be calculated by measuring the fluorescence and Cherenkov light. The GEANT4 Monte Carlo simulation toolkit is used to simulate the whole process occurring in the detector, whose optimum parameters are known. Analysis of the simulation results leads to a calculation method of neutron flux. This method is verified by calculating the neutron fluxes using pulsed n-γ mixed fields with different n/γ ratios, and the results show that the relative errors of all calculations are <5%. PMID:26844541

  7. Core design for use with precision composite reflectors

    NASA Technical Reports Server (NTRS)

    Porter, Christopher C. (Inventor); Jacoy, Paul J. (Inventor); Schmitigal, Wesley P. (Inventor)

    1992-01-01

    A uniformly flexible core, and method for manufacturing the same, is disclosed for use between the face plates of a sandwich structure. The core is made of a plurality of thin corrugated strips, the corrugations being defined by a plurality of peaks and valleys connected to one another by a plurality of diagonal risers. The corrugated strips are orthogonally criss-crossed to form the core. The core is particularly suitable for use with high accuracy spherically curved sandwich structures because undesirable stresses in the curved face plates are minimized due to the uniform flexibility characteristics of the core in both the X and Y directions. The core is self venting because of the open geometry of the corrugations. The core can be made from any suitable composite, metal, or polymer. Thermal expansion problems in sandwich structures may be minimized by making the core from the same composite materials that are selected in the manufacture of the curved face plates because of their low coefficients of thermal expansion. Where the strips are made of a composite material, the core may be constructed by first cutting an already cured corrugated sheet into a plurality of corrugated strips and then secondarily bonding the strips to one another or, alternatively, by lying a plurality of uncured strips orthogonally over one another in a suitable jig and then curing and bonding the entire plurality of strips to one another in a single operation.

  8. Limiting the Range of Terms Describing Matter in Supernova Cores and Proto Neutron Stars

    NASA Astrophysics Data System (ADS)

    Hudson, David

    2001-10-01

    Core collapse of supernovae and formation of proto neutron stars result in the rapid interralated variation of several physical quantities. However, in many studies, various parameters are fixed throughout the calculation. Particularly, the temperature is commonly set at zero. In this work, I specifically allow the temperature, effective nucleon mass, density and nuclear compression modulus to vary within ranges reported in the literature. Of special interest are the unusual results from a lattice calculation recently reported by Muller et. al. [1] showing a dramatic spike centered at a temperature of 14 MeV. Using and extending the analytic model of Baron, Cooperstein and Kahana [2] the influence on various factors such as pressure, adiabatic index and energy of compression are presented. The purpose of this work is to identify topics which may significantly reward more rigerous study. 1. Muller, Koonin, Seki, and van Kolck: Phys. Rev. C61, 044320 (2000). 2. Baron, Cooperstein, and Kahana: Nucl. Phys. A440, 744 (1985); Phys. Rev. Lett. 55, 126 (1985)

  9. 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.

  10. Spallation neutron source cryomodule heat loads and thermal design

    SciTech Connect

    E. F. Daly; V. Ganni; C. H. Rode; W. J. Schneider; K. M. Wilson; M. A. Wiseman

    2002-05-10

    When complete, the Spallation Neutron Source (SNS) will provide a 1 GeV, 2 MW beam for experiments. One portion of the machine's linac consists of over 80 Superconducting Radio Frequency (SRF) 805 MHz cavities housed in a minimum of 23 cryomodules operating at a saturation temperature of 2.1 K. Minimization of the total heat load is critical to machine performance and for efficient operation of the system. The total heat load of the cryomodules consists of the fixed static load and the dynamic load, which is proportional to the cavity performance. The helium refrigerator supports mainly the cryomodule loads and to a lesser extent the distribution system loads. The estimated heat loads and calculated thermal performance are discussed along with two unique features of this design: the helium heat exchanger housed in the cryomodule return end can and the helium gas cooled fundamental power coupler.

  11. 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.

  12. 17 CFR 37.1500 - Core Principle 15-Designation of chief compliance officer.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 17 Commodity and Securities Exchanges 1 2014-04-01 2014-04-01 false Core Principle 15-Designation of chief compliance officer. 37.1500 Section 37.1500 Commodity and Securities Exchanges COMMODITY FUTURES TRADING COMMISSION SWAP EXECUTION FACILITIES Designation of Chief Compliance Officer 37.1500 Core Principle 15Designation of chief...

  13. 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.

  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. IB: a Monte Carlo Simulation Tool for Neutron Scattering Instrument Design under Parallel Virtual Machine

    SciTech Connect

    Zhao, Jinkui

    2011-01-01

    IB is a Monte Carlo simulation tool for aiding neutron scattering instrument designs. It is written in C++ and implemented under Parallel Virtual Machine. The program has a few basic components, or modules, that can be used to build a virtual neutron scattering instrument. More complex components, such as neutron guides and multichannel beam benders, can be constructed using the grouping technique unique to IB. Users can specify a collection of modules as a group. For example, a neutron guide can be constructed by grouping four neutron mirrors together that make up the four sides of the guide. IB s simulation engine ensures that neutrons entering a group will be properly operated upon by all members of the group. For simulations that require higher computer speed, the program can be run in parallel mode under the PVM architecture. Initially, the program was written for designing instruments on pulsed neutron sources, it has since been used to simulate reactor based instruments as well.

  16. A NEW EQUATION OF STATE FOR NEUTRON STAR MATTER WITH NUCLEI IN THE CRUST AND HYPERONS IN THE CORE

    SciTech Connect

    Miyatsu, Tsuyoshi; Yamamuro, Sachiko; Nakazato, Ken'ichiro

    2013-11-01

    The equation of state for neutron stars in a wide-density range at zero temperature is constructed. The chiral quark-meson coupling model within relativistic Hartree-Fock approximation is adopted for uniform nuclear matter. The coupling constants are determined so as to reproduce the experimental data of atomic nuclei and hypernuclei. In the crust region, nuclei are taken into account within the Thomas-Fermi calculation. All octet baryons are considered in the core region, while only Ξ{sup –} appears in neutron stars. The resultant maximum mass of neutron stars is 1.95 M{sub ☉}, which is consistent with the constraint from the recently observed massive pulsar, PSR J1614-2230.

  17. 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.

  18. 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.

  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. 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.

  1. Design of an Aluminum Proton Beam Window for the Spallation Neutron Source

    SciTech Connect

    Janney, Jim G; McClintock, David A

    2012-01-01

    An aluminum proton beam window design is being considered at the Spallation Neutron Source primarily to increase the lifetime of the window, with secondary advantages of higher beam transport efficiency and lower activation. The window separates the core vessel, the location of the mercury target, from the vacuum of the accelerator, while withstanding the pass through of a proton beam of up to 2 MW with 1.0 GeV proton energy. The current aluminum alloy being investigated for the window material is 6061-T651 due to its combination of high strength, high thermal conductivity, and good resistance to aqueous corrosion, as well as demonstrated dependability in previous high-radiation environments. The window design will feature a thin plate with closely spaced cross drilled cooling holes. An analytical approach was used to optimize the dimensions of the window before finite element analysis was used to simulate temperature profiles and stress fields resulting from thermal and static pressure loading. The resulting maximum temperature of 60 C and Von Mises stress of 71 MPa are very low compared to allowables for Al 6061-T651. A significant challenge in designing an aluminum proton beam window for SNS is integrating the window with the current 316L SS shield blocks. Explosion bonding was chosen as a joining technique because of the large bonding area required. A test program has commenced to prove explosion bonding can produce a robust vacuum joint. Pending successful explosion bond testing, the aluminum proton beam window design will be proven acceptable for service in the Spallation Neutron Source.

  2. 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 temperature was kept the same for the BWR-HD and ABWR which resulted in 4 °K cooler core inlet temperature for the BWR-HD given that its feedwater makes up a larger fraction of total core flow. The stability analysis using the STAB and S3K codes showed satisfactory results for the hot channel, coupled regional out-of-phase and coupled core-wide in-phase modes. A RELAPS model of the ABWR system was constructed and applied to six transients for the BWR-HD and ABWR. The 6MCPRs during all the transients were found to be equal or less for the new design and the core remained covered for both. The lower void coefficient along with smaller core volume proved to be advantages for the simulated transients. Helical Cruciform Fuel (HCF) rods were proposed in prior MIT studies to enhance the fuel surface to volume ratio. In this work, higher fidelity models (e.g. CFD instead of subchannel methods for the hydraulic behaviour) are used to investigate the resolution needed for accurate assessment of the HCF design. For neutronics, conserving the fuel area of cylindrical rods results in a different reactivity level with a lower void coefficient for the HCF design. In single-phase flow, for which experimental results existed, the friction factor is found to be sensitive to HCF geometry and cannot be calculated using current empirical models. A new approach for analysis of flow crisis conditions for HCF rods in the context of Departure from Nucleate Boiling (DNB) and dryout using the two phase interface tracking method was proposed and initial results are presented. It is shown that the twist of the HCF rods promotes detachment of a vapour bubble along the elbows which indicates no possibility for an early DNB for the HCF rods and in fact a potential for a higher DNB heat flux. Under annular flow conditions, it was found that the twist suppressed the liquid film thickness on the HCF rods, at the locations of the highest heat flux, which increases the possibility of reaching early dryout. It was also shown that modeling the 3D heat and stress distribution in the HCF rods is necessary for accurate steady state and transient analyses. (Abstract shortened by UMI.) (Copies available exclusively from MIT Libraries, libraries.mit.edu/docs - docs mit.edu)

  3. Determination of the response function for two personal neutron dosemeter designs based on PADC.

    PubMed

    Mayer, S; Assenmacher, F; Boschung, M

    2014-10-01

    Since 1998 neutron dosimetry based on PADC (poly allyl diglycol carbonate) is done with a so-called original Paul Scherrer Institute (PSI) design at PSI. The original design (i.e. holder) was later changed. Both designs are optimised for use in workplaces around high-energy accelerators, where the neutron energy spectra are dominated by fast neutrons ranging up to some 100 MeV. In addition to the change of the dosemeter design a new evaluation method based on a microscope scanning technique has been introduced and the etching conditions have been optimised. In the present work, the responses obtained with the original and the new dosemeter designs are compared for fields of radionuclide sources and monoenergetic reference fields using the new evaluation method. The response curves in terms of the personal dose equivalent for normally incident neutrons were built as functions of the incident neutron energy. PMID:24179145

  4. Neutronic design of the APT Target/Blanket

    SciTech Connect

    Pitcher, E.J.; Russell, G.J.; Kidman, R.B.; Ferguson, P.D.

    1997-12-01

    The primary function of the Accelerator Production of Tritium Target/Blanket assembly is the safe and efficient production of tritium. The T/B accepts a 1.7-GeV, 100-mA proton beam and produces neutrons via the spallation process. These neutrons then react with {sup 3}He to produce tritium. Neutronic optimization of the T/B is achieved by efficiently using the proton beam to produce neutrons and then, once produced, assuring that they are captured mostly by {sup 3}He. This optimization must occur within the constraints imposed by engineering considerations such as heat flux limits, structural integrity, fabricability, and safe and reliable operation. The target/blanket achieves these goals with a neutron production rate that is 75% of that achievable with an ideal target, and a neutronic efficiency of 84%, leading to an overall tritium production rate that is 63% of the theoretical maximum.

  5. 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.

  6. Monte Carlo Calculation of Core Reactivity and Fluxes for the Development of the BNCT Neutron Source at the Kyiv Research Reactor

    NASA Astrophysics Data System (ADS)

    Gritzay, Olena; Kalchenko, Oleksandr; Klimova, Nataliya; Razbudey, Volodymyr; Sanzhur, Andriy; Binney, Stephen

    2005-05-01

    The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor safety. The general Monte Carlo radiation transport code MCNP, version 4B, has been used for these calculations.

  7. 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.

  8. 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.

  9. Detailed design of ex-vessel neutron yield monitor for ITER

    NASA Astrophysics Data System (ADS)

    Asai, K.; Iguchi, T.; Watanabe, K.; Kawarabayashi, J.; Nishitani, T.; Walker, C. I.

    2004-10-01

    Taking into consideration the latest design of the International Thermonuclear Experimental Reactor (ITER) main units, we have made the detailed design consideration for an ex-vessel neutron yield monitor to meet the ITER requirements. The monitoring system is constructed of four detector modules consisting of several 235U fission chambers with different sensitivities and graphite (or beryllium) neutron moderator. We also selected possible spaces in the diagnostic ports to install them at appropriate distances and neutron shielding effects from the plasma. Through Monte Carlo neutron transport calculations, it has been confirmed that the present system can cover all the neutron yields encountered in the ITER experiments including the in situ calibrations with a time resolution around 200 μs without detector replacement over the whole ITER experiments. This system can also be calibrated with 10% of required accuracies in a realistic 50 h of accumulation time using a DT neutron generator.

  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. Current directions in core-shell nanoparticle design

    NASA Astrophysics Data System (ADS)

    Schärtl, Wolfgang

    2010-06-01

    Ten years ago I wrote a review about the important field of core-shell nanoparticles, focussing mainly on our own work about tracer systems, and briefly addressing polymer-coated nanoparticles as fillers for homogeneous polymer-colloid composites. Since then, the potential use of core-shell nanoparticles as multifunctional sensors or potential smart drug-delivery vehicles in biology and medicine has gained more and more importance, affording special types of multi-functionalized and bio-compatible nanoparticles. In this new review article, I try to address the most important developments during the last ten years. This overview is mainly based on frequently cited and more specialized recent review articles from leaders in their respective field. We will consider a variety of nanoscopic core-shell architectures from highly fluorescent nanoparticles (NPs), protected magnetic NPs, multifunctional NPs, thermoresponsive NPs and biocompatible systems to, finally, smart drug-delivery systems.Ten years ago I wrote a review about the important field of core-shell nanoparticles, focussing mainly on our own work about tracer systems, and briefly addressing polymer-coated nanoparticles as fillers for homogeneous polymer-colloid composites. Since then, the potential use of core-shell nanoparticles as multifunctional sensors or potential smart drug-delivery vehicles in biology and medicine has gained more and more importance, affording special types of multi-functionalized and bio-compatible nanoparticles. In this new review article, I try to address the most important developments during the last ten years. This overview is mainly based on frequently cited and more specialized recent review articles from leaders in their respective field. We will consider a variety of nanoscopic core-shell architectures from highly fluorescent nanoparticles (NPs), protected magnetic NPs, multifunctional NPs, thermoresponsive NPs and biocompatible systems to, finally, smart drug-delivery systems. Dedicated to Professor Manfred Schmidt on the occasion of his 60th birthday

  12. 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

  13. 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, the two cores will likely merge to form a Thorne-Zytkow object. Implications for the origin of Cyg X-3, an X-ray source consisting of a Wolf-Rayet star and a compact companion, and for the fate of the remnant binary consisting of a helium star and a neutron star are briefly discussed.

  14. 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.

  15. A liquid-metal fast breeder reactor core design with seed blanket modular assemblies

    SciTech Connect

    Sehgal, B.R.; Fuller, E.L.; Lin, C.

    1982-05-01

    A liquid-metal fast breeder reactor core design employing seed-blanket-type modular assemblies is presented. The seed region contains PuO/sub 2/-UO/sub 2/ fuel and the blanket region contains depleted UO/sub 2/ fuel. These assemblies constitute the inner core while conventional mixed-oxide assemblies are employed in the outer core region. The results of design studies and analysis show that a large reduction in the sodium void reactivity can be obtained. The core fissile loading is larger than that for a homogeneous core. The analysis of an unprotected loss-of-flow (LOF) accident shows drastically reduced potential for a LOF-driven transient overpower accident. These results and conclusions are similar to those obtained for other heterogeneous core designs.

  16. 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.

  17. 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…

  18. 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

  19. 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…

  20. 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.

  1. 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...

  2. 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...

  3. 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/outer heat flux more effectively. The final minimized 10B loading for LEU case studies will have 0.635 g in the LEU fuel meat at the inner 2 fuel plates (1-2) and outer 2 fuel plates (18-19), which can achieve peak to average ratios similar to those for the ATR reference HEU case study. The investigation of this paper shows the optimized LEU Monolithic (U-10Mo) and Dispersed (U7Mo) cases can all meet the LEU conversion objectives.

  4. Finite element solution of the multi-group neutron diffusion equation applied to the hypothetical core disruptive accident

    SciTech Connect

    Baldasaro, P.F.

    1980-01-01

    A two-dimensional (r,z), space-time-energy dependent calculational method has been developed for use in LMFBR hypothetical core disruptive accident (HCDA) analyses which utilizes the Finite Element Method (subsequently referred to as FEM) for the necessary diffusion theory neutronics calculations.The hydrodynamic portion of the calculation is performed with a modified form of the VENUS-I computer program which does a thermodynamic energy balance and uses an energy-density-dependent equation-of-state to calculate the pressures which cause reactor disassembly. The reactivity feedback which results from material motion is based on a Hermite Finite Element neutronics method in (r,z) geometry. The FEM neutronic calculations are done using an Eulerian coordinate system while the VENUS calculations employ a Lagrangian coordinate system. A unique feature of the FEM calculations is that the use of Hermite finite elements allows the analytic construction of iteration matrices. The numerical solution of the resulting system of equations utilizes direct matrix inversion which helps to minimize computing time. The major disadvantage of this method, relative to the standard finite differencing method (FDM), is the additional computational logic which is required. This application of the FEM demonstrates the potential which the method has for providing fast, accurate, and reliable solutions to the multi-group neutron diffusion problem where the effects of local heterogeneities are minimal.

  5. 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.

