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Advanced neutron source final preconceptual reference core design
Microsoft Academic Search
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
G. L. Copeland; W. R. Gambill; R. M. Harrington; J. A. Johnson; F. J. Peretz; H. Reutler; J. M. Ryskamp; D. L. Selby; C. D. West; G. L. Yoder
Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system.
National Technical Information Service (NTIS)
The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assemb...
C. H. Kim S. C. Kim T. K. Kim J. Y. Cho H. C. Lee
McCARD for Neutronics Design and Analysis of Research Reactor Cores
NASA Astrophysics Data System (ADS)
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.
Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang
Optimizing a three-element core design for the Advanced Neutron Source Reactor
Source of neutrons in the proposed Advanced Neutron Source facility is a multipurpose research reactor providing 5-10 times the flux, for neutron beams, of the best existing facilities. Baseline design for the reactor core, based on the ``no new inventions`` rule, was an assembly of two annular fuel elements similar to those used in the Oak Ridge and Grenoble high
Detailed Core Design and Flow Coolant Conditions for Neutron Flux Maximization in Research Reactors
Following the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, we have already developed an asymmetric compact cylindrical core with an inner and outer reflector. The goal was to maximize the maximum thermal flux to power ratio, which is a desirable characteristic of a modern research reactor. This design, for a 10 MW power, was
F. E. Teruel; Rizwan Uddin
PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle
SciTech Connect
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)
Bi, G.; Liu, C.; Si, S. [Shanghai Nuclear Engineering Research and Design Inst., No. 29, Hongcao Road, Shanghai, 200233 (China)
The Pebble Bed Modular Reactor (PBMR) introduces several challenges for core neutronic methods. The particulate fuel is highly heterogeneous with a random distribution within the fuel pebbles and requires unique methods to calculate the effects of fuel resonance self shielding. In addition, the flow of fuel through the core is specific to the PBMR, again requiring specialised methods to model
Optimizing a three-element core design for the advanced neutron source reactor
The source of neutrons in the proposed Advanced Neutron Source facility is a multipurpose research reactor providing 5 to 10 times the flux, for neutron beams, of the best existing facilities. The project team constrained the design with the {open_quotes}no new inventions rule,{close_quotes} which states that the design should not rely on the development of new technology to meet the
Neutronics methods, models, and applications at the Idaho National Engineering Laboratory for the advanced neutron source reactor three-element core design
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.
Wemple, C.A.; Schnitzler, B.G.; Ryskamp, J.M. [Idaho National Engineering Lab., Idaho Falls, ID (United States)
Three-dimensional Core Design of High Temperature Supercritical-Pressure Light Water Reactor with Neutronic and Thermal-Hydraulic Coupling
The equilibrium core of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H) is designed by three-dimensional neutronic and thermal-hydraulic coupled core calculations. The average coolant core outlet temperature of 500�C is accurately evaluated for the first time in the development of the SCLWR-H.The average coolant core outlet temperature is one of the key parameters, which must be accurately determined in
Akifumi YAMAJI; Yoshiaki OKA; Seiichi KOSHIZUKA
Gfr Core Neutronics Studies at CEA
The Gas cooled Fast Reactor (GFR) is a high priority in the CEA R&D program on Future Nuclear Energy Systems. After preliminary neutronics and thermo-aerolic studies, a first He-cooled 2400MWth core design based on a series of carbide CERCER plates arranged in an hexagonal wrapper were selected. Although GFR subassembly and core design studies are still at an early stage of development, it is nonetheless possible to identify a number of nuclear data needs that could have some impact on the actual design: new materials, decay heat contributors�.
Bosq, J. C.; Brun-Magaud, V.; Rimpault, G.; Tommasi, J.; Conti, A.; Garnier, J. C.
RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design
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.
Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.
Core Vessel Insert Handling Robot for the Spallation Neutron Source
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.
Graves, Van B [ORNL; Dayton, Michael J [ORNL
Modeling and design of a new core-moderator assembly and neutron beam ports for the Penn State Breazeale Nuclear Reactor (PSBR)
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
Ucar, Dundar
Automated Core Design
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.
Kobayashi, Yoko; Aiyoshi, Eitaro
Annular core research reactor high flux neutron radiography facility
Sandia National Laboratories (SNL) has been performing neutron radiography since 1964. The radiography facilities have evolved from an aperture in a radiation exposure room in the now retired Sandia Engineering Reactor to a divergent collimator radiography facility adjacent to the core of the Annular Core Research Reactor (ACRR). The maximum thermal neutron flux achieved in these facilities has been limited to approximately 1 {times} 10{sup 7} n-cm{sup -2}-s{sup -1}. In order to perform high-resolution, real-time neutron radiography for transient events, higher neutron fluxes are required. In response to this need, Sandia is designing a new high-flux neutron radiography facility for the ACRR. The new facility uses the central irradiation cavity of the ACRR and consists of a collimator assembly, reactor control system, an experiment support structure, and an imaging system. This new facility is described in this paper. 2 refs., 2 figs.
McCrory, F.M.; Kelly, J.G.; Vernon, M.E.; Tichenor, D.A.
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.
Melatos, A., E-mail: email@example.com [School of Physics, University of Melbourne, Parkville, VIC 3010 (Australia)
DANDE: a linked code system for core neutronics\\/depletion analysis
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
R. J. LaBauve; T. R. England; D. C. George; R. E. MacFarlane; W. B. Wilson
Air Core Pulse Transformer Design.
Cylindrical air core pulse transformers capable of passing high voltage/energy pulse waveforms with high efficiency and low distortion require a much more delicate design balance of physical dimensions and electrical parameters than iron or ferrite core u...
J. P. O'Loughlin J. D. Sidler G. J. Rohwein
In-Core Neutron Flux Measurements at PARR Using Self Powered Neutron Detectors.
The report describes experimental reactor physics measurements at PARR using the in-core neutron detectors. Rhodium self powered neutron detectors (SPND) were used in the PARR core and several measurements were made aimed at detector calibration, response...
A. Hassain S. A. Ansari
Dynamics of dissipative multifluid neutron star cores
We present a Newtonian multifluid formalism for superfluid neutron star cores, focusing on the additional dissipative terms which arise when one takes into account the individual dynamical degrees of freedom associated with the coupled �fluids.� The problem is of direct astrophysical interest as the nature of the dissipative terms can have significant impact on the damping of the various oscillation modes of the star and the associated gravitational-wave signatures. A particularly interesting application concerns the gravitational-wave driven instability of f- and r-modes. We apply the developed formalism to two specific three-fluid systems: (i) a hyperon core in which both ? and ?- hyperons are present and (ii) a core of deconfined quarks in the color-flavor-locked phase in which a population of neutral K0 kaons is present. The formalism is, however, general and can be applied to other problems in neutron-star dynamics (such as the effect of thermal excitations close to the superfluid transition temperature) as well as laboratory multifluid systems.
Haskell, B.; Andersson, N.; Comer, G. L.
Design of an LEU core for the MIT reactor
A design of the MIT Reactor core using monolithic U-7Mo LEU fuel has been developed with the goal of maintaining thermal and fast neutron fluxes as well as increasing the flexibility for meeting the needs of in-core experiments. An optimum core was sought by varying the core materials, and fuel plate numbers and thicknesses, but maintaining the outside dimensions of a fuel element. A full-core model of the MITR by the Monte-Carlo transport code MCNP was used to calculate the neutron fluxes, reactivity and neutron spectrum available for experiments. The optimum reactor design consisted of the use of half-sized fuel elements made up of nine U-7Mo LEU fuel plates of 0.55 mm thickness with 0.25 mm finned aluminum cladding. This design also utilized solid beryllium fuel elements (dummies) with boron fixed absorbers or solid lead dummies, depending on the in-core experiment flux and spectrum needs. Because the new core design contains twice the amount of 235 U as does the existing HEU core, and produces much more Pu, its fuel cycle length is twice as long at the same power level. Preliminary thermal-hydraulic and neutronic safety evaluations indicate superior performance to the current HEU fuel. (authors)
Newton, T. [Massachusetts Inst. of Technology, Nuclear Reactor Laboratory, 138 Albany St., Cambridge, MA 02139 (United States); Kazimi, M.; Pilat, E. [Nuclear Science and Engineering Dept., 77 Massachusetts Ave., Cambridge, MA 02139 (United States)
Air core pulse transformer design
Cylindrical-air-core pulse transformers capable of passing high-voltage\\/high-energy pulse waveforms with high efficiency and low distortion require a much more delicate design balance of physical dimensions and electrical parameters than iron or ferrite core units. Special computer codes were written to evaluate their performance. The analysis includes calculation of the self and mutual inductances as determined by the dimensions and insulation
J. P. O'Loughlin; J. D. Sidler; Gerry J. Rohwein
HFIR cold neutron source moderator vessel design analysis
A cold neutron source capsule made of aluminum alloy is to be installed and located at the tip of one of the neutron beam tubes of the High Flux Isotope Reactor. Cold hydrogen liquid of temperature approximately 20 degree Kelvin and 15 bars pressure is designed to flow through the aluminum capsule that serves to chill and to moderate the incoming neutrons produced from the reactor core. The cold and low energy neutrons thus produced will be used as cold neutron sources for the diffraction experiments. The structural design calculation for the aluminum capsule is reported in this paper.
Chang, S.J.
Heat transfer and core neutronics considerations of the heat pipe cooled thermionic reactor
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
W. R. Determan; Brian Lewis
Sandia National Laboratories (SNL) has been performing neutron radiography since 1964. The radiography facilities have evolved from an aperture in a radiation exposure room in the now retired Sandia Engineering Reactor to a divergent collimator radiography facility adjacent to the core of the Annular Core Research Reactor (ACRR). The maximum thermal neutron flux achieved in these facilities has been limited
F. M. McCrory; J. G. Kelly; M. E. Vernon; D. A. Tichenor
Measurements of the thermal neutrons flux near the EAS core
The characteristics of the thermal neutrons flux have been measured near the EAS core at the "Carpet-2" EAS array. The thermal neutron detectors were placed on the floor of the tunnel of the Muon Detector (MD) and showers with a core near the MD were selected. Thermal neutrons multiplicity spectrum has been obtained for these showers. Measurements of the lateral distribution function of thermal neutrons at distances of 1-16 m from the shower axis have been performed. The mean number of the recorded thermal neutrons as a function of the number of hadrons crossing the MD has been measured.
Dzhappuev, D. D.; Volchenko, V. I.; Kudzhaev, A. U.; Mikhailova, O. I.; Petkov, V. B.; Stenkin, Yu V.; Tsyabuk, A. L.
Design of neutron detectors utilizing luminescent glass
The impetus for the development of new neutron detector designs has come from the worldwide commissioning of neutron spallation sources. The feature of these sources is their broadband neutron emission and the consequent potential which they provide for neutron investigation of all types of structures in materials ranging from biological molecules to steels. The primary advantage of conventional neutron detectors
A. R. Spowart
Fuel, Core Design and Subchannel Analysis of a Superfast Reactor
A compact supercritical water-cooled fast reactor (superfast reactor) core with a power of 700MWe is designed by using a three-dimensional neutronics thermal-hydraulic coupled method. The core consists of 126 seed assemblies and 73 blanket assemblies. In the seed assemblies, 251 fuel rods, consisting of MOX pellets, stainless steel (SUS304) cladding, and fission gas plenum are arranged into a tight triangle
Liangzhi CAO; Yoshiaki OKA; Yuki ISHIWATARI; Zhi SHANG
FabScalar: Automating Superscalar Core Design
Providing multiple superscalar core types on a chip, each tailored to different classes of instruction-level behavior, is an exciting direction for increasing processor performance and energy efficiency. Unfortunately, processor design and verification effort increases with each additional core type, limiting the microarchitectural diversity that can be practically implemented. FabScalar aims to automate superscalar core design, opening up processor design to
Niket K. Choudhary; Salil V. Wadhavkar; Tanmay A. Shah; Hiran Mayukh; Jayneel Gandhi; Brandon H. Dwiel; Sandeep Navada; Hashem H. Najaf-abadi; Eric Rotenberg
Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core
PubMed Central
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.
Lashkari, A.; Khalafi, H.; Kazeminejad, H.
Conceptual study of advanced PWR core design.
The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core c...
H. K. Joo K. K. Kim M. H. Chang Y. I. Kim Y. J. Kim
Optimization of IEC grid design for maximum neutron production
Two different, complementary approaches were taken to determine the effects of an Inertial Electrostatic Confinement (IEC) grid`s design on the neutron production rate of the device. A semi-empirical formula developed from experimental data predicts the neutron yield of an IEC device, given the chamber size, grid radius and transparency, and operating voltage and current. Results from the IXL computer program support some of the scalings found in the semi-empirical formula. A second formula was also developed that predicts the neutron yield of an IEC device using grid design parameters and the ion core radius. The SIMION computer program was used to calculate the ion core radius. These formulas are useful tools for designing grids that will maximize the neutron yield for IEC devices. 7 refs., 9 figs.
Miley, G.H.; DeMora, J.; Stubbers, R.; Tzonev, I.V. [Univ. of Illinois, Urbana, IL (United States); Anderl, R.A. [Lockheed Martin Idaho Technologies Company, Idaho Falls, ID (United States); Nadler, J.H. [Department of Energy, Idaho Falls, ID (United States); Nebel, R. [Los Alamos National Lab., NM (United States)
NASA Technical Reports Server (NTRS)
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.
Determan, W. R.; Lewis, Brian
Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor
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
K. C. Schulz; G. T. Yahr
Core design investigation for a SUPERSTAR small modular lead-cooled fast reactor demonstrator
In this paper a preconceptual neutronics design study for a SUstainable Proliferation-resistance Enhanced Refined Secure Transportable Autonomous Reactor (SUPERSTAR) demonstrator is presented. The main goal of achieving the highest realistic power level limited by natural circulation and transportability, while providing energy security and proliferation resistance thanks to a long core lifetime design has been satisfactorily attained. A preliminary core configuration
S. Bortot; A. Moisseytsev; J. J. Sienicki; Carlo Artioli
Neutronics of a Mixed-Flow Gas-Core Reactor.
The study was made to investigate the neutronic feasibility of a mixed-flow gas-core reactor. Three reactor concepts were studied: four- and seven-cell radial reactors and a seven-cell scallop reactor. The reactors were fueled with UF sub 6 (either U-233 ...
P. D. Soran G. E. Hansen
Neutronics properties of JMTR LEU core
The Japan materials testing reactor (JMTR) is a 50-MW light-water-moderated and -cooled tank-type reactor using engineering test reactor-type fuel. Core conversion from medium-enrichment uranium (MEU) aluminide fuel to low-enrichment uranium (LEU) silicide fuel of the JMTR is scheduled for 1993. The LEU silicide fuel element at 4.8 g U\\/cm[sup 3] with cadmium wires in the sideplates as burnable absorbers has
Yoshiharu Nagaoka; Minoru Saito; Yoshiaki Futamura
AHTR Mechanical, Structural, And Neutronic Preconceptual Design
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
Varma, Venugopal Koikal [ORNL; Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Bradley, Eric Craig [ORNL; Ilas, Dan [ORNL; Qualls, A L [ORNL; Zaharia, Nathaniel M [ORNL
Concept Design of Advanced Marine Reactor (1), Core Design. Design Study for Optimized Core.
We started the design study of the concept design of the Advanced Marine Reactor from FY 1987, and we researched and studied as for the optimization of the core and the components of the 100 MWt reactor plant that main research thema was minituarization a...
N. Ambo T. Yokomura
Heavy water breeder conceptual core design
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
Sensitivity of detecting in-core vibrations and boiling in pressurized water reactors using ex-core neutron detectors
Neutron transport and diffusion theory space- and energy-dependent reactor kinetics calculations were performed in the frequency domain to determine the sensitivity of an ex-core neutron detector to in-core vibrations and coolant boiling in a PWR. Results of these calculations indicate that the ex-core detectors are sensitive to neutron sources, to vibrations, and to boiling occurring over large regions of the
F. J. Sweeney; J. P. A. Renier
Core length testable reactor concept neutronic analysis
Development work on thermionic reactor systems has been ongoing in the US since the early 1950s. While significant successes were achieved, progress has been hampered by frequent changes in direction and funding instabilities (as has been true for many high technology initiatives). The recent Air Force thermionics initiative (1991) represents the latest in thermionics reactor development in the US. This Air Force initiative called for the development of thermionics reactors with the output power of about 40 kWe, and which incorporated the features of testability, fabricability, low development cost, high level of safety and reliability, and survivability. Several concepts were analyzed to define a design that would meet all the requirements set forth by the Air Force. This report describes the methodology used, the different designs analyzed and reasons for the evolution of the design, and presents the results for the different concepts.
Hanan, N.A.; Bhattacharyya, S.K.
NASA'S Chandra Finds Superfluid in Neutron Star's Core
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
DANDE: a linked code system for core neutronics/depletion analysis
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 cource 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 illustrated in this report by two sample problems. 25 refs.
LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.
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.
The impetus for the development of new neutron detector designs has come from the worldwide commissioning of neutron spallation sources. The feature of these sources is their broadband neutron emission and the consequent potential which they provide for neutron investigation of all types of structures in materials ranging from biological molecules to steels. The primary advantage of conventional neutron detectors filled with boron fluoride or helium 3 gas is their intrinsically low background count; their major drawback is a slow response time, high cost, and relative inflexibility in design for coping with large areas or complex geometry detection. This paper presents design strategies for glass neutron scintillators. These lithium silicate glasses, doped with Cesium ions, have high detection efficiencies for neutrons, fast decay times, total flexibility in shape and emission centered at 400 nm, and can be well matched to PM and image intensifier photocathodes. The neutron counting applications require one or a combination of the following systems: for neutron flux counting, a slice of glass of appropriate thickness coupled to a PM/photodiode feeding an operational amplifier and scaler; with no appreciable x-ray or other interfering radiation present, a slice of glass oc appropriate thickness coupled to a PM, a Schmitt trigger, and a window discriminator; for low signal-to-noise ratio, a sandwich layer construction.
Spowart, A.R.
Neutronic design of the ITER radial neutron camera
This paper summarizes the work, performed in the frame of various EFDA contracts during 2004�2005, on the design review and upgrade of the ITER radial neutron camera (RNC). The RNC, which should provide information on the spatial distribution and energy spectrum of the neutron emission, consists of an ex-vessel system (fan-like collimator with 12�3 lines of sights) and an in-vessel
L. Petrizzi; R. Barnsley; L. Bertalot; B. Esposito; H. Haskell; E. Mainardi; D. Marocco; S. Podda; C. Walker; S. Villari
Study of compact fast reactor core designs
A study is conducted to investigate conceptual liquid-metal reactor (LMR) core concepts, employing some unconventional design features for improved economics and safety. The unconventional design elements are used to supplement the conventional measures, which alone have apparently not led to an attractive LMR design for the 21st century. Better economics are obtained through simplicity and compactness of the core design. For simplicity, internal scattered blankets are omitted. Core compactness is achieved by maximum power flattening, resulting from axial and radial enrichment zones along with axial and radial (BeO) reflectors. To further enhance core compactness, the in-core compactness, the in-core control rods are replaced by reflector controls. For improved safety, the general objective is to reduce both coolant-void and burnup reactivities. However, even with the use of a wide spectrum of unconventional design features, such as burnable poisons, peripheral reflectors, and inner moderating regions, it is not possible to overcome the fact that both coolant-void and burnup reactivities cannot be reduced simultaneously to desirably low levels. The only resolution of this dilemma appears to be minimize coolant-void reactivity and to manage the burnup reactivities. However, even with the use of a wide spectrum of unconventional design features, such as burnable poisons, peripheral reflectors, and inner moderating regions, it is not possible to overcome the fact that both coolant-void and burnup reactivities cannot be reduced simultaneously to desirably low levels. The only resolution of this dilemma appears to be to minimize coolant-void reactivity and to manage the burnup reactivity losses, such that an accidental insertion of significant amounts of reactivity is mechanically not possible. A conceptual design with these characteristics is described.
Hamid, T.; Ott, K.O. (Purdue Univ., West Lafayette, IN (United States))
Neutron tube design study for boron neutron capture therapy application
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.
Verbeke, J.M.; Lee, Y.; Leung, K.N.; Vujic, J.; Williams, M.D.; Wu, L.K.; Zahir, N.
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.
Contribution of fuel vibrations to ex-core neutron noise during the first and second fuel cycles of the Sequoyah-1 pressurized water reactor
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.
Sweeney, F.J.; March-Leuba, J.; Smith, C.M.
Design of neutron beams for neutron capture therapy using a 300-kW slab TRIGA reactor
A design for a slab reactor to produce an epithermal neutron beam and a thermal neutron beam for use in neutron capture therapy (NCT) is described. A thin reactor with two large-area faces, a ``slab`` reactor, was planned using eighty-six 20% enriched TRIGA fuel elements and four B{sub 4}C control rods. Two neutron beams were designed: an epithermal neutron beam from one face and a thermal neutron beam from the other. The planned facility, based on this slab-reactor core with a maximum operating power of 300 kW, will provide an epithermal neutron beam of 1.8 {times} 10{sup 9} n{sub epi}/cm{sup 2}{center_dot}s intensity with low contamination by fast neutrons and gamma rays and a thermal neutron beam of 9.0 {times} 10{sup 9}n{sub th}/cm{sup 2}{center_dot}s intensity with low fast-neutron dose and gamma dose. Both neutron beams will be forward directed. Each beam can be turned on and off independently through its individual shutter. A complete NCT treatment using the designed epithermal or thermal neutron beam would take 30 or 20 min, respectively, under the condition of assuming 10{mu}g {sup 10}B/g in the blood. Such exposure times should be sufficiently short to maintain near-optimal target (e.g., {sup 10}B, {sup 157}Gd, and {sup 235}U) distribution in tumor versus normal tissues throughout the irradiation. With a low operating power of 300 kW, the heat generated in the core can be removed by natural convection through a pool of light water. The proposed design in this study could be constructed for a dedicated clinical NCT facility that would operate very safely.
Liu, H.B. [Brookhaven National Lab., Upton, NY (United States). Medical Dept.
Transport coefficients of nuclear matter in neutron star cores
We calculate thermal conductivity and shear viscosity of nucleons in the dense nuclear matter of neutron star cores in the nonrelativistic Brueckner-Hartree-Fock framework. The nucleon-nucleon interaction is described by the Argonne v18 potential with addition of the Urbana IX effective three-body forces. We find that this three-body force leads to a decrease of the kinetic coefficients with respect to the two-body case. The results of calculations are compared with electron and muon transport coefficients as well as with the results of other authors.
Shternin, P. S.; Baldo, M.; Haensel, P.
Finite Element Analysis of Neutron Diffusion Equations and Reactor Core Fuel Management.
The mathematical formulation for operation and reactor core fuel management was studied. Using time discrimination, the eigenvalue problem of neutron diffusion equations are governed by reactivity gain in the reactor. From the distribution of neutron flux...
A. Huang Q. Huang Y. Tang Z. Tu
Design of a silicon carbide neutron sensor
The radiation detectors based on the third era semiconductor material silicon carbide (SiC) with wide energy band gap are the most promising ionizing radiation detectors in high temperature and harsh radiation environment. This paper illustrated several important advantages of the SiC neutron detector and described its fabrication and detection principle briefly. Furthermore, we evaluated the neutron sensor's measurement performance when detecting the 2.5MeV and 14MeV neutrons under different conditions of sensor's active layer thickness and polyethylene converter film thickness based on MCNP simulation. According to the results of simulation, the sensor's optimal configuration was designed. For the sensor whose radius and depletion layer thickness are 3mm and 30?m respectively, the detection efficiency can reach 3.16�10-18 coulomb per neutron (c/n) and 1.80�10-17 c/n for 2.5MeV and 14MeV neutrons respectively. When adding a polyethylene converter film of 90?m thickness to the above sensor, the detection efficiency to 2.5MeV neutron will be 3.7 times that without neutron converter film; and with the converter film of 2mm thickness, the detection efficiency to 14MeV neutron will be improved by 246%.
Hu, Qing-Qing; Yang, Jun; Liu, GuoFu; Luo, Xiao-liang
Combined analysis of neutron and photon flux measurements for the Jules Horowitz reactor core mapping
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)
Fourmentel, D.; Villard, J. F.; Lyoussi, A. [DEN Reactor Studies Dept., French Nuclear Energy and Alternative Energies Commission, CEA Cadarache, 13108 Saint Paul-Lez-Durance (France); Reynard-Carette, C. [Laboratoire Chimie Provence LCP UMR 6264, Univ. of Provence, Centre St. Jerome, 13397 Marseille Cedex 20 (France); Bignan, G.; Chauvin, J. P.; Gonnier, C.; Guimbal, P.; Malo, J. Y. [DEN Reactor Studies Dept., French Nuclear Energy and Alternative Energies Commission, CEA Cadarache, 13108 Saint Paul-Lez-Durance (France); Carette, M.; Janulyte, A.; Merroun, O.; Brun, J.; Zerega, Y.; Andre, J. [Laboratoire Chimie Provence LCP UMR 6264, Univ. of Provence, Centre St. Jerome, 13397 Marseille Cedex 20 (France)
Experimental cores for benchmark experiments of effective delayed neutron fraction ? eff at FCA
Three cores were built for benchmark experiments of the effective delayed neutron fraction ?eff at the FCA facility of the Japan Atomic Energy Research Institute: the XIX-1 core fueled with highly enriched uranium; the XIX-2 core fueled with plutonium and natural uranium; the XIX-3 core fueled with the plutonium. Measurements were carried out for the criticality, the central fission rates,
Takeshi Sakurai; Shigeaki Okajima; Masaki Andoh; Toshitaka Osugi
Accurate simulation of thermal neutron filter effects in the design of research reactor beam applications
Thermal neutron filters are routinely used for spectral shaping in neutron beam applications. In many situations, the neutron source is the core of a research reactor that faces several beam ports. Consequently, the neutrons leaking into the beam port will have an energy spectrum that extends over several decades from the low energy (thermal) range to the high energy (fast) range that is typical of prompt fission neutrons. At the PULSTAR reactor of North Carolina State Univ., sapphire and bismuth filters are used within the beam ports to reduce the epi-thermal and fast neutron components and gamma-rays. The main objective is to produce thermal neutron beams with a high neutron-to-gamma ratio. The thermal beams are used to drive facilities for neutron imaging and neutron diffraction. To assure that these facilities are accurately designed, thermal neutron scattering cross section libraries are generated for sapphire and bismuth at 300 K. The libraries do not account for coherent elastic (i.e., 'Bragg') scattering, which is equivalent to assuming the use of a single crystal filter with a preferential orientation relative to the neutron beam. MCNP Monte Carlo simulations show that, using the created libraries, the phenomenon of neutron filtration can be captured and diagnosed to ensure accurate beam and facility designs. (authors)
Hawari, A. I.; Al-Qasir, I. I.; Mishra, K. K. [North Carolina State Univ., Dept. of Nuclear Engineering, Raleigh, NC 27695 (United States)
Neutronic and thermal design considerations for heat-pipe reactors
SABRE (Space-Arena Baseline Reactor) is a 100-kW/sub e/, heat-pipe-cooled, beryllium-reflected, fast reactor that produces heat at a temperature of 1500/sup 0/K and radiatively transmits it to high-temperature thermoelectric (TE) conversion elements. The use of heat pipes for core heat removal eliminates single-point failure mechanisms in the reactor cooling system, and provides minimal temperature drop radiative coupling to the TE array, as well as automatic, self-actuating removal of reactor afterheat. The question of how the failure of a fuel module heat pipe will affect neighboring fuel modules in the core is discussed, as is fission density peaking that occurs at the core/reflector interface. Results of neutronic calculations of the control margin available are described. Another issue that is addressed is that of helium generation in the heat pipes from neutron reactions in the core with the heat pipe fluid. Finally, the growth potential of the SABRE design to much higher powers is examined.
Ranken, W.A.; Koenig, D.R.
Advanced Neutron Sources: Plant Design Requirements
The Advanced Neutron Source (ANS) is a new, world class facility for research using hot, thermal, cold, and ultra-cold neutrons. At the heart of the facility is a 350-MW{sub th}, heavy water cooled and moderated reactor. The reactor is housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides fans out into a large guide hall, housing about 30 neutron research stations. Office, laboratory, and shop facilities are included to provide a complete users facility. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory at the end of the decade. This Plant Design Requirements document defines the plant-level requirements for the design, construction, and operation of the ANS. This document also defines and provides input to the individual System Design Description (SDD) documents. Together, this Plant Design Requirements document and the set of SDD documents will define and control the baseline configuration of the ANS.