  6. Design of an accelerator-based neutron source for neutron capture therapy.

    PubMed

    Terlizzi, R; Colonna, N; Colangelo, P; Maiorana, A; Marrone, S; Rain, A; Tagliente, G; Variale, V

    2009-07-01

    The boron neutron capture therapy is mainly suited in the treatment of some tumor kinds which revealed ineffective to the traditional radiotherapy. In order to take advantage of such a therapeutic modality in hospital environments, neutron beams of suitable energy and flux levels provided by compact size facilities are needed. The advantages and drawbacks of several neutron beams are here analysed in terms of therapeutic gains. In detail the GEANT-3/MICAP simulations show that high tumor control probability, with sub-lethal dose at healthy tissues, can be achieved by using neutron beams of few keV energy having a flux of about 10(9) neutrons/(cm(2)s). To produce such a neutron beam, the feasibility of a proton accelerator is investigated. In particular an appropriate choice of the radiofrequency parameters (modulation, efficiency of acceleration, phase shift, etc.) allows the development of relatively compact accelerators, having a proton beam current of 30 mA and an energy of 2 MeV, which could eventually lead to setting up of hospital-based neutron facilities. PMID:19406649

  7. Fast-neutron radiation effects in a silica-core optical fiber studied by a CCD-camera spectrometer

    SciTech Connect

    Griscom, D.L.; Gingerich, M.E.; Friebele, E.J. ); Putnam, M. ); Unruh, W. )

    1994-02-20

    A simple CCD-camera spectrometer was deployed at the Los Alamos Spallation Radiation Effects Facility to characterize fast-neutron irradiation effects in several silica-based optical fibers over the wavelength range [similar to]450--1100 nm. The experimental arrangement allowed optical loss spectra to be developed from remotely recovered frame grabs at various times during irradiation without it being necessary to resort to cutback methods. Data recorded for a pure-silica-core/F-doped-silica-clad fiber displayed a peculiar artifact, which is described and mathematically modeled in terms of leaky modes propagating in an optical cladding that is substantially less susceptible to radiation-induced optical attenuation than is the core. Evidence from optical time-domain reflectometry supports the postulate that mode leakage into the cladding may be a result of light scattering from the tracks of ions displaced by the 14-MeV neutrons. These results suggest that fibers with fluorine doping in the core, as well as in the cladding, would be relatively resistant to radiation-induced attenuation in the UV--visible spectral region.

  8. Considerations in the design of an improved transportable neutron spectrometer

    NASA Astrophysics Data System (ADS)

    Williams, A. M.; Spyrou, N. M.; Brushwood, J. M.; Beeley, P. A.

    2002-01-01

    The Transportable Neutron Spectrometer (TNS) has been used by the Ministry of Defence for over 15 years to characterise neutron fields in workplace environments and provide local correction factors for both area and personal dosimeters. In light of advances in neutron spectrometry, a programme to evaluate and improve TNS has been initiated. This paper describes TNS, presents its operation in known radioisotope fields and in a reactor environment. Deficiencies in the operation of the instrument are highlighted, together with proposals for updating the response functions and spectrum unfolding methodologies.

  9. A Workshop on Methods for Neutron Scattering Instrument Design. Introduction and Summary

    SciTech Connect

    Hjelm, Rex P.

    1996-12-31

    The future of neutron and x-ray scattering instrument development and international cooperation was the focus of the workshop on ``Methods for Neutron Scattering Instrument Design`` September 23-25 at the E.O. Lawrence Berkeley National Laboratory. These proceedings are a collection of a portion of the invited and contributed presentations.

  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 isotopes at the end than at the beginning of the irradiation) and the void coefficient would be negative. The addition of 1 cm of U-238 blanket at the top and bottom of the fuel would bring the fissile Pu mass balance from -7.5% to -6.5% but would also impact the void coefficient in the wrong way. In fact, it turns out that the void coefficient is so sensitive to the presence of axial blanket that it limits its size to only a few cm for driver fuel height of about 50-60 cm. For reference, the fissile Pu mass balance is about -35% in a standard PWR MOX fuel such as those used in France. In order to reduce the fissile Pu deficit and potentially reach a true breeding regime (i.e. a positive Pu mass balance), it would be necessary to make extensive use of radial blankets, both internal and external. Even though this was not addressed in detail here, it is reasonable to believe that at least as much U-238 blanket subassemblies as MOX driver fuel subassemblies would be necessary to breed enough Pu to compensate for the Pu deficit in the driver fuel. Hence, whereas a relatively simple D2O-cooled PWR core design makes it possible to obtain a near-breeder core, it may be necessary to more than double the mass of heavy metal in the core as well as the mass of heavy metal to reprocess per unit of energy produced in order to breed the few percents of Pu missing to reach a true breeding regime. It may be interesting to quantify these aspects further in the future.

  12. 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.

  13. 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)

  14. Use of albedo for neutron reflector regions in reactor core 3-D simulations

    NASA Astrophysics Data System (ADS)

    Mohanakrishnan, P.

    1989-10-01

    In this paper we present two new simplified schemes for the application of the albedo concept of replacing the reflector in 3-D reactor core simulations. Both involve the numerical derivation of albedoes from the fluxes at the core- (blanket-) reflector interface obtained from sample calculations including the reflector. Diffusion theory is used for core calculations in both cases. In the first scheme a new method for "diagonalising" the albedo matrix is demonstrated. This achieves easy applicability of the albedo parameters in core simulations of a fast breeder reactor core, resulting in significant savings in computing efforts. The second scheme, applied to light water reactors, achieves better accuracy in core periphery power predictions with the use of only uniform coarse meshes throughout the core and the numerically derived albedoes.

  15. 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. PMID:25445697

  16. System design considerations for fast neutron interrogation systems

    NASA Astrophysics Data System (ADS)

    Micklich, Bradley J.; Curry, B. P.; Fink, Charles L.; Smith, Donald L.; Yule, Thomas J.

    1994-03-01

    We are modelling a number of the fast-neutron interrogation techniques currently under consideration, to include fast neutron transmission spectroscopy, pulsed fast neutron analysis, and its variant, 14-MeV associated particle imaging. The goals of this effort are to determine the component requirements for each technique, identify trade-offs that system performance standards impose upon those component requirements, and assess the relative advantages and disadvantages of the different approaches. In determining the component requirements, we will consider how they are driven by system performance standards, such as image resolution, scanning time, and statistical uncertainty. In considering the trade-offs between system components, we concentrate primarily on those which are common to all approaches, for example: source characteristics versus detector array requirements. We will then use the analysis to propose some figures of merit that enable performance comparisons between the various fast-neutron systems under consideration. The status of this ongoing effort is presented.

  17. Preliminary detector design ST862-prototype neutron detector

    NASA Astrophysics Data System (ADS)

    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 CaF2Eu 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.

  18. Design for an accelerator-based orthogonal epithermal neutron beam for boron neutron capture therapy.

    PubMed

    Allen, D A; Beynon, T D; Green, S

    1999-01-01

    This paper is concerned with the proposed Birmingham accelerator-based epithermal neutron beam for boron neutron capture therapy (BNCT). In particular, the option of producing a therapy beam at an orthogonal direction to the incoming protons is considered. Monte Carlo radiation transport simulations, both with and without a head phantom, have shown that an orthogonal beam geometry is not only acceptable but is indeed beneficial, in terms of a lower mean neutron energy and an enhanced therapeutic ratio for the same useful neutron fluence in the therapy beam. Typical treatment times for various beam options have been calculated, and range from 20 to 48 min with a 5 mA beam of 2.8 MeV protons, if the maximum photon-equivalent dose delivered to healthy tissue is to be 12.6 Gy Eq. The effects of proton beam diameter upon the therapy beam parameters have also been considered. PMID:9949400

  19. 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.

  20. 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.

  1. 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 diffusion approximation, mostly. In MCNPX-2.7.0, a step-by-step approach was used, starting with a model in which the fissile rods were simulated on a homogeneous level, to finally integrate the actual heterogeneous description of the subassemblies. The code-to-code cell analysis performed between ECCO, SERPENT and MCNPX with different modern nuclear data library revealed that the results for the infinite multiplication factor between Monte Carlo and deterministic analysis are in good agreement ({Delta}p < 100 pcm). The differences between the results were observed to be larger for the neutron data libraries, with reactivity differences up to 350 pcm. (authors)

  2. 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.

  3. 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 transmuted; this burnup is slightly superior to that attainable in helium-cooled reactors. A preliminary analysis of the modular variant for the PB-AHTR investigated the triple heterogeneity of this design and determined its performance characteristics.

  4. Overview of the Conceptual Design of the Future VENUS Neutron Imaging Beam Line at the Spallation Neutron Source

    NASA Astrophysics Data System (ADS)

    Bilheux, Hassina; Herwig, Ken; Keener, Scott; Davis, Larry

    VENUS (Versatile Neutron Imaging Beam line at the Spallation Neutron Source) will be a world-class neutron-imaging instrument that will uniquely utilize the Spallation Neutron Source (SNS) time-of-flight (TOF) capabilities to measure and characterize objects across several length scales (mm to μm). When completed, VENUS will provide academia, industry and government laboratories with the opportunity to advance scientific research in areas such as energy, materials, additive manufacturing, geosciences, transportation, engineering, plant physiology, biology, etc. It is anticipated that a good portion of the VENUS user community will have a strong engineering/industrial research focus. Installed at Beam line 10 (BL10), VENUS will be a 25-m neutron imaging facility with the capability to fully illuminate (i.e., umbra illumination) a 20 cm x 20 cm detector area. The design allows for a 28 cm x 28 cm field of view when using the penumbra to 80% of the full illumination flux. A sample position at 20 m will be implemented for magnification measurements. The optical components are comprised of a series of selected apertures, T0 and bandwidth choppers, beam scrapers, a fast shutter to limit sample activation, and flight tubes filled with Helium. Techniques such as energy selective, Bragg edge and epithermal imaging will be available at VENUS.

  5. System design considerations for fast-neutron interrogation systems

    SciTech Connect

    Micklich, B.J.; Curry, B.P.; Fink, C.L.; Smith, D.L.; Yule, T.J.

    1993-10-01

    Nonintrusive interrogation techniques that employ fast neutrons are of interest because of their sensitivity to light elements such as carbon, nitrogen, and oxygen. The primary requirement of a fast-neutron inspection system is to determine the value of atomic densities, or their ratios, over a volumetric grid superimposed on the object being interrogated. There are a wide variety of fast-neutron techniques that can provide this information. The differences between the various nuclear systems can be considered in light of the trade-offs relative to the performance requirements for each system`s components. Given a set of performance criteria, the operational requirements of the proposed nuclear systems may also differ. For instance, resolution standards will drive scanning times and tomographic requirements, both of which vary for the different approaches. We are modelling a number of the fast-neutron interrogation techniques currently under consideration, to include Fast Neutron Transmission Spectroscopy (FNTS), Pulsed Fast Neutron Analysis (PFNA), and its variant, 14-MeV Associated Particle Imaging (API). The goals of this effort are to determine the component requirements for each technique, identify trade-offs that system performance standards impose upon those component requirements, and assess the relative advantages and disadvantages of the different approaches. In determining the component requirements, we will consider how they are driven by system performance standards, such as image resolution, scanning time, and statistical uncertainty. In considering the trade-offs between system components, we concentrate primarily on those which are common to all approaches, for example: source characteristics versus detector array requirements. We will then use the analysis to propose some figures-of-merit that enable performance comparisons between the various fast-neutron systems under consideration. The status of this ongoing effort is presented.

  6. 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 burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).

  7. Baseline design of a low energy neutron source at ESS-Bilbao

    NASA Astrophysics Data System (ADS)

    Sordo, F.; Fernandez-Alonso, F.; Gonzalez, Miguel A.; Ghiglino, A.; Magán, M.; Terrón, S.; Martínez, F.; de Vicente, J. P.; Vivanco, R.; Bermejo, F. J.; Perlado, J. M.

    2014-10-01

    This article briefly describes the basic design of the ESS-Bilbao neutron target station as well as its expected neutronic performance. The baseline engineering design, associated ancillary systems, and plant layout for the facility is now complete. A rotating target composed of twenty beryllium plates has been selected as the best choice in terms of both neutron yield and engineering complexity. It will provide neutron beams with a source term of 1015 n s-1 resulting from the direct 9Be(p, xn) reaction using a 75 mA proton beam at 50 MeV. The design envisages a target station equipped with two fully optimized moderators capable of withstanding a proton-beam power of 112 kW. This design is flexible enough to accommodate future upgrades in final proton energy. The envisaged neutron-beam brightness will enable several applications, including the use of cold and thermal neutrons for condensed matter research as well as fast-neutron irradiation studies. We close by discussing the role that this facility may play once the European Spallation Source becomes operational in Lund, Sweden.

  8. Design of a compact high-power neutron source—The EURISOL converter target

    NASA Astrophysics Data System (ADS)

    Samec, K.; Milenković, R. Ž.; Dementjevs, S.; Ashrafi-Nik, M.; Kalt, A.

    2009-07-01

    The EURISOL project, a multi-lateral initiative supported by the EU, aims to develop a facility to achieve high yields of isotopes in radioactive beams and extend the variety of these isotopes towards more exotic types. The neutron source at the heart of the projected facility is designed to generate isotopes by fissioning uranium carbide (UC) targets arranged around a 4 MW neutron source. For reasons of efficiency, it is essential that the neutron source be as compact as possible, to avoid losing neutrons by absorption whilst maximising the escaping neutron flux, thus increasing the number of fissions in the UC targets. The resulting configuration presents a challenge in terms of absorbing heat deposition rates of up to 8 kW/cm3 in the neutron source; it has led to the selection of liquid metal for the target material. The current paper presents the design of a compact high-power liquid-metal neutron source comprising a specially optimised beam window concept. The design is based on two-dimensional (2D) and three-dimensional (3D) computational fluid dynamics (CFD) numerical simulations for thermal hydraulics and hydraulic aspects, as well as finite-element method (FEM) for assessing thermo-mechanical stability. The resulting optimised design was validated by a dedicated hydraulic test under realistic flow conditions. A full-scale mock-up was built at the Paul Scherrer Institute (PSI) and was tested at the Institute of Physics of the University of Latvia (IPUL).

  9. 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.

  10. A review of irradiation effects on LWR core internal materials - neutron embrittlement.

    SciTech Connect

    Chopra, O. K.; Rao, A. S.

    2011-05-01

    Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of light water reactor (LWR) pressure vessels because of their relatively high strength, ductility, and fracture toughness. However, exposure to neutron irradiation for extended periods not only changes the microstructure and microchemistry of these steels, but also degrades their fracture properties. The existing data on irradiated austenitic SSs are reviewed to determine the effects of key parameters such as material type and condition and irradiation temperature, dose, and dose rate on neutron embrittlement. Differences in the radiation-induced degradation of fracture properties between LWR and fast-reactor irradiations are also discussed. The results are used to (a) define a threshold fluence above which irradiation effects on fracture toughness of the material are significant, (b) evaluate the potential of neutron embrittlement under LWR operating conditions, and (c) assess the potential effects of voids on fracture toughness.

  11. A review of irradiation effects on LWR core internal materials - Neutron embrittlement

    NASA Astrophysics Data System (ADS)

    Chopra, O. K.; Rao, A. S.

    2011-05-01

    Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of light water reactor (LWR) pressure vessels because of their relatively high strength, ductility, and fracture toughness. However, exposure to neutron irradiation for extended periods not only changes the microstructure and microchemistry of these steels, but also degrades their fracture properties. The existing data on irradiated austenitic SSs are reviewed to determine the effects of key parameters such as material type and condition and irradiation temperature, dose, and dose rate on neutron embrittlement. Differences in the radiation-induced degradation of fracture properties between LWR and fast-reactor irradiations are also discussed. The results are used to (a) define a threshold fluence above which irradiation effects on fracture toughness of the material are significant, (b) evaluate the potential of neutron embrittlement under LWR operating conditions, and (c) assess the potential effects of voids on fracture toughness.

  12. 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…

  13. Legal Protection on IP Cores for System-on-Chip Designs

    NASA Astrophysics Data System (ADS)

    Kinoshita, Takahiko

    The current semiconductor industry has shifted from vertical integrated model to horizontal specialization model in term of integrated circuit manufacturing. In this circumstance, IP cores as solutions for System-on-Chip (SoC) have become increasingly important for semiconductor business. This paper examines to what extent IP cores of SoC effectively can be protected by current intellectual property system including integrated circuit layout design law, patent law, design law, copyright law and unfair competition prevention act.

  14. 3He neutron detector design for active detection of cargo containers

    NASA Astrophysics Data System (ADS)

    McDevitt, Daniel B.; Eberhard, J. W.; Zelakiewicz, Scott; Maschinot, Aaron

    2008-04-01

    We report on the design of a neutron detector using industry standard 3He tubes to count delayed neutrons during the interrogation of cargo containers for the presence of Special Nuclear Material (SNM). Simulations of the detector design were run for delayed neutron spectra for a variety of cargos containing SNM using the Monte Carlo computer code COG. The simulations identified parameters crucial to optimize the detector design. These choices include moderating material type and thickness, tube spacing, tube pressure and number of tubes. An experimental prototype was also constructed based on the simulated design specifications. This paper discusses the parameters that lead up to the optimized detector design. It also compares the performance of the Monte Carlo simulated design and the experimental detector when exposed to a 239Pu-Be source.

  15. 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.

  16. Systematic Analysis for Distribution of Extra Neutrons and Core in Halo Nucleus

    NASA Astrophysics Data System (ADS)

    Matsumoto, T.; Sasabe, S.; Minomo, K.; Furutachi, N.; Ogata, K.; Yahiro, M.