Not Available
Stellar encounters involving neutron stars in globular cluster cores
Encounters between a 1.4 solar mass neutron star and a 0.8 solar mass red giant (RG) and between a 1.4 solar mass neutron star (NS) and an 0.8 solar mass main-sequence (MS) star have been successfully simulated. In the case of encounters involving an RG, bound systems are produced when the separation at periastron passage R(MIN) is less than about 2.5 R(RG). At least 70 percent of these bound systems are composed of the RG core and NS forming a binary engulfed in a common envelope of what remains of the former RG envelope. Once the envelope is ejected, a tight white dwarf-NS binary remains. For MS stars, encounters with NSs will produce bound systems when R(MIN) is less than about 3.5 R(MS). Some 50 percent of these systems will be single objects with the NS engulfed in a thick disk of gas almost as massive as the original MS star. The ultimate fate of such systems is unclear.
Davies, M. B.; Benz, W.; Hills, J. G.
Design and proposed utilization of the Sandia Annular Core Research Reactor (ACRR)
The Sandia ACRR became operational in 1978 and currently serves as the major in-pile fast reactor safety test facility for the US Nuclear Regulatory Commission. The ACRR is an upgrade of the Annular Core Pulse Reactor (ACPR) with the installation of a new flexible control system and a core of uniquely designed BeO-UO fuel elements for increasing the neutron fluence
J. V. Walker; J. A. Reuscher; P. S. Pickard
Advanced Neutron Source: Plant Design Requirements
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.
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.).
Energy Efficient Engine core design and performance report
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.
Stearns, E. Marshall
Using ex-core neutron detectors to estimate fuel quantities in the reactor vessel lower head
During the accident at Three Mile Island Unit 2 (TMI-2), a significant mass of core debris slumped to the bottom head of the reactor vessel. Defueling activities caused more core debris to relocate to the lower head region. The variations in the ex-core neutron detector, or source range monitor (SRM), readings gave evidence of this effect of defueling activity. Between
R. Rainisch; V. Fricke
Overview of core designs and requirements/criteria for core restraint systems
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.
Sutherland, W.H.
Neutron flux measurements in the side-core region of Hunterston B advanced gas-cooled reactor
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)
Allen, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Shaw, S.E. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom); Huggon, A.P.; Steadman, R.J.; Thornton, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Whiley, G.S. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom)
Global shielding analysis for the three-element core advanced neutron source reactor under normal operating conditions
Two-dimensional discrete ordinates radiation transport calculations were performed for a model of the three-element core Advanced Neutron Source reactor design under normal operating conditions. The core consists of two concentric upper elements and a lower element radially centered in the annulus between the upper elements. The initial radiation transport calculations were performed with the DORT two-dimensional discrete ordinates radiation transport code using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub 6} quadrature, and a P{sub 1} Legendre polynomial expansion of the cross sections to determine the fission neutron source distribution in the core fuel elements. These calculations were limited to neutron groups only. The final radiation transport calculations, also performed with DORT using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub l0} quadrature, and a P{sub 3} Legendre polynomial expansion of the cross sections, produced neutron and gamma-ray fluxes over the full extent of the geometry model. Responses (or activities) at various locations in the model were then obtained by folding the appropriate response functions with the fluxes at those locations. Some comparisons were made with VENTURE-calculated (diffusion theory) 20-group neutron fluxes that were summed into four broad groups. Tne results were in reasonably good agreement when the effects of photoneutrons were not included, thus verifying the physics model upon which the shielding model was based. Photoneutrons increased the fast-neutron flux levels deep within the D{sub 2}0 several orders of magnitude. Results are presented as tables of activity values for selected radial and axial traverses, plots of the radial and axial traverse data, and activity contours superimposed on the calculational geometry model.
Slater, C.O.; Bucholz, J.A.
Learn from the Core--Design from the Core
ERIC Educational Resources Information Center
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�
Ockerse, Thomas
Core refueling subsystem design description. Revision 1
The Core Refueling Subsystem of the Fuel Handling and Storage System provides the mechanisms and tools necessary for the removal and replacement of the hexagonal elements which comprise the reactor core. The Core Refueling Subsystem is not "safety-related." The Core Refueling Subsystem equipment is used to prepare the plant for element removal and replacement, install the machines which handle the elements, maintain control of air inleakage and radiation release, transport the elements between the core and storage, and control the automatic and manual operations of the machines. Much of the element handling is performed inside the vessel, and the entire exchange of elements between storage and core is performed with the elements in a helium atmosphere. The core refueling operations are conducted with the reactor module shutdown and the primary coolant pressure slightly subatmospheric. The subsystem is capable of accomplishing the refueling in a reliable manner commensurate with the plant availability requirements.
Anderson, J.K.; Harvey, E.C.
As part of the experimental programme on the first plutonium loading of ZENITH (Core 7) a series of measurements was carried out with the neutron chopper on a beam emerging from the core centre. The general experimental programme on the two ZENITH plutonium cores has been covered elsewhere. Core 7 had a carbon\\/Pu239 atomic ratio of 2666 and a steel\\/Pu239
F. R. Barclay; I. R. Cameron; H. H. W. Pitcher; C. R. Symons
Nuclear Calculation of the JMTR Core, (2). Revision of the Neutron Flux Calculation Method.
The method of nuclear calculation for JMTR core configuration is well established. Due to advance in the calculation techniques, however, the method has been partly revised, as follows. In calculating the fast neutron group constants, space dependent spec...
H. Ando H. Iida Y. Nagaoka R. Oyamada
/sup 18/O as a core plus two valence neutrons: A three-body Faddeev calculation
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.
Ueta, K.; Miyake, H.; Mizukami, A.
Sensitivity of ex-core neutron detectors to vibrations of PWR fuel assemblies
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
F. J. Sweeney; J. P. Renier
MPACT Fast Neutron Multiplicity System Design Concepts
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" � ? 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" � ? 3" detectors.
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
Self-powered in-core neutron detector assembly with uniform perturbation characteristics
Disclosed is a self-powered in-core neutron detector assembly in which a plurality of longitudinally extending self-powered detectors have neutron responsive active portions spaced along a longitudinal path. A low neutron absorptive extension extends from the active portions of the spaced active portions of the detectors in symmetrical longitudinal relationship with the spaced active detector portions of each succeeding detector. The
W. H. Todt; K. C. Playfoot
Neutron capture effects on samarium, europium, and gadolinium in Apollo 15 deep drill-core samples
The isotopic compositions of Sm and Gd in seven lunar samples from the Apollo 15 deep drill core were determined to discuss the effects of neutron capture near the lunar surface. Large isotopic deviations of 150Sm\\/149Sm, 156Gd\\/155Gd and 158Gd\\/157Gd derived from neutron capture effects were observed in all samples. Although neutron capture products in lunar samples were extensively in the
Hiroshi Hidaka; Mitsuru Ebihara; Shigekazu Yoneda
Development and preliminary verification of the 3D core neutronic code: COCO
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)
Lu, H.; Mo, K.; Li, W.; Bai, N.; Li, J. [Reactor Design and Fuel Management Research Center, China Nuclear Power Technology Research Inst., 47F/A Jiangsu Bldg., Yitian Road, Futian District, Shenzhen (China)
Experiences with Soft-Core Processor Design
Soft-core processors exploit the flexibility of Field Pro- grammable Gate Arrays (FPGAs) to allow a system de- signer to customize the processor to the needs of a target application. This paper describes the UT Nios implementa- tion of Altera's Nios architecture. A benchmark set appro- priate for soft-core processors is defined. Using the bench- mark set, the performance of UT
Franjo Plavec; Blair Fort; Zvonko G. Vranesic; Stephen Dean Brown
Core bit design reduces mud invasion, improves ROP
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.
Clydesdale, G. (Security DBS, Aberdeen (United Kingdom)); Leseultre, A.; Lamine, E. (Security DBS, Brussels (Belgium))
Physical design of target station and neutron instruments for China Spallation Neutron Source
The China Spallation Neutron Source (CSNS) is the first accelerator-based multidiscipline user facility to produce pulsed neutrons by tungsten target under collision of a pulsed proton beam with a beam power of 100 kW at a repetition rate of 25 Hz. In this paper, we focus on the physical design of CSNS target station and neutron instruments. Under optimized design, the flat tungsten target and the compact target-moderator-reflector coupling enhance effective cold and thermal neutron output from moderators. Three wing-type moderators supply four different characteristics of neutrons to 19 beamlines primarily for neutron scattering applications. Layout of neutron instruments are conceptually planned for total 20 beamlines, the configuration and specification have been determined for three day-one neutron instruments. All designs are optimized for the Phase I of 100 kW with a upgradable capacity to 500 kW.
Wang, FangWei; Liang, TianJiao; Yin, Wen; Yu, QuanZhi; He, LunHua; Tao, JuZhou; Zhu, Tao; Jia, XueJun; Zhang, ShaoYing
Measurement of Core Coolant Flow Velocities in PWRs Using Temperature: Neutron Noise Cross Correlation.
To study the relationship between the time delay inferred from this phase angle and core coolant flow velocities, noise measurements were performed at the Loss of Fluid Test Facility (LOFT) reactor and at a commercial PWR. In-core, self-powered neutron de...
F. J. Sweeney B. R. Upadhyaya
An intrinsically safe facility for forefront research and training on nuclear technologies � Core design
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.
Viberti, C. M.; Ricco, G.
Advanced BWR core component designs and the implications for SFD analysis
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.
Ott, L.J.
A miniature biaxial accelerometer was developed for vibration measurements in radioactive environments. The sensor is small enough to be assembled in Self-powered Neutron Detector (SPND) in-core instrument strings of PWRs or to be inserted into the travelling in-core probe system of BWRs. Two accelerometers were installed inside of the core of an operating power reactor (PWR, 350 MWel). The vibrations
J. Runkel; E. Laggiard; D. Stegemann; P. Heidemann; R. Blaser; F. Schmid; H. Reinmann
Annular core research reactor high flux neutron radiography facility.
Sandia National Laboratories (SNL) has been performing neutron radiography since 1964. The radiography facilities have evolved from an aperture in a radiation exposure room in the now retired Sandia Engineering Reactor to a divergent collimator radiograph...
F. M. McCrory J. G. Kelly M. E. Vernon D. A. Tichenor
Rapid design of application specific FFT cores
Methods are presented for developing synthesizable FFT cores. These are based on a modular approach in which parameterized commutator and processor blocks are cascaded to implement the computations required in many important FFT signal flow graphs. In addition, it is shown how the use of a digital serial data organization can be used to produce systems that offer 100% processor
Tiong Jiu Ding; John V. McCanny; Yi Hu
Determination of the kinetic parameters of the CALIBAN metallic core reactor from stochastic neutron measurements
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)
Casoli, P.; Authier, N.; Chapelle, A. [Commissariat a l'Energie Atomique et Aux Energies Alternatives, CEA, DAM, F-21120 Is sur Tille (France)
Evaluation of AGNI SFR core neutronics parameters with VESTA and ERANOS
This paper presents the calculation of core neutronics parameters for so called AGNI Sodium Fast Reactor (SFR) model performed with ERANOS code and Monte Carlo depletion interface software VESTA. The AGNI core has been developed at IRSN for its own R&D needs, i.e. to test performance of calculation codes for safety assessment of a generation IV SFR project. The ERANOS code is used as reference code for SFR core calculations at IRSN. In this work, VESTA calculations have been performed and compared with corresponding ERANOS results. These calculations have a double purpose: mastering the use of tools for the evaluation of SFR core static neutronics parameters and validate the use of VESTA for SFR cores.
Ecrabet, Fabrice; Haeck, Wim; Chaitanya Tadepalli, Sai
A Tight Lattice, Epithermal Core Design for the Integral PWR
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
J. G. B. Saccheri; N. E. Todreas; M. J. Driscoll
Designing systems-on-chip using cores
Leading-edge systems-on-chip (SoC) being designed today could reach 20 Million gates and 0.5 to 1 GHz operating frequency. In order to implement such systems, designers are increasingly relying on reuse of Intellectual property (IP) blocks. Since IP blocks are pre-designed and pre-verified, the designer can concentrate on the complete system without having to worry about the correctness or performance of
Reinaldo A. Bergamaschi; William R. Lee
Design and fabrication of integrated solenoid inductors with magnetic cores
Integrated solenoid inductors with magnetic core were designed and fabricated. An inductance above 70 nH was achieved while keeping the coil resistance below 1 Omega and the device area below 1 mm2. The inductance of the magnetic inductor was more than 30 times that of the air core inductor of the identical geometry, and the inductance density reached above 200
Dok Won Lee; Kyu-Pyung Hwang; Shan X. Wang
Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core
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.
Sterbentz, James W
Development of an inconel self powered neutron detector for in-core reactor monitoring
The paper describes the development and testing of an Inconel600 (2mm diameter�21cm long) self-powered neutron detector for in-core neutron monitoring. The detector has 3.5mm overall diameter and 22cm length and is integrally coupled to a 12m long mineral insulated cable. The performance of the detector was compared with cobalt and platinum detectors of similar dimensions. Gamma sensitivity measurements performed at
M. Alex; M. D. Ghodgaonkar
De novo design of the hydrophobic cores of proteins.
We have developed and experimentally tested a novel computational approach for the de novo design of hydrophobic cores. A pair of computer programs has been written, the first of which creates a "custom" rotamer library for potential hydrophobic residues, based on the backbone structure of the protein of interest. The second program uses a genetic algorithm to globally optimize for a low energy core sequence and structure, using the custom rotamer library as input. Success of the programs in predicting the sequences of native proteins indicates that they should be effective tools for protein design. Using these programs, we have designed and engineered several variants of the phage 434 cro protein, containing five, seven, or eight sequence changes in the hydrophobic core. As controls, we have produced a variant consisting of a randomly generated core with six sequence changes but equal volume relative to the native core and a variant with a "minimalist" core containing predominantly leucine residues. Two of the designs, including one with eight core sequence changes, have thermal stabilities comparable to the native protein, whereas the third design and the minimalist protein are significantly destabilized. The randomly designed control is completely unfolded under equivalent conditions. These results suggest that rational de novo design of hydrophobic cores is feasible, and stress the importance of specific packing interactions for the stability of proteins. A surprising aspect of the results is that all of the variants display highly cooperative thermal denaturation curves and reasonably dispersed NMR spectra. This suggests that the non-core residues of a protein play a significant role in determining the uniqueness of the folded structure.
Desjarlais, J. R.; Handel, T. M.
Influence of fuel vibration on PWR neutron noise associated with core barrel motion
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)
Sweeney, F.J.; March-Leuba, J.
We consider the pinning of superfluid (neutron) vortices to magnetic fluxtubes associated with a type II (proton) superconductor in neutron star cores. We demonstrate that core pinning affects the spin-down of the system significantly and discuss implications for regular radio pulsars and magnetars. We find that magnetars are likely to be in the pinning regime, whereas most radio pulsars are not. This suggests that the currently inferred magnetic field for magnetars may be overestimated. We also obtain a new timescale for the magnetic field evolution which could be associated with the observed activity in magnetars, provided that the field has a strong toroidal component.
Glampedakis, Kostas [Theoretical Astrophysics, University of Tuebingen, Tuebingen D-72076 (Germany); Andersson, Nils [School of Mathematics, University of Southampton, Southampton SO17 1BJ (United Kingdom)
Core and Refueling Design Studies for the Advanced High Temperature Reactor
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.
Holcomb, David Eugene [ORNL] [ORNL; Ilas, Dan [ORNL] [ORNL; Varma, Venugopal Koikal [ORNL] [ORNL; Cisneros, Anselmo T [ORNL] [ORNL; Kelly, Ryan P [ORNL] [ORNL; Gehin, Jess C [ORNL] [ORNL
Study of a new compact fast reactor core design
A study was conducted to investigate conceptual Liquid Metal Reactor (LMR) designs, employing some unconventional design features, for improved economics and safety. The unconventional design elements were used to supplement the conventional design measures, which alone did not lead to a truly competitive LMR design. Better economics was obtained through simplicity and compactness of core design. For simplicity of core design, internal blankets were omitted. Core compactness was achieved by maximum power flattening. This was done by employing axial and radial enrichment zones along with axial and radial (BeO) reflectors. To further enhance core compactness, the in-core control rods were replaced by reflector controls. For improved safety, the objective was to reduce both coolant void and burnup reactivities. However, even with the use of a wide spectrum of unconventional design features, such as burnable poisons, peripheral reflectors and inner moderating regions, it was not possible to overcome the classical known fact that both coolant void and burnup reactivities cannot be reduced simultaneously. The only resolution of this dilemma appeared to be to minimize coolant void reactivity, and to manage the burnup reactivity losses, such that an accidental insertion of significant amounts of reactivity is mechanically not possible. A conceptual design with these characteristics is described in this thesis.
Hamid, T.
Verifying IP-core based system-on-chip designs
We describe a methodology for verifying system-on-chip designs. In our methodology, the problem of verifying system-on-chip designs is decomposed into three tasks. First, we verify, once and for all, the standard bus interconnecting IP cores in the system. The next task is to verify the glue logic, which connects the IP cores to the buses. Finally, using the verified bus
Pankaj Chauhan; Edmund M. Clarke; Yuan Lu; Dong Wang
Reactor physics analyses of the advanced neutron source three-element core
A reactor physics analysis was performed for the Advanced Neutron Source reactor with a three-element core configuration. The analysis was performed with a two-dimensional r-z 20-energy-group finite-difference diffusion theory model of the 17-d fuel cycle. The model included equivalent r-z geometry representations of the central control rods, the irradiation and production targets, and reflector components. Calculated quantities include fuel cycle parameters, fuel element power distributions, unperturbed neutron fluxes in the reflector and target regions, reactivity perturbations, and neutron kinetics parameters.
Gehin, J.C.
TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.
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
Kurosawa, Masahiko
Designs for thin-film-coated semiconductor thermal neutron detectors
Thin film coated semiconductor detectors have been studied and used as neutron detectors for decades. Unfortunately, with front-irradiated devices, the basic design limits the thermal neutron detection efficiency to only 3.95% for 10B-coated devices and only 4.3% for 6LiF-coated devices. Presented in the following work are several straightforward methods to increase the thermal neutron detection efficiency for thin-film-coated semiconductor thermal
Douglas S. McGregor; Raymond T. Klann; Holly K. Gersch; Jeffrey D. Sanders
Neutron single-particle strength outside the N=50 core
The single-neutron properties of N = 51 nuclei have been studied with the (d,p) and (?,3He) reactions, at beam energies of 15 and 50 MeV respectively, on 88Sr, 90Zr, and 92Mo targets. The light reaction products were momentum analyzed using a conventional magnetic spectrometer. Additionally, the 2H(86Kr,p) reaction was measured at a beam energy of 10 MeV/u, where outgoing light ions were analyzed using a helical-orbit spectrometer. Absolute cross sections and angular distributions corresponding to the population of different final states in the heavy product were obtained for each reaction. Spectroscopic factors were extracted and centroids of the single-particle strength were deduced. The observations appear consistent with calculations based on an evolution of single-particle structure driven by the nucleon-nucleon forces acting between valence protons and neutrons.
Sharp, D. K.; Kay, B. P.; Thomas, J. S.; Freeman, S. J.; Schiffer, J. P.; Back, B. B.; Bedoor, S.; Bloxham, T.; Clark, J. A.; Deibel, C. M.; Hoffman, C. R.; Howard, A. M.; Lighthall, J. C.; Marley, S. T.; Mitchell, A. J.; Otsuka, T.; Parker, P. D.; Rehm, K. E.; Shetty, D. V.; Wuosmaa, A. H.
Advanced Neutron Source: The designer's perspective.
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, wit...
F. J. Peretz
Neutronics design of a medical therapy reactor
The development of a Medical Therapy Reactor (MTR) facility for the treatment of Glioblastoma Multiforme and other presently incurable cancer types is underway at the Idaho National Engineering Laboratory (INEL). This paper addresses the feasibility of utilizing existing reactor technologies to deliver therapeutic doses of epithermal neutrons with minimal fast neutron and gamma contaminants. The two primary fuel candidates are
W. A. Neuman; D. K. Parsons; J. A. Lake
Shielding design and analyses of KIPT neutron source facility
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. An electron accelerator drives a sub-critical facility (ADS) is used for generating the neutron source. The facility will be utilized for performing basic and applied nuclear researches, producing medical isotopes, and
Zhaopeng Zhong; Yousry Gohar
A small tokamak conceptual design for volumetric neutron source
The small tokamak conceptual design based on the fusion-fission hybrid reactor has been studied for a long time for breeding fuel and transmutation of high level waste. If poloidal distributions of neutron wall loading and flux are considered, the midplane equatorial port can be used to test some components, such as blanket modules at a neutron wall loading of 1
L. J Qiu; Y Wu; G. S Luan; Z. J Gou; Q Wu; P Duan; X. H Gao; L. L Liu; Q. Y Huang; B. J Xioa; Y. W Yang
Conceptual design of a medical reactor for neutron capture therapy
A conceptual design of a passively safe reactor facility for boron neutron capture therapy is presented. The facility configuration and its neutronic, thermal hydraulic, and safety issues are addressed in order to demonstrate the deployability of reactor technology for routine patient treatments and advanced research and dosimetry. The reactor has a power level of <10 MW (thermal) and is based
W. A. Neuman; J. L. Jones
Neutron Flux Measurements in the Full Power Leu Core of PARR-1.
Measurements of thermal and epithermal flux distributions were made in and around the first low enriched uranium (LEU) full power core of Pakistan Research Reactor-1(PARR-1). Neutron Flux measurements were made with the help of activation foils and the se...
L. Ali S. A. Ansari Q. D. Shami M. Iqbal
Water-level measurements using ex-core neutron detectors at Farley Unit One. Final report
This report documents the theory, tests, results, and analyses of water level measurement tests performed at the Farley Unit One nuclear power plant in November 1980. Measurements were made during a normal plant outage using an ex-core neutron detector system placed on top of the reactor vessel head. The data indicate that this system was able to indicate when the
S. Untermyer; W. K. Hagan; J. E. Jones
High temperature performance of in-core self-powered neutron detectors. II
The performance of the double emitter type self -powered neutron ; detector was investigated in a high-temperature reactor core. The noise current ; of the detector was influenced by an irradiation history correlated with ; temperature and radiation intensity. Even in the worst condition of high ; temperature and low radiation dose, the detector was able to operate up to
A. Sekiguchi; S. Takeuchi; M. Nakazawa
GFR fuel and core pre-conceptual design studies
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)
Chauvin, N.; Ravenet, A.; Lorenzo, D.; Pelletier, M.; Escleine, J.M.; Munoz, I.; Bonnerot, J.M. [CEA/DEN/DEC, CEA Cadarache, 13108 Saint-Paul-lez-Durance (France); Malo, J.Y.; Garnier, J.C.; Bertrand, F.; Bosq, J.C. [CEA/DEN/DER, CEA Cadarache, 13108 Saint-Paul-lez-Durance (France)
Proto-neutron Star Convection in the Post-bounce Epoch of Stellar Core Collapse
We present results of 2-D simulations of convective instabilities in proto-neutron stars in the immediate aftermath of stellar core collapse. The capture of electrons by protons during collapse and the subsequent post-bounce deleptonization sets up a strong gradient in the electron fraction near the proto-neutron star surface. The formation of a strong shock at the outer edge of the homologous core gives rise to a strong entropy gradient. Depending on the precise nature of these gradients, and the equation of state, there are several possible instabilities that can arise in the outer layers of the proto-neutron star. In this poster, we describe the results of our 2-D radiation-hydrodynamic simulations of the proto-neutron star. These simulations have revealed previously unseen beahvior, including stratified convection in the proto-neutron star and a rapid one-time deloptonization burst. We find that, in our models, vigorous proto-neutron star convection does not persist after destabilizing gradients have been eradicated. This work was performed at the State University of New York at Stony Brook as part of the TeraScale Supernova Initiative, and is funded by SciDAC grant DE-FC02-01ER41185 from the U.S. Dept. of Energy, Office of Science High-Energy, Nuclear, and Advanced Scientific Computing Research Programs. We gratefully acknowledge support of the National Energy Research Scientific Computing Center (NERSC) for computational and consulting support.
Swesty, F. D.; Myra, E. S.
Implementing a New Ion Chamber Design for Neutron Spin Rotation
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.
Gardiner, Hannah; Anderson, Eamon; Fry, Jason; Holley, Adam; Snow, Mike
Design analysis of core assemblies for supercritical pressure conditions
The increase of steam parameters to supercritical conditions could reduce the power generating costs of light water reactors significantly [Proceedings of SCR-2000 (2000) 1]. Core assemblies, however, will differ from current BWR or PWR design. In this context, this paper summarizes the main results related to a thermal-hydraulic design analysis of applicable fuel assemblies. Starting from a thorough literature survey
X. Cheng; T. Schulenberg; D. Bittermann; P. Rau
Design. Youth Training Scheme. Core Exemplar Work Based Project.
This trainer's guide is intended to assist supervisors of work-based career training projects in helping students understand the importance of industrial designers and the stages of the industrial design process. The guide is one in a series of core curriculum modules that is intended for use in combination on- and off-the-job programs to�
Further Education Staff Coll., Blagdon (England).
Characterization of Neutron Fields in the Experimental Fast Reactor Joyo Mk-Iii Core
In 2003, Joyo MK-III core was upgraded to increase the irradiation testing capability. This paper describes the details of distributions of neutron flux and reaction rate in the MK-III core that was measured by characterization tests during the first two operating cycles. The calculation accuracy of the core management codes HESTIA, TORT and MCNP, was also evaluated by the measured data. The calculated fission rates of 235U by HESTIA agreed well with the measured one within approximately 4% in the fuel region. MCNP could simulate within 6% in the central non-fuel irradiation test subassembly and the radial reflector region, while large discrepancies were obtained in TORT results. Hence, the precise geometry model was effective in evaluating the neutron spectrum and the flux at such locations.
Maeda, Shigetaka; Ito, Chikara; Ohkawachi, Yasushi; Sekine, Takashi; Aoyama, Takafumi
Effects of Shock Instability on Spin and Kick of Proto-Neutron Star in Supernova Cores
We have numerically studied the standing/spherical accretion shock instability (SASI) for a core-collapse supernova. The core-collapse supernova is an explosion of a massive star in the final stage of its evolution. Although this spectacular phenomenon is a key issue for astrophysics, the explosion mechanism has not been understood perfectly. Recently, SASI has widely been noticed since it may play an important role for the explosion mechanism of a core-collapse supernova. In addition to it, the latest studies suggest that SASI may also affect on rotation and kick of a pulsar which is regarded as a neutron star formed by the supernova explosion. The origin of a pulsar spin and kick has been vigorously investigated, but it is still controversial among astrophysicists. We report on the effects of SASI on spin and kick of the proto-neutron star with the results of three-dimensional simulations.
Iwakami, Wakana; Ohnishi, Naofumi; Kotake, Kei; Yamada, Shoichi; Sawada, Keisuke
A demonstration of a whole core neutron transport method in a gas cooled reactor
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)
Connolly, K. J.; Rahnema, F. [Nuclear and Radiological Engineering and Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA (United States)] [Nuclear and Radiological Engineering and Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA (United States)
Design, fabrication and experiment of iron-core passive compulsator
The paper presents design, fabrication and experiments of an iron-core, rotating field, two phases and modified non-salient pole rotor compulsator (compensated pulsed alternator), which is for comparison with air-core compulsator to be made in the near future. It includes parameters and special technics different from traditional generators. The decision of thickness of compensation shield is based on the simulation of
Shaopeng Wu; Shumei Cui; Liwei Song; Manlan Liu
Core design of the upgraded TREAT reactor
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.
Wade, D.C.; Bhattacharyya, S.K.; Lipinski, W.C.; Stone, C.C.
Maximum mass of neutron stars with quark matter core
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.
Takatsuka, Tatsuyuki; Hatsuda, Tetsuo; Masuda, Kota [Iwate University, Morioka 020-8550 (Japan); Department of Physics, University of Tokyo, Tokyo 113-0033, Japan and Theoretical Research Division, Nishina Center, RIKEN, Wako 351-0198 (Japan); Department of Physics, University of Tokyo, Tokyo 113-0033 (Japan)
Simultaneous measurement of neutron and gamma-ray radiation levels from a TRIGA reactor core using silicon carbide semiconductor detectors
The ability of a SiC detector to measure neutron and gamma radiation levels in a TRIGA reactor's mixed neutron\\/gamma field was demonstrated. Linear responses to an epicadmium neutron fluence rate (up to 3�107 cm-2 s-1) and to a gamma dose rate (0.6-234 krad-Si h-1) were obtained with the detector. Axial profiles of the reactor core's neutron and gamma-ray radiation levels
A. R. Dulloo; F. H. Ruddy; J. G. Seidel; C. Davison; T. Flinchbaugh; T. Daubenspeck
The ability of a silicon carbide radiation detector to measure neutron and gamma radiation levels in a TRIGA reactor's mixed neutron\\/gamma field was demonstrated. Linear responses to epicadmium neutron fluence rate (up to 3�107 cm-2 s-1) and to gamma dose rate (0.6-234 krad-Si h-1) were obtained with the detector. Axial profiles of the reactor core's neutron and gamma-ray radiation levels
Design and demonstration of a quasi-monoenergetic neutron source
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 filters such as vanadium and manganese are also explored and the possibility of studying the response of different materials to low-energy nuclear recoils using the resultant neutron beams is discussed.