    We demonstrate the series of analyses for 6He scattering to investigate two-neutron properties of 6He. Breakup, reaction cross sections, and neutron removal cross section are analyzed by the continuum discretized coupled-channels method and the eikonal reaction theory. Elastic cross sections for 6He scattering from a proton target at intermediate energies are estimated by the single folding model with the nucleon-nucleon t-matrix. This series of analyses is thus quite useful in studies of heavier halo nuclei, if the three measurements are performed for the same nucleus.

  17. 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.

  18. 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.

  19. 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

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

    SciTech Connect

    Bortolussi, S.; Altieri, S.

    2007-12-15

    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}{sub max}/{phi}{sub 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}{sub max}/{phi}{sub 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.

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

    SciTech Connect

    Kramer, K

    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 {micro}m of tungsten to mitigate x-ray damage. The first wall is cooled by Li{sub 17}Pb{sub 83} eutectic, chosen for its neutron multiplication and good heat transfer properties. The {sub 17}Pb{sub 83} flows in a jacket around the first wall to an extraction plenum. The main coolant injection plenum is immediately behind the Li{sub 17}Pb{sub 83}, separated from the Li{sub 17}Pb{sub 83} by a solid ODS wall. This main system coolant is the molten salt flibe (2LiF-BeF{sub 2}), 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 the same radial flibe flow that travels through perforated ODS walls to the reflector blanket. This reflector blanket is 75 cm thick comprised of 2 cm diameter graphite pebbles cooled by flibe. The flibe extraction plenum surrounds the reflector bed. Detailed neutronics designs studies are performed to arrive at the described design. The LFFH engine thermal power is controlled using a technique of adjusting the {sup 6}Li/{sup 7}Li enrichment in the primary and secondary coolants. The enrichment adjusts system thermal power in the design by increasing tritium production while reducing fission. To perform the simulations and design of the LFFH engine, a new software program named LFFH Nuclear Control (LNC) was developed in C++ to extend the functionality of existing neutron transport and depletion software programs. Neutron transport calculations are performed with MCNP5. Depletion calculations are performed using Monteburns 2.0, which utilizes ORIGEN 2.0 and MCNP5 to perform a burnup calculation. LNC supports many design parameters and is capable of performing a full 3D system simulation from initial startup to full burnup. It is able to iteratively search for coolant {sup 6}Li enrichments and resulting material compositions that meet user defined performance criteria. LNC is utilized throughout this study for time dependent simulation of the LFFH engine. Two additional methods were developed to improve the computation efficiency of LNC calculations. These methods, termed adaptive time stepping and adaptive mesh refinement were incorporated into a separate stand alone C++ library name the Adaptive Burnup Library (ABL). The ABL allows for other client codes to call and utilize its functionality. Adaptive time stepping is useful for automatically maximizing the size of the depletion time step while maintaining a desired level of accuracy. Adaptive meshing allows for analysis of fixed fuel configurations that would normally require a computationally burdensome number of depletion zones. Alternatively, Adaptive Mesh Refinement (AMR) adjusts the depletion zone size according to the variation in flux across the zone or fractional contribution to total absorption or fission. A parametric analysis on a fully mixed fuel core was performed using the LNC and ABL code suites. The resulting system parameters are found to optimize performance metrics using a 20 MT DU fuel load with a 20% TRISO packing and a 300 {micro}m kernel radius operated with a fusion input power of 500 MW and a fission blanket gain of 4.0. LFFH potentially offers a proliferation resistant technology relative to other nuclear energy systems primarily because of no need for fuel enrichment or reprocessing. A figure of merit of the material attractiveness is examined and it is found that the fuel is effectively contaminated to an unattractive level shortly after the system is started due to fission product and minor actinide build up.

  2. Design and experimental tests of a novel neutron spin analyzer for wide angle spin echo spectrometers

    SciTech Connect

    Fouquet, Peter; Farago, Bela; Andersen, Ken H.; Bentley, Phillip M.; Pastrello, Gilles; Sutton, Iain; Thaveron, Eric; Thomas, Frederic; Moskvin, Evgeny; Pappas, Catherine

    2009-09-15

    This paper describes the design and experimental tests of a novel neutron spin analyzer optimized for wide angle spin echo spectrometers. The new design is based on nonremanent magnetic supermirrors, which are magnetized by vertical magnetic fields created by NdFeB high field permanent magnets. The solution presented here gives stable performance at moderate costs in contrast to designs invoking remanent supermirrors. In the experimental part of this paper we demonstrate that the new design performs well in terms of polarization, transmission, and that high quality neutron spin echo spectra can be measured.

  3. 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.

  4. 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.

  5. 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 with detailed and accurate thermal-hydraulic models. The development of such reference high-fidelity coupled multi-physics scheme is described in this dissertation on the basis of MCNP5, NEM, NJOY and COBRA-TF (CTF) computer codes. This work presents results from studies performed and implemented at the Pennsylvania State University (PSU) on both accelerating Monte Carlo criticality calculations by using hybrid nodal diffusion Monte Carlo schemes and thermal-hydraulic feedback modeling in Monte Carlo core calculations. The hybrid MCNP5/CTF/NEM/NJOY coupled code system is proposed and developed in this dissertation work. The hybrid coupled code system contains a special interface developed to update the required MCNP5 input changes to account for dimension and density changes provided by the thermal-hydraulics feedback module. The interface has also been developed to extract the flux and reaction rates calculated by MCNP5 to later transform the data into the power feedback needed by CTF (axial and radial peaking factors). The interface is contained in a master program that controls the flow of the calculations. Both feedback modules (thermal-hydraulic and power subroutines) use a common internal interface to further accelerate the data exchange. One of the most important steps to correctly include the thermal hydraulic feedback into MCNP5 calculations begins with temperature dependent cross section libraries. If the cross sections used for the calculations are not at the correct temperature, the temperature feedback cannot be included into MCNP5 (referred to the effect of temperature on cross sections: Doppler boarding of resolve and unresolved resonances, thermal scattering and elastic scattering). The only method of considering the temperature effects on cross sections is through the generation (or as introduced in this dissertation through a novel interpolation mechanism) of continuous energy temperature-dependent cross section libraries. An automated methodology for generation of continuous energy temperature-dependent cross section libraries has been developed as part of the hybrid Monte Carlo-based coupled core studies at PSU. This tool is used together with the automated cross-section temperature interpolation capability for intermediate points. The automated methodology, combined with the interpolation capability, has considerably reduced the cross section generation time. A new methodology for generation and interpolation of temperature-dependent thermal scattering cross section libraries for MCNP5 is introduced as well. Using the interpolation methodology specially designed for thermal scattering cross sections, a thermal scattering grid at the desired temperature was generated. This gives the possibility of performing MCNP5 criticality calculations at the correct moderator temperature and improving the accuracy of the calculations. A cross section update methodology has been included, which efficiently reduces the time of the cross section libraries update. Several acceleration strategies are introduced and implemented in the hybrid coupled code system. The computation process is greatly accelerated by calculating the 3-D distributions of fission source and thermal-hydraulics parameters with the coupled NEM/CTF code and then using coupled MCNP5/CTF code to fine tune the results to obtain an increased accuracy. The PSU NEM code employs cross-sections generated by MCNP5 for pin-cell based nodal compositions. Finally, the hybrid coupled system is automated and enhanced in order to provide the user with an efficient and easy to use high accuracy modeling tool.

  6. 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.

  7. Protein structures by spallation neutron crystallography

    PubMed Central

    Langan, Paul; Fisher, Zoë; Kovalevsky, Andrii; Mustyakimov, Marat; Sutcliffe Valone, Amanda; Unkefer, Cliff; Waltman, Mary Jo; Coates, Leighton; Adams, Paul D.; Afonine, Pavel V.; Bennett, Brad; Dealwis, Chris; Schoenborn, Benno P.

    2008-01-01

    The Protein Crystallography Station at Los Alamos Neutron Science Center is a high-performance beamline that forms the core of a capability for neutron macromolecular structure and function determination. This capability also includes the Macromolecular Neutron Crystallography (MNC) consortium between Los Alamos (LANL) and Lawrence Berkeley National Laboratories for developing computational tools for neutron protein crystallography, a biological deuteration laboratory, the National Stable Isotope Production Facility, and an MNC drug design consortium between LANL and Case Western Reserve University. PMID:18421142

  8. Analysis of a shield design for a DT neutron generator test facility.

    PubMed

    Chichester, D L; Pierce, G D

    2007-10-01

    Independent numerical simulations have been performed using the MCNP5 and SCALE5 radiation transport codes to evaluate the effectiveness of a concrete facility designed to shield personnel from neutron radiation emitted from DT neutron generators. The analysis considered radiation source terms of 14.1 MeV monoenergetic neutrons located at three discrete locations within the two test vaults in the facility, calculating neutron and photon dose rates at 44 locations around the facility using both codes. In addition, dose rate contours were established throughout the facility using the MCNP5 mesh tally feature. Neutron dose rates calculated outside of the facility are predicted to be below 0.01 mrem/h at all locations when all neutron generator source terms are operating within the facility. Similarly, the neutron dose rate in one empty test vault when the adjacent test vault is being utilized is also less then 0.01 mrem/h. For most calculation locations outside the facility the photon dose rates were less then the neutron dose rates by a factor of 10 or more. PMID:17659876

  9. ETHERNES: A new design of radionuclide source-based thermal neutron facility with large homogeneity area.

    PubMed

    Bedogni, R; Sacco, D; Gómez-Ros, J M; Lorenzoli, M; Gentile, A; Buonomo, B; Pola, A; Introini, M V; Bortot, D; Domingo, C

    2016-01-01

    A new thermal neutron irradiation facility based on an (241)Am-Be source embedded in a polyethylene moderator has been designed, and is called ETHERNES (Extended THERmal NEutron Source). The facility shows a large irradiation cavity (45 cm × 45 cm square section, 63 cm in height), which is separated from the source by means of a polyethylene sphere acting as shadowing object. Taking advantage of multiple scattering of neutrons with the walls of this cavity, the moderation process is especially effective and allows obtaining useful thermal fluence rates from 550 to 800 cm(-2) s(-1) with a source having nominal emission rate 5.7×10(6) s(-1). Irradiation planes parallel to the cavity bottom have been identified. The fluence rate across a given plane is as uniform as 3% (or better) in a disk with 30 cm (or higher) diameter. In practice, the value of thermal fluence rate simply depends on the height from the cavity bottom. The thermal neutron spectral fraction ranges from 77% up to 89%, depending on the irradiation plane. The angular distribution of thermal neutrons is roughly isotropic, with a slight prevalence of directions from bottom to top of the cavity. The mentioned characteristics are expected to be attractive for the scientific community involved in neutron metrology, neutron dosimetry and neutron detector testing. PMID:26516990

  10. A design study for an accelerator-based epithermal neutron beam for BNCT.

    PubMed

    Allen, D A; Beynon, T D

    1995-05-01

    An achievable design concept for a boron neutron capture therapy (BNCT) facility, based on a high-current, low-energy proton accelerator, is described. Neutrons are produced within a thick natural lithium target, under bombardment from protons with an initial energy between 2.5 and 3.0 MeV. The proton current will be up to 10 mA. After gamma-ray filtering, the neutrons are partially moderated to epithermal energies within a heavy-water moderator, poisoned with 6Li to remove thermal neutrons. Monte Carlo modelling has been used to predict system performance in terms of neutron fluence rate and neutron and gamma-ray dose at the patient position. The relationship between the system performance and key parameters, such as proton energy, moderator depth and 6Li concentration, has been investigated. With a proton current of 10 mA, the facility is capable of providing a therapy beam with a useful neutron fluence rate of 10(9) cm-2 s-1 and a neutron dose per unit fluence of less than 6 x 10(-13) Gy cm2, with a gamma-ray contamination of the therapy beam of about 10(-13) Gy cm2. PMID:7652009

  11. 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.

  12. 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.

  13. Optimal Design of Measurement Network for Neutronic Activity Field Reconstruction by Data Assimilation

    NASA Astrophysics Data System (ADS)

    Bouriquet, Bertrand; Argaud, Jean-Philippe; Thual, Olivier

    Several studies show that data assimilation methods, by merging information both from model and measurement can be used to elaborate an optimal determination of neutronic activity field within a nuclear PWR reactor core. Here the problem addressed and solved is to determine an optimal repartition of the instruments, used to make measurements within the core, to get the best possible reconstructed field using a data assimilation procedure. The position optimization is based on simulated annealing, with a Metropolis-Hasting algorithm. To perform the method in the framework of data assimilation, algebraic optimisation related to computing time of data assimilation has been developed and is used here.

  14. Optimal design of measurement network for neutronic activity field reconstruction by data assimilation

    NASA Astrophysics Data System (ADS)

    Bouriquet, Bertrand; Argaud, Jean-Philippe; Cugnart, Romain

    2012-02-01

    An optimal reconstruction of the neutronic activity field can be determined for a nuclear reactor core using a data assimilation framework to merge information from models and measurements. In this paper, we focus on solving the inverse problem of determining an optimal repartition of the measuring instruments within the core, to get the best possible results from the data assimilation reconstruction procedure. The position optimisation is realised using a Simulated Annealing algorithm, based on the Metropolis-Hastings proposition. Moreover, algebraic improvements of data assimilation have been developed, optimising their computational performance, and are presented here.

  15. Design and construction of pulsed neutron diagnostic system for plasma focus device (SBUPF1)

    SciTech Connect

    Moghadam, Sahar Rajabi; Davani, Fereydoon Abbasi

    2010-07-15

    In this paper, two designs of pulsed neutron counter structure are introduced. To increase the activation counter efficiency, BC-400 plastic scintillator plates along with silver foils are utilized. Rectangular cubic and cylindrical geometries for activation counter cell are modeled using MCNP4C code. Eventually, an optimum length of 14 cm is calculated for the detector cell and optimum numbers of 20 silver foils for rectangular cubic geometry and ten foils for cylindrical geometry have been acquired. Due to the high cost of cutting, polishing of plastics, and etc., the rectangular cubic design is found to be more economical than the other design. In order to examine the functionality and ensure the detector output and corresponding designing, neutron yield of a 2.48 kJ plasma focus device (SBUPF1) in 8 mbar pressure with removal source method for calibration was measured (3.71{+-}0.32)x10{sup 7} neutrons per shot.

  16. LMFBR core design for low capital cost and low-sodium void

    SciTech Connect

    Fischer, G.J.

    1982-01-01

    The need to design LMFBR reactor cores as well as plants for lowest possible capital costs has been apparent internationally as well as in the US. At the same time it is also important to keep the sodium void reactivity gain as low as possible for safety reasons and it has always been important to assure a plant design which most effectively serves the operational needs of the utility. This paper describes a LMFBR core design which has evolved as a result of a recent effort to achieve these objectives.

  17. 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}).

  18. 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.

  19. 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.

  20. Design and fabrication of a novel core-suspended optic fiber for distributed gas sensor

    NASA Astrophysics Data System (ADS)

    Zhang, Tao; Ma, Lijia; Bai, Hongbo; Tong, Chengguo; Dai, Qiang; Kang, Chong; Yuan, Libo

    2014-06-01

    We designed a novel core-suspended capillary fiber that the core was suspended in the air hole and close to the inner surface of the capillary, and experimentally demonstrated its fabrication technology. In addition, a method for linking a single mode fiber and a core-suspended fiber was proposed based on splicing and tapering at the fusion point between the two fibers. By combining with the optical time domain reflectometer technology, we constructed a distributed gas sensor system to monitor greenhouse gas based on this novel fiber.

  1. 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

  2. Dynamical analysis of innovative core designs facing unprotected transients with the MAT5 DYN code

    SciTech Connect

    Darmet, G.; Massara, S.

    2012-07-01

    Since 2007, advanced Sodium-cooled Fast Reactors (SFR) are investigated by CEA, AREVA and EDF in the framework of a joint French collaboration. A prototype called ASTRID, sets out to demonstrate progress made in SFR technology, is due to operate in the years 2020's. The modeling of unprotected transients by computer codes is one of the key safety issues in the design approach to such SFR systems. For that purpose, the activity on CATHARE, which is the reference code for the transient analysis of ASTRID, has been strengthened during last years by CEA. In the meantime, EDF has developed a simplified and multi-channel code, named MAT5 DYN, to analyze and validate innovative core designs facing protected and unprotected transients. First, the paper consists in a description of MAT5 DYN: a code based on the existing code MAT4 DYN including major improvements on geometry description and physical modeling. Second, two core designs based on the CFV core design developed at CEA are presented. Then, the dynamic response of those heterogeneous cores is analyzed during unprotected loss of flow (ULOF) transient and unprotected transient of power (UTOP). The results highlight the importance of the low void core effect specific to the CFV design. Such an effect, when combined with a sufficient primary pump halving time and an optimized cooling group scheme, allows to delay (or, possibly, avoid) the sodium boiling onset during ULOF accidents. (authors)

  3. 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. PMID:22410232

  4. Design for a focusing high-resolution neutron crystal diffractometer

    NASA Astrophysics Data System (ADS)

    Ionita, I.; Stoica, A. D.; Popovici, M.; Popa, N. C.

    1999-07-01

    A new concept of high-resolution focusing configuration begins to be accepted as an alternative solution to the existing conventional configurations. Among the earliest work performed in this direction is that performed at the Institute for Nuclear Research, Pitesti. These results are presented below. The experimentally determined resolution properties for two focusing configurations obtained at TRIGA reactor Pitesti and at VVRS reactor Bucharest are given in order to be compared with those obtained for the conventional ones. The principles to get focusing in crystal neutron diffractometry are presented. The main characteristics for a focusing instrument are given.