Joshi, T. H.; Sangiorgio, S.; Mozin, V.; Norman, E. B.; Sorensen, P.; Foxe, M.; Bench, G.; Bernstein, A.
Design of neutron beams for neutron capture therapy using a 300kW slab TRIGA reactor
A design for a slab reactor to produce an epithermal neutron beam and a thermal neutron beam for use in neutron capture therapy (NCT) is described. A thin reactor with two large-area faces, a ``slab`` reactor, was planned using eighty-six 20% enriched TRIGA fuel elements and four B�C control rods. Two neutron beams were designed: an epithermal neutron beam from
Electron capture in core-collapse supernovae investigated through configuration mixing in neutron-rich nuclei
Electron capture on neutron-rich medium-mass nuclei is a key process where the electrons that impede the collapse of the core of massive stars are captured, thereby producing very neutron-rich nuclei. As the core collapses, the supernova is then initiated. For the electron capture to proceed, however, the allowed Gamow-Teller (GT) transition must be unblocked either by thermal excitations or by mixing of proton configurations from a higher-lying shell into the ground-state configuration of the nucleus. The present paper presents an experiment performed at the National Superconducting Cyclotron Laboratory at Michigan State University, in which we study the configuration mixing in the neutron-rich76Zn isotope. The experiment utilised single-proton and single-neutron knockout with detection of reaction-residue ? rays and measurement of the parallel momentum of the residue. Through this we investigate the proton components of the 76Zn ground state, with a particular interest in ?-g9/2, which may unblock the GT electron capture even at low temperatures and thereby open a new pathway for the initiation of the collapse of the pre-supernova stellar core.
Diget, C. Aa; Baugher, T.; Bazin, D.; Bentley, M. A.; Berryman, J. S.; Brown, J. R.; Chipps, K. A.; Davies, P. J.; Fulton, B. R.; Gade, A.; Iwasaki, H.; Laird, A. M.; Lemasson, A.; McDaniel, S.; Ratkiewicz, A.; Scruton, L.; Shore, A.; Stroberg, S. R.; Weisshaar, D.; Wimmer, K.; Winkler, R.
Transport properties of a quark-hadron Coulomb lattice in the cores of neutron stars
Already more that 40 years ago, it has been suggested that because of the enormous mass densities in the cores of neutron stars, the hadrons in the centers of neutron stars may undergo a phase transition to deconfined quark matter. In this picture, neutron stars could contain cores made of pure (up, down, strange) quark matter which are surrounded by a mixed phase of quarks and hadrons. More than that, because of the competition between the Coulomb and the surface energies associated with the positively charged regions of nuclear matter and negatively charged regions of quark matter, the mixed phase may develop geometrical structures similarly to what is expected of the subnuclear liquid-gas phase transition. In this paper we restrict ourselves to considering the formation of rare phase blobs in the mixed quark-hadron phase. The influence of rare phase blobs on the thermal and transport properties of neutron star matter is investigated. The total specific heat cV, thermal conductivity ?, and electron-blob bremsstrahlung neutrino emissivities ??,BR, of quark-hybrid matter are computed, and the results are compared with the associated thermal and transport properties of standard neutron star matter. Our results show that the contribution of rare phase blobs to the specific heat is negligibly small. This is different for the neutrino emissivity from electron-blob bremsstrahlung scattering, which turns out to be of the same order of magnitude as the total contributions from other bremsstrahlung processes for temperatures below about 108K.
Na, Xuesen; Xu, Renxin; Weber, Fridolin; Negreiros, Rodrigo
Core-collapse Supernova Equations of State Based on Neutron Star Observations
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 ? 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.
Steiner, A. W.; Hempel, M.; Fischer, T.
Massive neutron stars with a hyperonic core: A case study with the IUFSU relativistic effective interaction
The recent discoveries of massive neutron stars, such as PSR J0348+0432 and PSR J1614-2230, have raised questions about the existence of exotic matter such as hyperons in the neutron star core. The validity of many established equations of states (EoSs) like GM1 and FSUGold are also questioned. We investigate the existence of hyperonic matter in the central regions of massive neutron stars by using relativistic mean field (RMF) theory with the recently proposed Indiana University Florida State University (IUFSU) model. The IUFSU model is extended by including hyperons to study the neutron star in ? equilibrium. The effect of different hyperonic potentials, namely ? and ? potentials, on the EoS and hence the maximum mass of neutron stars has been studied. We have also considered the effect of stellar rotation since the observed massive stars are pulsars. It has been found that a maximum mass of 1.93M?, which is within the 3? limit of the observed mass of PSR J0348+0432, can be obtained for rotating stars, with certain choices of the hyperonic potentials. The said star contains a fair amount of hyperons near the core.
Bhowmick, Bipasha; Bhattacharya, Madhubrata; Bhattacharyya, Abhijit; Gangopadhyay, G.
Effects of lumbar artificial disc design on intervertebral mobility: in vivo comparison between mobile-core and fixed-core
Although in theory, the differences in design between fixed-core and mobile-core prostheses should influence motion restoration,\\u000a in vivo kinematic differences linked with prosthesis design remained unclear. The aim of this study was to investigate the\\u000a rationale that the mobile-core design seems more likely to restore physiological motion since the translation of the core\\u000a could help to mimic the kinematic effects
Jo�l Del�crin; J�r�me Allain; Jacques Beaurain; Jean-Paul Steib; Jean Huppert; Herv� Chataigner; Marc Ameil; Lucie Aubourg; Jean-Michel Nguyen
Optimization of a neutron detector design using adjoint transport simulation
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)
Yi, C.; Manalo, K.; Huang, M.; Chin, M.; Edgar, C.; Applegate, S.; Sjoden, G. [Georgia Inst. of Technology, Gilhouse Boggs Bldg., 770 State St, Atlanta, GA 30332-0745 (United States)
Safety and core design of large liquid-metal cooled fast breeder reactors
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.
Qvist, Staffan Alexander
Analysis and Design of ITER 1 MV Core Snubber
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.
Wang, Haitian; Li, Ge
Design and analysis of a nuclear reactor core for innovative small light water reactors
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.
Soldatov, Alexey I.
During the accident at Three Mile Island Unit 2 (TMI-2), a significant mass of core debris slumped to the bottom head of the reactor vessel. Defueling activities caused more core debris to relocate to the lower head region. The variations in the ex-core neutron detector, or source range monitor (SRM), readings gave evidence of this effect of defueling activity. Between October 1986 and November 1987, as a result of defueling the lower core region, increases in SRM rates were noted as 50% for NI-1 and 90% for NI-2. Analysis of these increases shows that they correspond to lower head rubble bed mass increases of 30 to 50%, or between 6 and 8 tonnes of rubble. The above yields a total lower head inventory of between 21 and 28 tonnes of rubble; this equates to 15.5 to 21 tonnes of UO/sub 2/.
Rainisch, R.; Fricke, V.
Influence Of Low Boron Core Design On PWR Transient Behavior
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)
Aleksandrov Papukchiev, Angel; Yubo Liu [Technical University Munich, Arcisstrasse 21, 80333 Muenchen (Germany); Schaefer, Anselm [ISaR Institute for Safety and Reliability, Walther-Meissner-Str. 2 85748 Garching (Germany)
GCRA review and appraisal of HTGR reactor-core-design program. [HTGR-SC, -R, -NHSDR
The reactor-core-design program has as its principal objective and responsibility the design and resolution of major technical issues for the reactor core and core components on a schedule consistent with the plant licensing and construction program. The task covered in this review includes three major design areas: core physics, core thermal and hydraulic performance fuel element design, and in-core fuel performance evaluation.
Physics design of a cold neutron source for KIPT neutron source facility.
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.
Zhong, Z.; Gohar, Y.; Kellogg, R.; Nuclear Engineering Division
Design Studies on Iron-Core Synchronously Operating Linear Motors.
This report concerns the design of iron-core synchronously-operating linear motors with passive rail track. Claw-pole and homopolar inductor motors were the 2 basic types considered. The principal conclusions were the homopolar inductor motor is the prefe...
E. Levi
Core Design with Respect to the Safety Concept.
In the present paper the following topics are dealt with: principles of reactor core design and optimization, fuel management and safety concept for higher cycles and results of risk analyses (e.g. rod ejection, steam line break etc.). (Atomindex citation...
W. Kollmar
Are Neutron Stars with Crystalline Color-Superconducting Cores Relevant for the LIGO Experiment?
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.
Haskell, B.; Andersson, N.; Jones, D. I.; Samuelsson, L. [School of Mathematics, University of Southampton, Southampton SO17 1BJ (United Kingdom)
Are neutron stars with crystalline color-superconducting cores relevant for the LIGO experiment?
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
Haskell, B; Andersson, N; Jones, D I; Samuelsson, L
Castilla-La Mancha neutron monitor: situation and design
This work shows the present status of the Castilla-La Mancha Neutron Monitor (CaLMa). Thanks to the founds recently agree for the Castilla-La Mancha Comunity governement. The new neutron monitor station will be building and placed in the Alcal� University campus in Guadalajara. Being operative in a provisional site at the end of 2010. The station is based on 18NM64 of LND 25373 3He detectors integrated into the Neutron Monitor Data Base (NMDB). Details about geomagnetic conditions, design and schedule are presented.
Medina, Jose; Jos� Blanco, Juan; Garc�a, Oscar
Advanced Microstructured Semiconductor Neutron Detectors: Design, Fabrication, and Performance
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/?) 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/? 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.
Bellinger, Steven Lawrence
On the optimum design of a white neutron source
The conceptual design of a neutron source for time-of-flight (TOF) measurements with good energy resolution was presented in a previous paper, aiming at its implementation on a high energy electron Linac. There is a growing interest in nuclear data worldwide and the existing neutron sources are clearly insufficient.The distinguished feature of this source is the very small size of the
S. Bartalucci; V. Angelov
Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR
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.
Hanson, A.L.; Diamond, D.
Track retaining properties of quartz for high temperature in-core neutron fluence measurements
Track registration and retention characteristics of natural quartz and its synthetic counterpart Vitreosil (fused silica) have been studied. Tests have been carried out at high temperatures (200�1000�C) and the detectors subjected to intense ?-ray fields of up to 250 MRads, with a view to establishing their suitability for high temperature in-core neutron fluence measurements. Both the major crystallographic axes of
A. Aframian
Active Neutron Veto Shield Design for SuperCDMS-SNOLAB
Protection against neutron backgrounds is one of the key issues for the next generation SuperCDMS direct dark matter detection experiment that aims at exploring the 10-46 cm^2 cross section region for spin-independent interactions of Weakly Interacting Massive Particles (WIMPs) with nucleons. Estimation of the background from radioactive processes is a crucial task for the current experiment stage and for designing future large-scale detectors. An active neutron veto would make the next generation of dark matter experiment more robust, improving the credibility of a dark matter detection claim based on the observation of a few recoil events. SuperCDMS is investigating the feasibility of adding a liquid scintillator active neutron veto, loaded with high neutron capture cross section isotopes, to the SuperCDMS SNOLAB shield design. This system not only will allow us to reject neutron-induced backgrounds, but also will provide an in-situ measurement of the neutron flux near the detector. I will report the status of ongoing studies related to the design and construction of such a veto.
Qiu, Hang
Optimization of the core configuration design using a hybrid artificial intelligence algorithm for research reactors
To successfully carry out material irradiation experiments and radioisotope productions, a high thermal neutron flux at irradiation box over a desired life time of a core configuration is needed. On the other hand, reactor safety and operational constraints must be preserved during core configuration selection. Two main objectives and two safety and operational constraints are suggested to optimize reactor core
Afshin Hedayat; Hadi Davilu; Ahmad Abdollahzadeh Barfrosh; Kamran Sepanloo
The EUROBALL neutron wall � design and performance tests of neutron detectors
The mechanical design of the EUROBALL neutron wall and neutron detectors, and their performance measured with a 246,248Cm fission source are described. The array consists of 15 pseudohexaconical detector units subdivided into three, 149mm high, hermetically separated segments and a smaller central pentagonal unit subdivided into five segments. The detectors are filled with Bicron BC501A liquid scintillator. Each section of
�. Skeppstedt; H. A. Roth; L. Lindstr�m; R. Wadsworth; I. Hibbert; N. Kelsall; D. Jenkins; H. Grawe; M. G�rska; M. Moszynski; Z. Sujkowski; D. Wolski; M. Kapusta; M. Hellstr�m; S. Kalogeropoulos; D. Oner; A. Johnson; J. Cederk�ll; W. Klamra; J. Nyberg; M. Weiszflog; J. Kay; R. Griffiths; J. Garces Narro; C. Pearson; J. Eberth
Design considerations for neutron activation and neutron source strength monitors for ITER
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.
Barnes, C.W. [Los Alamos National Lab., NM (United States); Jassby, D.L.; LeMunyan, G.; Roquemore, A.L. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Walker, C. [ITER Joint Central Team, Garching (Germany)
Design of composite flywheel rotors with soft cores
A flywheel is an inertial energy storage system in which the energy or momentum is stored in a rotating mass. Over the last twenty years, high-performance flywheels have been developed with significant improvements, showing potential as energy storage systems in a wide range of applications. Despite the great advances in fundamental knowledge and technology, the current successful rotors depend mainly on the recent developments of high-stiffness and high-strength carbon composites. These composites are expensive and the cost of flywheels made of them is high. The ultimate goal of the study presented here is the development of a cost-effective composite rotor made of a hybrid material. In this study, two-dimensional and three-dimensional analysis tools were developed and utilized in the design of the composite rim, and extensive spin tests were performed to validate the designed rotors and give a sound basis for large-scale rotor design. Hybrid rims made of several different composite materials can effectively reduce the radial stress in the composite rim, which is critical in the design of composite rims. Since the hybrid composite rims we studied employ low-cost glass fiber for the inside of the rim, and the result is large radial growth of the hybrid rim, conventional metallic hubs cannot be used in this design. A soft core developed in this study was successfully able to accommodate the large radial growth of the rim. High bonding strength at the shaft-to-core interface was achieved by the soft core being molded directly onto the steel shaft, and a tapered geometry was used to avoid stress concentrations at the shaft-to-core interface. Extensive spin tests were utilized for reverse engineering of the design of composite rotors, and there was good correlation between tests and analysis. A large-scale composite rotor for ground transportation is presented with the performance levels predicted for it.
Kim, Taehan
Accelerator shield design of KIPT neutron source facility
Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the design development of a neutron source facility at KIPT utilizing an electron-accelerator-driven subcritical assembly. Electron beam power is 100 kW, using 100 MeV electrons. The facility is designed to perform basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The biological shield of the accelerator building is designed to reduce the biological dose to less than 0.5-mrem/hr during operation. The main source of the biological dose is the photons and the neutrons generated by interactions of leaked electrons from the electron gun and accelerator sections with the surrounding concrete and accelerator materials. The Monte Carlo code MCNPX serves as the calculation tool for the shield design, due to its capability to transport electrons, photons, and neutrons coupled problems. The direct photon dose can be tallied by MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is less than 0.01 neutron per electron. This causes difficulties for Monte Carlo analyses and consumes tremendous computation time for tallying with acceptable statistics the neutron dose outside the shield boundary. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were developed for the study. The generated neutrons are banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron and secondary photon doses. The weight windows variance reduction technique is utilized for both neutron and photon dose calculations. Two shielding materials, i.e., heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total dose outside the shield boundary at less than 0.5-mrem/hr. The shield configuration and parameters of the accelerator building have been determined and are presented in this paper. (authors)
Zhong, Z.; Gohar, Y. [Argonne National Laboratory, 9700 South Cass Avenue, Lemont, IL 60439 (United States)] [Argonne National Laboratory, 9700 South Cass Avenue, Lemont, IL 60439 (United States)
Study of the neutron beam line shield design for JSNS.
The JSNS, a spallation neutron source of J-PARC (JAERI-KEK Joint Project of the High Intensity Proton Accelerator) has 23 neutron beam lines. In the present study, a database was formulated for an optimum shielding design using the MCNP-X code. The calculations involved two steps. In the first step, the neutron distributions were created in the typical neutron beam line with a model that included the spallation neutron source target. The neutron currents evaluated flowed from the duct into the duct wall which was the boundary source for the bulk shield surrounding the beam line. In the second step, bulk-shield calculations were performed for the various shielding materials (iron, concrete, heavy concrete and so on) used and their composites up to thicknesses of 3 m. The results were compared with each other. Composite material shields of iron and such hydrogeneous materials as polyethylene or concrete were more effective. A typical design was prepared for a beam line within 25 m distance from a moderator, as a sample. PMID:16381789
Kawai, M; Saito, K; Sanami, T; Nakao, N; Maekawa, F
Advanced Neutronics Tools for BWR Design Calculations
This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method Of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007 BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy. (authors)
Santamarina, A.; Hfaiedh, N.; Letellier, R.; Sargeni, A.; Vaglio, C. [CEA-Cadarache, 13108 St Paul lez Durance Cedex (France); Marotte, V. [AREVA NP SAS (France); Misu, S. [AREVA NP GmbH (Germany); Zmijarevic, I. [CEA-Saclay, 91191 Gif-sur-Yvette Cedex (France)
Designing an Institutional Web-based Core Facility Management System
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.
Tabarini, D.; Clisham, S.; John, D.; Hagen, A.
Neutron Tube Design Study for Boron Neutron Capture TherapyApplication
Burning of two-flavor quark matter into strange matter in neutron stars and in supernova cores
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
J. D. Anand; Ashok Goyal; V. K. Gupta; S. Singh
Small Angle Neutron-Scattering Studies of the Core Structure of Intact Neurosecretory Vesicles.
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.
Krueger, Susan Takacs
A low resolution model for the chromatin core particle by neutron scattering
Neutron scattering studies have been applied to chromatin core particles in solution, using the contrast variation technique. On the basis of the contrast dependance of the radius of gyration and the radial distribution function it is shown that the core particle consists of a core containing most of the histone around which is wound the DNA helix,following a path with a mean radius of 4.5 nm,in association with a small proportion of the histones. Separation of the shape from the internal structure, followed by model calculations shows that the overall shape of the particle is that of a flat cylinder with dimensions ca. 11�11�6 nm. Further details of the precise folding of the DNA cannot be deduced from the data, but detailed model calculations support concurrent results from crystallographic studies25. Images
Suau, Pedro; Kneale, G.Geoff; Braddock, Gordon W.; Baldwin, John P.; Bradbury, E.Morton
Test Wrapper Design and Optimization Under Power Constraints for Embedded Cores With Multiple Clock Domains
Even though many embedded cores contain several clock domains, most published methods for wrapper design have been limited to single-frequency cores. Cumbersome and invasive design techniques, such as insertion of test points, are needed to make these methods applicable to current-generation embedded cores. This paper presents a new method for designing test wrappers for embedded cores with multiple clock domains.
Qiang Xu; Nicola Nicolici; Krishnendu Chakrabarty
Neutronics analyses in support of the conceptual design of the MAPS NTP reactor
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.}
Raepsaet, X.; Lenain, R. [Commissariat a l`Energie Atomique, DRN/DMT CEN Saclay, F-91191 Gif-sur-Yvette Cedex (France); Naury, S. [Altran Technologies, 58 Boulvard Gouvion Saint Cyr, 75858 PARIS Cedex 17 (France)
Phenix Power Plant Decommissioning Project - Removal of Core and Neutron Blanket Components
Phenix is a sodium-cooled fast neutron reactor located at the CEA's Rhone Valley Center where it was commissioned in 1974. It has an electric power rating of 250 MW and is operated jointly by the CEA and EDF. Its primary role today is to investigate the transmutation of long-lived radioactive waste into shorter-lived wasteform. Its final shutdown is scheduled for the beginning of 2009 in accordance with the reactor operating program with a mean availability of 80%. In this context the Phenix Power Plant Decommissioning Project was initiated in 2003. It covers the definitive cessation of plant operation and the dismantling (D and D) operations together with the final shutdown preparatory phase. The final shutdown phase corresponds to the initial operations undertaken to remove the maximum of hazardous materials. During this phase, subject to the standard operating methodological procedures, all the fuel in the plant will be removed and most of the other reactor core elements and other removable components (primary pumps, intermediate heat exchangers, control rod mechanisms, etc.) will be also dismantled and packaged in accordance with the available disposition routes. The core elements include fissile and fertile subassemblies, steel reflectors, and lateral neutron shielding rods. About half the lateral neutron shielding rods are accessible using the existing handling equipment; specific development work will be necessary to remove the others. Multiple engineering studies are in progress concerning the removal of the core elements. They address safety concerns taking into account thermal and seismic stresses, waste management including radiological characterization of the objects to be removed and the specification of suitable disposal routes (interim storage or disposal facilities), and scenario studies with the definition of each stage of the waste removal process. All these aspects are discussed in this paper. The feasibility of removing the core elements has been demonstrated with regard to handling and safety. A new handling arm is now being developed to remove the inaccessible core elements. The new earthquake safety calculations completed in 2006 did not reveal any insurmountable difficulties. Major work remains to characterize the core elements and determine the capacity of the disposition routes. The characterization program now in progress will validate the core element activation levels, but must be supplemented by a program to characterize the surface contamination from the primary sodium system. Concerning the disposition routes, several specific containers and shipping casks must be developed; engineering studies are now in progress to obtain the necessary approvals.
Moitrier, C. [CEA /Marcoule DDCO/SDSP BP 17171 302078 Bagnols Sur Ceze (France)
Core compressor exit stage study. 1: Aerodynamic and mechanical design
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.
Burdsall, E. A.; Canal, E., Jr.; Lyons, K. A.
Neutrino pair emission due to scattering of electrons off fluxoids in superfluid neutron star cores.
We study the emission of neutrinos, resulting from the scattering of electrons off magnetic flux tubes (fluxoids) in the neutron star cores with superfluid (superconducting) protons. In the absence of proton superfluidity (T>=T_cp_), this process transforms into the well known electron synchrotron emission of neutrino pairs in a locally uniform magnetic field B, with the neutrino energy loss rate Q proportional to B^2^ T^5^. For temperatures T not much below T_cp_, the synchrotron regime (Q{prop.to}T^5^) persists and the emissivity Q can be amplified by several orders of magnitude due to the appearance of the fluxoids and associated enhancement of the field within them. For lower T, the synchrotron regime transforms into the bremsstrahlung regime (Q{prop.to}T^6^) similar to the ordinary neutrino-pair bremsstrahlung of electrons which scatter off atomic nuclei. We calculate Q numerically and represent our results through a suitable analytic fit. In addition, we estimate the emissivities of two other neutrino-production mechanisms which are usually neglected - neutrino-pair bremsstrahlung processes due to electron-proton and electron-electron collisions. We show that the electron-fluxoid and electron-electron scattering can provide the main neutrino production mechanisms in the neutron star cores with highly superfluid protons and neutrons at T<~5x10^8^K. The electron-fluxoid scattering is significant if the initial, locally uniform magnetic field B>10^13^G.
Kaminker, A. D.; Yakovlev, D. G.; Haensel, P.
New casing and backfill design for neutron logging access boreholes
In an effort to enhance the usefulness of neutron logging for environmental applications, a new combination of backfill and casing materials for access boreholes has been developed. The combination of acrylic casing and polyurethane foam backfill has been tested under laboratory and field conditions. Acrylic casing does not significantly attenuate fluxes of either high energy or thermal neutrons, in contrast with polyvinyl chloride casing which reduces the thermal neutron flux by more than 40% due to neutron absorption by chlorine. Polyurethane foam, which is inert, hydrophobic, and insoluble in water, adheres well to both dry and wetted soils, sediments, and rocks. It can be formed in situ at a low, but controllable, bulk density. At a bulk density of 0.08 g cm{sup {minus}3}, and in combination with acrylic casing, polyurethane foam increases the thermal neutron count by less than 5% in a saturated sand, relative to background. In addition to its small effect on the neutron flux, polyurethane foam, unlike bentonite or cement, does not affect the moisture content of the surrounding formation during installation. Furthermore, because it is a closed-cell foam, its moisture content does not change under varying formation moisture conditions. As was shown in related field tests, polyurethane foam is especially well suited for backfilling boreholes in fractured rocks because of its fast set time which minimizes penetration into fractures. The design proved to be convenient and durable under rugged field conditions.
Zawislanski, P.T.; Faybishenko, B. [Lawrence Berkeley National Lab., CA (United States). Earth Sciences Div.
In an effort to enhance the usefulness of neutron logging for environmental applications, a new combination of backfill and casing materials for access boreholes has been developed. The combination of acrylic casing and polyurethane foam backfill has been tested under laboratory and field conditions. Acrylic casing does not significantly attenuate fluxes of either high energy or thermal neutrons, in contrast with polyvinyl chloride casing which reduces the thermal neutron flux by more than 40 percent due to neutron absorption by chlorine, Polyurethane foam, which is inert, hydrophobic, and insoluble in water, adheres well to both dry and wetted soils, sediments, and rocks. It can be formed in situ at a low but controllable, bulk density. At a bulk density of 0.08 g cm(-3), and in combination with acrylic casing, polyurethane foam increases the thermal neutron count by less than 5 percent in a saturated sand, relative to background. In addition to its small effect on the neutron flux, polyurethane foam, unlike bentonite or cement, does not affect the moisture content of the surrounding formation during installation. Furthermore, because it is a closed-cell foam, its moisture content does not change under varying formation moisture conditions. As was shown in related field tests, polyurethane foam is especially well suited for backfilling boreholes in fractured rocks because of its fast set time which minimizes penetration into fractures. The design proved to be convenient and durable under rugged field conditions.
Zawislanski, Peter T.; Faybishenko, Boris
Effects of core excess reactivity and coolant average temperature on maximum operable time of NIRR-1 miniature neutron source reactor
We appraised in this study the effects of core excess reactivity and average coolant temperature on the operable time of the Nigeria Research Reactor-1 (NIRR-1), which is a miniature neutron source reactor (MNSR). The duration of the reactor operating time and fluence depletion under different operation mode as well as change in core excess reactivity with temperature coefficient was investigated
Y. A. Ahmed; I. B. Mansir; I. Yusuf; G. I. Balogun; S. A. Jonah
High Flux Isotope Reactor cold neutron source reference design concept
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.
Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others
Preliminary design study of advanced multistage axial flow core compressors
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.
Wisler, D. C.; Koch, C. C.; Smith, L. H., Jr.
Design of a boron neutron capture enhanced fast neutron therapy assembly
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 multiforme (GBM). A boron neutron capture enhanced fast neutron therapy assembly has been designed for the Fermilab Neutron Therapy Facility (NTF). This assembly uses a tungsten filter and collimator near the patient's head, with a graphite reflector surrounding the head to significantly increase the dose due to boron neutron capture reactions. The assembly was designed using Monte Carlo radiation transport code MCNP version 5 for a standard 20x20 cm2 treatment beam. The calculated boron dose enhancement at 5.7-cm depth in a water-filled head phantom in the assembly with a 5x5 cm2 collimation was 21.9% per 100-ppm 10B for a 5.0-cm tungsten filter and 29.8% for a 8.5-cm tungsten filter. The corresponding dose rate for the 5.0-cm and 8.5-cm thick filters were 0.221 and 0.127 Gy/min, respectively; about 48.5% and 27.9% of the dose rate of the standard 10x10 cm2 fast neutron treatment beam. To validate the design calculations, a simplified BNCEFNT assembly was built using four lead bricks to form a 5x5 cm2 collimator. Five 1.0-cm thick 20x20 cm2 tungsten plates were used to obtain different filter thicknesses and graphite bricks/blocks were used to form a reflector. Measurements of the dose enhancement of the simplified assembly in a water-filled head phantom were performed using a pair of tissue-equivalent ion chambers. One of the ion chambers is loaded with 1000-ppm natural boron (184-ppm 10B) to measure dose due to boron neutron capture. The measured 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 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.
Wang, Zhonglu
Neutronic design of a Liquid Salt-cooled Pebble Bed Reactor (LSPBR)
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)
De Zwaan, S. J.; Boer, B.; Lathouwers, D.; Kloosterman, J. L. [Delft Univ. of Technology (TUD), Mekelweg 15, 2629 JB Delft (Netherlands)
Shield design for next-generation, low-neutron-fluence, superconducting tokamaks
A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of built-up, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components.