  5. Measurements of actinide-fission product yields in Caliban and Prospero metallic core reactor fission neutron fields

    SciTech Connect

    Casoli, P.; Authier, N.; Laurec, J.; Bauge, E.; Granier, T.

    2011-07-01

    In the 1970's and early 1980's, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on {sup 233}U, {sup 235}U, and {sup 239}Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energy causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments. (authors)

  6. The advanced neutron source--designing to meet the needs of the user community

    SciTech Connect

    Peretz, F.J. )

    1989-01-01

    The Advanced Neutron Source (ANS) is to be a multi-purpose neutron research center, constructed around a high-flux reactor now being designed at the Oak Ridge National Laboratory (ORNL). Its primary purpose is to place the United States in the forefront of neutron scattering in the twenty-first century. Other research programs include nuclear and fundamental physics, isotopes production, materials irradiation, and analytical chemistry. The Advanced Neutron Source will be a unique and invaluable research tool because of the unprecedented neutron flux available from the high intensity research reactor. But that reactor would be ineffective without world-class research facilities that allow the fullest utilization of the available neutrons. And, in turn, those research facilities will not produce new and exciting science without a broad population of users coming from all parts of the nation, and the world, placed in a simulating environment in which experiments can be effectively conducted, and in which scientific exchange is encouraged. This paper discusses the measures being taken to ensure that the design of the ANS focuses not only on the reactor, but on providing the experiment and user support facilities needed to allow its effective use. 5 refs., 4 figs.

  7. 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.

  8. 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.

  9. 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.

  10. Design and realization of the IP control core in field controllers for LAMOST spectroscopes

    NASA Astrophysics Data System (ADS)

    Wang, Jianing; Han, Zhongyi; Zeng, Yizhong; Dai, Songxin; Hu, Zhongwen; Zhu, Yongtian; Wang, Lei; Hou, Yonghui

    2010-07-01

    The China-made telescope, LAMOST, consists of 16 spectroscopes to detect stellar spectra via 4000 optical fibers. In each spectroscope, many movable parts work in phase. Those parts are real-time controlled and managed by field controllers based on FPGA. This paper mainly introduces how to use DSP Builder module library in MATLAB / Simulink to construct the IP control core on FPGA chip. This method can also be used to design the control core of PID arithmetic, to carry out arithmetic simulation and generate VHDL language file, as well as to integrate it into SOPC developing environment so as to repeatedly use. In this way, the design period of the control system may be shortened and design process simplified. Finally due to the reversibility and programmability of the IP control core ,a system on a chip for field controllers of spectroscope is realized, which meets astronomical control requirements, providing an effective scheme for embedded system in astronomical instrument applications.

  11. 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 prediction tools for integrated core assemblies as well as and strategies for noise reduction and control is needed in order to meet the NASA N+3 noise goals. The NASA Fundamental Aeronautics Program has the principal objective of overcoming today's national challenges in air transportation. The SFW Reduced-Perceived-Noise Technical Challenge aims to develop concepts and technologies to dramatically reduce the perceived aircraft noise outside of airport boundaries. This reduction of aircraft noise is critical to enabling the anticipated large increase in future air traffic.

  12. 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)

  13. Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

    SciTech Connect

    Clarno, Kevin T; Palmtag, Scott; Davidson, Gregory G; Salko, Robert K; Evans, Thomas M; Turner, John A; Belcourt, Kenneth; Hooper, Russell; Schmidt, Rodney

    2014-01-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly cases are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.

  14. 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)

  15. Design and development of position sensitive detectors for neutron scattering instruments at National Facility for Neutron Beam Research in India

    NASA Astrophysics Data System (ADS)

    Desai, Shraddha S.

    2014-07-01

    Various neutron scattering instruments at Dhruva reactor, BARC, are equipped with indigenously developed neutron detectors. Range of detectors includes proportional counters, beam monitors and linear position sensitive detectors (PSD). One of the instruments is recently upgraded with multi-PSD system of high efficiency and high resolution PSDs arranged in stacking geometry. These efforts have resulted in improving the throughput of the instrument and reducing experiment time. Global scarcity of 3He has made essential to explore other options like BF3 gas and 10B coatings. PSDs with coaxial geometry using BF3 gas and 10B coating (90% enriched) are fabricated and characterized successfully. These PSDs are used as the alternative to 3He PSD in equivalent geometry. Though efficiency of PSDs in similar dimensions is lower than that with 3He, these large numbers of PSDs can be arranged in multi-PSD system. The PSD design is optimized for reasonable efficiency. An array of 60 BF3 filled PSDs (1 m long) is under development for the Time of Flight Instrument at Dhruva. Further improvement in efficiency can be obtained with novel designs with complex anode-cathode geometry. Various challenges arise for long term operation of PSDs with BF3 gas, in addition to complexity of data acquisition electronics. Study of gas aging with detector fabrication materials has been carried out. PSDs with 10B coating show advantage of non toxic nature but have low efficiency. Multiple 10B layers intercepting neutron beam are used to increase the efficiency. PSD designed with small anode- cathode spacing and array of multiwire grids placed between double sided 10B coated plates are being fabricated. Assembly is arranged in curvilinear geometry with zero parallax. Overview of these developments is presented.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Core design and safety studies for a small modular fast reactor

    SciTech Connect

    Yang, W. S.; Cahalan, J. E.; Dunn, F. E.

    2006-07-01

    The paper describes the core design and performance characteristics and the safety analysis results for a 50 MWe small modular fast reactor design that was developed jointly by ANL, CEA, and JNC as an international collaborative effort. The main goal in the core design was to achieve a 30-year lifetime with no refueling. In order to minimize the burnup reactivity swing, metal fuel with a high heavy metal volume fraction was selected. To enhance the proliferation resistance and actinide transmutation, all the transuranic (TRU) elements recovered from light water reactor spent fuel were used in a ternary alloy form of U-TRU-10Zr. A 125 MWt core design was developed, for which the burnup reactivity swing was only 1.6$ over the 30-year core lifetime. The average discharge burnup was 87 MWd/kg, and the maximum sodium void worth was 4.65$. The evaluated reactivity coefficients provided sufficient negative feedbacks. Shutdown margins of control systems were confirmed. Steady-state thermal-hydraulic analysis results showed that peak 2{sigma} cladding inner-wall and fuel centerline temperatures were less than design limits with sufficient margins. Detailed transient analyses for the total loss of power to reactor and intermediate coolant pumps showed that no fuel damage or cladding failure would occur, even when multiple safety systems were assumed to malfunction. (authors)

  1. 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.

  2. Sensitivity Analysis of Neutron Cross-Sections Considered for Design and Safety Studies of Lfr and SFR Generation IV Systems

    NASA Astrophysics Data System (ADS)

    Tucek, Kamil; Carlsson, Johan; Wider, Hartmut

    2006-04-01

    We evaluated the sensitivity of several design and safety parameters with regard to five different nuclear data libraries, JEF2.2, JEFF3.0, ENDF/B-VI.8, JENDL3.2, and JENDL3.3. More specifically, the effective multiplication factor, burn-up reactivity swing and decay heat generation in available LFR and SFR designs were estimated. Monte Carlo codes MCNP and MCB were used in the analyses of the neutronic and burn-up performance of the systems. Thermo-hydraulic safety calculations were performed by the STAR-CD CFD code. For the LFR, ENDF/B-VI.8 and JEF2.2 showed to give a harder neutron spectrum than JEFF3.0, JENDL3.2, and JENDL3.3 data due to the lower inelastic scattering cross-section of lead in these libraries. Hence, the neutron economy of the system becomes more favourable and keff is higher when calculated with ENDF/B-VI.8 and JEF2.2 data. As for actinide cross-section data, the uncertainties in the keff values appeared to be mainly due to 239Pu, 240Pu and 241Am. Differences in the estimated burn-up reactivity swings proved to be significant, for an SFR as large as a factor of three (when comparing ENDF/B-VI.8 results to those of JENDL3.2). Uncertainties in the evaluation of short-term decay heat generation showed to be of the order of several per cent. Significant differences were, understandably, observed between decay heat generation data quoted in literature for LWR-UOX and those calculated for an LFR (U,TRU)O2 spent fuel. A corresponding difference in calculated core parameters (outlet coolant temperature) during protected total Loss-of-Power was evaluated.

  3. Designing a recombinant Bacmid construct of HCV core+1 in Baculovirus expression system

    PubMed Central

    Safarnezhad Tameshkel, Fahimeh; Rahimi, Pooneh; Khataminejad, Mohammad Reza

    2015-01-01

    Background and Objectives: Hepatitis C virus (HCV) chronically infects around 200 million people worldwide and frequently causes liver cirrhosis and hepatocellular carcinoma. Rapid detection of this virus results in decreasing the distance between infection and initiation the anti-viral treatment, and may prevent most of the undesirable consequences. The new detected HCV protein “Core+1” made from the ribosomal frame shift in Core region is an important candidate for diagnostic tools. This study was conducted to design a recombinant Bacmid plasmid expressing the HCV 1a Core+1 sequence in the Baculovirus expression system for further diagnostic applications. Materials and Methods: The HCV Core +1 gene was amplified by PCR using the pcDNA-HAF recombinant vector that contained the Core+1 sequence from HCV genotype 1a as a template, and the specific primers with 2 restriction sites for Nco I and Xba I restriction enzymes. The PCR product was cloned in XbaI/NcoI restriction sites of the linearized pFastBac-HTB vector and evaluated by using those restriction enzymes and sequencing. Then the recombinant pFastBac-HTB vector was transformed in DH10Bac and the result was screened and confirmed by X-Gal discrimination and PCR. Results: The HCV 1a Core+1 was successfully amplified and the PCR product was confirmed by using the related restriction enzymes and sequencing. Cloning of pFastBac vector with the purified PCR product of HCV Core+1 was confirmed. Finally, the recombinant Bacmid was successfully transformed in DH10Bac. Conclusion: The recombinant Bac-Core+1 expression vector is considered as an important tool to transfect the sf9 cell line and expression the Core+1 protein. PMID:26697162

  4. Design studies for a high-resolution, transportable neutron radiography/radioscopy system

    SciTech Connect

    Gillespie, G.H.; Micklich, B.J.; McMichael, G.E.

    1996-09-30

    A preliminary design has been developed for a high-resolution, transportable neutron radiology system (TNRS) concept. The primary system requirement is taken to be a thermal neutron flux of 10[sup 6] n/(cm[sup 2]-sec) with a L/D ratio of 100. The approach is to use an accelerator-driven neutron source, with a radiofrequency quadrupole (RFQ) as the primary accelerator component. Initial concepts for all of the major components of the system have been developed,and selected key parts have been examined further. An overview of the system design is presented, together with brief summaries of the concepts for the ion source, low energy beam transport (LEBT), RFQ, high energy beam transport (HEBT), target, moderator, collimator, image collection, power, cooling, vacuum, structure, robotics, control system, data analysis, transport vehicle, and site support. The use of trade studies for optimizing the TNRS concept are also described.

  5. The design of a proton recoil telescope for 14MeV neutron spectrometry

    NASA Astrophysics Data System (ADS)

    Hawkes, N. P.; Bond, D. S.; Croft, S.; Jarvis, O. N.; Sherwood, A. C.

    2002-01-01

    As part of the design effort for a 14MeV neutron spectrometer for the Joint European Torus (JET), computer codes were developed to calculate the response of a proton recoil telescope comprising a proton radiator film mounted in front of a proton detector. The codes were used to optimise the geometrical configuration in terms of efficiency and resolution, bearing in mind the constraints imposed by the proposed application as a JET neutron diagnostic for the Deuterium-Tritium phase. A prototype instrument was built according to the optimised design, and tested with monoenergetic 14MeV neutrons from the Harwell 500keV Van de Graaff accelerator. The measured energy resolution and absolute efficiency were found to be in acceptable agreement with the calculations. Based on this work, a multi-radiator production version of the spectrometer has now been constructed and successfully deployed at JET.

  6. Design status of an intense 14 MeV neutron source for cancer therapy

    NASA Astrophysics Data System (ADS)

    Yao, Ze-En; Su, Tong-Ling; Cheng, Shang-Wen; Jia, Wen-Bao

    2002-09-01

    Design and development of an intense 14 MeV neutron source for cancer therapy is in progress at the Institute of Nuclear Research of Lanzhou University. The neutrons from the T(d,n) 4He reaction are produced by bombarding a rotating titanium tritide target with a 40 mA deuteron beam at 600 keV. The designed neutron yield is 8×10 12 n/s and the maximum dose rate at a 100 cm source-to-skin distance is 25 cGy/min. The HV terminal, accelerating column and HV power supply are enclosed inside a stainless steel pressure vessel containing 6 atm SF 6 gas to provide the electrical insulation.

  7. 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 discharged hydride fuel is more proliferation resistant. Preliminary feasibility assessment indicates that by replacing some of the ZrH1.6 by ThH2 it will be possible to further improve the plutonium incineration capability of PWR’s. Other possibly promising applications of hydride fuel were identified but not evaluated in this work. A number of promising oxide fueled PWR core designs were also found as spin-offs of this study: (1) The optimal oxide fueled PWR core design features smaller fuel rod diameter of D=6.5 mm and a larger pitch-to-diameter ratio of P/D=1.39 than presently practiced by industry – 9.5mm and 1.326. This optimal design can provide a 30% increase in the power density and a 24% reduction in the cost of electricity (COE) provided the PWR could be designed to have the coolant pressure drop across the core increased from the reference 29 psia to 60 psia. (2) Using wire wrapped oxide fuel rods in hexagonal fuel assemblies it is possible to design PWR cores to operate at 54% higher power density than the reference PWR design that uses grid spacers and a square lattice, provided 60 psia coolant pressure drop across the core could be accommodated. Uprating existing PWR’s to use such cores could result in 40% reduction in the COE. The optimal lattice geometry is D = 8.08 mm and P/D = 1.41. The most notable advantages of wire wraps over grid spacers are their significant lower pressure drop, higher critical heat flux and improved vibrations characteristics.

  8. 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. PMID:15353726

  9. 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. PMID:24478306

  10. Design of a new IRSN thermal neutron field facility using Monte-Carlo simulations.

    PubMed

    Lacoste, V

    2007-01-01

    The Institute for Radiological Protection and Nuclear Safety owns a graphite-moderated AmBe neutron field facility, SIGMA, that has to be reconstructed. Monte-Carlo simulations were performed to study the design of a new thermal facility based on IRSN existing facilities. Studies related to an update version of SIGMA concerned the enhancement of the thermal neutrons contribution to the dose equivalent. Calculations were mainly performed for a (252)Cf neutron source distribution located at the centre of a graphite moderator block. A quasi-pure thermal neutron field was obtained with a 2.4 x 2.4 x 2.4-m(3) block of graphite. A second acceptable neutron field was obtained with 3.3-MeV mono-energetic neutrons created by a 400-kV accelerator coupled to a graphite assembly of 1.5 x 1.5 x 1.5 m(3). The characteristics of the studied thermal fields with the requirement for a reference calibration field are compared, and the advantages and drawbacks of the different producing methods are discussed. PMID:17578877

  11. Design of a GEM-based detector for the measurement of fast neutrons

    NASA Astrophysics Data System (ADS)

    Esposito, B.; Murtas, F.; Villari, R.; Angelone, M.; Marocco, D.; Pillon, M.; Puddu, S.

    2010-05-01

    A novel neutron detector has been developed and tested in collaboration between LNF-INFN and ENEA-Frascati. The aim is to obtain a versatile system that can be employed for the simultaneous measurement of the neutron flux in various energy bands from 1 to 20 MeV. The main drive for this development is the need of neutron detectors with low sensitivity to γ-rays and high count rate capability for operation in the neutron flux environment ~3×108 n/cm2 s expected in future controlled thermonuclear fusion reactors. In these devices the fusion power is assessed through the measurement of the 2.5 and 14 MeV neutrons emitted by the plasma. A multilayer detector architecture, including a proton recoil converter, a proton absorber and a triple Gas Electron Multiplier (GEM), has been adopted. The detector read-out system consists of 128 pads (12.3×6 mm2) in a 8×16 matrix. The work on the detector design and optimization carried out with the MCNPX code and the experimental tests at the Frascati Neutron Generator (FNG) on a detector prototype for 2.5 and 14 MeV measurements are presented.

  12. BetaCore, a designed water soluble four-stranded antiparallel β-sheet protein

    PubMed Central

    Carulla, Natàlia; Woodward, Clare; Barany, George

    2002-01-01

    BetaCore is a designed ∼50-residue protein in which two BPTI-derived core modules, CM I and CM II, are connected by a 22-atom cross-link. At low temperature and pH 3, homo- and heteronuclear NMR data report a dominant folded (`f') conformation with well-dispersed chemical shifts, i, i+1 periodicity, numerous long-range NOEs, and slowed amide hydrogen isotope exchange patterns that is a four-stranded antiparallel β-sheet with nonsymmetrical and specific association of CM I and CM II. BetaCore `f' conformations undergo reversible, global, moderately cooperative, non-two-state thermal transitions to an equilibrium ensemble of unfolded `u' conformations. There is a significant energy barrier between `f' and `u' conformations. This is the first designed four-stranded antiparallel β-sheet that folds in water. PMID:12021452

  13. An Effective Instructional Design at Iowa Lakes Community College to Make Hard Core Unemployed People Employable.

    ERIC Educational Resources Information Center

    Schorzmann, Eugene F.