Lee, V.D.; Gohar, Y.
Lee, V. D.; Gohar, Y.
Proton vs. neutron captures in the neutrino winds of core-collapse supernovae
Recent one-dimensional (1D) hydrodynamical simulations of core-collapse supernovae (CCSNe) with a sophisticated treatment of neutrino transport indicate the neutrino-driven winds being proton-rich all the way until the end of their activity. This seems to exclude all possibilities of neutron-capture nucleosynthesis, but provide ideal conditions for the ?p-process, in neutrino winds. New 2D explosion simulations of electron-capture supernovae (ECSNe; a subset of CCSNe) exhibit, however, convective neutron-rich lumps, which are absent in the 1D case. Our nucleosynthesis calculations indicate that these neutron-rich lumps allow for interesting production of elements between iron group and N = 50 nuclei (Zn, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, with little Ga). Our models do not confirm ECSNe as sources of the strong r-process (but possibly of a weak r-process up to Pd, Ag, and Cd in the neutron-rich lumps) nor of the ?p-process in the subsequent proton-rich outflows. We further study the ?p-process with semi-analytic models of neutrino winds assuming the physical conditions for CCSNe. We also explore the sensitivities of some key nuclear reaction rates to the nucleosynthetic abundances. Our result indicates that the ?/p-process in CCSNe (other than ECSNe) can be the origin of p-nuclei up to A = 108, and even up to A = 152 in limiting conditions.
Wanajo, S.; Janka, H.-T.; M�ller, B.; Kubono, S.
3D Core-Collapse Supernova Simulations: Neutron Star Kicks and Nickel Distribution
We perform a set of neutrino-driven core-collapse supernova (CCSN) simulations studying the hydrodynamical neutron star kick mechanism in three-dimensions. Our simulations produce neutron star (NS) kick velocities in a range between ~100-600 km/s resulting mainly from the anisotropic gravitational tug by the asymmetric mass distribution behind the supernova shock. This stochastic kick mechanism suggests that a NS kick velocity of more than 1000 km/s may as well be possible. An enhanced production of heavy elements in the direction roughly opposite to the NS recoil direction is also observed as a result of the asymmetric explosion. This large scale asymmetry might be detectable and can be used to constrain the NS kick mechanism.
Wongwathanarat, Annop; Janka, Hans-Thomas; M�ller, Ewald
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.
Design of a tritium breeding blanket for volumetric neutron source
A water-cooled and austenitic stainless-steel-structured breeding blanket system is designed for a volumetric neutron source (VNS), based on a steady-state tokamak device. The designed VNS with super-conducting coils has a 4.5 m main radius and a total of 300 MW in fusion power. It yields the tritium consumption of approximately 10 kg per year with 50% availability. It is unrealistic
Y. Asaoka; Y. Ogawa; K. Okano; N. Inoue; Y. Murakami; K. Tomabechi; T. Yamamoto; T. Yoshida
Linac design study for an intense neutron-source driver
The 1-MW spallation-neutron source under design study at Los Alamos is driven by a linac-compressor-ring scheme that utilizes a large portion of the existing Los Alamos Meson Physics Facility (LAMPF) linac, as well as the facility infrastructure. The project is referred to as the National Center for Neutron Research (NCNR). A second phase of the proposal will upgrade the driver power to 5 MW. A description of the 1-MW scheme is given in this paper. In addition, the upgrade path to the substantial increase of beam power required for the 5 MW scenario is discussed.
Lynch, M.T.; Browman, A.; DeHaven, R.; Jameson, R.; Jason, A.; Neuschaefer, G.; Tallerico, P.; Regan, A.
Monte Carlo Neutronics and Thermal Hydraulics Analysis of Reactor Cores with Multilevel Grids
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.
Bernnat, W.; Mattes, M.; Guilliard, N.; Lapins, J.; Zwermann, W.; Pasichnyk, I.; Velkov, K.
Neutronic design studies for an unattended, low power reactor
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.
Palmer, R. G.; Durkee, J. W., Jr.
Design of active-neutron fuel rod scanner
An active-neutron fuel rod scanner has been designed for the assay of fissile materials in mixed oxide fuel rods. A 252Cf source is located at the center of the scanner very near the through hole for the fuel rods. Spontaneous fission neutrons from the californium are moderated and induce fissions within the passing fuel rod. The rod continues past a combined gamma-ray and neutron shield where delayed gamma rays above 1 MeV are detected. We used the Monte Carlo code MCNP to design the scanner and review optimum materials and geometries. An inhomogeneous beryllium, graphite, and polyethylene moderator has been designed that uses source neutrons much more efficiently than assay systems using polyethylene moderators. Layers of borated polyethylene and tungsten are used to shield the detectors. Large NaI(Tl) detectors were selected to measure the delayed gamma rays. The enrichment zones of a thermal reactor fuel pin could be measured to within 1% counting statistics for practical rod speeds. Applications of the rod scanner include accountability of fissile material for safeguards applications, quality control of the fissile content in a fuel rod, and the verification of reactivity potential for mixed oxide fuels.
Griffith, G. W.; Menlove, H. O.
Characterization of core debris/concrete interactions for the Advanced Neutron Source
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.
Hyman, C.R.; Taleyarkhan, R.P.
Neutron transport analysis for nuclear reactor design
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
The Spallation Neutron Source accelerator systems will provide a 1 GeV, 1.44 MW proton beam to a liquid mercury target for neutron production. Beam tuning dumps are provided at the end of the linac (the Linac Dump) and in the Ring-to-Target transport line (the Extraction Dump) [1]. Thin windows are required to separate the accelerator vacuum from the poor vacuum upstream of the beam dump. There are several challenging engineering issues that have been addressed in the window design. Namely, handling of the high local power density deposited by the stripped electrons from the H-beam accelerated in the linac, and the need for low-exposure removal and replacement of an activated window. The thermal design of the linac dump window is presented, as is the design of a vacuum clamp and mechanism that allows remote removal and replacement of the window.
Coupled full core neutron transport/CFD simulations of pressurized water reactors
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)
Kochunas, B.; Stimpson, S.; Collins, B.; Downar, T. [Dept. of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, 2200 Bonisteel Blvd, Ann Arbor, MI 48104 (United States); Brewster, R.; Baglietto, E. [CD-adapco, 60 Broadhollow Road, Melville, NY 11747 (United States); Yan, J. [Westinghouse Electric Company LLC, Columbia, SC (United States)
Core damage frequency (reactor design) perspectives based on IPE results
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.
Camp, A.L.; Dingman, S.E.; Forester, J.A. [and others
Pre-conceptual design and preliminary neutronic analysis of the proposed National Spallation Neutron Source (NSNS)
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.
Johnson, J.O.; Barnes, J.M.; Charlton, L.A. [Oak Ridge National Lab., TN (United States). Computational Physics and Engineering Div.
Advanced Neutron Source: Plant Design Requirements. Revision 4
The Double Pulsar: Evidence for Neutron Star Formation without an Iron Core-collapse Supernova
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� � 11� � 5�. Furthermore, we find a mean 95% upper limit on the misalignment between its spin and orbital angular momentum axes of 3.�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.
Ferdman, R. D.; Stairs, I. H.; Kramer, M.; Breton, R. P.; McLaughlin, M. A.; Freire, P. C. C.; Possenti, A.; Stappers, B. W.; Kaspi, V. M.; Manchester, R. N.; Lyne, A. G.
Isoscalar-vector interaction and hybrid quark core in massive neutron stars
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-Nambu�Jona-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.
Shao, G. Y.; Colonna, M.; Di Toro, M.; Liu, Y. X.; Liu, B.
The Mystery of CEMPs+r Stars and the Dual Core-Flash Neutron Superburst
Carbon-enhanced metal-poor (CEMPs+r) stars show large enhancements of elements produced both by the slow and the rapid neutron capture processes (the s and r process, respectively) and represent a relatively large fraction, 30% to 50%, of the CEMP population. Many scenarios have been proposed to explain this peculiar chemical composition and most of them involve a binary companion producing the s-process elements during its Asymptotic Giant Branch (AGB) phase. The problem is that none of the proposed explanations appears to be able to account for all observational constraints, hence, alternatives are needed to be put forward and investigated. In this spirit, we propose a new scenario for the formation of CEMPs+r stars based on S. W. Campbell's finding that during the `dual core flash' in low-mass stars of extremely low metallicity, when protons are ingested in the He-flash convective zone, a `neutron superburst' is produced. Further calculations are needed to verify if this neutron superburst could make the r-process component observed in CEMPs+r, as well as their Fe abundances. The s-process component would then be produced during the following AGB phase.
Lugaro, M.; Campbell, S. W.; de Mink, S. E.
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.
Ferdman, R. D.; Kramer, M.; Stappers, B. W.; Lyne, A. G. [School of Physics and Astronomy, University of Manchester, Jodrell Bank Centre for Astrophysics, Alan Turing Building, Oxford Road, Manchester M13 9PL (United Kingdom)] [School of Physics and Astronomy, University of Manchester, Jodrell Bank Centre for Astrophysics, Alan Turing Building, Oxford Road, Manchester M13 9PL (United Kingdom); Stairs, I. H. [Department of Physics and Astronomy, University of British Columbia, Vancouver, BC V6T 1Z1 (Canada)] [Department of Physics and Astronomy, University of British Columbia, Vancouver, BC V6T 1Z1 (Canada); Breton, R. P. [School of Physics and Astronomy, University of Southampton, Highfield, Southampton SO17 1BJ (United Kingdom)] [School of Physics and Astronomy, University of Southampton, Highfield, Southampton SO17 1BJ (United Kingdom); McLaughlin, M. A. [Department of Physics, West Virginia University, Morgantown, WV 26505 (United States)] [Department of Physics, West Virginia University, Morgantown, WV 26505 (United States); Freire, P. C. C. [Max-Planck-Institut fuer Radioastronomie, Auf dem Huegel 69, D-53121 Bonn (Germany)] [Max-Planck-Institut fuer Radioastronomie, Auf dem Huegel 69, D-53121 Bonn (Germany); Possenti, A. [INAF-Osservatorio Astronomico di Cagliari, Loc. Poggio dei Pini, I-09012 Capoterra (Italy)] [INAF-Osservatorio Astronomico di Cagliari, Loc. Poggio dei Pini, I-09012 Capoterra (Italy); Kaspi, V. M. [Department of Physics, McGill University, Ernest Rutherford Physics Building, 3600 University Street, Montreal, QC H3A 2T8 (Canada)] [Department of Physics, McGill University, Ernest Rutherford Physics Building, 3600 University Street, Montreal, QC H3A 2T8 (Canada); Manchester, R. N., E-mail: firstname.lastname@example.com [CSIRO Astronomy and Space Science, Australia Telescope National Facility, Epping, NSW 1710 (Australia)
Natural nuclear reactor at Oklo and variation of fundamental constants: Computation of neutronics of a fresh core
Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities, and neutron fluxes were calculated. We also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of {sub 62}{sup 149}Sm and its dependence on the shift of a resonance position E{sub r} (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73{<=}{delta}E{sub r}{<=}62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant {alpha}. We obtain new, more accurate limits of -4x10{sup -17}{<=}{alpha}{center_dot}/{alpha}{<=}3x10{sup -17} yr{sup -1}. Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress.
Petrov, Yu. V.; Nazarov, A. I.; Onegin, M. S.; Petrov, V. Yu.; Sakhnovsky, E. G. [St. Petersburg Nuclear Physics Institute, Gatchina, RU-188-300, St. Petersburg (Russian Federation)
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/ ?/cm 2-s/cm), -5.2�10 -18 A/R/h/cm (-1.133�10 -23 A/ ?/cm 2-s/cm) and 34�10 -18 A/R/h/cm (7.14�10 -23 A/ ?/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.
Alex, M.; Ghodgaonkar, M. D.
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.
Vujic, Jasmina L. (Lisle, IL)
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.
Vujic, J.L.
Design and Testing of a Position-Sensitive Plastic Scintillator Detector for Fast Neutron Imaging
We describe the design and performance of a position sensitive scintillator detector developed for neutron measurements. Several of these detectors are to be used in the assembly of the Fast Neutron Imaging Telescope (FNIT), an instrument with imaging and energy measurement capabilities, sensitive to neutrons in the 2-20 MeV energy range. FNIT was initially conceived to study solar neutrons as
Ulisse Bravar; Paul J. Bruillard; Erwin O. Fl�ckiger; John R. Macri; Mark L. McConnell; Michael R. Moser; James M. Ryan; Richard S. Woolf
The Spallation Neutron Source (SNS) conceptual design shielding analysis
The shielding design is important for the construction of an intense high-energy accelerator facility like the proposed Spallation Neutron Source (SNS) due to its impact on conventional facility design, maintenance operations, and since the cost for the radiation shielding shares a considerable part of the total facility costs. A calculational strategy utilizing coupled high energy Monte Carlo calculations and multi-dimensional discrete ordinates calculations, along with semi-empirical calculations, was implemented to perform the conceptual design shielding assessment of the proposed SNS. Biological shields have been designed and assessed for the proton beam transport system and associated beam dumps, the target station, and the target service cell and general remote maintenance cell. Shielding requirements have been assessed with respect to weight, space, and dose-rate constraints for operating, shutdown, and accident conditions. A discussion of the proposed facility design, conceptual design shielding requirements calculational strategy, source terms, preliminary results and conclusions, and recommendations for additional analyses are presented.
Johnson, J.O.; Odano, N.; Lillie, R.A.
Shielding Design of the Spallation Neutron Source (SNS)
The shielding design is important for the construction of an intense high-energy accelerator facility like the proposed Spallation Neutron Source (SNS) due to its impact on conventional facility design, maintenance operations, and since the cost for the radiation shielding shares a considerable part of the total facility costs. A calculational strategy utilizing coupled high energy Monte Carlo calculations and multi-dimensional discrete ordinates calculations, along with semi-empirical calculations, was implemented to perform the conceptual design shielding assessment of the proposed SNS. Biological shields have been designed and assessed for the proton beam transport system and associated beam dumps, the target station, and the target service cell and general remote maintenance cell. Shielding requirements have been assessed with respect to weight, space, and dose-rate constraints for operating, shutdown, and accident conditions. A discussion of the proposed facility design, conceptual design shielding requirements, calculational strategy, source terms, preliminary results and conclusions, and recommendations for additional analyses are presented.
Johnson, J.O.
Development of neutron chambers with improved electrode design
A boron-lined neutron chamber has been developed with improved electrode design. The use of suitably machined PEEK spacers at the ends of the electrodes resulted in better performance of the ion chamber. The performance of the ion chamber was compared with a similar detector developed earlier with alumina discs. Neutron sensitivity tests on both types of chambers at Apsara Thermal Column chamber showed a sharper knee in the case of the chamber with PEEK spacers due to well-defined volume and elimination of edge and fringe effects. Tests with a calibrated 60Co Teletherapy source at gamma intensity levels ranging from 900 to 3000 R/h showed an average gamma compensation factor of 3.8%. The corresponding value for chambers with the alumina discs was 11.5%. A plot of the compensation factor as a function of gamma flux, showed a much more pronounced variation in the gamma compensation factor in the earlier design using alumina discs.
Alex, Mary; Prasad, D. N.; Prasad, K. R.; Kataria, S. K.
GEANT4 used for neutron beam design of a neutron imaging facility at TRIGA reactor in Morocco
Neutron imaging has a broad scope of applications and has played a pivotal role in visualizing and quantifying hydrogenous masses in metallic matrices. The field continues to expand into new applications with the installation of new neutron imaging facilities. In this scope, a neutron imaging facility for computed tomography and real-time neutron radiography is currently being developed around 2.0MW TRIGA MARK-II reactor at Maamora Nuclear Research Center in Morocco (Reuscher et al., 1990 [1]; de Menezes et al., 2003 [2]; Deinert et al., 2005 [3]). The neutron imaging facility consists of neutron collimator, real-time neutron imaging system and imaging process systems. In order to reduce the gamma-ray content in the neutron beam, the tangential channel was selected. For power of 250 kW, the corresponding thermal neutron flux measured at the inlet of the tangential channel is around 3�10 11 ncm 2/s. This facility will be based on a conical neutron collimator with two circular diaphragms with diameters of 4 and 2 cm corresponding to L/D-ratio of 165 and 325, respectively. These diaphragms' sizes allow reaching a compromise between good flux and efficient L/D-ratio. Convergent-divergent collimator geometry has been adopted. The beam line consists of a gamma filter, fast neutrons filter, neutron moderator, neutron and gamma shutters, biological shielding around the collimator and several stages of neutron collimator. Monte Carlo calculations by a fully 3D numerical code GEANT4 were used to design the neutron beam line ( http://www.info.cern.ch/asd/geant4/geant4.html[4]). To enhance the neutron thermal beam in terms of quality, several materials, mainly bismuth (Bi) and sapphire (Al 2O 3) were examined as gamma and neutron filters respectively. The GEANT4 simulations showed that the gamma and epithermal and fast neutron could be filtered using the bismuth (Bi) and sapphire (Al 2O 3) filters, respectively. To get a good cadmium ratio, GEANT 4 simulations were used to define the design of the moderator in the inlet of the radiation channel. A graphite block of 22 cm thickness seems to be the optimal neutron moderator. The results showed that the combination of 5 cm of bismuth with 5 cm of sapphire permits the filtration of gamma-rays, epithermal neutrons as well as fast neutrons in a considerable way without affecting the neutron thermal flux.
Ouardi, A.; Machmach, A.; Alami, R.; Bensitel, A.; Hommada, A.
Documented Safety Analysis Addendum for the Neutron Radiography Reactor Facility Core Conversion
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.
Boyd D. Christensen
3D Neutron Transport PWR Full-core Calculation with RMC code
Nowadays, there are more and more interests in the use of Monte Carlo codes to calculate the detailed power density distributions in full-core reactors. With the Inspur TS1000 HPC Server of Tsinghua University, several calculations have been done based on the EDF 3D Neutron Transport PWR Full-core benchmark through large-scale parallelism. To investigate and compare the results of the deterministic method and Monte Carlo method, EDF R&D and Department of Engineering Physics of Tsinghua University are having a collaboration to make code to code verification. So in this paper, two codes are used. One is the code COCAGNE developed by the EDF R&D, a deterministic core code, and the other is the Monte Carlo code RMC developed by Department of Engineering Physics in Tsinghua University. First, the full-core model is described and a 26-group calculation was performed by these two codes using the same 26-group cross-section library provided by EDF R&D. Then the parallel and tally performance of RMC is discussed. RMC employs a novel algorithm which can cut down most of the communications. It can be seen clearly that the speedup ratio almost linearly increases with the nodes. Furthermore the cell-mapping method applied by RMC consumes little time to tally even millions of cells. The results of the codes COCAGNE and RMC are compared in three ways. The results of these two codes agree well with each other. It can be concluded that both COCAGNE and RMC are able to provide 3D-transport solutions associated with detailed power density distributions calculation in PWR full-core reactors. Finally, to investigate how many histories are needed to obtain a given standard deviation for a full 3D solution, the non-symmetrized condensed 2-group fluxes of RMC are discussed.
Qiu, Yishu; She, Ding; Fan, Xiao; Wang, Kan; Li, Zeguang; Liang, Jingang; Leroyer, Hadrien
Initial global 2-D shielding analysis for the Advanced Neutron Source core and reflector
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.
Bucholz, J.A.
Development of a neutronics calculation method for designing commercial type Japanese sodium-cooled fast reactor
Under the R and D project to improve the modeling accuracy for the design of fast breeder reactors the authors are developing a neutronics calculation method for designing a large commercial type sodium- cooled fast reactor. The calculation method is established by taking into account the special features of the reactor such as the use of annular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verification and Validation, and Uncertainty Qualification (V and V and UQ) of the calculation method is being performed by using measured data from the prototype FBR Monju. The results of this project will be used in the design and analysis of the commercial type demonstration FBR, known as the Japanese Sodium fast Reactor (JSFR). (authors)
Takeda, T.; Shimazu, Y.; Hibi, K.; Fujimura, K. [Research Inst. of Nuclear Engineering, Univ. of Fukui, 1cho-me 2gaiku 4, Kanawa-cho, Tsuruga-shi, Fukui 914-0055 (Japan)
Fuel management strategy for the new equilibrium silicide core design of RSG GAS (MPR30)
The design procedure and fuel management strategy are described for converting the oxide core of the 30 MWth RSG GAS (MPR-30) to the new equilibrium silicide core using a higher uranium loading. A procedure to directly search for the equilibrium core has been devised and implemented in an in-core fuel management code developed for RSG GAS. Compared to the present
Liem Peng Hong; Bakri Arbie; T. M. Sembiring; P. Prayoto; R. Nabbi
Physics design of the National Spallation Neutron Source linac
The National Spallation Neutron Source (NSNS) requires a linac that accelerates a H{sup {minus}} beam to 1.0 GeV. The linac starts with a radio-frequency quadrupole (RFQ) accelerator, which is followed by a drift-tube linac (DTL), a coupled-cavity drift-tube linac (CCDTL), and a conventional coupled-cavity linac (CCL). In this paper, the authors focus on the DTL, CCDTL, and CCL parts of the accelerator. They discuss the linac design parameters and beam dynamics issues. The design rationale of no separate matching sections between different accelerating sections maintains the current independence of beam behavior.
Takeda, H.; Billen, J.H.; Nath, S.
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}
Anand, J.D.; Goyal, A.; Gupta, V.K. [Department of Physics and Astrophysics, University of Delhi, Delhi-110 007 (India)] [Department of Physics and Astrophysics, University of Delhi, Delhi-110 007 (India); Singh, S. [Department of Physics and Astrophysics, University of Delhi, Delhi-110 007 (India)] [Department of Physics and Astrophysics, University of Delhi, Delhi-110 007 (India); [Deshbandhu College, University of Delhi-110 019 (India)
Design Core commonalities: A study of the College of Design at Iowa State University
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 that is necessary to all of
Jane E. Venes
A neutronic feasibility study of the AP1000 design loaded with fully ceramic micro-encapsulated fuel
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)
Liang, C.; Ji, W. [Department of Mechanical, Aerospace, and Nuclear Engineering Rensselaer, Polytechnic Institute, 110 8th Street, Troy, NY 12180 (United States)] [Department of Mechanical, Aerospace, and Nuclear Engineering Rensselaer, Polytechnic Institute, 110 8th Street, Troy, NY 12180 (United States)
Design Optimization and the path towards a 2 MW Spallation Neutron Source
The Spallation Neutron Source (SNS) is designed to ultimately reach an average proton beam power of 2 MW for pulsed neutron production. The SNS physics groups analyze the machine performance within the hardware constraints, optimize the accelerator design, and establish the best path towards a 2 MW and higher spallation neutron source.
M. Blaskiewicz; N. Catalan-Lasheras; D. Davino; A. Fedotov; Y. Lee; N. Malitsky; Y. Papaphilippou; D. Raparia; A. Shishlo; N. Tsoupas; J. Wei; W. Weng; S. Zhang; J. Billen; S. Kurennoy; S. Nath; J. Stovall; H. Takeda; L. Young; R. Keller; J. Staples; A. Aleksandrov; Y. Cho; P. Chu; S. Cousineau; V. Danilov; M. Doleans; J. Galambos; J. Holmes; N. Holtkamp; D. Jeon; S. Kim; R. Kustom; E. Tanke; W. Wan; R. Sundelin
78 FR 32988 - Core Principles and Other Requirements for Designated Contract Markets; Correction
Federal Register 2010, 2011, 2012, 2013
...COMMISSION 17 CFR Part 38 RIN 3038-AD09 Core Principles and Other Requirements for Designated...Register release of the final rule regarding Core Principles and Other Requirements for Designated...Register release of the final rule regarding Core Principles and Other Requirements for...
Design trade-off study for a large volume short pulse neutron assembly.
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 p...
P. J. Griffin J. D. Miller G. A. Harms E. J. Parma R. L. Coats
Tokamak Fusion Core Experiment: design studies based on superconducting and hybrid toroidal field coils. Design overview
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.
Flanagan, C.A. (ed.)
A Lightweight Pneumatic Coring Device: Design and Field Test.
A lightweight pneumatic coring device for use from relatively small research vessels was developed and field tested. The device consists of an aluminum frame supporting a core barrel surmounted by a pneumatic industrial vibrator. Tests of a number of pair...
J. A. Fuller E. P. Meisburger
Improved core design of the high temperature supercritical-pressure light water reactor
A new coolant flow scheme has been devised to raise the average coolant core outlet temperature of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530�C.In previous studies, the average coolant core outlet temperature
A. Yamaji; K. Kamei; Y. Oka; S. Koshizuka
Core design studies for a 1000 MW{sub th} advanced burner reactor.
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.
Kim, T. K.; Yang, W. S.; Grandy, C.; Hill, R.; Nuclear Engineering Division
Neutron transport with the method of characteristics for 3-D full core boiling water reactor applications
The Numerical Nuclear Reactor (NNR) is a code suite that is being developed to provide high-fidelity multi-physics capability for the analysis of light water nuclear reactors. The focus of the work here is to extend the capability of the NNR by incorporation of the neutronics module, DeCART, for Boiling Water Reactor (BWR) applications. The DeCART code has been coupled to the NNR fluid mechanics and heat transfer module STAR-CD for light water reactor applications. The coupling has been accomplished via an interface program, which is responsible for mapping the STAR-CD and DeCART meshes, managing communication, and monitoring convergence. DeCART obtains the solution of the 3-D Boltzmann transport equation by performing a series of 2-D modular ray tracing-based method of characteristics problems that are coupled within the framework of 3-D coarse-mesh finite difference. The relatively complex geometry and increased axial heterogeneity found in BWRs are beyond the modeling capability of the original version of DeCART. In this work, DeCART is extended in three primary areas. First, the geometric capability is generalized by extending the modular ray tracing scheme and permitting an unstructured mesh in the global finite difference kernel. Second, numerical instabilities, which arose as a result of the severe axial heterogeneity found in BWR cores, have been resolved. Third, an advanced nodal method has been implemented to improve the accuracy of the axial flux distribution. In this semi-analytic nodal method, the analytic solution to the transverse-integrated neutron diffusion equation is obtained, where the nonhomogeneous neutron source was first approximated by a quartic polynomial. The successful completion of these three tasks has allowed the application of the coupled DeCART/STAR-CD code to practical BWR problems.
Thomas, Justin W.
FabScalar: composing synthesizable RTL designs of arbitrary cores within a canonical superscalar template
A growing body of work has compiled a strong case for the single-ISA heterogeneous multi-core paradigm. A single-ISA heterogeneous multi-core provides multiple, differently-designed superscalar core types that can streamline the execution of diverse programs and program phases. No prior research has addressed the 'Achilles' heel of this paradigm: design and verification effort is multiplied by the number of different core
Prompt-gamma neutron activation analysis system design: Effects of DT versus D-D neutron generator source selection
Prompt-gamma neutron activation (PGNA) analysis is used for the non-invasive measurement of human body composition. Advancements\\u000a in portable, compact neutron generator design have made those devices attractive as neutron sources. Two distinct generators\\u000a are available: D-D with 2.5 MeV and D-T with 14.2 MeV neutrons. To compare the performance of these two units in our present\\u000a PGNA system, we performed
R. J. Shypailo; K. J. Ellis
Designing for safety in the conceptual design of the Advanced Neutron Source
The Advanced Neutron Source is a major new research facility proposed by the Department of Energy for construction over the next six years. The unique set of nuclear safety features selected to give the recently completed conceptual design a high degree of safety are identified and discussed.
Harrington, R.M.; West, C.D.
The object model at the core of the IDEA+ design environment
** The IDEA+ project aims at developing an Integrated Design Environment for Architect designers, in which design tools and computational tests are gathered around and make use of one and the same core object model. This paper focuses on three different aspects in which the IDEA+ core model differs from many other product modelling research initiatives: the systematic approach in
Ann Hendricx; Herman Neuckermans
Core design for use with precision composite reflectors
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.
Porter, Christopher C. (inventor); Jacoy, Paul J. (inventor); Schmitigal, Wesley P. (inventor)
New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions
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 � 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.
Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Br�aud, S.; Oriol, L.; Villard, J.-F.
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.
Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F. [CEA, DEN, Cadarache, SPEx/LDCI, F-13108 Saint-Paul-lez-Durance (France); Vermeeren, L. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Lopez, A. Legrand [CEA, DEN, Saclay, SIREN/LECSI, F-91400 Saclay (France)
New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions.
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 � 10(20) n?cm(2). A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement. PMID:21456734
Geslot, B; Vermeeren, L; Filliatre, P; Lopez, A Legrand; Barbot, L; Jammes, C; Br�aud, S; Oriol, L; Villard, J-F
Design study of a medical proton linac for neutron therapy
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.