    In order to meet the educational needs of the hard-core unemployed, an instructional program known as the "Alternative Career Center" was designed at Iowa Lakes Community College to prepare those identified by the Comprehensive Employment and Training Act for employment. After reviewing the literature on educational programs for the hard-core…

  14. Design issues for a single core transformer thyristor controlled phase-angle regulator

    SciTech Connect

    Nyati, S.; Eitzmann, M.; Kappenman, J.; VanHouse, D.; Mohan, N.; Edris, A.

    1995-10-01

    Power electronics have become increasingly reliable and cost effective at the transmission voltage level. New applications of FACTS devices are just emerging, Thyristor Controlled Phase-Angle Regulator (TCPAR) being one of them. This paper identifies and discusses several gate control and system issues for application of single core transformer Thyristor Controlled Phase-Angle Regulator. Alternative TCPAR designs are proposed.

  15. 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.

  16. Design Guideline of Hollow-Core Fibres with Cobweb Cladding Structure

    NASA Astrophysics Data System (ADS)

    Huo, Liang; Yu, Rong-Jin; Zhang, Bing; Chen, Ming-Yang; Li, Bing-Xin

    2006-08-01

    By using a plane wave expansion method, some important parameters of designing the hollow-core fibre with cobweb cladding structure are analysed. Taking a dielectric material PMMA, for example, the tolerance of the parameters is discussed. The results show that the parameters of the structure possess oneself of a regularity and limit, and have a larger tolerance for the structural parameters in fabrication.

  17. 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…

  18. Designing high frequency ac inductors using ferrite and Molypermalloy Powder Cores (MPP)

    NASA Technical Reports Server (NTRS)

    Mclyman, W. T.; Wagner, A. P.

    1985-01-01

    The major considerations in the design of high frequency ac inductors are reviewed. Two methods for designing the inductor: the area product method and the core geometry method, are presented. The two major effects of the inductor air gap, fringing flux power loss and increase of inductance, are discussed. Equations for the inductor design and a step-by-step design procedure are given. The use of a lumped air gap or a distributed air gap are discussed and a comparison of the losses resulting from these gaps, together with experimental results are presented.

  19. Design and simulation of a neutron source based on an electron linear accelerator for BNCT of skin melanoma.

    PubMed

    Pazirandeh, Ali; Torkamani, Ali; Taheri, Ali

    2011-05-01

    The PNS project is devoted to the design and simulation of a Photo-Neutron Source for BNCT using a Varian 2300C/D electron accelerator. This paper describes the production of the high-energy gamma-rays followed by neutron production in (gamma,n) reaction. To optimize the whole setup and maximize the neutron flux, the FLUKA code is employed to simulate the system. The results show the neutron flux creation in the order of 10?(n/cm s) with a neutron spectrum that is practical for superficial cancers treatment using BNCT. PMID:21334211

  20. 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)

  1. The SNL100-02 blade : advanced core material design studies for the Sandia 100-meter blade.

    SciTech Connect

    Griffith, Daniel

    2013-11-01

    A series of design studies are performed to investigate the effects of advanced core materials and a new core material strategy on blade weight and performance for large blades using the Sandia 100-meter blade designs as a starting point. The initial core material design studies were based on the SNL100-01 100- meter carbon spar design. Advanced core material with improved performance to weight was investigated with the goal to reduce core material content in the design and reduce blade weight. A secondary element of the core study was to evaluate the suitability of core materials from natural, regrowable sources such as balsa and recyclable foam materials. The new core strategy for the SNL100-02 design resulted in a design mass of 59 tons, which is a 20% reduction from the most recent SNL100-01 carbon spar design and over 48% reduction from the initial SNL100-00 all-glass baseline blade. This document provides a description of the final SNL100-02 design, includes a description of the major design modifications, and summarizes the pertinent blade design information. This document is also intended to be a companion document to the distribution of the NuMAD blade model files for SNL100-02 that are made publicly available.

  2. 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 graphite reflected (2 inches or less) experiments also using the same set of highly enriched uranium metal parts are evaluated in HEU MET FAST 071. Polyethylene-reflected configurations are evaluated in HEU-MET-FAST-076. A stack of highly enriched metal discs with a thick beryllium top reflector is evaluated in HEU-MET-FAST-069, and two additional highly enriched uranium annuli with beryllium cores are evaluated in HEU-MET-FAST-059. Both detailed and simplified model specifications are provided in this evaluation. Both of these fast neutron spectra assemblies were determined to be acceptable benchmark experiments. The calculated eigenvalues for both the detailed and the simple benchmark models are within ~0.26 % of the benchmark values for Configuration 1 (calculations performed using MCNP6 with ENDF/B-VII.1 neutron cross section data), but under-calculate the benchmark values by ~7s because the uncertainty in the benchmark is very small: ~0.0004 (1s); for Configuration 2, the under-calculation is ~0.31 % and ~8s. Comparison of detailed and simple model calculations for the potassium worth measurement and potassium mass coefficient yield results approximately 70 – 80 % lower (~6s to 10s) than the benchmark values for the various nuclear data libraries utilized. Both the potassium worth and mass coefficient are also deemed to be acceptable benchmark experiment measurements.

  3. Probabilistic risk assessment in the design of the Advanced Neutron Source

    SciTech Connect

    Harrington, R.M.; Ramsey, C.T.; Fullwood, R.R.

    1994-09-01

    Probabilistic risk assessment (PRA) has been used extensively in the design of the Advanced Neutron Source (ANS) reactor to safety risk and enhance availability. Design feautures incorporated to minimize risk include submerged primary coolant piping, circulation cooling capability, and dual independent and diverse shutdown systems. The recently completed Level I, Phase 1 PRA shows that the risk dominating event sequence initiator is now blockage; a program to minimize this identified risk is described.

  4. 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).

  5. 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

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

    DOE PAGESBeta

    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

  7. Physics Analyses in the Design of the HFIR Cold Neutron Source

    SciTech Connect

    Bucholz, J.A.

    1999-09-27

    Physics analyses have been performed to characterize the performance of the cold neutron source to be installed in the High Flux Isotope Reactor at the Oak Ridge National Laboratory in the near future. This paper provides a description of the physics models developed, and the resulting analyses that have been performed to support the design of the cold source. These analyses have provided important parametric performance information, such as cold neutron brightness down the beam tube and the various component heat loads, that have been used to develop the reference cold source concept.

  8. Systematic approach for designing zero-DGD coupled multi-core optical fibers

    NASA Astrophysics Data System (ADS)

    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 then be used to determine the characteristics of the remaining waveguide elements. Using this technique, a number of pertinent geometrical configurations are investigated in order to minimize intermodal dispersion.

  9. Design issues of 40-Gbit/s WDM systems for metro and core network application

    NASA Astrophysics Data System (ADS)

    Freund, Ronald; Molle, Lutz; Hanik, Norbert; Richter, Andre

    2005-02-01

    Deployment of 40-Gbit/s technology in metro and core networks is still attractive, to bring down costs and to increase transmission capacity. This paper summarises design issues of 40-Gbit/s WDM systems for their application in wavelength division multiplexed metro and core networks. Parameter tolerances and transmission distances for different modulation formats are numerically and experimentally investigated. Based on Deutsche Telekom's fibre infrastructure, upgrade studies show that polarisation mode dispersion will be the main obstacle when installing 40-Gbit/s technology in deployed fibre infrastructure.

  10. Fuel and Core Design for Long Operating Cycle Simplified BWR (LSBWR)

    SciTech Connect

    Noriyuki Yoshida; Kouji Hiraiwa; Mikihide Nakamaru; Hideaki Heki

    2002-07-01

    This paper describes an innovative core concept currently being developed for long operating cycle simplified BWR (LSBWR). The LSBWR adopts the long cycle operation (15 years) for the elimination of the fuel pool and the refueling machines and for the capacity usage ratio improvement. To achieve long cycle operation, a combination of enriched gadolinium and 0.7- times sized small bundle with peripheral-positioned gadolinium rod is adopted as a key design concept. A nuclear design for fuel bundle has been determined based on three dimensional nuclear and thermal hydraulic calculation. A core performance has been evaluated based on this bundle design and thermal performance and reactivity characteristics indicated preferable value. (authors)

  11. 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.

  12. 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.

  13. Design of the radiation shielding for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamaka)

    NASA Astrophysics Data System (ADS)

    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.

  14. 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.

  15. Small angle neutron scattering study of nano sized microstructure in Fe-Cr ODS steels for gen IV in-core applications.

    PubMed

    Han, Young-Soo; Mao, Xiadong; Jang, Jinsung

    2013-11-01

    The nano-sized microstructures in Fe-Cr oxide dispersion strengthened steel for Gen IV in-core applications were studied using small angle neutron scattering. The oxide dispersion strengthened steel was manufactured through hot isostatic pressing with various chemical compositions and fabrication conditions. Small angle neutron scattering experiments were performed using a 40 m small angle neutron scattering instrument at HANARO. Nano sized microstructures, namely, yttrium oxides and Cr-oxides were quantitatively analyzed by small angle neutron scattering. The yttrium oxides and Cr-oxides were also observed by transmission electron microscopy. The microstructural analysis results from small angle neutron scattering were compared with those obtained by transmission electron microscopy. The effects of the chemical compositions and fabrication conditions on the microstructure were investigated in relation to the quantitative microstructural analysis results obtained by small angle neutron scattering. The volume fraction of Y-oxide increases after fabrication, and this result is considered to be due to the formation of non-stochiometric Y-Ti-oxides. PMID:24245285

  16. 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)

  17. 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. PMID:26242556

  18. The design of the inelastic neutron scattering mode for the Extreme Environment Diffractometer with the 26 T High Field Magnet

    NASA Astrophysics Data System (ADS)

    Bartkowiak, Maciej; Stüßer, Norbert; Prokhnenko, Oleksandr

    2015-10-01

    The Extreme Environment Diffractometer is a neutron time-of-flight instrument, designed to work with a constant-field hybrid magnet capable of reaching fields over 26 T, unprecedented in neutron science; however, the presence of the magnet imposes both spatial and technical limitations on the surrounding instrument components. In addition to the existing diffraction and small-angle neutron scattering modes, the instrument will operate also in an inelastic scattering mode, as a direct time-of-flight spectrometer. In this paper we present the Monte Carlo ray-tracing simulations, the results of which illustrate the performance of the instrument in the inelastic-scattering mode. We describe the focussing neutron guide and the chopper system of the existing instrument and the planned design for the instrument upgrade. The neutron flux, neutron spatial distribution, divergence distribution and energy resolution are calculated for standard instrument configurations.

  19. 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 previous insert designs. A casting process for manufacturing the v.3 inserts was developed. The developed casting process, when producing more than 13 inserts, becomes more economical than machining. An exploratory study was conducted looking at the effects of dynamic loading on the v.3 insert performance. The results of this study highlighted areas for improving dynamic testing of foam-core sandwich structure inserts. Correlations were developed relating design variables such as face-sheet thickness and insert diameter to a failure load for different load cases. This was done through simulations using Computer Aided Engineering (CAE) software, and experimental testing. The resulting correlations were integrated into a computer program which outputs the required insert dimensions given a set of design parameters, and load values.

  20. The design of a multisource americium-beryllium (Am-Be) neutron irradiation facility using MCNP for the neutronic performance calculation.

    PubMed

    Sogbadji, R B M; Abrefah, R G; Nyarko, B J B; Akaho, E H K; Odoi, H C; Attakorah-Birinkorang, S

    2014-08-01

    The americium-beryllium neutron irradiation facility at the National Nuclear Research Institute (NNRI), Ghana, was re-designed with four 20 Ci sources using Monte Carlo N-Particle (MCNP) code to investigate the maximum amount of flux that is produced by the combined sources. The results were compared with a single source Am-Be irradiation facility. The main objective was to enable us to harness the maximum amount of flux for the optimization of neutron activation analysis and to enable smaller sample sized samples to be irradiated. Using MCNP for the design construction and neutronic performance calculation, it was realized that the single-source Am-Be design produced a thermal neutron flux of (1.8±0.0007)×10(6) n/cm(2)s and the four-source Am-Be design produced a thermal neutron flux of (5.4±0.0007)×10(6) n/cm(2)s which is a factor of 3.5 fold increase compared to the single-source Am-Be design. The criticality effective, k(eff), of the single-source and the four-source Am-Be designs were found to be 0.00115±0.0008 and 0.00143±0.0008, respectively. PMID:24792123

  1. Design and Construction of the Ultracold Neutron Source at the NC State PULSTAR Research Reactor

    NASA Astrophysics Data System (ADS)

    Palmquist, Grant R.

    An ultracold neutron (UCN) source using solid deuterium is being constructed at the 1MWPULSTAR nuclear reactor on the campus of North Carolina State University. The final stages of assembly and commissioning are underway. The overall design, status of construction, and benchmarking measurements are presented. The UCN source design is based on detailed simulations including MCNP, UCN transport Monte Carlo, and computational fluid dynamics (CFD) simulation of the cryogenic systems. The source will be useful for developing UCN technologies, including guides and detectors, and in support of current projects including measurements of neutron beta-decay asymmetry coefficients and the electric dipolemoment of the neutron. The facility will also be available for testing new techniques using UCN in material and surface physics, as well as new fundamental physics measurements such as neutron lifetime and beta decay measurements. The expected experimental density of UCN/cm3 in a storage volume will be competitive with currently available sources, including those at significantly more powerful reactors.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Design, development and characterization of a novel neutron and x-ray combined computed tomography system

    NASA Astrophysics Data System (ADS)

    Sinha, Vaibhav

    Visualizing the three dimensional structure of objects (e.g. nuclear fuel, nuclear materials, explosives and bio materials) and phenomena (e.g. particle tracking) can be very important in nondestructive testing applications. Computed tomography systems are indispensable tools for these types of applications because they provide a versatile non-destructive technique for analysis. A novel neutron and X-ray combined computed tomography (NXCT) system has been designed and developed at the Missouri University of Science & Technology. The neutron and X-ray combined computed tomography system holds much promise for non-destructive material detection and analysis where multiple materials having similar atomic number and differing thermal cross section or vice versa may be present within an object, exclusive neutron or X-ray analysis may exhibit shortcomings in distinguishing interfaces. However, fusing neutron image and X-ray image offers the strengths of both and may provide a superior method of analysis. In addition, a feasible design of a sample positioning system which allows the user to remotely and automatically manipulate the objects makes the NXCT system viable for commercial applications. Moreover, characterization of the newly developed digital imaging system is imperative to the performance evaluation, as well as for describing the associated parameters. The performance of a combined neutron/X-ray digital imaging system was evaluated in terms of modulation transfer function (MTF), noise power spectrum (NPS) and detective quantum efficiency (DQE). This dissertation is a complete overview of the design of the NXCT system, operation, algorithms, performance evaluation and results.

  7. Informed Design: A Contemporary Approach to Design Pedagogy as the Core Process in Technology

    ERIC Educational Resources Information Center

    Burghardt, M. David; Hacker, Michael

    2004-01-01

    In classroom settings, most problems are usually well defined, so students have little experience with open-ended problems. Technological design problems, however, are seldom well defined. The design process begins with broad ideas and concepts and continues in the direction of ever-increasing detail, resulting in an acceptable solution. So using…

  8. Hardware-Software Co-design of QRD-RLS Algorithm with Microblaze Soft Core Processor

    NASA Astrophysics Data System (ADS)

    Lodha, Nupur; Rai, Nivesh; Dubey, Rahul; Venkataraman, Hrishikesh

    This paper presents the implementation of QR Decomposition based Recursive Least Square (QRD-RLS) algorithm on Field Programmable Gate Arrays (FPGA). The design is based on hardware-software co-design. The hardware part consists of a custom peripheral that solves the part of the algorithm with higher computational costs and the software part consists of an embedded soft core processor that manages the control functions and rest of the algorithm. The use of Givens Rotation and Systolic Arrays make this architecture suitable for FPGA implementation. Moreover, the speed and flexibility of FPGAs render them viable for such computationally intensive application. The system has been implemented on Xilinx Spartan 3E FPGA with Microblaze soft core processor using Embedded Development Kit (EDK). The paper also presents the implementation results and their analysis.

  9. 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)

  10. 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 description of the implementation of the recommended model and Chapter 5 discusses the testing which could be done to verify the design and implementation of the model.

  11. 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.

  12. Design, synthesis and photochemical properties of the first examples of iminosugar clusters based on fluorescent cores

    PubMed Central

    Lepage, Mathieu L; Mirloup, Antoine; Ripoll, Manon; Stauffert, Fabien; Bodlenner, Anne

    2015-01-01

    Summary The synthesis and photophysical properties of the first examples of iminosugar clusters based on a BODIPY or a pyrene core are reported. The tri- and tetravalent systems designed as molecular probes and synthesized by way of Cu(I)-catalysed azide–alkyne cycloadditions are fluorescent analogues of potent pharmacological chaperones/correctors recently reported in the field of Gaucher disease and cystic fibrosis, two rare genetic diseases caused by protein misfolding. PMID:26124868

  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. 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.

  15. Design manufacture and test of a cryo-stable Offner relay using aluminum foam core optics

    NASA Astrophysics Data System (ADS)

    McClelland, Ryan S.; Content, David A.

    2001-12-01

    Aluminum foam core optics have the desirable characteristics of being lightweight, cryo-stable, and low cost. The availability of high quality aluminum foam and a bare aluminum super-polishing process have allowed high performance foam core optics made entirely of aluminum to be produced. Mirrors with integral mounts were designed for minimum surface error induced by self-weight deflection, thermal gradients, and mounting stresses. The design of the optics was extensively optimized using Finite Element Analysis (FEA) and Geometric Element Analysis (GEA) to determine the effects of design parameters on mirror performance under the anticipated operating environments. A unique manufacturing process was developed to accommodate the aluminum brazing process used to install the aluminum foam while maintaining dimensional stability. Aluminum foam core optics have the additional advantage of being fabricated from a common aerospace structural material. An Offner relay using all aluminum optics and structure will be manufactured and tested with the goal of demonstrating that an all aluminum optical system can be aligned at room temperature and maintain alignment at cryogenic temperatures due to near zero CTE mismatch between all system components. If successful, an all aluminum Offner relay has potential uses for NGST, specifically in the testing of micro-mirror arrays.

  16. 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.

  17. Lunar in-core thermionic nuclear reactor power system conceptual design

    SciTech Connect

    Mason, L.S. ); Schmitz, P.C. ); Gallup, D.R. )

    1991-01-05

    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 Explortion 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.

  18. 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.

  19. Design of thick aperture for fine-resolution neutron penumbral imaging

    SciTech Connect

    Ress, D.

    1989-10-19

    Compact sources of 14-MeV neutrons have been imaged with a penumbral-coded aperture at a two-point resolution of 80{mu}m. We desire to improve the penumbral-aperture microscope to obtain resolutions as fine as 10{mu}m. In penumbral-coded-aperture imaging, the resolution is ultimately limited by the sharpness of the aperture point-spread function. I present a design for a thick penumbral aperture that provides the desired sharpness over a field of view of 150{mu}m. The point-spread function of these apertures is sufficiently isoplanatic and distortion-free to allow linear reconstruction of complex source distributions. The designs is generally appropriate for similar imaging techniques, such as fine-resolution neutron or gamma-ray pinhole imaging. 5 refs., 5 figs.

  20. 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

    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.

  1. CHINA SPALLATION NEUTRON SOURCE PROJECT: DESIGN ITERATIONS AND R AND D STATUS.

    SciTech Connect

    WEI,J.

    2006-09-21

    The China Spallation Neutron Source (CSNS) is an accelerator based high power project currently under preparation in China. The accelerator complex is based on an H{sup -} linear accelerator and a rapid cycling proton synchrotron. During the past year, the design of most accelerator systems went through major iterations, and initial research and developments were started on the prototyping of several key components. This paper summarizes major activities of the past year.

  2. 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 to maintain the biological dose equivalent during operation {le} 0.5 mrem/h inside the subcritical hall, which is five times less than the allowable dose for working forty hours per week for 50 weeks per year. This study analyzed and designed the thickness and the shape of the radial and top shields of the neutron source based on the biological dose equivalent requirements inside the subcritical hall during operation. The Monte Carlo code MCNPX is selected because of its capabilities for transporting electrons, photons, and neutrons. Mesh based weight windows variance reduction technique is utilized to estimate the biological dose outside the shield with good statistics. A significant effort dedicated to the accurate prediction of the biological dose equivalent outside the shield boundary as a function of the shield thickness without geometrical approximations or material homogenization. The building wall was designed with ordinary concrete to reduce the biological dose equivalent to the public with a safety factor in the range of 5 to 20.

  3. 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; Leslie Dutton, P; 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

  4. 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.

  5. 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 similar to those obtained with the traditional UO2 nuclear fuel.

  6. 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.

  7. The new vertical neutron beam line at the CERN n_TOF facility design and outlook on the performance

    NASA Astrophysics Data System (ADS)

    Weiß, C.; Chiaveri, E.; Girod, S.; Vlachoudis, V.; Aberle, O.; Barros, S.; Bergström, I.; Berthoumieux, E.; Calviani, M.; Guerrero, C.; Sabaté-Gilarte, M.; Tsinganis, A.; Andrzejewski, J.; Audouin, L.; Bacak, M.; Balibrea-Correa, J.; Barbagallo, M.; Bécares, V.; Beinrucker, C.; Belloni, F.; Bečvář, F.; Billowes, J.; Bosnar, D.; Brugger, M.; Caamaño, M.; Calviño, F.; Cano-Ott, D.; Cerutti, F.; Colonna, N.; Cortés, G.; Cortés-Giraldo, M. A.; Cosentino, L.; Damone, L.; Deo, K.; Diakaki, M.; Domingo-Pardo, C.; Dupont, E.; Durán, I.; Dressler, R.; Fernández-Domínguez, B.; Ferrari, A.; Ferreira, P.; Finocchiaro, P.; Frost, R.; Furman, V.; Ganesan, S.; Gheorghe, A.; Glodariu, T.; Göbel, K.; Gonçalves, I. F.; González-Romero, E.; Goverdovski, A.; Griesmayer, E.; Gunsing, F.; Harada, H.; Heftrich, T.; Heinitz, S.; Hernández-Prieto, A.; Heyse, J.; Jenkins, D. G.; Jericha, E.; Kadi, Y.; Käppeler, F.; Katabuchi, T.; Kavrigin, P.; Ketlerov, V.; Khryachkov, V.; Kimura, A.; Kivel, N.; Kokkoris, M.; Krtička, M.; Leal-Cidoncha, E.; Lederer, C.; Leeb, H.; Lerendegui, J.; Licata, M.; Lo Meo, S.; López, D.; Losito, R.; Macina, D.; Marganiec, J.; Martínez, T.; Massimi, C.; Mastinu, P. F.; Mastromarco, M.; Matteucci, F.; Mendoza, E.; Mengoni, A.; Milazzo, P. M.; Mingrone, F.; Mirea, M.; Montesano, S.; Musumarra, A.; Nolte, R.; Palomo Pinto, R.; Paradela, C.; Patronis, N.; Pavlik, A.; Perkowski, J.; Porras, I.; Praena, J.; Quesada, J. M.; Rauscher, T.; Reifarth, R.; Riego-Perez, A.; Robles, M. S.; Rubbia, C.; Ryan, J.; Saxena, A.; Schillebeeckx, P.; Schmidt, S.; Schumann, D.; Sedyshev, P.; Smith, G.; Stamatopoulos, A.; Steinegger, P.; Suryanarayana, S. V.; Tagliente, G.; Tain, J. L.; Tarifeño-Saldivia, A.; Tassan-Got, L.; Valenta, S.; Vannini, G.; Variale, V.; Vaz, P.; Ventura, A.; Vlastou, R.; Wallner, A.; Warren, S.; Weigand, M.; Wright, T.; Žugec, P.

    2015-11-01

    At the neutron time-of-flight facility n_TOF at CERN a new vertical beam line was constructed in 2014, in order to extend the experimental possibilities at this facility to an even wider range of challenging cross-section measurements of interest in astrophysics, nuclear technology and medical physics. The design of the beam line and the experimental hall was based on FLUKA Monte Carlo simulations, aiming at maximizing the neutron flux, reducing the beam halo and minimizing the background from neutrons interacting with the collimator or back-scattered in the beam dump. The present paper gives an overview on the design of the beam line and the relevant elements and provides an outlook on the expected performance regarding the neutron beam intensity, shape and energy resolution, as well as the neutron and photon backgrounds.

  8. 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

  9. Conceptual design for a fast neutron ionization chamber for fusion reactor plasma diagnostics

    SciTech Connect

    Sailor, W.C.; Barnes, C.W.

    1994-06-01

    A conceptual design for a radiation-hard ``pointing`` fast neutron ionization chamber that is capable of delivering a 1 MHz countrate of T(D,n) events at ITER is given. The detector will use a {approximately}1 cm{sup 3} volume of CO{sub 2} fill gas at 0.1 bar pressure in a 500 V/cm electric field. The pulse widths will be {approximately}10 ns, enabling it to operate in a flux of {approximately} 6 {times} 10{sup 13} DT n/cm{sup 2}/sec. A special collimator design is used, giving an estimated angular resolution of 4.5 degrees HWHM.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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. PMID:25362410

  15. 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 asymmetric absorber rods), surface-dependent discontinuity factors based on nodal equivalence theory have been introduced into the nodal diffusion theory option of the DIF3D code (DIF3D-nodal) to improve modeling accuracy. Additionally, the discontinuity factors based on the Simplified Equivalence Theory (SET) have been incorporated as an alternative and may be employed for both the DIF3D-nodal and DIF3D-VARIANT (nodal transport) solution options. Two- and three-dimensional core calculations have been performed using the routines developed and modified in this work, along with cross sections generated from single fuel block and one-dimensional or two-dimensional fuel-reflector model. Generally, REBUS-3/DIF3D results for the core multiplication factor and power distribution are found to be in good agreement with reference results (generated with MCNP continuous energy calculations) particularly when discontinuity factors are applied. The DIF3D-VARIANT option was found to provide a more accurate solution in its diffusion approximation than the DIF3D-nodal option. Control rod worths can be estimated with acceptably small errors compared to MCNP results. However, estimation of the core power tilt needs to be improved by introducing the surface-dependent discontinuity factor capability in DIF3D-VARIANT.

  16. 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.

  17. 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.

  18. 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

  19. Thoria-based cermet nuclear fuel : neutronics fuel design and fuel cycle analysis.

    SciTech Connect

    Downar, T.J.; McDeavitt, S.M.; Revankar, S.T.; Solomon, A.A.; Kim, T.K.

    2002-01-09

    Cermet nuclear fuel has been demonstrated to have significant potential to enhance fuel performance because of low internal fuel temperatures and low stored energy. The combination of these benefits with the inherent proliferation resistance, high burnup capability, and favorable neutronic properties of the thorium fuel cycle produces intriguing options for advanced nuclear fuel cycles. This paper describes aspects of a Nuclear Energy Research Initiative (NERI) project with two primary goals: (1) Evaluate the feasibility of implementing the thorium fuel cycle in existing or advanced reactors using a zirconium-matrix cermet fuel, and (2) Develop enabling technologies required for the economic application of this new fuel form. This paper will first describes the fuel thermal performance model developed for the analysis of dispersion metal matrix fuels. The model is then applied to the design and analysis of thorium/uranium/zirconium metal-matrix fuel pins for light-water reactors using neutronic simulation methods.

  20. Exploration of Adiabatic Resonance Crossing Through Neutron Activator Design for Thermal and Epithermal Neutron Formation in (99)Mo Production and BNCT Applications.

    PubMed

    Khorshidi, Abdollah

    2015-10-01

    A feasibility study was performed to design thermal and epithermal neutron sources for radioisotope production and boron neutron capture therapy (BNCT) by moderating fast neutrons. The neutrons were emitted from the reaction between (9)Be, (181)Ta, and (184)W targets and 30 MeV protons accelerated by a small cyclotron at 300 μA. In this study, the adiabatic resonance crossing (ARC) method was investigated by means of (207)Pb and (208)Pb moderators, graphite reflector, and boron absorber around the moderator region. Thermal/epithermal flux, energy, and cross section of accumulated neutrons in the activator were examined through diverse thicknesses of the specified regions. Simulation results revealed that the (181)Ta target had the highest neutron yield, and also tungsten was found to have the highest values in both surface and volumetric flux ratio. Transmutation in the (98)Mo sample through radiative capture was investigated for the natural lead moderator. When the sample radial distance from the target was increased inside the graphite region, the production yield had the greatest value of activity. The potential of the ARC method is a replacement or complements the current reactor-based supply sources of BNCT purposes. PMID:26397967

  1. Preliminary core design studies for the advanced burner reactor over a wide range of conversion ratios.

    SciTech Connect

    Hoffman, E. A.; Yang, W. S.; Hill, R. N.; Nuclear Engineering Division

    2008-05-05

    A consistent set of designs for 1000 MWt commercial-scale sodium-cooled Advance Burner Reactors (ABR) have been developed for both metal and oxide-fueled cores with conversion ratios from breakeven (CR=1.0) to fertile-free (CR=0.0). These designs are expected to satisfy thermal and irradiation damage limits based on the currently available data. The very low conversion ratio designs require fuel that is beyond the current fuel database, which is anticipated to be qualified by and for the Advanced Burned Test Reactor. Safety and kinetic parameters were calculated, but a safety analysis was not performed. Development of these designs was required to achieve the primary goal of this study, which was to generate representative fuel cycle mass flows for system studies of ABRs as part of the Global Nuclear Energy Partnership (GNEP). There are slight variations with conversion ratio but the basic ABR configuration consists of 144 fuel assemblies and between 9 and 22 primary control assemblies for both the metal and oxide-fueled cores. Preliminary design studies indicated that it is feasible to design the ABR to accommodate a wide range of conversion ratio by employing different assembly designs and including sufficient control assemblies to accommodate the large reactivity swing at low conversion ratios. The assemblies are designed to fit within the same geometry, but the size and number of fuel pins within each assembly are significantly different in order to achieve the target conversion ratio while still satisfying thermal limits. Current irradiation experience would allow for a conversion ratio of somewhat below 0.75. The fuel qualification for the first ABR should expand this experience to allow for much lower conversion ratios and higher bunrups. The current designs were based on assumptions about the performance of high and very high enrichment fuel, which results in significant uncertainty about the details of the designs. However, the basic fuel cycle performance trends such as conversion ratio and mass flow parameters are less sensitive to these parameters and the current results should provide a good basis for static and dynamic system analysis. The conversion ratio is fundamentally a ratio of the macroscopic cross section of U-238 capture to that of TRU fission. Since the microscopic cross sections only change moderately with fuel design and isotopic concentration for the fast reactor, a specific conversion ratio requires a specific enrichment. The approximate average charge enrichment (TRU/HM) is 14%, 21%, 33%, 56%, and 100% for conversion ratios of 1.0, 0.75, 0.50, 0.25, and 0.0 for the metal-fueled cores. The approximate average charge enrichment is 17%, 25%, 38%, 60%, and 100% for conversion ratios of 1.0, 0.75, 0.50, 0.25, and 0.0 for the oxide-fueled core. For the split batch cores, the maximum enrichment will be somewhat higher. For both the metal and oxide-fueled cores, the reactivity feedback coefficients and kinetics parameters seem reasonable. The maximum single control assembly reactivity faults may be too large for the low conversion ratio designs. The average reactivity of the primary control assemblies was increased, which may cause the maximum reactivity of the central control assembly to be excessive. The values of the reactivity coefficients and kinetics parameters show that some values appear to improve significantly at lower conversion ratios while others appear far less favorable. Detailed safety analysis is required to determine if these designs have adequate safety margins or if appropriate design modifications are required. Detailed system analysis data has been generated for both metal and oxide-fueled core designs over the entire range of potential burner reactors. Additional data has been calculated for a few alternative fuel cycles. The systems data has been summarized in this report and the detailed data will be provided to the systems analysis team so that static and dynamic system analyses can be performed.

  2. 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.

  3. 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.

  4. 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…

  5. 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. Preliminary design studies looking at fast ignition and hot spot ignition fusion options are documented, along with limited scoping studies performed to investigate other options of interest that surfaced during the main design effort. Lastly, side studies that were not part of the main design effort but may alter future work performed on LIFE engine designs are shown. The majority of all work reported in this document was performed during the Molten Salt Fast Ignition Moderator Study (MSFIMS) which sought to optimize the amount of moderator mixed into the molten salt region in order to produce the most compelling design. The studies in this report are of a limited scope and are intended to provide a preliminary neutronics analysis of the design concepts described herein to help guide decision processes and explore various options that a LIFE engine with a molten salt blanket might enable. None of the designs shown in this report, even reference cases selected for detailed description and analysis, have been fully optimized. The analyses were performed primarily as a neutronics study, though some consultation was made regarding thermal-hydraulic and structural concerns during both scoping out an initial model and subsequent to identifying a neutronics-based reference case to ensure that the design work contained no glaring mechanical or thermal issues that would preclude its feasibility. Any analyses and recommendations made in this report are either primarily or solely from the point of view of LIFE neutronics and ignore other fundamental issues related to molten salt fuel blankets such as chemical processing feasibility and political feasibility of a molten salt system.

  6. 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…

  7. Gigawatt, Closed Cycle, Vapor Core-Mhd Space Power System Conceptual Design Study

    NASA Astrophysics Data System (ADS)

    Wetch, Joseph R.; Rhee, Hyop S.; Koester, J. Kent; Goodman, Julius; Maya, Issac

    1988-04-01

    A conceptual design study for a closed cycle gigawatt electric space power system has been conducted. The closed cycle static operation reduces power system interaction effects upon the space craft. This system utilizes a very high temperature (5500 K) plasma core reactor and a magnetohydrodynamic (MHD) power conversion subsystem to provide a power density of about 8 kWe/kg (0.13 kg/kWe) for several kilo-seconds. Uranium vapor is the fuel. Candidate working fluids are metal vapors such as lithium or calcium. The system is based on a Rankine cycle to minimize the electromagnetic pumping power requirement. The fission fragment induced nonequilibrium ionization in the plasma in the MHD power duct provides the plasma electric conductivity for gigawatt power generation. Waste heat is rejected utilizing lithium heat pipes at temperatures just below 2000 K, thus minimizing the radiator area requirement. Key technology issues are identified, including the containment of the 5500 K 'sun-liken plasma at 4 to 0 MPa In a reflector moderated, gas/vapor filled cavity core reactor. A promising scheme to protect the refractory metal reactor inner wall is presented, together with a heating load analysis in the wall. This scheme utilizes an ablating film of liquid lithium/calcium that evaporates into the cavity core to become the working fluid of the cycle.