Machida, S.; Raparia, D.
Design study of a medical linac for neutron therapy
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 milliamps peak current at 60 Hz and is one meter long. The 18-meter-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. 5 refs., 4 figs., 3 tabs.
Raparia, D.; Machida, S.
Monte Carlo design for a new neutron collimator at the ENEA Casaccia TRIGA reactor
The TRIGA RC-1 1MW reactor operating at ENEA Casaccia Center is currently being developed as a second neutron imaging facility that shall be devoted to computed tomography as well as neutron tomography. In order to reduce the gamma-ray content in the neutron beam, the reactor tangential piercing channel was selected. A set of Monte Carlo simulation was used to design
N Burgio; R Rosa
New Core Design for Use with Precision Composite Reflectors.
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 v...
C. C. Porter P. J. Jacoy W. P. Schmitigal
Optimization Design and Finite Element Analysis of Core Cutter
The hydro-hammer sampler is a new type of sampler compared with traditional ones. An important part of this new offshore sampler is that the structure of the core cutter has a significant effect on penetration and core recovery. In our experiments, a commercial finite element code with a capability of simulating large-strain frictional contact between two or more solid bodies
Pin-lu CAO; Kun YIN; Jian-ming PENG; Jian-lin LIU
Conceptual Design of a Modular Island Core Fast Breeder Reactor \\
A metal fueled modular island core sodium cooled fast breeder reactor concept RAPID-M to improve reactor per- formance and proliferation resistance and to accommodate various power requirements has been demonstrated. The essential feature of the RAPID-M concept is that the reactor core consists of integrated fuel assemblies (IFAs) instead of conventional fuel subassemblies. The RAPID concept enables quick and simplified
Mitsuru KAMBE
Visibility evaluation of a neutron grating interferometer operated with a polychromatic thermal neutron beam
Visibility evaluation of neutron gratings for a polychromatic thermal neutron beam was performed for a neutron grating interferometer. Four sets of neutron gratings designed for neutron wavelengths of 2.0 �, 2.7 �, 3.5 �, and 4.4 � were fabricated and tested to find the neutron grating interferometer setup with maximum visibility. The measurements were carried out at the thermal neutron beamline of the Ex-core Neutron irradiation Facility (ENF) of the High-flux Advanced Neutron Application Reactor. The maximum visibility was obtained at the neutron grating set designed for a neutron wavelength of 2.7 � among the four sets, and the visibility was 9.7%. The experimental data can be the basis for an optimization of the neutron grating interferometer at the thermal neutron beamline, and can be further optimized for neutron dark-field imaging with high spatial resolution and a shorter data acquisition time.
Kim, Jongyul; Lee, Seung Wook; Cho, Gyuseong
Facing Challenges for Monte Carlo Analysis of Full PWR Cores : Towards Optimal Detail Level for Coupled Neutronics and Proper Diffusion Data for Nodal Kinetics
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.
Nuttin, A.; Capellan, N.; David, S.; Doligez, X.; El Mhari, C.; M�plan, O.
Magnetorotational collapse of massive stellar cores to neutron stars: Simulations in full general relativity
We study magnetohydrodynamic (MHD) effects arising in the collapse of magnetized, rotating, massive stellar cores to proto-neutron stars (PNSs). We perform axisymmetric numerical simulations in full general relativity with a hybrid equation of state. The formation and early evolution of a PNS are followed with a grid of 2500�2500 zones, which provides better resolution than in previous (Newtonian) studies. We confirm that significant differential rotation results even when the rotation of the progenitor is initially uniform. Consequently, the magnetic field is amplified both by magnetic winding and the magnetorotational instability (MRI). Even if the magnetic energy EEM is much smaller than the rotational kinetic energy Trot at the time of PNS formation, the ratio EEM/Trot increases to 0.1 0.2 by the magnetic winding. Following PNS formation, MHD outflows lead to losses of rest mass, energy, and angular momentum from the system. The earliest outflow is produced primarily by the increasing magnetic stress caused by magnetic winding. The MRI amplifies the poloidal field and increases the magnetic stress, causing further angular momentum transport and helping to drive the outflow. After the magnetic field saturates, a nearly stationary, collimated magnetic field forms near the rotation axis and a Blandford-Payne type outflow develops along the field lines. These outflows remove angular momentum from the PNS at a rate given by J?�?EEMCB, where ? is a constant of order �0.1 and CB is a typical ratio of poloidal to toroidal field strength. As a result, the rotation period quickly increases for a strongly magnetized PNS until the degree of differential rotation decreases. Our simulations suggest that rapidly rotating, magnetized PNSs may not give rise to rapidly rotating neutron stars.
Shibata, Masaru; Liu, Yuk Tung; Shapiro, Stuart L.; Stephens, Branson C.
Simulation of in-core neutron noise measurements for axial void profile reconstruction in boiling water reactors
A possibility to reconstruct the axial void profile from the simulated in-core neutron noise which is caused by density fluctuations in a Boiling Water Reactor (BWR) heated channel is considered. For this purpose, a self-contained model of the two-phase flow regime is constructed which has quantitatively and qualitatively similar properties to those observed in real BWRs. The model is subsequently used to simulate the signals of neutron detectors induced by the corresponding perturbations in the flow density. The bubbles are generated randomly in both space and time using Monte-Carlo techniques. The axial distribution of the bubble production is chosen such that the mean axial void fraction and void velocity follow the actual values of BWRs. The induced neutron noise signals are calculated and then processed by the standard signal analysis methods such as Auto-Power Spectral Density (APSD) and Cross-Power Spectral Density (CPSD). Two methods for axial void and velocity profiles reconstruction are discussed: the first one is based on the change of the break frequency of the neutron auto-power spectrum with axial core elevation, while the second refers to the estimation of transit times of propagating steam fluctuations between different axial detector positions. This paper summarizes the principles of the model and presents a numerical testing of the qualitative applicability to estimate the required parameters for the reconstruction of the void fraction profile from the neutron noise measurements. (authors)
Dykin, V.; Pazsit, I. [Chalmers Univ. of Technology, Div. of Nuclear Engineering, Dept. of Applied Physics, SE-412 96 Gothenburg (Sweden)
Thermomechanical and neutron lifetime modelling and design of Be pebbles in the neutron multiplier for the LIFE engine
Concept designs for the laser inertial fusion\\/fission energy (LIFE) engine include a neutron multiplication blanket containing Be pebbles flowing in a molten salt coolant. These pebbles must be designed to withstand the extreme irradiation and temperature conditions in the blanket to enable a reliable and cost-effective operation of LIFE. In this work, we develop design criteria for spherical Be pebbles
P. DeMange; J. Marian; M. Caro; A. Caro
Development of Optimized Core Design and Analysis Methods for High Power Density BWRs
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
Shirvan, Koroush
Spallation Neutron Source (SNS) accelerator includes a nominally 1000 MeV, 2 mA average current linac consisting of a radio frequency quadrapole (RFQ), drift tube linac (DTL), coupled cavity linac (CCL), a medium and high beta super conducting (SC) linac, and two buncher cavities for beam transport to the ring. Los Alamos is responsible for the RF systems for all sections of the linac. The SNS linac is a pulsed proton linac and the RF system must support a 1 msec beam pulse at up to a 60 Hz repetition rate. The RFQ and DTL utilize seven, 2.5 MW klystrons and operate at 402.5 MHz. The CCL, SC, and buncher cavities operate at 805 MHz. Six, 5 MW klystrons are utilized for the CCL and buncher cavities while eighty-one 550 kW klystrons are used for the SC cavities. All of the RF hardware for the SNS linac is currently in production. This paper will present details of the RF system-level design as well as specific details of the SNS RF equipment. The design parameters will be discussed. One of the design challenges has been achieving a reasonable cost with the very large number of high-power klystrons. The approaches we used to reduce cost and the resulting design compromises will be discussed.
Equation of state of neutron star cores and spin down of isolated pulsars
We study possible impact of a softening of the equation of state by a phase transition, or appearance of hyperons, on the spin evolution of isolated pulsars. Numerical simulations are performed using exact 2-D simulations in general relativity. The equation of state of dense matter at supranuclear densities is poorly known. Therefore, the accent is put on the general correlations between evolution and equation of state, and mathematical strictness. General conjectures referring to the structure of the one-parameter families of stationary configurations are formulated. The interplay of the back bending phenomenon and stability with respect to axisymmetric perturbations is described. Changes of pulsar parameters in a corequake following instability are discussed, for a broad choice of phase transitions predicted by different theories of dense matter. The energy release in a corequake, at a given initial pressure, is shown to be independent of the angular momentum of collapsing configuration. This result holds for various types of phases transition, with and without metastability. We critically review observations of pulsars that could be relevant for the detection of the signatures of the phase transition in neutron star cores.
Haensel, P.; Zdunik, J. L.
A new paradigm for local-global coupling in whole-core neutron transport.
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.
Lewis, E.; Smith, M.; Palmiotti, G,; Nuclear Engineering Division; Northwestern Univ.; INL
Voltage and frequency island optimizations for many-core\\/networks-on-chip designs
Many-core chips interconnected by networks-on-chip (NoC) are increasingly challenged by the tight power consumption constraints. The concept of voltage and frequency island (VFI) which has been recently introduced for achieving fine-grain core-level power management fits well with an NoC design style. This paper will discuss some recent advancement of VFI optimizations for many-core\\/NoC designs. We will also discuss other research
Wooyoung Jang; Duo Ding; David Z. Pan
Direct access test scheme-design of block and core cells for embedded ASICs
Intel requires the use of a direct-access test scheme in embedded-core or block-based ASIC (application-specific integrated-circuit) designs. This scheme provides for separate testing of individual block or core cells using proven test vectors. The authors discuss the design modifications for block cells with low pin counts, user application blocks, and large cores with high pin counts. The implementation and verification
V. Immaneni; S. Raman
System design considerations for fast-neutron interrogation systems.
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 a...
B. J. Micklich B. P. Curry C. L. Fink D. L. Smith T. J. Yule
This paper aims to present the design aspects of electrical motors with soft magnetic composite (SMC) core. Combined classical and modern analysis procedures are proposed for developing SMC motors. A permanent magnet claw pole motor using SMC material as the stator core was firstly designed by the equivalent magnetic circuit method. Three- dimensional finite element magnetic field analysis was conducted
Y. G. Guo; J. G. Zhu; W. Wu
Spallation neutron source cryomodule heat loads and thermal design
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.
E. F. Daly; V. Ganni; C. H. Rode; W. J. Schneider; K. M. Wilson; M. A. Wiseman
Design of 6Mev linear accelerator based pulsed thermal neutron source: FLUKA simulation and experiment.
The 6MeV LINAC based pulsed thermal neutron source has been designed for bulk materials analysis. The design was optimized by varying different parameters of the target and materials for each region using FLUKA code. The optimized design of thermal neutron source gives flux of 3�10(6)ncm(-2)s(-1) with more than 80% of thermal neutrons and neutron to gamma ratio was 1�10(4)ncm(-2)mR(-1). The results of prototype experiment and simulation are found to be in good agreement with each other. PMID:21908197
Patil, B J; Chavan, S T; Pethe, S N; Krishnan, R; Bhoraskar, V N; Dhole, S D
Error Assessment of Homogenized Cross Sections Generation for Whole Core Neutronic Calculation
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.
Hursin, Mathieu; Kochunas, Brendan; Downar, Thomas J.
On a simplified effective neutron-proton interaction including core-polarization effects for odd-odd deformed nuclei
It is well known that the description of odd-odd deformed nuclei is not satisfactory if the usual one-parameter delta-force is used as the residual neutron-proton interaction. In this letter we show that, starting with a simple delta force, an improvement can be obtained which takes into account polarization effects of the core, as well as possible tensor and spin-orbit force
J. P. Boisson; W. Ogle; R. Piepenbring
Performance of prompt- and delayed-responding self-powered in-core neutron detectors in a pressurized water reactor
An assembly of self-powered in-core neutron detectors has been tested for 6 yr over four fuel cycles in the Oconee 2 pressurized water reactor. The assembly contained both prompt-responding ytterbium and delayed-responding rhodium detectors. Two ytterbium detectors were paired with two rhodium detectors in the assembly. The experiment was conducted to define the long-term performance characteristics of the ytterbium detectors.
H. D. Warren; M. F. Sulcoski
Design and construction of a facility for neutron activation analysis using the 14 MeV neutron generator at HMS Sultan
The potential for using a small, sealed tube, DT neutron generator for neutron activation analysis has been well documented\\u000a but not well demonstrated, except for 14 MeV activation analysis. This paper describes the design, construction and characterization\\u000a of a neutron irradiation facility incorporating a small sealed tube DT neutron generator producing 14 MeV neutrons with fluence\\u000a rates of 2�108 s?1
S. E. Jarman; J. Pinchin; J. M. Brushwood; T. McCarthy; M. Bray; P. A. Beeley
Design of an Aluminum Proton Beam Window for the Spallation Neutron Source
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.
Janney, Jim G [ORNL; McClintock, David A [ORNL
Risk minimization using PRA (probabilistic risk analysis) in preconceptual design for the advanced neutron source reactor
The advanced neutron source (ANS) reactor being designed at Oak Ridge National Laboratory (ORNL) for the 1990s will be the worlds best source of low-energy neutrons for materials studies, physics research, transplutonium production, and radiation effects. Probabilistic risk analysis (PRA) is one of the tools being used for safety and operational optimization. The ANS is currently in preconceptual design. This
R. Fullwood; W. Shier
Alternative Design Concepts for the ITER Core Ion-Temperature Diagnostics
Measurements of the ion temperature and plasma flow velocities in the ITER core must be made with spatial and temporal resolutions of 10 cm and 10 ms, respectively, over the range r/a=0-0.85. These requirements can be met by Doppler measurements of the spectral lines of neon- or helium-like, ion of tungsten, iron, and krypton with high-resolution x-ray imaging crystal spectrometers consisting of one spherically bent crystal and an array of two-dimensional pixilated detectors in a Johann configuration. One detector dimension displays spectral information and the other displays spatial information in a direction perpendicular to the toroidal magnetic field. It is challenging to implement this type of spectrometer on ITER due to neutron and gamma streaming through the viewing apertures. This paper discusses the feasibility of two alternative design concepts: a spectrometer with two concentric, spherically bent (convex and concave) crystals and a new von Hamos type spectrometer with one spherically bent (concave) crystal.
Bitter, Manfred; Delgado-Aparicio, Luis; Efthimion, Philip; Feder, Russell; Hill, Kenneth; Johnson, David; Pablant, Novimir; Stratton, Brent; Young, Kenneth; Beiersdorfer, Peter; Wang, Eric; Barnsley, Robin
Results of a Neutronic Simulation of HTR-Proteus Core 4.2 Using PEBBED and Other INL Reactor Physics Tools: FY09 Report.
The Idaho National Laboratorys 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-...
H. D. Gougar
Teaching to the Common Core by Design, Not Accident
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�
Phillips, Vicki; Wong, Carina
Core Curriculum Analysis: A Tool for Educational Design
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�
Levander, Lena M.; Mikkola, Minna
Monte Carlo design for a new neutron collimator at the ENEA Casaccia TRIGA reactor.
The TRIGA RC-1 1MW reactor operating at ENEA Casaccia Center is currently being developed as a second neutron imaging facility that shall be devoted to computed tomography as well as neutron tomography. In order to reduce the gamma-ray content in the neutron beam, the reactor tangential piercing channel was selected. A set of Monte Carlo simulation was used to design the neutron collimator, to determine the preliminary choice of the materials to be employed in the collimator design. PMID:15246415
Burgio, N; Rosa, R
Pressurized water reactor steam line break analysis by means of coupled three-dimensional neutronic, three-dimensional core thermohydraulic, and fast running system codes
The steam line break (SLB) accident in pressurized water reactors is characterized by a large asymmetric cooling of the core, asymmetric stuck control rods, and large primary coolant flow variations. Because of these space- and time-dependent neutronic and thermal-hydraulic conditions in the core, former SLB analyses that used simplified core models were usually performed with many conservative assumptions. To clarify the complicated behavior of the core, the three-dimensional neutronic code CRONOS-2, the three-dimensional core thermal-hydraulic code FLICA-4, and the system code FLICA-S are completely coupled. The results obtained from the coupled codes indicate that the local thermal-hydraulic feedback effects are important in mitigating neutronic power excursions during SLBs.
Paik, H.J.; Raymond, P. (French Atomic Energy Commission, Gif-sur-Yvette (France))
Neutron micro-beam design simulation by Monte Carlo
Over the last two decades neutron micro-beam has increasingly been developing in view of various applications in molecular activation analysis, micro-radiography in space and aviation and in radiation induced bystander effects in bio-cells. In this paper the structure and simulation of a neutron micro-beam is presented. The collimator for micro-beam is made of a polyethylene cylinder with a small hole along the centerline of the cylinder. The hole is filled with very thin needles in triangular or rectangular arrangement. The neutron source was reactor neutrons or a spontaneous Cf-252 neutron source falling on the top side of the collimator. The outgoing thermal and epithermal neutron fluxes were calculated.
Pazirandeh, Ali; Taheri, Ali
Neutronic design of the APT Target/Blanket
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.
Pitcher, E.J.; Russell, G.J.; Kidman, R.B.; Ferguson, P.D. [and others
Neutron transport benchmark examples with web-based AGENT
The AGENT (Arbitrary GEometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two- or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) mathematical theory
Tatjana Jevremovic; Shanjie Xiao; Nader Satvat; Godfree Gert; John Hopkins
Charged-partricle and neutron-capture processes in the high-entropy wind of core-collapse supernovae.
The astrophysical site of the r-process is still uncertain, and a full exploration of the systematics of this process in terms of its dependence on nuclear properties from stability to the neutron drip-line within realistic stellar environments has still to be undertaken. Sufficiently high neutron-to-seed ratios can only be obtained either in very neutron-rich low-entropy environments or moderately neutron-rich high-entropy environments, related to neutron star mergers (or jets of neutron star matter) and the high-entropy wind of core-collapse supernova explosions. As chemical evolution models seem to disfavor neutron star mergers, we focus here on high-entropy environments characterized by entropy S, electron abundance Y{sub e}, and expansion velocity V{sub exp}. We investigate the termination point of charged-particle reactions, and we define a maximum entropy S{sub final} for a given V{sub exp} and Y{sub e}, beyond which the seed production of heavy elements fails due to the very small matter density. We then investigate whether an r-process subsequent to the charged-particle freeze-out can in principle be understood on the basis of the classical approach, which assumes a chemical equilibrium between neutron captures and photodisintegrations, possibly followed by a {beta}-flow equilibrium. In particular, we illustrate how long such a chemical equilibrium approximation holds, how the freeze-out from such conditions affects the abundance pattern, and which role the late capture of neutrons originating from {beta}-delayed neutron emission can play. Furthermore, we analyze the impact of nuclear properties from different theoretical mass models on the final abundances after these late freeze-out phases and {beta}-decays back to stability. As only a superposition of astrophysical conditions can provide a good fit to the solar r-abundances, the question remains how such superpositions are attained, resulting in the apparently robust r-process pattern observed in low metallicity stars.
Farouqi, K.; Kratz, K.-L.; Pfeiffer, B.; Rauscher, T.; Thielemann, F.-K.; Truran, J. W.; Physics; Univ. of Chicago; Joint Inst. for Nuclear Astrophysics; Univ. Mainz; Virtual Inst. for Nuclear Structure and Astrophysics; Max-Planck-Inst. fur Chemie; Univ. of Basel
Charged-particle and neutron-capture processes in the high-entropy wind of core-collapse supernovae.
Farouqi, K.; Kratz, K.-L.; Pfeiffer, B.; Rauscher, T.; Thielemann, F.-K.; Truran, J.W.; Physics; Univ. of Chicago; Joint Inst. for Nuclear Astrophysics; Univ. Mainz; Virtual Inst. for Nuclear Structure and Astrophysics; Max-Planck-Insti. fur Chemie; Univ. of Basel
Design of a Large-Area Fast Neutron Directional Detector
A large-area fast-neutron double-scatter directional detector and spectrometer is being constructed using 1-meter-long plastic scintillator paddles with photomultiplier tubes at both ends. The scintillators detect fast neutrons by proton recoil and also gamma rays by Compton scattering. The paddles are arranged in two parallel planes so that neutrons can be distinguished from muons and gamma rays by time of flight
Peter E. Vanier; Leon Forman; Cynthia Salwen; Istvan Dioszegi
Preliminary Design of Neutron Flux and Spectrum Diagnostics in NT-TBM
A special neutron diagnostic system is proposed that facilitates the measurement of neutron fluxes and spectra in the neutronics and tritium production-test blanket module (NT-TBM) without interrupting the operation of the International Thermal-nuclear Experimental Reactor (ITER), for studying the multiplication rate in the neutron multiplier and breeding ratio of tritium in the breeder. This system includes an encapsulated foil activation system, micro-fission chamber detectors (MFC), and a compact neutron spectrometer using a natural diamond detector (NDD). A helium coolant loop with a reasonable diameter is designed carefully for every measurement channel that ensures that the neutron detectors and preamplifiers would work well under a high temperature scenario and that the filling rates of the neutron multiplier (beryllium pebble) and tritium breeder material (Li4SiO4) would not decrease excessively (the expected value>=80%) due to the dimensions of the helium coolant loop.
Yang, Jinwei; Feng, Kaiming; Cheng, Zhi
Iron-core superconducting magnet design and test results for Maglev application
Design and test results are presented for a superconducting electromagnet for levitating and propelling Maglev vehicles at high velocities. A U-shaped iron core carries a superconducting magnet around its back leg and a normal control coil around each leg of the U-core. The open side of the U-core is bridged by an iron rail through a large airgap between the
S. Kalsi; M. Proise; T. Schultheiss; B. Dawkins; K. Herd
The performance of 3500 MWth homogeneous and heterogeneous metal fueled core designs
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.
Turski, R.; Yang, Shi-tien
WCDMA multiprocessor on chip: design methodology using soft IP cores
The implementation of the physical layer of W-CDMA on embedded devices requires optimizing the resources required due to the limited space and energy allowed. Although general purpose processors will eventually be embedded they are still lacking performance and more importantly they are not tailored to the computation requirements. We propose ion this paper a methodology based on multiobjective genetic algorithms to tailor soft IP processor cores for the purpose of embedding W-CDMA.
Ghali, K.; Hammami, Omar
Effective Optimistic-Checker Tandem Core Design through Architectural Pruning
Design complexity is rapidly becoming a limiting fac- tor in the design of modern, high-performance micro- processors. This paper introduces an optimization tech- nique to improve the efficiency of complex processors. Us- ing a new metric (�Utilization), the designer can identify infrequently-used functionality which contributes little to performance and then systematically \\
Francisco J. Mesa-martinez; Jose Renau
Design complexity is rapidly becoming a limiting factor in the design of modern, high-performance microprocessors. This paper introduces an optimization technique to improve the efficiency of complex processors. Using a new metric (^Utilization), the designer can identify infrequently-used functionality which contributes little to performance and then systematically \\
F. J. Mesa-Martinez; J. Renau
A practical neutron shielding design based on data-base interpolation
Neutron shielding design is an important part of the construction of nuclear reactors and high-energy accelerators. Neutron shielding design is also indispensable in the packaging and storage of isotopic neutron sources. Most efforts in the development of neutron shielding design have been concentrated on nuclear reactor shielding because of its huge mass and strict requirement of accuracy. Sophisticated computational tools, such as transport and Monte Carlo codes and detailed data libraries have been developed. In principle, now, neutron shielding, in spite of its complexity, can be designed in any detail and with fine accuracy. However, in most practical cases, neutron shielding design is accomplished with simplified methods. Unlike practical gamma-ray shielding design, where exponential attenuation coupled with buildup factors has been applied effectively and accurately, simplified neutron shielding design, either by using removal cross sections or by applying charts or tables of transmission factors such as the National Council on Radiation Protection and Measurements (NCRP) 38 (Ref. 1) for general neutron protection or to NCRP 51 (Ref. 2) for accelerator neutron shielding, is still very primitive and not well established. The available data are limited in energy range, materials, and thicknesses, and the estimated results are only roughly accurate. It is the purpose of this work to establish a simple, convenient, and user-friendly general-purpose computational tool for practical preliminary neutron shielding design that is reasonably accurate. A wide-range (energy, material, and thickness) data base of dose transmission factors has been generated by applying one-dimensional transport calculations in slab geometry.
Jiang, S.H.; Sheu, R.J. (National Tsing-Hua Univ., Hsinchu (Taiwan, Province of China))
Double core evolution. 7: The infall of a neutron star through the envelope of its massive star companion
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.
Terman, James L.; Taam, Ronald E.; Hernquist, Lars
Design and analysis of a thermal core for a high performance light water reactor
The High Performance Light Water Reactor is a Generation IV light water reactor concept, operated at a supercritical pressure of 25MPa with a core outlet temperature of 500�C. A thermal core design for this reactor has been worked out by a consortium of Euratom member states within the 6th European Framework Program. Aiming at peak cladding temperatures of less than
T. Schulenberg; C. Mar�czy; J. Heinecke; W. Bernnat
2ND Reactor Core of the NS Otto Hahn. Design, Operation Experience, Developments.
Details of the design of the 2nd reactor core are given, followed by a brief report summarising the operating experience gained with this 2nd core, as well as by an evaluation of measured data and statements concerning the usefulness of the knowledge gain...
H. J. Manthey H. Kracht
Conceptual design studies of GDT-based neutron source
An analysis shows that the existing nuclear technology data bases are not sufficient to arrive at a DEMO with reasonable confidence in achieving target availability levels. In order to provide these data, a testing facility capable of simulating neutron environment in a fusion reactor is needed. A number of proposals for plasma-type neutron sources have been made recently to meet
A. A. Ivanov; E. P. Kruglyakov; Yu. A. Tsidulko; V. G. Krasnoperovt; V. V. Korshakovt
Modeling and Design of a Gadolinium Based Neutron Detector
The goal is to measure the partial cross sections for (gamma, xn) and (gamma, f) on the actinides with the ultimate objective of developing and refining a method for the gamma-ray interrogation of fissionable material. These measurements require the construction of a highly segmented neutron detector to record multiple neutrons emitted following photodisintegration and fission. Each detector segment will use
David Ticehurst; Jeromy Tompkins; Hugon Karwowski
Results and Interpretation of Noise Measurements Using in-Core Self Powered Neutron Detector Strings at Unit 2 of the Paks Nuclear Power Plant.
In-core neutron noise and fuel assembly outlet temperature noise measurements were performed at Unit 2 of Paks Nuclear Power Plant. Characteristics of the reactor and the noise measuring equipment are briefly described. The in-core Rhodium emitter selfpow...
O. Gloeckler G. Por J. Valko
Junction device between the delivery duct of a primary pump and a duct joined to the core support of a fast neutron nuclear reactor
A junction device between the delivery duct of a primary pump and a duct joined to the core support of a liquid metal cooled fast neutron nuclear reactor comprises a frusto-conical sleeve widening out towards the inlet of the duct joined to the core support, a connecting member joined to the sleeve, a sealing device interposed between the outlet end
M. Thevenin; G. Jullien
Explication of the Graphite Structural Design Code of core components for the High Temperature Engineering Test Reactor.
The integrity evaluation of the core graphite components for the High Temperature Engineering Test Reactor (HTTR) will be carried out based upon the Graphite Structural Design Code for core components. In the application of this design code, it is necessa...
T. Iyoku M. Ishihara J. Toyota S. Shiozawa
Neutronics and Thermal Hydraulic Coupling Methods for the Nuclear Reactor Core
In this paper an overview of the coupling methodologies between reactor physics (Neutronics) and thermal hydraulic have been studied by taken into account the previous and ongoing research. The current study includes the neutronics and thermal hydraulic model and the coupling methodology between the two phenomenons which is inter related with computational approach. Since the coupling involves some computer codes
Khan Salah Ud-Din; Minjun Peng; Muhammad Zubair
Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor
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).
B. Boer; A. M. Ougouag
Core Design and Operation Optimization Methods Based on Time-Dependent Perturbation Theory.
A general approach for the optimization of nuclear reactor core design and operation is outlined; it is based on two cornerstones: a newly developed time-dependent (or burnup-dependent) perturbation theory for nonlinear problems and a succesive iteration ...
E. Greenspan
76 FR 14825 - Core Principles and Other Requirements for Designated Contact Markets
...3038-AD09 Core Principles and Other Requirements for Designated Contact Markets AGENCY: Commodity Futures Trading Commission. ACTION...site at http://www.cftc.gov. FOR FURTHER INFORMATION CONTACT: Nancy Markowitz, Assistant Deputy Director,...