  8. Thermal Design of an Ultrahigh Temperature Vapor Core Reactor Combined Cycle Nuclear Power Plant

    SciTech Connect

    Bays, Samuel E.; Anghaie, Samim; Smith, Blair; Knight, Travis

    2004-07-01

    Current work modeling high temperature compact heat exchangers may demonstrate the design feasibility of a Vapor Core Reactor (VCR) driven combined cycle power plant. For solid nuclear fuel designs, the cycle efficiency is typically limited by a metallurgical temperature limit which is dictated by fuel and structural melting points. In a vapor core, the gas/vapor phase nuclear fuel is uniformly mixed with the topping cycle working fluid. Heat is generated homogeneously throughout the working fluid thus extending the metallurgical temperature limit. Because of the high temperature, magnetohydrodynamic (MHD) generation is employed for topping cycle power extraction. MHD rejected heat is transported via compact heat exchanger to a conventional Brayton gas turbine bottoming cycle. High bottoming cycle mass flow rates are required to remove the waste heat because of low heat capacities for the bottoming cycle gas. High mass flow is also necessary to balance the high Uranium Tetrafluoride (UF{sub 4}) mass flow rate in the topping cycle. Heat exchanger design is critical due to the high temperatures and corrosive influence of fluoride compounds and fission products existing in VCR/MHD exhaust. Working fluid compositions for the topping cycle include variations of Uranium Tetrafluoride, Helium and various electrical conductivity seeds for the MHD. Bottoming cycle working fluid compositions include variations of Helium and Xenon. Some thought has been given to include liquid metal vapor in the bottoming cycle for a Cheng or evaporative cooled design enhancement. The NASA Glenn Lewis Research Center code Chemical Equilibrium with Applications (CEA) is utilized for evaluating chemical species existing in the gas stream. Work being conducted demonstrates the compact heat exchanger design, utilization of the CEA code, and assessment of different topping and bottoming working fluid compositions. (authors)

  9. Is it possible to use "twin cores" as a unique sedimentary record? An experimental design based on sediment color

    NASA Astrophysics Data System (ADS)

    Veiga-Pires, C.; Mestre, N. C.

    2009-01-01

    Sedimentary cores are widely used for studying Quaternary records. However, the amount of sediment that is available is proportional to the diameter of the core, which is rarely bigger than 15 cm. One way to obtain more sediment is to use two cores retrieved from almost the same location and use them as if they represent a unique sedimentary record. In the present work, an experimental design has been applied to verify if "twin cores" from an estuary can be considered as representing the same sedimentary record with twice the amount of sediment to study. Because sediment can be characterized based on its color, the variables used as replicates in the experimental design are the three Lab CIE colors acquired with a X-Rite Colortron spectrophotometer. Sediment cores were retrieved from the upper saltmarsh of Gilão River's estuary, southern Portugal. Twin cores, with in between distances of 50 cm, 100 cm and 200 cm, from two different sites were analysed. Results from a nested ANOVA show that even for the closest twin cores (50 cm apart) there is at least one color variable that shows significant variations between the profiles of both cores. These results clearly show that "twin cores" cannot be used as a unique sedimentary record without any previous testing, at least in such transitional regions.

  10. Soft error rate simulation and initial design considerations of neutron intercepting silicon chip (NISC)

    NASA Astrophysics Data System (ADS)

    Celik, Cihangir

    Advances in microelectronics result in sub-micrometer electronic technologies as predicted by Moore's Law, 1965, which states the number of transistors in a given space would double every two years. The most available memory architectures today have submicrometer transistor dimensions. The International Technology Roadmap for Semiconductors (ITRS), a continuation of Moore's Law, predicts that Dynamic Random Access Memory (DRAM) will have an average half pitch size of 50 nm and Microprocessor Units (MPU) will have an average gate length of 30 nm over the period of 2008-2012. Decreases in the dimensions satisfy the producer and consumer requirements of low power consumption, more data storage for a given space, faster clock speed, and portability of integrated circuits (IC), particularly memories. On the other hand, these properties also lead to a higher susceptibility of IC designs to temperature, magnetic interference, power supply, and environmental noise, and radiation. Radiation can directly or indirectly affect device operation. When a single energetic particle strikes a sensitive node in the micro-electronic device, it can cause a permanent or transient malfunction in the device. This behavior is called a Single Event Effect (SEE). SEEs are mostly transient errors that generate an electric pulse which alters the state of a logic node in the memory device without having a permanent effect on the functionality of the device. This is called a Single Event Upset (SEU) or Soft Error . Contrary to SEU, Single Event Latchup (SEL), Single Event Gate Rapture (SEGR), or Single Event Burnout (SEB) they have permanent effects on the device operation and a system reset or recovery is needed to return to proper operations. The rate at which a device or system encounters soft errors is defined as Soft Error Rate (SER). The semiconductor industry has been struggling with SEEs and is taking necessary measures in order to continue to improve system designs in nano-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 results show that soft errors in the memories increase proportionally with the neutron flux, whereas they decrease with increasing the supply voltages. NISC design considerations include the effects of device scaling, 10B content in the BPSG layer, incoming neutron energy, and critical charge of the node for this dissertation. NISCSAT simulations were performed with various memory node models to account these effects. Device scaling simulations showed that any further increase in the thickness of the BPSG layer beyond 2 mum causes self-shielding of the incoming neutrons due to the BPSG layer and results in lower detection efficiencies. Moreover, if the BPSG layer is located more than 4 mum apart from the depletion region in the node, there are no soft errors in the node due to the fact that both of the reaction products have lower ranges in the silicon or any possible node layers. Calculation results regarding the critical charge indicated that the mean charge deposition of the reaction products in the sensitive volume of the node is about 15 fC. It is evident that the NISC design should have a memory architecture with a critical charge of 15 fC or less to obtain higher detection efficiencies. Moreover, the sensitive volume should be placed in close proximity to the BPSG layers so that its location would be within the range of alpha and 7Li particles. Results showed that the distance between the BPSG layer and the sensitive volume should be less than 2 mum to increase the detection efficiency of the NISC. Incoming neutron energy was also investigated by simulations and the results obtained from these simulations showed that NISC neutron detection efficiency is related with the neutron cross-sections of 10B (n,alpha) 7Li reaction, e.g., ratio of the thermal (0.0253 eV) to fast (2 MeV) neutron detection efficiencies is approximately equal to 8000:1. Environmental conditions and their effects on the NISC performance were also studied in this research. Cosmic rays were modeled and simulated via NISCSAT to investigate detection reliability of the NISC. Simulation results show that cosmic rays account for less than 2 % of the soft errors for the thermal neutron detection. On the other hand, fast neutron detection by the NISC, which already has a poor efficiency due to the low neutron cross-sections, becomes almost impossible at higher altitudes where the cosmic ray fluxes and their energies are higher. NISCSAT simulations regarding soft error dependency of the NISC for temperature and electromagnetic fields show that there are no significant effects in the NISC detection efficiency. Furthermore, the detection efficiency of the NISC decreases with both air humidity and use of moderators since the incoming neutrons scatter away before reaching the memory surface.

  11. 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 the same radial flibe flow that travels through perforated ODS walls to the reflector blanket. This reflector blanket is 75 cm thick comprised of 2 cm diameter graphite pebbles cooled by flibe. The flibe extraction plenum surrounds the reflector bed. Detailed neutronics designs studies are performed to arrive at the described design. The LFFH engine thermal power is controlled using a technique of adjusting the 6Li/7Li enrichment in the primary and secondary coolants. The enrichment adjusts system thermal power in the design by increasing tritium production while reducing fission. To perform the simulations and design of the LFFH engine, a new software program named LFFH Nuclear Control (LNC) was developed in C++ to extend the functionality of existing neutron transport and depletion software programs. Neutron transport calculations are performed with MCNP5. Depletion calculations are performed using Monteburns 2.0, which utilizes ORIGEN 2.0 and MCNP5 to perform a burnup calculation. LNC supports many design parameters and is capable of performing a full 3D system simulation from initial startup to full burnup. It is able to iteratively search for coolant 6Li enrichments and resulting material compositions that meet user defined performance criteria. LNC is utilized throughout this study for time dependent simulation of the LFFH engine. Two additional methods were developed to improve the computation efficiency of LNC calculations. These methods, termed adaptive time stepping and adaptive mesh refinement were incorporated into a separate stand alone C++ library name the Adaptive Burnup Library (ABL). The ABL allows for other client codes to call and utilize its functionality. Adaptive time stepping is useful for automatically maximizing the size of the depletion time step while maintaining a desired level of accuracy. Adaptive meshing allows for analysis of fixed fuel configurations that would normally require a computationally burdensome number of depletion zones. Alternatively, Adaptive Mesh Refinement (AMR) adjusts the depletion zone size according to the variation in flux across the zone or fractional contribution to total absorption or fission. A parametric analysis on a fully mixed fuel core was performed using the LNC and ABL code suites. The resulting system parameters are found to optimize performance metrics using a 20 MT DU fuel load with a 20% TRISO packing and a 300 im kernel diameter operated with a fusion input power of 500 MW and a fission blanket gain of 4.0. LFFH potentially offers a proliferation resistant technology relative to other nuclear energy systems primarily because of no need for fuel enrichment or reprocessing. A figure of merit of the material attractiveness is examined and it is found that the fuel is effectively contaminated to an unattractive level shortly after the system is started due to fission product and minor actinide build up.

  12. Reactor core design and modeling of the MIT research reactor for conversion to LEU

    SciTech Connect

    Newton, Thomas H. Jr.; Olson, Arne P.; Stillman, John A.

    2008-07-15

    Feasibility design studies for conversion of the MIT Research Reactor (MITR) to LEU are described. Because the reactor fuel has a rhombic cross section, a special input processor was created in order to model the reactor in great detail with the REBUS-PC diffusion theory code, in 3D (triangular-z) geometry. Comparisons are made of fuel assembly power distributions and control blade worth vs. axial position, between REBUS-PC results and Monte Carlo predictions from the MCNP code. Results for the original HEU core at zero burnup are also compared with measurement. These two analysis methods showed remarkable agreement. Ongoing fuel cycle studies are summarized. A status report will be given as to results thus far that affect key design decisions. Future work plans and schedules to achieve completion of the conversion are presented. (author)

  13. Statistical core design methodology using the VIPRE thermal-hydraulics code

    SciTech Connect

    Lloyd, M.W.; Feltus, M.A.

    1994-12-31

    This Penn State Statistical Core Design Methodology (PSSCDM) is unique because it not only includes the EPRI correlation/test data standard deviation but also the computational uncertainty for the VIPRE code model and the new composite box design correlation. The resultant PSSCDM equation mimics the EPRI DNBR correlation results well, with an uncertainty of 0.0389. The combined uncertainty yields a new DNBR limit of 1.18 that will provide more plant operational flexibility. This methodology and its associated correlation and uniqe coefficients are for a very particular VIPRE model; thus, the correlation will be specifically linked with the lumped channel and subchannel layout. The results of this research and methodology, however, can be applied to plant-specific VIPRE models.

  14. 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.

  15. 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.

  16. Conceptual design of a high-intensity positron source for the Advanced Neutron Source

    SciTech Connect

    Hulett, L.D.; Eberle, C.C.

    1994-12-01

    The Advanced Neutron Source (ANS) is a planned new basic and applied research facility based on a powerful steady-state research reactor that provides neutrons for measurements and experiments in the fields of materials science and engineering, biology, chemistry, materials analysis, and nuclear science. The useful neutron flux will be at least five times more than is available in the world`s best existing reactor facility. Construction of the ANS provides a unique opportunity to build a positron spectroscopy facility (PSF) with very-high-intensity beams based on the radioactive decay of a positron-generating isotope. The estimated maximum beam current is 1000 to 5000 times higher than that available at the world`s best existing positron research facility. Such an improvement in beam capability, coupled with complementary detectors, will reduce experiment durations from months to less than one hour while simultaneously improving output resolution. This facility will remove the existing barriers to the routine use of positron-based analytical techniques and will be a giant step toward realization of the full potential of the application of positron spectroscopy to materials science. The ANS PSF is based on a batch cycle process using {sup 64}Cu isotope as the positron emitter and represents the status of the design at the end of last year. Recent work not included in this report, has led to a proposal for placing the laboratory space for the positron experiments outside the ANS containment; however, the design of the positron source is not changed by that relocation. Hydraulic and pneumatic flight tubes transport the source material between the reactor and the positron source where the beam is generated and conditioned. The beam is then transported through a beam pipe to one of several available detectors. The design presented here includes all systems necessary to support the positron source, but the beam pipe and detectors have not been addressed yet.

  17. 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 were learned from this effect. This paper addresses these issues and the resulting methodology. PMID:15353738

  18. 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 several obsolete components of the current analytical tool set used for ATR neutronics support. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). It will also greatly facilitate the LEU conversion effort, since the upgraded computational capabilities are now at a stage where they can be, and in fact have been, used for the required physics analysis from the beginning. In this context, extensive scoping neutronics analyses were completed for six preconceptual candidate LEU fuel element designs for the ATR (and for its companion critical facility, ATRC). Of these, four exhibited neutronics performance in what is believed to be an acceptable range. However, there are currently some concerns with regard to fabricability and mechanical performance that have emerged for one of the four latter concepts. Thus three concepts have been selected for more comprehensive conceptual design analysis during the upcoming fiscal year.

  19. 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.

  20. 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.

  1. 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.

  2. Photoelastic stress and fracture analysis of two neutron tube designs. Final report

    SciTech Connect

    Parks, V.J.; Sanford, R.J.

    1980-10-15

    Photoelastic stress analysis and fracture stress analysis were carried out on two-dimensional models of the cross-sections of the M5N and M19N designs of a neutron tube under axial load. Of special interest was the metal-ceramic interface, where failure had been experienced. Both types of analysis showed the M5N to be superior to the M19N for the given loading in the area of failure. Tangential stresses along the free surfaces, and mixed-mode stress intensity factors for various crack lengths along the interface are reported.

  3. A conceptual design of a beam-shaping assembly for boron neutron capture therapy based on deuterium-tritium neutron generators.

    PubMed

    Martn, 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. PMID:15191299

  4. 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. PMID:25430294

  5. Optimal design at inner core of the shaped pyramidal truss structure

    NASA Astrophysics Data System (ADS)

    Lee, Sung-Uk; Yang, Dong-Yol

    2013-12-01

    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.

  6. 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.

  7. Design and fabrication of a novel self-powered solid-state neutron detector

    NASA Astrophysics Data System (ADS)

    LiCausi, Nicholas

    There is a strong interest in intercepting special nuclear materials (SNM) at national and international borders and ports for homeland security applications. Detection of SNM such as U and Pu is often accomplished by sensing their natural or induced neutron emission. Such detector systems typically use thermal neutron detectors inside a plastic moderator. In order to achieve high detection efficiency gas filled detectors are often used; these detectors require high voltage bias for operation, which complicates the system when tens or hundreds of detectors are deployed. A better type of detector would be an inexpensive solid-state detector that can be mass-produced like any other computer chip. Research surrounding solid-state detectors has been underway since the late 1990's. A simple solid-state detector employs a planar solar-cell type p-n junction and a thin conversion material that converts incident thermal neutrons into detectable alpha-particles and 7Li ions. Existing work has typically used 6LiF or 10B as this conversion layer. Although a simple planar detector can act as a highly portable, low cost detector, it is limited to relatively low detection efficiency (˜10%). To increase the efficiency, 3D perforated p-i-n silicon devices were proposed. To get high efficiency, these detectors need to be biased, resulting in increased leakage current and hence detector noise. In this research, a new type of detector structure was proposed, designed and fabricated. Among several detector structures evaluated, a honeycomb-like silicon p-n structure was selected, which is filled with natural boron as the neutron converter. A silicon p+-n diode formed on the thin silicon wall of the honeycomb structure detects the energetic alpha-particles emitted from the boron conversion layer. The silicon detection layer is fabricated to be fully depleted with an integral step during the boron filling process. This novel feature results in a simplified fabrication process. Three key advantages of the novel devices are theoretical neutron detection efficiency of ˜48%, a self-passivating structure that reduces leakage current and detector operation with no bias resulting in extremely low device noise. Processes required to fabricate the 3D type detector were explored and developed in this thesis. The detector capacitance and processing steps have been simulated with MEDICI and TSuprem-4, respectively. Lithography masks were then designed using Cadence. The fabrication process development was conducted in line with standard CMOS grade integrated circuit processing to allow for simple integration with existing fabrication facilities. A number of new processes were developed including the low pressure chemical vapor deposition of conformal boron films using diborane on very high aspect-ratio trenches and holes. Development also included methods for "wet" chemical etching and "dry" reactive ion etching of the deposited boron films. Fabricated detectors were characterized with the transmission line method, 4-point probe, I-V measurements and C-V measurements. Finally the detector response to thermal neutrons was studied. Characterization has shown significant reduction in reverse leakage current density to ˜8x10-8 A/cm2 (nearly 4 orders of magnitude over the previously published data). Results show that the fabrication process developed is capable of producing efficient (˜22.5%) solid-state thermal neutron detectors.

  8. Preliminary probabilistic design accident evaluation of the cold source facilities of the advanced neutron source

    SciTech Connect

    Harrington, R.M.; Ramsey, C.T.

    1995-08-01

    Consistent with established Advanced Neutron Source (ANS) project policy for the use of probabilistic risk assessment (PRA) in design, a task has been established to use PRA techniques to help guide the design and safety analysis of the ANS cold sources. The work discussed in this report is the first formal output of the cold source PRA task. The major output at this stage is a list of design basis accidents, categorized into approximate frequency categories. This output is expected to focus attention on continued design work to define and optimize the design such that design basis accidents are better defined and have acceptable outcomes. Categorizing the design basis events (DBEs) into frequency categories should prove helpful because it will allow appropriate acceptance criteria to be applied. Because the design of the cold source is still proceeding, it is beyond the scope of this task to produce detailed event probability calculations or even, in some cases, detailed event sequence definitions. That work would take place as a logically planned follow-on task, to be completed as the design matures. Figure 1.1 illustrates the steps that would typically be followed in selecting design basis accidents with the help of PRA. Only those steps located above the dashed line on Fig. 1.1 are included in the scope of the present task. (Only an informal top-level failure modes and effects analysis was done.) With ANS project closeout expected in the near future, the scope of this task has been abbreviated somewhat beyond the state of available design information on the ANS cold sources, or what could be achieved in a reasonable time. This change was necessary to ensure completion before the closeout and because the in-depth analytical support necessary to define fully some of the accidents has already been curtailed.