The 'virtual density' principle of neutronics: Toward rapid computation of reactivity effects in practical core distortion scenarios
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)
Reed, M.; Smith, K.; Forget, B. [Massachusetts Institute of Technology - MIT, 77 Massachusetts Ave., Cambridge, MA 02139 (United States)] [Massachusetts Institute of Technology - MIT, 77 Massachusetts Ave., Cambridge, MA 02139 (United States)
Rapid design of DSP ASIC cores using hierarchical VHDL libraries
Methods are presented for the rapid design of DSP ASICs based on the use of a series of hierarchical VHDL libraries which are portable across many silicon foundries. These allows complex DSP silicon systems to be developed in a small fraction of the time normally required. Resulting designs are highly competitive with those developed using more conventional methods. The approach
J. V. McCanny; Y. Hu; T. J. Ding; D. Trainor; D. Ridge
An integrated design of SM2C core motor for vehicular applications
This paper focuses on the design and the evaluation of a three-phase permanent magnet synchronous machine (PMSM) with a stator core made of Soft Magnetic Mouldable Composite - SM2C. A radial flux PM machine with modular winding is designed for a high-speed low-voltage automotive application. Manufacturability and a high winding fill factor are vital for a low permeability stator core
Avo Reinap; Conny H�gmark; Mats Alak�la; Tord Cedell; Mats Andersson
Architecting voltage islands in core-based system-on-a-chip designs
Voltage islands enable core-level power optimization for System-on-Chip (SoC) designs by utilizing a unique supply voltage for each core. Architecting voltage islands involves island partition creation, voltage level assignment and floorplanning. The task of island partition creation and level assignment have to be done simultaneously in a floorplanning context due to the physical constraints involved in the design process. This
Jingcao Hu; Youngsoo Shin; Nagu R. Dhanwada; Radu Marculescu
What makes a protein a protein? Hydrophobic core designs that specify stability and structural properties
Here we describe how the systematic redesign of a protein's hydrophobic core alters its structure and stability. We have repacked the hydrophobic core of the four-helix-bundle protein, Rop, with altered packing patterns and various side chain shapes and sizes. Several designs reproduce the structure and native-like properties of the wild- type, while increasing the thermal stability. Other designs, either with
Mary Munson; Suganthi Balasubramanian; Karen G. Fleming; Athena D. Nagi; Ronan O'Brien; Julian M. Sturtevant; Lynne Regan
Shielding analysis and design of the KIPT experimental neutron source facility of Ukraine
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
Z. Zhong; M. Y. A. Gohar; D. Naberezhnev; J. Duo
The Spallation Neutron Source (SNS) is a second generation pulsed neutron source and is presently in the fourth year of a seven-year construction cycle at Oak Ridge National Laboratory. A collaboration of six national laboratories (ANL, BNL, LANL, LBNL, ORNL, TJNAF) is responsible for the design and construction of the various subsystems. The operation of the facility will begin in
Use of albedo for neutron reflector regions in reactor core 3-D simulations
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.
Mohanakrishnan, P.
Design and Evaluation of an Enhanced In-Vessel Core Catcher
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.
Joy L. Rempe
Design and fabrication of ultra-low crosstalk and low-loss multi-core fiber
We designed and fabricated a multi-core fiber (MCF) in which seven identical trench-assisted pure-silica cores were arranged hexagonally. To design MCF, the relation among the crosstalk, fiber parameters, and fiber bend was derived using a new approximation model based on the coupled-mode theory with the equivalent index model. The mean values of the statistical distributions of the crosstalk were observed to be extremely low and estimated to be less than -30 dB even after 10,000-km propagation because of the trench-assisted cores and utilization of the fiber bend. The attenuation of each core was very low for MCFs (0.175--0.181 dB/km at 1550 nm) because of the pure-silica cores. Both the crosstalk and attenuation values are the lowest achieved in MCFs.
Hayashi, Tetsuya; Taru, Toshiki; Shimakawa, Osamu; Sasaki, Takashi; Sasaoka, Eisuke
Design of large-core single-mode Yb3+-doped photonic crystal fiber
The effective index of the cladding fundamental space-filling mode in photonic crystal fiber (PCF) is simulated by the effective index method. The variation of the effective index with the structure parameters of the fiber is achieved. For the first time, the relations of the V parameter of Yb3+-doped PCF with the refractive index of core and the structure parameters of the fiber are provided. The single-mode characteristics of large-core Yb3+-doped photonic crystal fibers with 7 and 19 missing air holes in the core are analyzed. The large-core single-mode Yb3+-doped photonic crystal fibers with core diameters of 50 ?m, 100 ?m and 150 ?m are designed. The results provide theory instruction for the design and fabrication of fiber.
Zhao, Xing-tao; Zheng, Yi; Liu, Xiao-xu; Zhou, Gui-yao; Liu, Zhaolun; Hou, Lan-tian
Design of low-loss and highly birefringent hollow-core photonic crystal fiber.
A practical hollow-core photonic crystal fiber design suitable for attaining low-loss propagation is analyzed. The geometry involves a number of localized elliptical features positioned on the glass ring that surrounds the air core and separates the core and cladding regions. The size of each feature is tuned so that the composite core-surround geometry is antiresonant within the cladding band gap, thus minimizing the guided mode field intensity both within the fiber material and at material/air interfaces. A birefringent design, which involves a 2-fold symmetric arrangement of the features on the core-surround ring, gives rise to wavelength ranges where the effective index difference between the polarization modes is larger than 10(-4). At such high birefringence levels, one of the polarization modes retains favorable field exclusion characteristics, thus enabling low-loss propagation of this polarization channel. PMID:19529102
Roberts, P J; Williams, D P; Sabert, H; Mangan, B J; Bird, D M; Birks, T A; Knight, J C; Russell, P St J
Design and fabrication of ultra-low crosstalk and low-loss multi-core fiber.
We designed and fabricated a multi-core fiber (MCF) in which seven identical trench-assisted pure-silica cores were arranged hexagonally. To design MCF, the relation among the crosstalk, fiber parameters, and fiber bend was derived using a new approximation model based on the coupled-mode theory with the equivalent index model. The mean values of the statistical distributions of the crosstalk were observed to be extremely low and estimated to be less than -30 dB even after 10,000-km propagation because of the trench-assisted cores and utilization of the fiber bend. The attenuation of each core was very low for MCFs (0.175-0.181 dB/km at 1550 nm) because of the pure-silica cores. Both the crosstalk and attenuation values are the lowest achieved in MCFs. PMID:21935022
Development and Design of the EPR� Core Catcher
\\u000a The EPR� is an evolutionary pressurized water reactor in the thermal range of 4,500 MWth, designed and marketed by AREVA NP.\\u000a Currently, there are four EPR� plants under design and construction: Olkiluoto-3 (OL3) in Finland, Flamanville-3 (FA3) in\\u000a France, and Taishan 1&2 (TSN) in the People�s Republic of China.
Dietmar Bittermann; Manfred Fischer
Development and applications of methodologies for the neutronic design of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR)
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.
Fratoni, Massimiliano
Spatial and spectral characteristics of a compact system neutron beam designed for BNCT facility.
The development of suitable neutron sources and neutron beam is critical to the success of Boron Neutron Capture Therapy (BNCT). In this work a compact system designed for BNCT is presented. The system consists of (252)Cf fission neutron source and a moderator/reflector/filter/shield assembly. The moderator/reflector/filter arrangement has been optimized to maximize the epithermal neutron component which is useful for BNCT treatment of deep seated tumors with the suitably low level of beam contamination. The MCMP5 code has been used to calculate the different components of neutrons, secondary gamma rays originating from (252)Cf source and the primary gamma rays emitted directly by this source at the exit face of the compact system. The fluence rate distributions of such particles were also computed along the central axis of a human head phantom. PMID:19168369
Ghassoun, J; Chkillou, B; Jehouani, A
Characterization of core debris/concrete interactions for the Advanced Neutron Source. ANS Severe Accident Analysis Program
The goal is to measure the partial cross sections for (?, xn) and (?, f) on the actinides with the ultimate objective of developing and refining a method for the ?-ray interrogation of fissionable material. These measurements require the construction of a highly segmented neutron detector to record multiple neutrons emitted following photodisintegration and fission. Each detector segment will use a gadolinium-loaded liquid organic scintillator optically coupled to a photomultiplier tube. Gadolinium has one of the highest thermal neutron capture cross sections (49 kb), therefore its presence in the detector will greatly enhance neutron detection efficiency. A prototype detector of 3 L volume has been constructed and modeled using Geant4, a Monte Carlo based program. The detector model and conclusions developed from it along with the results from testing the prototype detector will be presented. S.Agostinelli et al., Nucl. Instr. and Meth. A506, 250, (2003).
Ticehurst, David; Tompkins, Jeromy; Karwowski, Hugon
Transmission of very slow neutrons through material foils and its influence on the design of ultracold neutron sources
At the Paul Scherrer Institute (PSI), a very intense source of ultracold neutrons (UCN) is being built. The UCN converter of solid deuterium must be contained in a vessel. Produced UCN leave that vessel through its top lid. To decide on the design of the vessel and the top lid, we have measured the transmission of neutrons with velocities between 3 and 20 m/s through different material foils. Contrary to expectations, we found that transmission through aluminium and aluminium alloys is equal or even higher compared to zirconium and reactor-grade zirconium alloys, respectively.
Atchison, F.; Blau, B.; Bollhalder, A.; Daum, M.; Fierlinger, P.; Geltenbort, P.; Hampel, G.; Kasprzak, M.; Kirch, K.; K�chli, S.; Kuczewski, B.; Leber, H.; Locher, M.; Meier, M.; Ochse, S.; Pichlmaier, A.; Plonka, C.; Reiser, R.; Ulrich, J.; Wang, X.; Wiehl, N.; Zimmer, O.; Zsigmond, G.
Beam dump window design for the Spallation Neutron Source
The Spallation Neutron Source accelerator systems will provide a 1 GeV, 1.44 MW proton beam to a liquid mercury target for neutron production. Beam tuning dumps are provided at the end of the linac (the Linac Dump) and in the Ring-to-Target transport line (the Extraction Dump). Thin windows are required to separate the accelerator vacuum from the poor vacuum upstream
G. Murdoch; A. Decarlo; S. Henderson; S. Kim; K. Potter; T. Roseberry; J. Rank; D. Raparia
Design Verification Report Neutron Radiography Facility (NRF) TRIGA Fuel Storage Systems
This report outlines the methods, procedures, and outputs developed during the Neutron Radiography Facility (NRF) Training, Research and Isotope Production, General Atomics (TRIGA) fuel storage system design and fabrication.
Design Constrution and Testing of a Self-Powered Neutron Detector.
The design, construction and testing of a self-powered neutron detector (SPN) and associated electronics are described. Several tests were performed giving information about dielectrical properties od detector and cable, gamma spectra induced in the detec...
R. F. Correa
Design, Construction and Testing of a Self-Powered Neutron Detector.
The design, construction and testing of a self-powered neutron detector (SPN) and associated electronics are described. Several tests were performed giving information about dielectrical properties of detector and cable, gamma spectra induced in the detec...
Design, construction, and demonstration of a neutron beamline and a neutron imaging facility at a Mark-I TRIGA reactor
The fleet of research and training reactors is aging, and no new research reactors are planned in the United States. Thus, there is a need to expand the capabilities of existing reactors to meet users' needs. While many research reactors have beam port facilities, the original design of the United States Geological Survey TRIGA Reactor (GSTR) did not include beam ports. The MInes NEutron Radiography (MINER) facility developed by this thesis and installed at the GSTR provides new capabilities for both researchers and students at the Colorado School of Mines. The facility consists of a number of components, including a neutron beamline and beamstop, an optical table, an experimental enclosure and associated interlocks, a computer control system, a multi-channel plate imaging detector, and the associated electronics. The neutron beam source location, determined through Monte Carlo modeling, provides the best mixture of high neutron flux, high thermal neutron content, and low gamma radiation content. A Monte Carlo n-Particle (MCNP) model of the neutron beam provides researchers with a tool for designing experiments before placing objects in the neutron beam. Experimental multi-foil activation results, compared to calculated multi-foil activation results, verify the model. The MCNP model predicts a neutron beamline flux of 2.2*106 +/- 6.4*105 n/cm2-s based on a source particle rate determined from the foil activation experiments when the reactor is operating at a power of 950 kWt with the beam shutter fully open. The average cadmium ratio of the beamline is 7.4, and the L/D of the neutron beam is approximately 200+/-10. Radiographs of a sensitivity indicator taken using both the digital detector and the transfer foil method provide one demonstration of the radiographic capabilities of the new facility. Calibration fuel pins manufactured using copper and stainless steel surrogate fuel pellets provide additional specimens for demonstration of the new facility and offer a comparison between digital and film radiography at the new facility. Comparison of the radiographs taken by the two methods reveals that the digital detector does not produce high quality images when compared to film radiography.
Craft, Aaron E.
The convex hull of two core capacitated network design problems
The network loading problem (NLP) is a specialized capacitated network design problem in which prescribed point-to-point demand between various pairs of nodes of a network must be met by installing (loading) a capacitated facility. We can load any number of units of the facility on each of the arcs at a specified arc dependent cost. The problem is to determine
Thomas L. Magnanti; Prakash Mirchandani; Rita Vachani
Reactor Physics Parameters of Alternate Fueled FBR Core Designs.
Nuclear non-proliferation considerations have resulted in renewed interest in the thorium fuel cycle. Reactor physics parameters of a typical 1200-MW(e) Fast Breeder Reactor (FBR) design were compared when U-233 is substituted for Pu as a fissile fuel and...
D. R. Haffner, R. W. Hardie
Modified Y-TZP Core Design Improves All-ceramic Crown Reliability
This study tested the hypothesis that all-ceramic core-veneer system crown reliability is improved by modification of the core design. We modeled a tooth preparation by reducing the height of proximal walls by 1.5 mm and the occlusal surface by 2.0 mm. The CAD-based tooth preparation was replicated and positioned in a dental articulator for core and veneer fabrication. Standard (0.5 mm uniform thickness) and modified (2.5 mm height lingual and proximal cervical areas) core designs were produced, followed by the application of veneer porcelain for a total thickness of 1.5 mm. The crowns were cemented to 30-day-aged composite dies and were either single-load-to-failure or step-stress-accelerated fatigue-tested. Use of level probability plots showed significantly higher reliability for the modified core design group. The fatigue fracture modes were veneer chipping not exposing the core for the standard group, and exposing the veneer core interface for the modified group.
Silva, N.R.F.A.; Bonfante, E.A.; Rafferty, B.T.; Zavanelli, R.A.; Rekow, E.D.; Thompson, V.P.; Coelho, P.G.
Modified Y-TZP core design improves all-ceramic crown reliability.
This study tested the hypothesis that all-ceramic core-veneer system crown reliability is improved by modification of the core design. We modeled a tooth preparation by reducing the height of proximal walls by 1.5 mm and the occlusal surface by 2.0 mm. The CAD-based tooth preparation was replicated and positioned in a dental articulator for core and veneer fabrication. Standard (0.5 mm uniform thickness) and modified (2.5 mm height lingual and proximal cervical areas) core designs were produced, followed by the application of veneer porcelain for a total thickness of 1.5 mm. The crowns were cemented to 30-day-aged composite dies and were either single-load-to-failure or step-stress-accelerated fatigue-tested. Use of level probability plots showed significantly higher reliability for the modified core design group. The fatigue fracture modes were veneer chipping not exposing the core for the standard group, and exposing the veneer core interface for the modified group. PMID:21057036
Silva, N R F A; Bonfante, E A; Rafferty, B T; Zavanelli, R A; Rekow, E D; Thompson, V P; Coelho, P G
Exploration and assesment of design windows for a Tokamak-based volumetric neutron source
Design options for a Tokamak volumetric neutron source (VNS) for fusion nuclear technology testing are explored using well-defined and consistent sets of requirements and constraints. Optimum designs with a peak value of the neutron wall load and minimum cost are obtained with intermediate-to-high aspect ratios of about 3.4�3.6. Sensitivity analysis shows that the nuclear technology testing requirements can be achieved
S. K. Ho; M. A. Abdou
System design considerations for fast-neutron interrogation systems
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.
Micklich, B.J.; Curry, B.P.; Fink, C.L.; Smith, D.L.; Yule, T.J.
Determination of core design thermal safety limits for a two-loop pressurized water reactor
Results are given of independent research of core thermal design limits for the Nuklearna Elektrarna Krsko (NEK) nuclear power plant; procedures for two-loop pressurized water reactor plant core design safety limit calculation are used. Emphasis is placed on researching the vessel exit boiling and the hot-channel exit quality limits and their impact on the maximum available design safety operating range and thermal operating margin of the NEK reactor core. For this purpose, the LIMITS computer code is developed. Based on the modified, well-tried COBRA-IV-I computer code, the departure of nuclear boiling ratio core safety limits are calculated. The original results complement well those of the NEK Final Safety Analysis Report. The procedures and the methods for determining the reactor core design thermal limits are successfully proven despite the unavailability of proprietary data, different models, and computer codes. In addition to the acquired capability of in-house independent checking of the vendor`s results, the bases are set for further independent analyses of the limiting safety system settings for the NEK core.
Kostadinov, V.
Increasing sequence diversity with flexible backbone protein design: the complete redesign of a protein hydrophobic core.
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, 1-2��, were needed to entirely mutate the core. The redesigned protein, DRNN, is exceptionally stable (melting point >140�C). An NMR and X-ray crystal structure show that the side chains and backbone were accurately modeled (all-atom RMSD�= 1.3��). PMID:22632833
Murphy, Grant S; Mills, Jeffrey L; Miley, Michael J; Machius, Mischa; Szyperski, Thomas; Kuhlman, Brian
Energy efficient engine. Core engine bearings, drives and configuration: Detailed design report
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.
Broman, C. L.
Multi-frequency wrapper design and optimization for embedded cores under average power constraints
This paper presents a new method for designing test wrappers for embedded cores with multiple clock domains. By exploiting the use of multiple shift frequencies, the proposed method improves upon a recent wrapper design method that requires a common shift frequency for the scan elements in the different clock domains. We present an integer linear programming (ILP) model that can
Design/Operations review of core sampling trucks and associated equipment
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.
Shrivastava, H.P.
Expanding the core: a new approach for the design of single-mode waveguides
A new approach is proposed for the design of single-mode waveguides. The approach is based on slicing the core of the waveguide into a uniform array of subwaveguides and expanding the separation between these waveguides up to the point where the array is still supporting a single supermode. By appropriate selection of the array design parameters, it is possible to
Tarek A. Ramadan
MLA fiber injection for a square core fiber optic beam delivery system: design versus prototype results
The paper details a square core fiber beam delivery design utilizing a unique micro lens array launch method. The paper includes the resulting performance of a prototype created to verify the design and its stability with an emphasis on homogenization as the fiber is articulated.
Lizotte, Todd; Dickey, Fred
Design and fabrication of the instrumented fuel elements for the Annular Core Research Reactor (ACRR)
This report describes the design and fabrication techniques for the instrumented fuel elements of the Annular Core Research Reactor (ACRR). The thermocouple assemblies were designed and fabricated at Sandia Laboratories while the instrumented elements were assembled at Los Alamos Scientific Laboratory. In order to satisfy the ACRR's Technical Specifications, the thermocouples are required to measure temperature in excess of 1800°C
T. R. Schmidt; D. J. Sasmor; J. T. Martin; F. Gonzalez; D. N. Cox
Shielding Analyses for Design of the Upgraded JRR-3 Research Reactor, 2. Shielding of Neutron Beam Holes.
Shielding analyses of neutron beam holes have been presented for the shield design of the upgraded JRR-3 research reactor. Description is given about the calculational procedures and results for the standard beam hole, the beam hole for neutron radiograph...
T. Ise T. Maruo Y. Miyasaka M. Isshiki T. Kumai
Boron tracedrug design for neutron dynamic therapeutics for LDL.
We describe our solution for removal of the low-density lipoprotein (LDL) depot contained in proteins and lipids as a 'druggable' target for atherosclerotic cardiovascular diseases by neutron dynamic therapy (NDT), which we developed using boron tracedrugs for NDT against bovine serum albumin as a model protein. Thus, we examined, among our developed boron tracedrugs, a boron-containing curcuminoid derivative UTX-51, to destroy freshly isolated human LDL dynamically under irradiated thermal neutron to obtain a decreased intensity of band of LDL treated with UTX-51 and thermal neutron irradiation in their SDS-PAGE and electrophoresis analysis. These results suggest that UTX-51 might be a novel candidate of 'beyond chemical' therapeutic agents for atherosclerotic cardiovascular disease. PMID:23852519
Hori, Hitoshi; Nazumi, Yoshijiro; Uto, Yoshihiro
Development of a thermal-hydraulic design methodology for an advanced reactor core with vertical parallel channels.
A thermal-hydraulic design methodology has been developed for the analysis of an advanced reactor core and compared the results with existing analysis data. On the basis of one dimensional core design methodology for the reference core, a computer code CO...
D. H. Hwang Y. J. Yoo K. K. Kim M. Chang
Fuel and Core Design Verification for Extended Power Up-rate in Ringhals Unit 3
Vattenfall's Westinghouse 3-loop PWR Ringhals 3 at the western coast of Sweden is scheduled for an extended power up-rate from 2783 to 3160 MWt in 2007, in the frame of the so called GREAT-project. The project will realize an up-rating initially planned and analysed back in 1995, but with a number of significant improvements outlined in this paper. For the licensing of the up-rated power level, a complete revision of the safety analyses, radiological analyses and systems verifications in FSAR is being performed by Westinghouse Electrics Belgium. The work is performed in close cooperation with Vattenfall in the areas of core calculations and input data. For more than a decade, Vattenfall has performed all core design and reload safety evaluations (RSE) for Ringhals, independent of fuel vendors and safety analysts. In GREAT all core parameters in the safety analysis checklist (SAC) used for the safety analyses are determined based upon a set of nine reference loading patterns designed by Vattenfall covering a wide range of fuel and core designs and extreme cycle-to-cycle variations. To facilitate the calculation of SAC parameters Westinghouse has provided a Reload Safety Evaluation Procedure report (RSEP) with detailed specifications for the calculation of all core parameters used in the analyses. The procedure has been automatized by Vattenfall in a set of scripts executing 3D core simulator calculations and extracting the key results. The same tools will be used in Vattenfall's future RSE for Ringhals 3. This approach is taken to obtain consistency between core designs and core calculations for the safety analyses and the cycle specific calculations, to minimize the risk for future violations of the safety analyses. (authors)
Gabrielsson, Petter; Stepniewski, Marek; Almberger, Jan [Vattenfall Bransle AB, Jaemtlandsg. 99, 162 87 Stockholm (Sweden)
Design and VLSI implementation of a digital audio-specific DSP core for MP3\\/AAC
We present a digital audio-specific DSP core designed for a dual decoder for MPEG\\/audio layer-3 (MP3) and MPEG-2 advanced audio coding (AAC). The processing core is a 20-bit fixed-point programmable DSP having an architecture suitable for audio signal processing. It supports special instructions such as UNPACK and Huffman as well as general arithmetic and logical instructions including pipelined-MAC. All instructions
Kyoung Ho Bang; Nam Hun Jeong; Joon Seok Kim; Young Cheol Park; Dae Hee Youn
Design and fabrication of a novel core-suspended optic fiber for distributed gas sensor
we design a novel core-suspended capillary fiberhat the core is suspened in the air hole and close to the inner surface of the capillary, and experimentally demonstrate its fabrication technology. In addition, a method for linking a single mode fiber and a core-suspended fiber is proposed based on splicing and tapering at the fusion point between the two fibers. By combining with the optical time domain reflectometer technology, we construct a distributed gas sensor system to monitor greenhouse gas based on this novel fiber.
Zhang, T.; Ma, L. J.; Bai, H. B.; Tong, Ch. G.; Dai, Q.; Yang, J.; Yuan, L. B.
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.
Zhang, Tao; Ma, Lijia; Bai, Hongbo; Tong, Chengguo; Dai, Qiang; Kang, Chong; Yuan, Libo
The RF system design for the Spallation Neutron Source
Spallation Neutron Source (SNS) accelerator includes a nominally 1000 MeV, 2 mA average current linac consisting of a radio frequency quadrupole (RFQ), drift tube linac (DTL), coupled cavity linac (CCL), a medium and high beta superconducting (SC) linac, and two buncher cavities for beam transport to the ring. Los Alamos is responsible for the RF systems for all sections of
D. Rees; P. Tallerico; W. Roybal; K. Cummings; T. Hardek; J. Bradley; A. Regan
We have numerically studied the standing\\/spherical accretion shock instability (SASI) for a core-collapse supernova. The core-collapse supernova is an explosion of a massive star in the final stage of its evolution. Although this spectacular phenomenon is a key issue for astrophysics, the explosion mechanism has not been understood perfectly. Recently, SASI has widely been noticed since it may play an
Wakana Iwakami; Naofumi Ohnishi; Kei Kotake; Shoichi Yamada; Keisuke Sawada
Neutronics analysis of an open-cycle high-impulse gas core reactor concept
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.
Whitmarsh, C. L., Jr.
A review of irradiation effects on LWR core internal materials � Neutron embrittlement
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
O. K. Chopra; A. S. Rao
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
Design and preparation of ethyl cellulose microcapsules of gadopentetate dimeglumine for neutron-capture therapy using the Wurster process.
Microcapsules of hygroscopic, highly water-soluble gadopentetate dimeglumine (Gd-DTPA-DM) for use in preliminary in vivo experiments for neutron-capture therapy were designed. They were prepared with such properties as a particle size small enough to be suspended and injected through a syringe, a negligible release of Gd-DTPA-DM, and a high drug content by means of the Wurster process, a spray coating method using a spouted bed with a draft tube. They were composed of lactose cores of 53-63 microm, an undercoat of ethyl cellulose (EC) and polyvinylpyrrolidone (PVP), a drug-layer of Gd-DTPA-DM, EC and PVP, a waterproof coat and a release-sustaining overcoat of EC and cholesterol (1:1), and a surface treated with hydrogenated egg lecithin. By curing at 110 degrees C for 30 min after mixing with 20% pulverized mannitol powder, the 20% overcoating suppressed the release of Gd-DTPA-DM from 75-106 microm microcapsules to less than 10% for the first 20 min, which was the period required to prepare a suspension, inject it and irradiate the neutron. The microcapsules could be used to confirm that the intracellular presence of Gd is not critical in gadolinium neutron-capture therapy. PMID:8370113
Fukumori, Y; Ichikawa, H; Tokumitsu, H; Miyamoto, M; Jono, K; Kanamori, R; Akine, Y; Tokita, N
System design and radiation field characteristics of the High Flux Neutron Radiography Facility (HFNRF) at Sandia National Laboratories
The High Flux Neutron Radiography Facility (HFNRF) has been fabricated, tested and used successfully. This paper describes the final design features and presents the results of the radiation measurements made during the characterization tests. A complete neutron spectrum (from 10{sup {minus}8} to 15 MeV) has been measured with the aid of activation foils. Gold foils were used to determine a thermal neutron flux of 9.4 {times} 10{sup 7} n{center_dot}cm{sup {minus}2}s{sup {minus}1} at the image plane in the Sandia Annular Core Research Reactor (ACRR) operating at 1.0 MW. The reactor can be operated at 20 MW for 10-second periods with the radiography tube in the central cavity. In pulse mode with peak power of 24,000 MW and a 7.6 ms FWHM {approximately}1.9 {times} 10{sup 10} n{center_dot}cm{sup {minus}2} will be generated with a peak flux of 2.3 {times} 10{sup 12} n{center_dot}cm{sup {minus}2}s{sup {minus}1}.
Kelly, J.G.; McCrory, F.M. [Sandia National Labs., Albuquerque, NM (United States); Cooper, P.J. [Science and Engineering Associates, Inc., Albuquerque, NM (United States)
The High Flux Neutron Radiography Facility (HFNRF) has been fabricated, tested and used successfully. This paper describes the final design features and presents the results of the radiation measurements made during the characterization tests. A complete neutron spectrum (from 10{sup {minus}8} to 15 MeV) has been measured with the aid of activation foils. Gold foils were used to determine a thermal neutron flux of 9.4 {times} 10{sup 7} n{center dot}cm{sup {minus}2}s{sup {minus}1} at the image plane in the Sandia Annular Core Research Reactor (ACRR) operating at 1.0 MW. The reactor can be operated at 20 MW for 10-second periods with the radiography tube in the central cavity. In pulse mode with peak power of 24,000 MW and a 7.6 ms FWHM {approximately}1.9 {times} 10{sup 10} n{center dot}cm{sup {minus}2} will be generated with a peak flux of 2.3 {times} 10{sup 12} n{center dot}cm{sup {minus}2}s{sup {minus}1}.