  9. 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.

  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. 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 of: (1) comparison to an analytical solution for the dam break problem, (2) water spreading tests in a 1/10 linear scale model of the Mark I containment by Theofanous et al., and (3) steel spreading tests by Suzuki et al. that were also conducted in a geometry similar to the Mark I. The objective of this work was to utilize the MELTSPREAD code to check the assumption of uniform melt spreading in the EPR core catcher design. As a starting point for the project, the code was validated against the worldwide melt spreading database that emerged after the code was originally written in the very early 1990's. As part of this exercise, the code was extensively modified and upgraded to incorporate findings from these various analytical and experiment programs. In terms of expanding the ability of the code to analyze various melt simulant experiments, the options to input user-specified melt and/or substrate material properties was added. The ability to perform invisicid and/or adiabatic spreading analysis was also added so that comparisons with analytical solutions and isothermal spreading tests could be carried out. In terms of refining the capability to carry out reactor material melt spreading analyses, the code was upgraded with a new melt viscosity model; the capability was added to treat situations in which solid fraction buildup between the liquidus-solidus is non-linear; and finally, the ability to treat an interfacial heat transfer resistance between the melt and substrate was incorporated. This last set of changes substantially improved the predictive capability of the code in terms of addressing reactor material melt spreading tests. Aside from improvements and upgrades, a method was developed to fit the model to the various melt spreading tests in a manner that allowed uncertainties in the model predictions to be statistically characterized. With these results, a sensitivity study was performed to investigate the assumption of uniform spreading in the EPR core catcher that addressed parametric variations in: (1) melt pour mass, (2) melt composition, (3) melt pour rate, (4) pour configuration (i.e., homogeneous vs. stratified metal-oxide phases), (5) melt temperature, (6) cavity condition (wet vs. dry), (7) spreading channel inclination angle, and finally (8) uncertainties in the melt viscosity correlation that are based on comparisons with the reactor material melt spreading database. Although differences were found in the rate of spreading and the degree to which the sacrificial concrete in the spreading room is ablated during the transients, in all cases the melt eventually (over a period of minutes) spreads to a uniform depth in the system.

  12. Design, Synthesis, and Characterization of Bent-Core Mesogen-Jacketed Liquid Crystalline Polymers

    SciTech Connect

    Chen,X.; Tenneti, K.; Li, C.; Bai, Y.; Zhou, R.; Wan, X.; Fan, X.; Zhou, Q.

    2006-01-01

    We report the design, synthesis, and characterization of a series of mesogen-jacketed liquid crystalline polymers with bent-core liquid crystals (BCLCs). For the first time, BCLC mesogens were directly side-attached to the polymer backbone and bent-core mesogen-jacketed liquid crystalline polymers (BMJLCPs) were achieved. Both three-ring and five-ring mesogens were employed. The n-alkoxy substituent lengths for the three-ring and five-ring BMJLCPs were controlled as n = 1-5 and n = 6-16, respectively. Various characterization techniques such as differential scanning calorimetry, wide-angle X-ray diffraction, and polarized light microscopy were used to study their mesomorphic phase behavior. The monomers of five-ring BMJLCPs with relatively long tails showed mesophase behavior. Columnar liquid crystalline phase was observed in both three-ring and five-ring BMJLCPs. Columnar rectangular ({phi}{sub R}) phase was observed in the three-ring system. In the five-ring BMJLCPs, relatively short-tail homologues possess {phi}{sub R} phase, while columnar hexagonal phase was observed in the long-tail samples. The differences in the phase structures were attributed to the 'softness' of the macromolecular BMJLCP column surface.

  13. 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

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

    SciTech Connect

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

    2012-07-01

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

  15. Design Optimization of Nuclear Vapor Thermal Rocket Core - A Thermo-Mechanical Study

    NASA Astrophysics Data System (ADS)

    Keshavmurthy, Shyam P.; Watanabe, Yoichi; Dugan, Edward T.; Diaz, Nils J.

    1994-07-01

    Fuel structural materials for the Nuclear Vapor Thermal Rocket (NVTR) are exposed to very high temperature vapor fuel in the fuel channel and to high temperature but cooler propellant in the coolant channel. This temperature difference leads to thermal stress in the fuel element. There is also a mismatch in the value of coefficients of thermal expansion between the fuel element material and the coating material that could lead to failure of the coating. The stress in the coating and the fuel element material is dependent on the power density of the core and also on the arrangement of fuel and coolant channels. In order to achieve higher power density, the fuel element design has to be optimized to yield lower stress. Analytical studies found that carbon/carbon composite hexagonal fuel elements employing a square lattice arrangement of multiple UF4 fuel and hydrogen coolant channels yield maximum stress intensities well below fuel element materials stress limit.

  16. SAHA-based novel HDAC inhibitor design by core hopping method.

    PubMed

    Zang, Lan-Lan; Wang, Xue-Jiao; Li, Xiao-Bo; Wang, Shu-Qing; Xu, Wei-Ren; Xie, Xian-Bin; Cheng, Xian-Chao; Ma, Huan; Wang, Run-Ling

    2014-11-01

    The catalytic activity of the histone deacetylase (HDAC) is directly relevant to the pathogenesis of cancer, and HDAC inhibitors represented a promising strategy for cancer therapy. SAHA (suberoanilide hydroxamic acid), an effective HDAC inhibitor, is an anti-cancer agent against T-cell lymphoma. However, SAHA has adverse effects such as poor pharmacokinetic properties and severe toxicities in clinical use. In order to identify better HDAC inhibitors, a compound database was established by core hopping of SAHA, which was then docked into HDAC-8 (PDB ID: 1T69) active site to select a number of candidates with higher docking score and better interaction with catalytic zinc ion. Further ADMET prediction was done to give ten compounds. Molecular dynamics simulation of the representative compound 101 was performed to study the stability of HDAC8-inhibitor system. This work provided an approach to design novel high-efficiency HDAC inhibitors with better ADMET properties. PMID:25241128

  17. 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.

  18. 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 materialsLinotype 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.

  19. 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.

  20. 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 are good correlations between the theoretical model and experimental data. Both simulation and experiment confirm large variations of the magnetic bearing characteristics due to air gap growth. Therefore, the gap growth effect should be considered in the magnetic bearing system design. Additionally, the magnetic bearing control system will be compared to other design methods using not only parameter design but H-infinity optimal control and mu synthesis.

  1. Analytical applications for delayed neutrons

    SciTech Connect

    Eccleston, G.W.

    1983-01-01

    Analytical formulations that describe the time dependence of neutron populations in nuclear materials contain delayed-neutron dependent terms. These terms are important because the delayed neutrons, even though their yields in fission are small, permit control of the fission chain reaction process. Analytical applications that use delayed neutrons range from simple problems that can be solved with the point reactor kinetics equations to complex problems that can only be solved with large codes that couple fluid calculations with the neutron dynamics. Reactor safety codes, such as SIMMER, model transients of the entire reactor core using coupled space-time neutronics and comprehensive thermal-fluid dynamics. Nondestructive delayed-neutron assay instruments are designed and modeled using a three-dimensional continuous-energy Monte Carlo code. Calculations on high-burnup spent fuels and other materials that contain a mix of uranium and plutonium isotopes require accurate and complete information on the delayed-neutron periods, yields, and energy spectra. A continuing need exists for delayed-neutron parameters for all the fissioning isotopes.

  2. The design, construction and performance of a variable collimator for epithermal neutron capture therapy beams

    NASA Astrophysics Data System (ADS)

    Riley, K. J.; Binns, P. J.; Ali, S. J.; Harling, O. K.

    2004-05-01

    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.

  3. Neutron Stars

    NASA Technical Reports Server (NTRS)

    Cottam, J.

    2007-01-01

    Neutron stars were discovered almost 40 years ago, and yet many of their most fundamental properties remain mysteries. There have been many attempts to measure the mass and radius of a neutron star and thereby constrain the equation of state of the dense nuclear matter at their cores. These have been complicated by unknown parameters such as the source distance and burning fractions. A clean, straightforward way to access the neutron star parameters is with high-resolution spectroscopy. I will present the results of searches for gravitationally red-shifted absorption lines from the neutron star atmosphere using XMM-Newton and Chandra.

  4. The design of dual emitting cores for green thermally activated delayed fluorescent materials.

    PubMed

    Cho, Yong Joo; Jeon, Sang Kyu; Chin, Byung Doo; Yu, Eunsun; Lee, Jun Yeob

    2015-04-20

    Dual emitting cores for thermally activated delayed fluorescent (TADF) emitters were developed. Relative to the corresponding TADF emitter with a single emitting core the TADF emitter with a dual emitting core, 3,3',5,5'-tetra(carbazol-9-yl)-[1,1'-biphenyl]-2,2',6,6'-tetracarbonitrile, showed enhanced light absorption accompanied by a high photoluminescence quantum yield. The quantum and power efficiencies of the TADF devices were enhanced by the dual emitting cores. PMID:25727757

  5. Design and development of an in-line sputtering system and process development of thin film multilayer neutron supermirrors

    NASA Astrophysics Data System (ADS)

    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.

  6. 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.

  7. 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.

  8. Design and Operating Characteristics of a Lead-lined Hexagonal Neutron-Multiplicity Counter

    SciTech Connect

    Mueller, W.F.; Croft, S.; McElroy, R.D.; Venkataraman, R.; Zhu, H.

    2006-07-01

    A high-efficiency Hexagonal Neutron-Multiplicity Counter (HNMC) has been developed for measurement of non-contact handleable power plant decommissioning and dismantling wastes. The HNMC counter was originally designed for nuclear stockpile safeguards measurements, but the unique challenges of the present application required significant adaptations to the design. The counter consists of six detector modules, each containing two parallel rows of 11 He-3 tubes embedded in high-density polyethylene and arranged in a hexagonal pattern around a central cavity. Graphite reflectors are located above and below the cavity to improve the axial linearity of the response and increase the overall efficiency, which is measured to be 39% for a centrally located Cf-252 source. The HNMC also includes an add-a-source feature by which a known Cf-252 source can be introduced near the bottom of the cavity. The perturbation of the count rate due to the Cf-252 is used to estimate the efficiency loss due to the moderator content of the drum matrix. Unusual to similar large volume multiplicity counters is the addition of a 6.4 mm lead lining on the inner walls of the cavity. This addition results in a suitably low value for the minimum detectable Pu-240-effective mass in drums with as much as 10 mSv.h{sup -1} gamma-ray dose rates when operated in a properly shielded bunker to suppress the background production of neutrons in the lead by cosmic rays. The details of the design and operating characteristics of the HNMC are presented and discussed. (authors)

  9. 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.

  10. 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.

  11. Design, implementation, and testing of a cryogenic loading capability on an engineering neutron diffractometer

    NASA Astrophysics Data System (ADS)

    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.

  12. Conceptual design of a high-frame-rate fast neutron radiography detector

    NASA Astrophysics Data System (ADS)

    Zhang, Fa-qiang; Li, Zheng-hong; Yang, Jian-lun; Guo, Cun; Yang, Hong-qiong; Ye, Fan; Wang, Zhen; Ying, Chun-tong; Liu, Guang-jun

    2007-01-01

    Fast neutron radiography offers means to inspect thick hydrogenous materials because of high penetration depth of fast neutrons. Further more, quasi monoenergetic neutrons is relatively easy to obtain by neutron generators and it is helpful for density inversion of targets, which has many difficulties in flash radiography. In order to investigate dynamic processes, an intense repetitive pulsed neutron source will be used. Efficient detection of fast neutrons is one of the hardest problems for fast neutron imaging detectors. In the system, a scintillating fiber array is employed to obtain a detection efficiency of about 20% for DT neutrons. High-performance ICCDs and large aperture lens are taken into account to increase the conversion efficiency and the collective efficiency. The properties of the detector are charaterized in this paper.

  13. Design and Characterization of a Collimated Neutron Beam User Facility at SUNY Geneseo

    NASA Astrophysics Data System (ADS)

    Krieger, Michael; Padalino, Stephen; Russ, Megan; Polsin, Danae; Bienstock, Mollie; Ellison, Drew; Simone, Angela

    2012-10-01

    The Collimated Neutron Beam (CNB) Facility at SUNY Geneseo provides users an opportunity to perform neutron experiments that require a low neutron background. Neutrons with energies up to 10 MeV are produced by a Plutonium-Beryllium (Pu-Be) source and are collimated to form a well characterized beam. A six foot high, 18 inch thick shielding wall made of water-bricks was built to reduce neutron background in the target area. Neutron and gamma radiation were extensively mapped throughout the facility using a calibrated Bonner sphere, Geiger counter, plastic scintillator and an HPGe detector. Potential uses for the CNB include neutron activation, time-of-flight, attenuation and neutron detector calibration experiments. A detailed description and layout of the facility will be displayed on the poster. Funded in part by a grant from the DOE through the Laboratory for Laser Energetics.

  14. Designing of the 14 MeV neutron moderator for BNCT

    NASA Astrophysics Data System (ADS)

    Cheng, Dao-Wen; Lu, Jing-Bin; Yang, Dong; Liu, Yu-Min; Wang, Hui-Dong; Ma, Ke-Yan

    2012-09-01

    In boron neutron capture therapy (BNCT), the ratio of the fast neutron flux to the neutron flux in the tumor (RFNT) must be less than 3%. If a D-T neutron generator is used in BNCT, the 14 MeV neutron moderator must be optimized to reduce the RFNT. Based on the neutron moderation theory and the simulation results, tungsten, lead and diamond were used to moderate the 14 MeV neutrons. Satisfying RFNT of less than 3%, the maximum neutron flux in the tumor was achieved with a three-layer moderator comprised of a 3 cm thick tungsten layer, a 14 cm thick lead layer and a 21 cm thick diamond layer.

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

    SciTech Connect

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

    1997-12-01

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

  16. Design of a subcritical multiplying low-enriched uranium externally driven neutron assembly at the Los Alamos neutron science center

    SciTech Connect

    Gavron, Victor

    2008-01-01

    The Sandia Pulsed Reactor (SPR) has had a long history of supplying short intense neutron bursts to test the survivability of critical weapons components. The most recent version, SPR III, was resurrected in 2005 in support of critical survivability testing for W-76 Life Extension Program and other testing for the Qualification Alternatives to SPR program. SPR has now been decommissioned. The SPR had a cylindrical shape, with an internal cylindrical cavity where samples were placed for testing. The approximate dimensions of the cavity were 16.5 cm radius, and 38.1 cm high. The uniformity of the fluence over the entire volume was {+-} 25%. The nominal fluence was 5.4 {center_dot} 10{sup 14} n/cm{sup 2}, 1 MeV Si equivalent (Si equivalent implies that the neutron spectrum, at energies other than 1 MeV, is weighted by its relative damage potential in silicon), with a pulse duration of 55 microseconds, FWHM. We propose a new facility as a replacement to the SPR pulsed neutron capabilities, utilizing the proton linear accelerator at the Los Alamos Neutron Science Center (LANSCE). Currently, LANSCE provides neutron beams generated by the 800 MeV proton beam, over 13 decades of energy, to five different facilities. LANSCE is in the process of being refurbished; the refurbishment project (LANSCE-R) is funded and due to be completed by 2014. In parallel to the refurbishment, the experimental user program will continue and provide 3000 hours of beam per year. LANSCE is also considering upgrade options in the framework of MaRIE (Materials and Radiation Interactions in Extreme). MaRIE is planned to be the new signature facility for Los Alamos for the foreseeable future. Hence, they expect LANSCE to continue to operate and provide beams for decades t come. They propose to use the 800 MeV proton beam, extracted from the Proton Storage Ring (PSR), to initiate a neutron pulse in a sub-critical multiplying assembly, fabricated from Low-Enriched Uranium (LEU). The reason for using LEU (rather than highly-enriched uranium) is that it is not considered an attractive material that requires a significant investment in security and materials protection.

  17. 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).

  18. 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.

  19. Enhancing the Practicum Experience for Pre-service Chemistry Teachers Through Collaborative CoRe Design with Mentor Teachers

    NASA Astrophysics Data System (ADS)

    Hume, Anne; Berry, Amanda

    2013-01-01

    This paper reports findings from an ongoing study exploring how the Content Representation (CoRe) design can be used as a tool to help chemistry student teachers begin acquiring the professional knowledge required to become expert chemistry teachers. Phase 2 of the study, reported in this paper, investigated how collaboration with school-based mentors (associate teachers) on teaching practice (practicum) might impact on this process and student teachers' development of their pedagogical content knowledge (PCK). The collaboration involved identifying and discussing pedagogical issues related to a practicum-teaching topic using a student teacher's draft CoRe as a starting point and ongoing focus for the professional dialogue. Practicum offered an opportunity for aspects of student teachers' PCK, as embodied in their draft CoRes, to be explored and expanded upon in classroom programmes with the support and input of associate teachers. The findings were influenced by different contextual factors; however, the student teachers found their CoRes to be very useful frameworks for engaging in focussed professional dialogue with their teaching mentors. They valued the expertise, currency of knowledge and mentoring of their associates and reported positively about the contribution this support made to their PCK development via the CoRe design process and the transformation of the CoRe into classroom teaching.

  20. 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.