Kelly, J.G.; McCrory, F.M. (Sandia National Labs., Albuquerque, NM (United States)); Cooper, P.J. (Science and Engineering Associates, Inc., Albuquerque, NM (United States))
Effective Domain Partitioning for Multi-Clock Domain IP Core Wrapper Design under Power Constraints
The rapid advancement of VLSI technology has made it possible for chip designers and manufacturers to embed the components of a whole system onto a single chip, called System-on-Chip or SoC. SoCs make use of pre-designed modules, called IP-cores, which provide faster design time and quicker time-to-market. Furthermore, SoCs that operate at multiple clock domains and very low power requirements
Thomas Edison Yu; Tomokazu Yoneda; Danella Zhao; Hideo Fujiwara
Digital Data Processing Peripheral Design for an Embedded Application based on the Microblaze Soft Core
In this paper we present a design of a peripheral for microblaze soft core processor as part of a R+D project carried out in cooperation with three different companies. The objective of the project consisted in the development of an embedded system with a SoC implemented on a FPGA custom-designed board. This work addresses the design of a digital data
E. Ostua; J. Viejo; M. J. Bellido; A. Millan; J. Juan; A. Munoz
Preliminary core design studies for the advanced burner reactor over a wide range of conversion ratios
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
E. A. Hoffman; W. S. Yang; R. N. Hill
Design of air-gapped magnetic-core inductors for superimposed direct and alternating currents
Using data on standard magnetic-material properties and standard core sizes for air-gap-type cores, an algorithm designed for a computer solution is developed which optimally determines the air-gap length and locates the quiescent point on the normal magnetization curve so as to yield an inductor design with the minimum number of turns for a given ac voltage and frequency and with a given dc bias current superimposed in the same winding. Magnetic-material data used in the design are the normal magnetization curve and a family of incremental permeability curves. A second procedure, which requires a simpler set of calculations, starts from an assigned quiescent point on the normal magnetization curve and first screens candidate core sizes for suitability, then determines the required turns and air-gap length.
Ohri, A. K.; Wilson, T. G.; Owen, H. A., Jr.
Analysis of Stainless Steel Sandwich Panels with a Metal Foam Core for Lightweight Fan Blade Design
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.
Min, James B.; Ghosn, Louis J.; Lerch, Bradley A.; Raj, Sai V.; Holland, Frederic A., Jr.; Hebsur, Mohan G.
Core Noise: Implications of Emerging N+3 Designs and Acoustic Technology Needs
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.
Hultgren, Lennart S.
Thermal neutron irradiation field design for boron neutron capture therapy of human explanted liver
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.
Bortolussi, S.; Altieri, S. [Department of Nuclear and Theoretical Physics, University of Pavia, INFN Section of Pavia, via Bassi 6, 27100 Pavia (Italy)
Thermal neutron irradiation field design for boron neutron capture therapy of human explanted liver.
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
Bortolussi, S; Altieri, S
Thermo-mechanical and neutron lifetime modelling and design of Be pebbles in the neutron multiplier for the LIFE engine
Concept designs for the laser inertial fusion/fission energy (LIFE) engine include a neutron multiplication blanket containing Be pebbles flowing in a molten salt coolant. These pebbles must be designed to withstand the extreme irradiation and temperature conditions in the blanket to enable a reliable and cost-effective operation of LIFE. In this work, we develop design criteria for spherical Be pebbles on the basis of their thermo-mechanical behaviour under continued neutron exposure. We consider the effects of high fluence and fast fluxes on the elastic, thermal and mechanical properties of nuclear-grade Be. Our results suggest a maximum pebble diameter of 30 mm to avoid tensile failure, coated with an anti-corrosive, high-strength metallic shell to avoid failure by pebble contact. Moreover, we find that the operation temperature must always be kept above 450 �C to enable creep to relax the stresses induced by swelling. Under these circumstances, we estimate the pebble lifetime to be at least 16 months if uncoated, and up to six years when coated. We identify the sources of uncertainty on the properties used and discuss the advantages of new intermetallic beryllides and their use in LIFE's neutron multiplier. To establish Be-pebble lifetimes with improved confidence, reliable experiments to measure irradiation creep must be performed.
DeMange, P.; Marian, J.; Caro, M.; Caro, A.
A novel design approach for a neutron measurement station for burnt fuel
The design and characterization of a passive neutron measurement station for highly burnt fuel has been undertaken at the Paul Scherrer Institute (PSI). The measurement station aims at the determination of the total neutron emission rate of full-length light water reactor (LWR) fuel rods, as also the corresponding axial distributions. It is intended that the measurement station be introduced into the hot cells available at PSI to allow measuring the neutron emission of spent fuel rods provided by the Swiss nuclear power plants. In addition, the neutron emission of a large set of burnt fuel samples that have been previously characterized by post-irradiation examination (PIE) will be measured, in order to relate neutron emission to the burnup and isotopic composition of different fuel types. The design of the measurement station is presented in this article. A post-processing algorithm is introduced to improve the spatial resolution of the "measured" axial profile. In order to quantify the accuracy of the reconstructed neutron source distribution, a figure-of-merit (FOM) is defined and adapted to the detection procedure. With the optimized measurement station and procedure, it is estimated that the neutron emission distribution of a highly burnt, full-length fuel rod would be measurable with acceptable accuracy in about 20 min.
Dietler, Rodolfo; Hursin, Mathieu; Perret, Gregory; Jordan, Kelly; Chawla, Rakesh
Dependence of neutron-induced radioactivity in fusion reactors on geometric design parameters
Although the neutron-induced activation in a fusion reactor is a non-linear problem whose solution requires the use of neutron transport codes and neutron activation and decay codes, a number of simple arguments can be made which give useful scaling laws for the total radioactivity in a fusion reactor. Because these laws rely heavily on assumptions of linearity and the smallness of second-order effects, we have compared them to the results of computer experiments designed to investigate their validity over the range of operating parameters typical of fusion reactors.
Lasche, G.P.; Blink, J.A.
The design and performance of IceCube DeepCore
The IceCube neutrino observatory in operation at the South Pole, Antarctica, comprises three distinct components: a large buried array for ultrahigh energy neutrino detection, a surface air shower array, and a new buried component called DeepCore. DeepCore was designed to lower the IceCube neutrino energy threshold by over an order of magnitude, to energies as low as about 10 GeV. DeepCore is situated primarily 2100 m below the surface of the icecap at the South Pole, at the bottom center of the existing IceCube array, and began taking physics data in May 2010. Its location takes advantage of the exceptionally clear ice at those depths and allows it to use the surrounding IceCube detector as a highly efficient active veto against the principal background of downward-going muons produced in cosmic-ray air showers. DeepCore has a module density roughly five times higher than that of the standard IceCube array, and uses photomultiplier tubes with a new photocathode featuring a quantum efficiency about 35% higher than standard IceCube PMTs. Taken together, these features of DeepCore will increase IceCube's sensitivity to neutrinos from WIMP dark matter annihilations, atmospheric neutrino oscillations, galactic supernova neutrinos, and point sources of neutrinos in the northern and southern skies. In this paper we describe the design and initial performance of DeepCore.
Abbasi, R.; Abdou, Y.; Abu-Zayyad, T.; Ackermann, M.; Adams, J.; Aguilar, J. A.; Ahlers, M.; Allen, M. M.; Altmann, D.; Andeen, K.; Auffenberg, J.; Bai, X.; Baker, M.; Barwick, S. W.; Bay, R.; Bazo Alba, J. L.; Beattie, K.; Beatty, J. J.; Bechet, S.; Becker, J. K.; Becker, K.-H.; Benabderrahmane, M. L.; BenZvi, S.; Berdermann, J.; Berghaus, P.; Berley, D.; Bernardini, E.; Bertrand, D.; Besson, D. Z.; Bindig, D.; Bissok, M.; Blaufuss, E.; Blumenthal, J.; Boersma, D. J.; Bohm, C.; Bose, D.; B�ser, S.; Botner, O.; Brown, A. M.; Buitink, S.; Caballero-Mora, K. S.; Carson, M.; Chirkin, D.; Christy, B.; Clevermann, F.; Cohen, S.; Colnard, C.; Cowen, D. F.; Cruz Silva, A. H.; D'Agostino, M. V.; Danninger, M.; Daughhetee, J.; Davis, J. C.; De Clercq, C.; Degner, T.; Demir�rs, L.; Descamps, F.; Desiati, P.; de Vries-Uiterweerd, G.; DeYoung, T.; D�az-V�lez, J. C.; Dierckxsens, M.; Dreyer, J.; Dumm, J. P.; Dunkman, M.; Eisch, J.; Ellsworth, R. W.; Engdeg�rd, O.; Euler, S.; Evenson, P. A.; Fadiran, O.; Fazely, A. R.; Fedynitch, A.; Feintzeig, J.; Feusels, T.; Filimonov, K.; Finley, C.; Fischer-Wasels, T.; Fox, B. D.; Franckowiak, A.; Franke, R.; Gaisser, T. K.; Gallagher, J.; Gerhardt, L.; Gladstone, L.; Gl�senkamp, T.; Goldschmidt, A.; Goodman, J. A.; G�ra, D.; Grant, D.; Griesel, T.; Gro�, A.; Grullon, S.; Gurtner, M.; Ha, C.; Haj Ismail, A.; Hallgren, A.; Halzen, F.; Han, K.; Hanson, K.; Heinen, D.; Helbing, K.; Hellauer, R.; Hickford, S.; Hill, G. C.; Hoffman, K. D.; Hoffmann, B.; Homeier, A.; Hoshina, K.; Huelsnitz, W.; H�l�, J.-P.; Hulth, P. O.; Hultqvist, K.; Hussain, S.; Ishihara, A.; Jacobi, E.; Jacobsen, J.; Japaridze, G. S.; Johansson, H.; Kampert, K.-H.; Kappes, A.; Karg, T.; Karle, A.; Kenny, P.; Kiryluk, J.; Kislat, F.; Klein, S. R.; K�hne, J.-H.; Kohnen, G.; Kolanoski, H.; K�pke, L.; Koskinen, D. J.; Kowalski, M.; Kowarik, T.; Krasberg, M.; Kroll, G.; Kurahashi, N.; Kuwabara, T.; Labare, M.; Laihem, K.; Landsman, H.; Larson, M. J.; Lauer, R.; L�nemann, J.; Madsen, J.; Marotta, A.; Maruyama, R.; Mase, K.; Matis, H. S.; Meagher, K.; Merck, M.; M�sz�ros, P.; Meures, T.; Miarecki, S.; Middell, E.; Milke, N.; Miller, J.; Montaruli, T.; Morse, R.; Movit, S. M.; Nahnhauer, R.; Nam, J. W.; Naumann, U.; Nygren, D. R.; Odrowski, S.; Olivas, A.; Olivo, M.; O'Murchadha, A.; Panknin, S.; Paul, L.; P�rez de los Heros, C.; Petrovic, J.; Piegsa, A.; Pieloth, D.; Porrata, R.; Posselt, J.; Price, P. B.; Przybylski, G. T.; Rawlins, K.; Redl, P.; Resconi, E.; Rhode, W.; Ribordy, M.; Richman, M.; Rodrigues, J. P.; Rothmaier, F.; Rott, C.; Ruhe, T.; Rutledge, D.; Ruzybayev, B.; Ryckbosch, D.; Sander, H.-G.; Santander, M.; Sarkar, S.; Schatto, K.; Schmidt, T.; Sch�nwald, A.; Schukraft, A.; Schultes, A.; Schulz, O.; Schunck, M.; Seckel, D.; Semburg, B.; Seo, S. H.; Sestayo, Y.; Seunarine, S.; Silvestri, A.; Spiczak, G. M.; Spiering, C.; Stamatikos, M.; Stanev, T.; Stezelberger, T.; Stokstad, R. G.; St��l, A.; Strahler, E. A.; Str�m, R.; St�er, M.; Sullivan, G. W.; Swillens, Q.; Taavola, H.; Taboada, I.; Tamburro, A.; Tepe, A.; Ter-Antonyan, S.; Tilav, S.; Toale, P. A.; Toscano, S.; Tosi, D.; van Eijndhoven, N.; Vandenbroucke, J.; Van Overloop, A.; van Santen, J.; Vehring, M.; Voge, M.; Walck, C.; Waldenmaier, T.; Wallraff, M.; Walter, M.; Weaver, Ch.; Wendt, C.; Westerhoff, S.; Whitehorn, N.; Wiebe, K.; Wiebusch, C. H.; Williams, D. R.; Wischnewski, R.; Wissing, H.; Wolf, M.; Wood, T. R.; Woschnagg, K.; Xu, C.; Xu, D. L.; Xu, X. W.; Yanez, J. P.; Yodh, G.; Yoshida, S.; Zarzhitsky, P.; Zoll, M.
Validation of the Nuclear Design Method for MOX Fuel Loaded LWR Cores
The actual batch loading of mixed-oxide (MOX) fuel in light water reactors (LWRs) is now ready to start in Japan. One of the efforts that have been devoted to realizing this batch loading has been validation of the nuclear design methods calculating the MOX-fuel-loaded LWR core characteristics. This paper summarizes the validation work for the applicability of the CASMO-4/SIMULATE-3 in-core fuel management code system to MOX-fuel-loaded LWR cores. This code system is widely used by a number of electric power companies for the core management of their commercial LWRs. The validation work was performed for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. Each validation consists of two parts: analyses of critical experiments and core tracking calculations of operating plants. For the critical experiments, we have chosen a series of experiments known as the VENUS International Program (VIP), which was performed at the SCK/CEN MOL laboratory in Belgium. VIP consists of both BWR and PWR fuel assembly configurations. As for the core tracking calculations, the operating data of MOX-fuel-loaded BWR and PWR cores in Europe have been utilized.
Saji, E.; Inoue, Y.; Mori, M.; Ushio, T.
Conceptual design of a reversed-field pinch fusion neutron source
The conceptual design of an ohmically-heated, reversed-field pinch (RFP) operating with a 5-MWm/sup 2/ steady-state DT fusion neutron wall loading while generating /approximately/100-MW total fusion power is presented. These results are also useful in projecting the development of an economic source of DT neutrons for large-volume (/approximately/10 m/sup 3/) fusion nuclear testing. 6 refs., 4 figs., 5 tabs.
Bathke, C.G.; Krakowski, R.A.; Manzanares, R.G.; Miller, R.L.; Werley, K.A.
Conceptual design of an intense neutron source for time-of-flight measurements
Among the accelerator-based neutron sources, the ones that are driven by electron Linacs still appear quite attractive, notably in the case of cross-section measurements with the time-of-flight method. This is due to their better beam quality and economy aspects, which make them complementary, rather than inferior to the hadron (protons, deuterons)-driven spallation facilities.A conceptual design study of a powerful neutron
S. Bartalucci; V. Angelov; K. Drozdowicz; D. Dworak; G. Tracz
Reactor physics design for an epithermal neutron beam at the power burst reactor facility
An epithermal neutron beam facility has been designed for the Power Burst reactor Facility (PBF). The beam will have a broadband epithermal (1.0 eV - 10 keV) spectrum with extraordinary intensity and collimation with minimum fast neutron and gamma contamination. The PBF reactor is located at the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho, USA. This facility has
F. J. Wheeler; B. L. Rushton; D. K. Parsons; D. W. Nigg
Design of a steady-state tokamak device with superconducting coils for a volumetric neutron source
We designed a volumetric neutron source for testing large-scale blanket components, based on a steady-state tokamak device with superconducting coils. It is found that a neutron flux of approximately 1.0 MW m?2 is available in the medium-size device (R=4.5 m, a=1.0 m, ?=1.8, Ip=5.6 MA) under the conditions of H?2 and ?N?3 with a neutral beam injection (NBI) power of
Y. Ogawa; K. Okano; N. Inoue; T. Amano; Y. Asaoka; R. Hiwatari; Y. Murakami; K. Takemura; K. Tokimatsu; K. Tomabechi; T. Yamamoto; T. Yoshida
Designing with advanced composites; Report on the European Core Conference, 1st, Zurich, Switzerland, Oct. 20, 21, 1988, Conference Papers
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.
Design and construction of pulsed neutron diagnostic system for plasma focus device (SBUPF1).
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(7) neutrons per shot. PMID:20687711
Moghadam, Sahar Rajabi; Davani, Fereydoon Abbasi
Design and construction of pulsed neutron diagnostic system for plasma focus device (SBUPF1)
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.
Moghadam, Sahar Rajabi; Davani, Fereydoon Abbasi [Department of Radiation Application, Shahid Beheshti University, 1983963113 Tehran (Iran, Islamic Republic of)
Research advances in polymer emulsion based on "core-shell" structure particle design.
In recent years, quite many studies on polymer emulsions with unique core-shell structure have emerged at the frontier between material chemistry and many other fields because of their singular morphology, properties and wide range of potential applications. Organic substance as a coating material onto either inorganic or organic internal core materials promises an unparalleled opportunity for enhancement of final functions through rational designs. This contribution provides a brief overview of recent progress in the synthesis, characterization, and applications of both inorganic-organic and organic-organic polymer emulsions with core-shell structure. In addition, future research trends in polymer composites with core-shell structure are also discussed in this review. PMID:23726300
Ma, Jian-zhong; Liu, Yi-hong; Bao, Yan; Liu, Jun-li; Zhang, Jing
Modeling and design of a reload PWR core for a 48-month fuel cycle
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.
McMahon, M.V.; Driscoll, M.J.; Todreas, N.E. [Massachusetts Inst. of Tech., Cambridge, MA (United States)
Theory of deep inelastic neutron scattering: Hard-core perturbation theory
Details are presented of a new many-body theory for deep inelastic neutron scattering (DINS) experiments to measure momentum distributions in quantum fluids and solids. The high-momentum and energy-transfer scattering law in helium is shown to be a convolution of the impulse approximation with a final-state broadening function which depends on the scattering phase shifts and the radial distribution function. The
Richard N. Silver
Carbon-enhanced metal-poor (CEMPs+r) stars show large enhancements of elements produced both by the slow and the rapid neutron capture processes (the s and r process, respectively) and represent a relatively large fraction, 30% to 50%, of the CEMP population. Many scenarios have been proposed to explain this peculiar chemical composition and most of them involve a binary companion producing the
M. Lugaro; S. W. Campbell; S. E. de Mink
The Mystery of CEMP s + r Stars and the Dual Core-Flash Neutron Superburst
Carbon-enhanced metal-poor (CEMP)-s+r stars show large enhancements of elements produced both by the slow and the rapid neutron capture processes (the s and r process, respectively) and represent a relatively large fraction, 30% to 50%, of the CEMP population. Many scenarios have been proposed to explain this peculiar chemical composition and most of them involve a binary companion producing the
Neutrino emissivity of 3P2-3F2 superfluid cores in neutron stars
The influence of the admixture of the 3F2 state onto collective spin oscillations and neutrino emission processes in the triplet superfluid neutron liquid is studied in the BCS approximation. The eigenmode of spin oscillations with ??58/35? is predicted to exist in the triplet superfluid neutron condensate besides the already known mode ???/5. Excitation of the high-frequency spin oscillations in the condensate occurs through the tensor interactions between quasiparticles. Neutrino energy losses through neutral weak currents are found to consist of three separate contributions caused by a recombination of broken Cooper pairs and by weak decays of the collective modes of spin oscillations. Neutrino decays of the low-frequency spin waves can play an important role in the cooling scenario of neutron stars. Weak decays of the high-frequency oscillations that occur only if the tensor forces are taken into account in the pairing interactions does not modify substantially the total energy losses. Simple expressions are suggested for the total neutrino emissivity.
Leinson, L. B.
Electric dipole response of He6: Halo-neutron and core excitations
Electric dipole (E1) response of He6 is studied with a fully microscopic six-body calculation. The wave functions for the ground and excited states are expressed as a superposition of explicitly correlated Gaussians. Final state interactions of three-body decay channels are explicitly taken into account. The ground state properties and the low-energy E1 strength are obtained consistently with observations. Two main peaks as well as several small peaks are found in the E1 strength function. The peak at the high-energy region indicates a typical macroscopic picture of the giant dipole resonance, the out-of-phase proton-neutron motion. The transition densities of the lower-lying peaks exhibit in-phase proton-neutron motion in the internal region, out-of-phase motion near the surface region, and spatially extended neutron oscillation, indicating a soft-dipole mode and its vibrationally excited mode. The compressional dipole strength is also examined in relation to the soft-dipole mode.
Mikami, D.; Horiuchi, W.; Suzuki, Y.
Safety of fast neutron reactor power units under operation and design in Russia
Fast neutron reactors have been designed and operated in Russia for over 40 years. The paper provides the story of the regulatory basis applied for designing, constructing and operating fast reactors. The increasing extent of the requirements of the supervisory authorities, and the issuing of modified regulatory documents, has increasingly led to the necessity of implementing difficult and expensive measures
Yu. E. Bagdassarov
Turbulent Magnetic Field Amplification from Spiral SASI Modes: Implications for Core-Collapse Supernovae and Proto-Neutron Star Magnetization
We extend our investigation of magnetic field evolution in three-dimensional flows driven by the stationary accretion shock instability (SASI) with a suite of higher-resolution idealized models of the post-bounce core-collapse supernova environment. Our magnetohydrodynamic simulations vary in initial magnetic field strength, rotation rate, and grid resolution. Vigorous SASI-driven turbulence inside the shock amplifies magnetic fields exponentially; but while the amplified fields reduce the kinetic energy of small-scale flows, they do not seem to affect the global shock dynamics. The growth rate and final magnitude of the magnetic energy are very sensitive to grid resolution, and both are underestimated by the simulations. Nevertheless our simulations suggest that neutron star magnetic fields exceeding $10^{14}$~G can result from dynamics driven by the SASI, \\emph{even for non-rotating progenitors}.
Endeve, Eirik [ORNL; Cardall, Christian Y [ORNL; Budiardja, Reuben D [ORNL; Beck, Sam [University of Tennessee, Knoxville (UTK); Bejnood, Alborz [ORNL; Toedte, Ross J [ORNL; Blondin, J. M. [North Carolina State University; Mezzacappa, Anthony [ORNL
We extend our investigation of magnetic field evolution in three-dimensional flows driven by the stationary accretion shock instability (SASI) with a suite of higher-resolution idealized models of the post-bounce core-collapse supernova environment. Our magnetohydrodynamic simulations vary in initial magnetic field strength, rotation rate, and grid resolution. Vigorous SASI-driven turbulence inside the shock amplifies magnetic fields exponentially; but while the amplified fields reduce the kinetic energy of small-scale flows, they do not seem to affect the global shock dynamics. The growth rate and final magnitude of the magnetic energy are very sensitive to grid resolution, and both are underestimated by the simulations. Nevertheless, our simulations suggest that neutron star magnetic fields exceeding 10{sup 14} G can result from dynamics driven by the SASI, even for non-rotating progenitors.
Endeve, Eirik; Mezzacappa, Anthony [Computer Science and Mathematics Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6354 (United States); Cardall, Christian Y.; Budiardja, Reuben D. [Physics Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6354 (United States); Beck, Samuel W.; Bejnood, Alborz [Joint Institute for Heavy Ion Research, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6374 (United States); Toedte, Ross J. [National Center for Computational Sciences, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6354 (United States); Blondin, John M., E-mail: firstname.lastname@example.com [Physics Department, North Carolina State University, Raleigh, NC 27695-8202 (United States)
Measurements of actinide-fission product yields in Caliban and Prospero metallic core reactor fission neutron fields
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)
Casoli, P.; Authier, N. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Laurec, J.; Bauge, E.; Granier, T. [CEA, Centre DIF, 91297 Arpajon (France)
Core Noise: Implications of Emerging N+3 Designs and Acoustic Technology Needs.
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 Turboelectr...
L. S. Hultgren
Design and Analysis of a Claw Pole Permanent Magnet Motor With Molded Soft Magnetic Composite Core
Soft magnetic composite (SMC) materials and SMC electromagnetic devices have undergone substantial development in the past decade. Much work has been conducted on designing and prototyping various types of electrical machine. However, the iron cores were often made by cutting existing SMC preforms that were formed by compacting SMC powder in simple cylinder or bar-shape molds, and the magnetic properties
Youguang Guo; Jianguo Zhu; D. G. Dorrell
Active Learning: Effects of Core Training Design Elements on Self-Regulatory Processes, Learning, and Adaptability
This research describes a comprehensive examination of the cognitive, motivational, and emotional processes underlying active learning approaches, their effects on learning and transfer, and the core training design elements (exploration, training frame, emotion-control) and individual differences (cognitive ability, trait goal orientation, trait anxiety) that shape these processes. Participants (N = 350) were trained to operate a complex computer-based simulation. Exploratory
Bradford S. Bell; Steve W. J. Kozlowski
Using Domain Partitioning in Wrapper Design for IP Cores Under Power Constraints
This paper presents a novel design method for power-aware test wrappers targeting embedded cores with multiple clock domains. We show that effective partitioning of clock domains combined with bandwidth conversion and gated-clocks would yield shorter test times due to greater flexibility when determining optimal test schedules especially under tight power constraints
Design and test of a torsional vibratory core sampler for marine sediments
Studies of marine sediments for geological, biological or engineering purposes is an area of increasing importance as more investigations are directed toward surveys of submerged bottoms of oceans, lakes, and rivers. This paper summarizes a project initiated in Jan. 1969, to design, develop, and test prototype coring equipment for marine environments. The torsional system of vibration offers several specific advantages
R. J. Harker; R. M. Shah
Design and Implementation of the National Institute of Environmental Health Sciences Dublin Core Metadata Schema
The National Institute of Environmental Health Sciences (NIEHS) has formed a team to design and implement a Dublin Core-based metadata schema to enhance the public's ability to retrieve pertinent public health information on the organization's Web site. The team decided to use the DC schema because it is a de facto standard and because of its flexibility. With a little
W. Davenport Robertson; Ellen M. Leadem; Jed Dube; Jane Greenberg
Estimation of hot spot temperature in distribution transformer considering core design using FEM
Power transformers represent the largest portion of capital investment in transmission and distribution substations. One of the most important parameters governing a transformerpsilas life expectancy is the hot-spot temperature. Therefore design of transformer is so important that should be considered. Core is one of parts of transformer that is effective on hot-spot temperature. This paper presents a methodology to assess
Sh. Taheri; A. Vahedi; A. Gholami; H. Taheri
Development of design options for the port plug components of the ITER core CXRS diagnostic
The paper presents component concepts developed for the ITER core charge exchange recombination spectroscopy (cCXRS). They are based on the cCXRS layout of 2009, named as reference design option. It includes an outer shell, carrying a blanket shield module and a shielding cassette. The cassette, attached to the outer shell, carries secondary mirrors and a retractable tube. The tube holds
Yu. Krasikov; T. Baross; W. Biel; A. Litnovsky; N. Hawkes; G. Kiss; J. F. F. Klinkhamer; J. F. Koning; A. Krimmer; O. Neubauer; A. Panin
Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core.
A new non-TRISO fuel and clad design concept is proposed for the prismatic, helium cooled 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 r...
J. W. Sterbentz
Designing high frequency ac inductors using ferrite and Molypermalloy Powder Cores (MPP)
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.
Mclyman, W. T.; Wagner, A. P.
Hardware-Software Co-design of QRD-RLS Algorithm with Microblaze Soft Core Processor
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
Nupur Lodha; Nivesh Rai; Rahul Dubey; Hrishikesh Venkataraman
Effectiveness of Horizontal Stirrups in Joint Core for Exterior Beam-Column Joints with Nonseismic Design
Reversed cyclic-load tests on full-scale reinforced concrete exterior beam-column joints with nonseismic design to the British standard BS 8110 were carried out. The seismic behaviour of these non-seismically designed beamcolumn joints was investigated, and the effectiveness of horizontal stirrups in joint core on the seismic performance and shear strength of the joints were emphasised. It was found that horizontal stirrups
J. S. Kaung; H. F. Wong
Thermally Activated Post-glitch Response of the Neutron Star Inner Crust and Core. I. Theory
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 ??, pinning introduces a delay in the post-glitch response time. The delay time is td = 7(t sd/104 yr)((??/?)/10�6) d, where t sd is the spin-down age; td 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.
Link, Bennett
Production of gamma ray bursts from asymmetric core combustion of magnetized young neutron stars
Many works in the past have explored the idea that the conversion of hadronic matter into strange quark matter in neutron stars may be an energy source for GRBs (see references in Lugones et al. 2002, Lugones and Horvath 2003). These models addressed essentially spherically symmetric conversions of the whole neutron star rendering isotropic gamma emission. Accumulating observational evidence suggests that at least ''long'' GRBs are strongly asymmetric, jet-like outflows. The ''short'' burst subclass is not obviously asymmetric, and they may actually be spherically symmetric if the sources are close enough. A new potentially important feature recently recognized (Lugones et al. 2002) is that if a conversion to strange quark matter actually begins near the center of a neutron star, the presence of a magnetic field with intensity B � 1013 G (see also Ghezi, de Gouveia Dal Pino & Horvath 2004) will originate a prompt collimated gamma emission, which may be observed as a short, beamed GRB after the recovery of a fraction of the neutrino energy via ? {bar?} ? e+e- ? ??. The calculations show that the neutrino luminosity is � 1053 erg/sec and that the e+e- luminosity is about two orders of magnitude smaller ( tet{Lugones2002grb}). We find that 90 % of the e+e- pairs are injected inside small cylinders located just above the polar caps (with radius ? and height 0.4 R) in a timescale of ?i ? 0.2 s almost independently of the initial temperature. This provides an interesting suitable explanation for the inner engine of short gamma ray bursts.
de Gouveia dal Pino, E. M.; Lugones, G.; Horvath, J. E.; Ghezzi, C. R.
The SNL100-02 blade : advanced core material design studies for the Sandia 100-meter blade.
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.
Griffith, Daniel Todd
The behavior of reactor power and flux resulting from changes in core-coolant temperature for a miniature neutron source reactor
In this work, measurements were performed to verify the theoretical predictions of reactor power and flux parameters that result from changes in core inlet temperature (Tin) and the temperature difference between the coolant inlet and outlet (?T) in the Nigeria Research Reactor-1 (NIRR-1), which is a Miniature Neutron Source Reactor (MNSR). The measured data shows that there is a strong
Y. A. Ahmed; G. I. Balogun; S. A. Jonah; I. I. Funtua
Design, construction and testing of a cold neutron beam facility for the Cornell TRIGA reactor
This study is to design and construct cold neutron sources and guide tubes at medium power university research reactors and demonstrate the feasibility of operation without complexity, costliness and potential hazard. The goal is a neutron beam that is for basic and applied research in which low backgrounds are especially important. In the design we have chosen mesitylene as the cold moderator, located in a chamber inside a beam tube within the reflector region of the reactor, and cooled by copper cold fingers connected to a cryogenic refrigerator outside the biological shield of the reactor, avoiding the circulation of cryogenic liquids or gases within the shield. In a series of in-pile tests, design parameters relating to nuclear heating, radiation effects, and moderating efficiency of the source geometry are being measured and interpreted in term of simple models including neutron transport calculations. Low-power trials were carried out to measure the neutron spectrum at the exit of a temporarily installed guide of five meters total length. A slow chopper was mounted at the exit, and a Li-6 glass scintillator neutron detector was placed two meters downstream from the chopper. The spatial distribution across the exit was measured with BF{sub 3} counter. Both the shift in the neutron energy spectrum and the spatial distribution agreed within experimental error with calculated values corresponding to the moderator temperature and the guide tube length that were used. Monte Carlo simulations were used for our particular geometry for both the transmission of the guide, and the spatial distribution across the exit aperture as function of neutron energy. The final system design such as the size and shape of the mesitylene chamber and of the nuclear heating shield will be optimized using simulation calculation and the results of the low-power trials.
Emoto, Takashi.
The advanced neutron source--designing to meet the needs of the user community
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.
Peretz, F.J. (Oak Ridge National Lab., TN (USA))
Design and performance considerations for perforated semiconductor thermal-neutron detectors
Perforated silicon structures backfilled with either B10 or LiF6 are presently under construction as high efficiency thermal-neutron detectors. Although many perforated structures are possible, there are three fundamental designs that are studied in the present work, namely (a) cylindrical perforations where holes are filled with neutron reactive material, (b) pillar configurations where semiconductor pillars are surrounded by neutron reactive material, and (c) trench structures where the trenches are filled with neutron reactive material. From results of Monte Carlo simulations, it is found that the trench structure affords the best efficiencies, exceeding 20% using conservative geometries, and over 35% for more aggressive structures. Simulated spectroscopic features and manufacturing constraints are presented and discussed.
Shultis, J. K.; McGregor, D. S.
Target station design for a 1 MW pulsed spallation neutron source
Target stations are vital components of the 1 MW, next generation spallation neutron source proposed for LANSCE. By and large, target stations design determines the overall performance of the facility. Many traditional concepts will probably have to be rethought, and many new concepts will have to be put forward to meet the 1 MW challenge. This article gives a brief overview of the proposed neutron spallation source from the target station viewpoint, as well as the general philosophy adopted for the design of the LANSCE-II target stations. Some of the saliant concepts and features envisioned for LANSCE-II are briefly described.
Russell, G.J.; Baker, G.D.; Brewton, R.J. [and others
Development of a coupled neutronic\\/thermal-hydraulic tool with multi-scale capabilities and applications to HPLWR core analysis
The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper.The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced
Lanfranco Monti; J�rg Starflinger; Thomas Schulenberg
Fuel cycles and advanced core designs for the gas cooled fast-breeder reactor
This report summarizes the fuel cycle and advanced core design analysis for the gas-cooled fast breeder reactor (GCFR) performed in conjunction with the Nonproliferation Alternative Systems Assessment Program (NASAP) and the International Nuclear Fuel Cycle Evaluation (INFCE) between 1976 and 1980. The report contrasts traditional fast reactor fuel cycles (plutonium/uranium) with alternative (uranium/thorium) cycles in an effort to define fuel systems which might reduce nuclear weapons proliferation risks without incurring extreme resource or economic disadvantages. It studies symbiotic reactor systems involving fast and thermal reactors and provides basic GCFR mass flow information for NASAP and INFCE. It defines an improved fuel cycle performance index (i.e., energy potential) for reactors and reactor systems. This improved index is combined with sensitivity studies of core materials and configurations to select the basic core and primary system characteristics for future core designs. This report establishes and characterizes advanced designs which may represent targets for future commercial-size fast reactors.
Hamilton, C.J.; Perkins, R.G.
A particle-bed gas cooled fast reactor core design for waste minimization.
The issue of waste minimization in advanced reactor systems has been investigated using the Particle-Bed Gas-Cooled Fast Reactor (PB-GCFR) design being developed and funded under the U.S. Department of Energy Nuclear Energy Research Initiative (USDOE NERI) Program. Results indicate that for the given core power density and constraint on the maximum TRU enrichment allowable, the lowest amount of radiotoxic transuranics to be processed and hence sent to the repository is obtained for long-life core designs. Calculations were additionally done to investigate long-life core designs using LWR spent fuel TRU and recycle TRU, and different feed, matrix and reflector materials. The recycled TRU and LWR spent TRU fuels give similar core behaviors, because of the fast spectrum environment which does not significantly degrade the TRU composition. Using light elements as reflector material was found to be unattractive because of power peaking problems and large reactivity swings. The application of a lead reflector gave the longest cycle length and lowest TRU processing requirement. Materials compatibility and performance issues require additional investigation.
Hoffman, E. A.; Taiwo, T. A.; Yang, W. S.; Fatone, M.
Sensitivity Analysis of Neutron Cross-Sections Considered for Design and Safety Studies of Lfr and SFR Generation IV Systems
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.
Tucek, Kamil; Carlsson, Johan; Wider, Hartmut
Neutronics and thermal design analyses of US solid breeder blanket for ITER
The US Solid Breeder Blanket is designed to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Safety, low tritium inventory, reliability, flexibility cost, and minimum R D requirements are the other design criteria. To satisfy these criteria, the produced tritium is recovered continuously during operation and the blanket coolant operates at low pressure. Beryllium multiplier material is used to control the solid-breeder temperature. Neutronics and thermal design analyses were performed in an integrated manner to define the blanket configuration. The reference parameters of ITER including the operating scenarios, the neutron wall loading distribution and the copper stabilizer are included in the design analyses. Several analyses were performed to study the impact of the reactor parameters, blanket dimensions, material characteristics, and heat transfer coefficient at the material interfaces on the blanket performance. The design analyses and the results from the different studies are summarized. 6 refs., 3 figs., 3 tabs.
Gohar, Y.; Billone, M.; Attaya, H. (Argonne National Lab., IL (USA)); Sawan, M. (Wisconsin Univ., Madison, WI (USA))
This report describes the design and fabrication techniques for the instrumented fuel elements of the Annular Core Research Reactor (ACRR). The thermocouple assemblies were designed and fabricated at Sandia Laboratories while the instrumented elements were assembled at Los Alamos Scientific Laboratory. In order to satisfy the ACRR's Technical Specifications, the thermocouples are required to measure temperature in excess of 1800/sup 0/C under rapid heating conditions. Because of the potentially high failure rates for thermocouples in such environments, the instrumented fuel elements are designed so that the thermocouples can be replaced easily.
Schmidt, T.R.; Sasmor, D.J.; Martin, J.T.; Gonzalez, F.; Cox, D.N.
Shielding design of a treatment room for an accelerator-based neutron source for BNCT
For several years, research has been ongoing in the Ohio State University (OSU) Nuclear Engineering Program toward the development of an accelerator-based irradiation facility (ANIF) neutron source for boron neutron capture therapy (BNCT). The ANIF, which is planned to be built in a hospital, has been conceptually designed and analyzed. After Qu, an OSU researcher, determined that the shielding design of a 6-MV X-ray treatment room was inadequate to protect personnel from an accelerator neutron source operating at 30 mA, we decided to analyze and determine the shielding requirements of a treatment room for an ANIF. We determined the amount of shielding that would be sufficient to protect facility personnel from excessive radiation exposure caused by operation of the accelerator at 30 mA.
Evans, J.F.; Blue, T.E. [Ohio State Univ., Columbus, OH (United States)
Design of a hohlraum-driven exploding pusher capsule experiment for NIF neutron diagnostic calibration
Neutron diagnostics are a critical component of the National Ignition Facility and measure key parameters of capsule performance such as neutron yield, ion temperature, neutron bang time, and down scattered ratio. Therefore, accurate calibration of these diagnostics is essential. Such calibration requires high DT yield (greater than 1.e14) with low ?R as well as a symmetric implosion -- an implosion with small non-radial velocities remaining in the fuel during the burn phase. In order to meet these requirements, we have used the radiation-hydrodynamics code HYDRA to design an indirect-drive exploding pusher capsule experiment, driven with a 2.5 ns pulse in a vacuum hohlraum. Features of this design will be presented as well as its feasibility for symmetry control.
Berzak Hopkins, L.; Callahan, D.; Divol, L.; Le Pape, S.; Meezan, N.; Masse, L.
A low overhead design for testability and test generation technique for core-based systems-on-a-chip
In a fundamental paradigm shift in system design, entire systems are being built on a single chip, using multiple embedded cores. Though the newest system design methodology has several advantages in terms of time-to-market and system cost, testing such core-based systems is difficult, mainly due to the problem of justifying test sequences at the inputs of a core embedded deep
Indradeep Ghosh; Niraj K. Jha; Sujit Dey
The design and installation of a core discharge monitor for CANDU-type reactors
A new type of surveillance systems that monitors neutron and gamma radiation in a reactor containment is being installed at the Ontario Hydro Darlington Nuclear Generating Station A, Unit 2. Unlike video or film surveillance that monitors mechanical motion, this system measures fuel-specific radiation emanating from irradiated fuel as it is pushed from the core of CANDU-type reactors. Proof-of-principle measurements have been carried out at Bruce Nuclear Generating Station A, Unit 3. The system uses ({gamma},n) threshold detectors and ionization detectors. A microprocessor-based electronics package, GRAND-II (Gamma Ray and Neutron Detector electronics package), provides detector bias, preamplifier power, and signal processing. Firmware in the GRAND-2 controls the surveillance activities, including data acquisition and a level of detector authentication, and it handles authenticated communication with a central data logging computer. Data from the GRAND-II are transferred to an MS-DOS-compatible computer and stored. These data are collected and reviewed for fuel-specific radiation signatures from the primary detector and proper ratios of signals from secondary detectors. 5 figs.
Halbig, J.K. (Los Alamos National Lab., NM (USA)); Monticone, A.C.; Ksiezak, L. (International Atomic Energy Agency, Vienna (Austria)); Smiltnieks, V. (International Atomic Energy Agency, Toronto, ON (Canada). Regional Office)
An approach for catalyst design in artificial photosynthetic systems: focus on nanosized inorganic cores within proteins.
Some enzymes can be considered as a catalyst having a nanosized inorganic core in a protein matrix. In some cases, the metal oxide or sulfide clusters, which can be considered as cofactors in enzymes, may be recruited for use in other related reactions in artificial photosynthetic systems. In other words, one approach to design efficient and environmentally friendly catalysts in artificial photosynthetic systems for the purpose of utilizing sunlight to generate high energy intermediates or useful material is to select and utilize inorganic cores of enzymes. For example, one of the most important goals in developing artificial photosynthesis is hydrogen production. However, first, it is necessary to find a "super catalyst" for water oxidation, which is the most challenging half reaction of water splitting. There is an efficient system for water oxidation in cyanobacteria, algae, and plants. Published data on the Mn-Ca cluster have provided details on the mechanism and structure of the water oxidizing complex as a Mn-Ca nanosized inorganic core in photosystem II. Progress has been made in introducing Mn-Ca oxides as efficient catalysts for water oxidation in artificial photosynthetic systems. Here, in the interest of designing efficient catalysts for other important reactions in artificial photosynthesis, a few examples of our knowledge of inorganic cores of proteins, and how Nature used them for important reactions, are discussed. PMID:23377954
Najafpour, Mohammad Mahdi
Design of an energy-independent spherical-type neutron dose monitor
The thickness of a spherical-type neutron dose monitor has been optimized in order to obtain improved dose response compared to that of existing instruments. The monitor structure includes four layers of the spherical shell. The outermost layer is composed of a poly-methyl methacrylate (PMMA) shell. A boron nitride shell is used as the second layer while two polyethylene shells are used as the third and the core layers. Twelve radial directional thermo-luminescent detectors (TLD) detectors are arranged between the layers at two different depths and one in the center. Considering the reaction rate of 6Li(n,?) T for the TLD detectors, a series of calculations were performed using MCNP5 by varying the thickness of the layers. Neutron cross-section libraries based on JENDL-3.3 were applied for the calculations. To evaluate the dose response of the monitor, the spectrum of the D 2O-moderated 252Cf source was used as mentioned in ISO 8529. The monitor is sensitive to measurements of three different energy groups at three depths of the moderator. The neutron dose was calculated from the responses of a linear combination of TLD group detectors. The dose response was improved and well agree with expected dose, H*(10), compared to that of existing instruments, especially for intermediate energies of neutrons. The total ambient dose equivalent was varied only 3% for two different incident directions, hence the direction distribution was found small for dose measurement.
Bhuiya, Sariful Haque; Yamanishi, Hirokuni; Uda, Tatsuhiko
Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor
The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.
Abdullah, Y.; Hamid, N. A.; Mansor, M. A.; Ahmad, M. H. A. R. M.; Yusof, M. R.; Yazid, H.; Mohamed, A. A.
Multipurpose Intense 14 MeV Neutron Source at Bratislava: Design Study.
The present state of design of the multipurpose intense 14 MeV neutron source based on a D exp + ion beam and a metal tritide target is reported. It is essentially a 300 keV electrostatic air insulated accelerator capable to accelerate a deuterium ion bea...
J. Pivarc S. Hlavac J. Kral P. Oblozinsky I. Ribansky
Los Alamos National Laboratory is responsible for the design of the room-temperature linac for the Spallation Neutron Source (SNS). This linac consists of a Coupled-Cavity Linac (CCL) and a Drift Tube Linac (DTL). During normal operation, about 80% of the Radio Frequency (RF) power is dissipated in the DTL cavity walls. This waste heat causes the cavities to expand, causing
Design and simulation of a neutron source based on an electron linear accelerator for BNCT of skin melanoma.
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
Pazirandeh, Ali; Torkamani, Ali; Taheri, Ali
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.
Lodha, Nupur; Rai, Nivesh; Dubey, Rahul; Venkataraman, Hrishikesh
Optimum design of imploded core plasma for effective fast ignition at GXII
In the implosion phase of the fast ignition scheme, most critical issues are breakup of the cone tip and the formation of high ?-R core plasma to improve its heating efficiency. For the integrated fast ignition experiment at ILE Osaka University, robust and reliable implosion must be redesign. In this paper, feasible target design under the constraint condition of existing GXII and LFEX facilities is studied using two-dimensional radiation hydrodynamic simulations, and an optimum target design based on low velocity implosion is proposed. The advantages of low velocity implosion are low adiabat, robust against Rayleigh-Taylor instability, which are verified. Also longer life time of compressed core plasma which is preferable for fast ignition is confirmed in this study.
Nagatomo, H.; Johzaki, T.; Sunahara, A.; Sakagami, H.; Yanagawa, T.; Mima, K.
Neutronics Design and Fuel Cycle Analysis of a High Conversion BWR with Pu-Th Fuel
As part of the U.S. Department of Energy's (DOE) Nuclear Energy Research Initiative (NERI), a 'Generation IV' high conversion Boiling Water Reactor design is being investigated at Purdue University and Brookhaven National Laboratory. One of the primary innovative design features of the core proposed here is the use of Thorium as fertile material. In addition to the advantageous nonproliferation and
Yunlin Xu; T. J. Downar; H. Takahashi
Preliminary design report for SCDAP/RELAP5 lower core plate model
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.
Coryell, E.W. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.; Griffin, F.P. [Oak Ridge National Lab., TN (United States)
Energy Efficient Engine integrated core/low spool design and performance report
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.
A Microwave Radiometer for Close Proximity Core Body Temperature Monitoring: Design, Development, and Experimentation
Presented is a radiometric sensor and associated electromagnetic propagation models, developed to facilitate non-invasive core body temperature extraction. The system has been designed as a close-proximity sensor to detect thermal emissions radiated from deep-seated tissue 1 cm � 3 cm inside the human body. The sensor is intended for close proximity health monitoring applications, with potential implications for deployment into
Quenton Bonds
Analytical Results for Design Space Exploration of Many-Core Processors for Sound Synthesis of Guitar
In this paper, we present a design space exploration of optimal many-core processors for the physics-based sound synthesis of guitar by quantitatively evaluating the impact of a sample-per-processing element (SPE) ratio - the amount of sample data directly mapped to each processing element (or varying the number of processing elements for a fixed sample size) on system performance and efficiency
Myeongsu Kang; Jiwon Choi; Jong-Myon Kim
Conceptual Design Study of 180 MWt Small-Sized Reduced-Moderation Water Reactor Core
Conceptual design of a Small-sized Reduced-Moderation Water Reactor (S-RMWR) core, which has the thermal output of 180 MW, the conversion ratio of 1.0 and the void reactivity coefficient of negative value, has been constructed. S-RMWR is a technology demonstration reactor which also conducts material and fuel testing for commercial use of Reduced-Moderation Water Reactor (RMWR) in large-scale power plants. It
Yoshihiro NAKANO; Tsutomu OKUBO; Sadao UCHIKAWA
Dual optical tweezers integrated in a four-core fiber: design and simulation
We propose a novel dual optical fiber tweezers integrated in a four-core fiber which can trap, rotate and orient a micro particle immersed in a fluid medium. We design the structures and the functions of this dual optical fiber tweezers, and simulate the optical trapping forces, optical torques exerting on the micro particle. We also give out the experimental setup and the controlling method of this integrated dual optical fiber tweezers.
Zhang, Yu; Yuan, Libo; Liu, Zhihai; Yang, Jun
Integrating IMS Learning Design and IMS Question and Test Interoperability using CopperCore Service Integration
Please, cite this publication as: Vogten, H., Martens, H., Nadolski, R., Tattersall, C., van Rosmalen, P., & Koper, R. (2006). Integrating IMS Learning Design and IMS Question and Test Interoperability using CopperCore Service Integration. Proceedings of International Workshop in Learning Networks for Lifelong Competence Development, TENCompetence Conference. March 30th-31st, Sofia, Bulgaria: TENCompetence. Retrieved June 30th, 2006, from http:\\/\\/dspace.learningnetworks.org
Hubert Vogten; Harrie Martens; Rob Nadolski; Colin Tattersall; Peter Van Rosmalen; Rob Koper
The reversed-field-pinch (RFP) fusion neutron source: A conceptual design
The conceptual design of an ohmically heated, reversed-field pinch (RFP) operating at /approximately/5-MW/m/sup 2/ steady-state DT fusion neutron wall loading and /approximately/124-MW total fusion power is presented. These results are useful in projecting the development of a cost effective, low input power (/approximately/206 MW) source of DT neutrons for large-volume (/approximately/10 m/sup 3/), high-fluence (3.4 MW yr/m/sup 2/) fusion nuclear materials and technology testing. 19 refs., 15 figs., 9 tabs.
Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Werley, K.A.
Design of a neutron converter for fission studies at the IGISOL facility
The upgraded IGISOL facility with JYFLTRAP, at the accelerator laboratory of the University of Jyv�skyl�, has been supplied with a new cyclotron which will provide proton or deuteron beams of the order of 100 ?A with up to 30 MeV energy. This makes it an ideal place for measurements of neutron-induced fission fragments from various actinides, in view of proposed future nuclear fuel cycles. In the present paper, some considerations for the design of a neutron converter, based on simulations with the Monte Carlo codes MCNPX and FLUKA, are described.
Lantz, M.; Gorelov, D.; Mattera, A.; Penttil�, H.; Pomp, S.; Rados, D.; Ryzhov, I.; the IGISOL Group
Lunar in-core thermionic nuclear reactor power system conceptual design
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.
Mason, Lee S.; Schmitz, Paul C.; Gallup, Donald R.
Design of the neutron imaging pinhole for use at the national ignition facility
The Neutron Imaging (NI) diagnostic is designed to be used at the National Ignition Facility (NIF). This instrument will be used to image both primary (14MeV neutrons) and down scattered (6-8MeV neutrons). The pinhole body sits 225mm from the target, while the scintillator and recording systems are located 28m from the target. The diagnostic uses port 90, 315 and the recording system is located in a specifically built room located outside of switchyard I. The location of the pinhole and the recording system combine to give a magnification of 104. The recording of both the primary and downscattered image is done by recording the image from both the front and back side of the scintillator.
Fatherley, Valerie E [Los Alamos National Laboratory; Day, Robert D [Los Alamos National Laboratory; Garcia, Felix P [Los Alamos National Laboratory; Grim, Gary P [Los Alamos National Laboratory; Oertel, John A [Los Alamos National Laboratory; Wilde, Carl H [Los Alamos National Laboratory; Wilke, Mark D [Los Alamos National Laboratory
Comments on ``design optimization of a small-angle neutron scattering spectrometer''
Marga�a, Falc�o, Salgado and Carvahlo [Nucl. Instr. and Meth. A274 (1989) 606, and Proc. Int. Conf. on Neutron Scattering, Grenoble, Physica B156/157 (1989) 608] have recently discussed the design parameters of a small-angle scattering spectrometer. They claim to have shown that the instrument can be optimized for constant angular resolution and a fixed neutron source area, such that the detector count rate is higher than that obtained with a spectrometer optimized for a fixed total length and having the same angular resolution. They also claim that the relative gain in intensity depends upon the size of the neutron emitting surface available to the spectrometer. The initial assumptions made by the authors are too constrained to be generally applicable and therefore misleading conclusions could be drawn, and in fact the equal path length arrangement is optimum.
Mildner, D. F. R.
Small angle neutron scattering study of nano sized microstructure in Fe-Cr ODS steels for gen IV in-core applications.
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
Han, Young-Soo; Mao, Xiadong; Jang, Jinsung
Design and performance of a pulse transformer based on Fe-based nanocrystalline core
A dry-type pulse transformer based on Fe-based nanocrystalline core with a load of 0.88 nF, output voltage of more than 65 kV, and winding ratio of 46 is designed and constructed. The dynamic characteristics of Fe-based nanocrystalline core under the impulse with the pulse width of several microseconds were studied. The pulse width and incremental flux density have an important effect on the pulse permeability, so the pulse permeability is measured under a certain pulse width and incremental flux density. The minimal volume of the toroidal pulse transformer core is determined by the coupling coefficient, the capacitors of the resonant charging circuit, incremental flux density, and pulse permeability. The factors of the charging time, ratio, and energy transmission efficiency in the resonant charging circuit based on magnetic core-type pulse transformer are analyzed. Experimental results of the pulse transformer are in good agreement with the theoretical calculation. When the primary capacitor is 3.17 ?F and charge voltage is 1.8 kV, a voltage across the secondary capacitor of 0.88 nF with peak value of 68.5 kV, rise time (10%-90%) of 1.80 ?s is obtained.
Yi, Liu; Xibo, Feng; Lin, Fuchang
Design and operation of the core topography data acquisition system for TMI-2
Development of effective procedures for recovery from the 1979 accident at the Three Mile Island 2 nuclear station requires a detailed and quantitative description of the postaccident configuration of the core. This report describes the techniques, equipment, and procedures used for making precise ultrasonic, sonar-like measurements of the cavity left in the upper core region as a result of the accident and details the primary results of the measurements. The system developed for the measurements uses computer techniques for the command and control of remote mechanical and electronic equipment, and for data acquisition and reduction. The system was designed, fabricated, and tested; procedures developed; and personnel trained in 4-1/2 months. The primary results are detailed topographic maps of the cavity. A variety of visual aids was developed to supplement the maps and aid in interpreting companion videotape surveys. The measurements reveal a cavity of 9.3 m/sup 3/, approximately 26% of the total core volume. The cavity occupies most of the full diameter of the core to an average depth of about 1.5 m and approaches 2 m in places.
Beller, L.S.; Brown, H.L.
Design and performance of a pulse transformer based on Fe-based nanocrystalline core.
A dry-type pulse transformer based on Fe-based nanocrystalline core with a load of 0.88 nF, output voltage of more than 65 kV, and winding ratio of 46 is designed and constructed. The dynamic characteristics of Fe-based nanocrystalline core under the impulse with the pulse width of several microseconds were studied. The pulse width and incremental flux density have an important effect on the pulse permeability, so the pulse permeability is measured under a certain pulse width and incremental flux density. The minimal volume of the toroidal pulse transformer core is determined by the coupling coefficient, the capacitors of the resonant charging circuit, incremental flux density, and pulse permeability. The factors of the charging time, ratio, and energy transmission efficiency in the resonant charging circuit based on magnetic core-type pulse transformer are analyzed. Experimental results of the pulse transformer are in good agreement with the theoretical calculation. When the primary capacitor is 3.17 ?F and charge voltage is 1.8 kV, a voltage across the secondary capacitor of 0.88 nF with peak value of 68.5 kV, rise time (10%-90%) of 1.80 ?s is obtained. PMID:21895262
Conceptual design of thorium-fuelled Mitrailleuse accelerator-driven subcritical reactor using D-Be neutron source
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)
Kokubo, Y. [Quan Japan Company Limited, 3-9-15 Sannomiya-cho, Chuo-ku, Kobe, Hyogo, 650-0021 (Japan); Kamei, T. [Research Inst. for Applied Sciences, 49 Tanaka Ohicho, Sakyo-ku, Kyoto-shi, Kyoto, 606-8202 (Japan)
Dosimetric comparison of four new design {sup 103}Pd brachytherapy sources: Optimal design using silver and copper rod cores
Four new brachytherapy sources, IRA1-{sup 103}Pd, IRA2-{sup 103}Pd, IRA3-{sup 103}Pd, and IRA4-{sup 103}Pd, have been developed at Agricultural, Medical, and Industrial Research School and are designed for permanent implant application. With the goal of determining an optimal design for a {sup 103}Pd source, this article compares the dosimetric properties of these sources with reference to the authors' earlier IRA-{sup 103}Pd source. The four new sources differ in end cap configuration and thickness and in the core material, silver or copper, that carries the adsorbed {sup 103}Pd. Dosimetric data derived from the authors' Monte Carlo simulation results are reported in accordance with the updated AAPM Task Group No. 43 report (TG-43U1). For each source, the authors obtained detailed results for the dose rate constant {Lambda}, the radial dose function g(r), the anisotropy function F(r,{theta}), and the anisotropy factor {phi}{sub an}(r). In this study, the optimal source IRA3-{sup 103}Pd provides the most isotropic dose distribution in water with the dose rate constant of 0.678({+-}0.1%) cGy h{sup -1} U{sup -1}. The IRA3-{sup 103}Pd design has a silver rod core combined with thin-wall, concave end caps. Finally, the authors compared the results for their optimal source with published results for those of other source manufacturers.
Hosseini, S. Hamed; Sadeghi, Mahdi; Ataeinia, Vahideh [Agricultural, Medical and Industrial Research School, Nuclear Science and Technology Research Institute, P.O. Box 31485-498, Karaj, Tehran (098)21 (Iran, Islamic Republic of)