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1

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

2014-06-01

2

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)

2012-07-01

3

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

SciTech Connect

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.

1996-02-01

4

Core Vessel Insert Handling Robot for the Spallation Neutron Source  

SciTech Connect

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

2011-01-01

5

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

NASA Astrophysics Data System (ADS)

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

6

Automated Core Design  

SciTech Connect

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

2005-07-15

7

CORE DESIGN CHARACTERISTICS OF THE HYPER SYSTEM  

Microsoft Academic Search

In Korea, an accelerator-driven system (ADS) called HYPER (Hybrid Power Extraction Reactor) is being studied for the transmutation of the radioactive wastes. HYPER is a 1 000 MWth lead- bismuth eutectic (LBE)-cooled ADS. In this paper, the neutronic design characteristics of HYPER are described and its transmutation performances are assessed for an equilibrium cycle. The core is loaded with a

Yonghee Kim; Won-Seok Park; R. N. Hill

8

FAST FOSSIL ROTATION OF NEUTRON STAR CORES  

SciTech Connect

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: amelatos@unimelb.edu.au [School of Physics, University of Melbourne, Parkville, VIC 3010 (Australia)

2012-12-10

9

Shear viscosity in neutron star cores  

E-print Network

We calculate the shear viscosity $\\eta = \\eta_{e\\mu}+\\eta_{n}$ in a neutron star core composed of nucleons, electrons and muons ($\\eta_{e\\mu}$ being the electron-muon viscosity, mediated by collisions of electrons and muons with charged particles, and $\\eta_{n}$ the neutron viscosity, mediated by neutron-neutron and neutron-proton collisions). Deriving $\\eta_{e\\mu}$, we take into account the Landau damping in collisions of electrons and muons with charged particles via the exchange of transverse plasmons. It lowers $\\eta_{e\\mu}$ and leads to the non-standard temperature behavior $\\eta_{e\\mu}\\propto T^{-5/3}$. The viscosity $\\eta_{n}$ is calculated taking into account that in-medium effects modify nucleon effective masses in dense matter. Both viscosities, $\\eta_{e\\mu}$ and $\\eta_{n}$, can be important, and both are calculated including the effects of proton superfluidity. They are presented in the form valid for any equation of state of nucleon dense matter. We analyze the density and temperature dependence of $\\eta$ for different equations of state in neutron star cores, and compare $\\eta$ with the bulk viscosity in the core and with the shear viscosity in the crust.

P. S. Shternin; D. G. Yakovlev

2008-08-21

10

Simplified cut core inductor design  

NASA Technical Reports Server (NTRS)

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

Mclyman, W. T.

1974-01-01

11

The Thermal Mass limit of Neutron Cores  

E-print Network

Static thermal equilibrium of a quantum self-gravitating ideal gas in General Relativity is studied at any temperature, taking into account the Tolman-Ehrenfest effect. Thermal contribution to the gravitational stability of static neutron cores is quantified. The curve of maximum mass with respect to temperature is reported. At low temperatures is recovered the Oppenheimer-Volkoff calculation, while at high temperatures is recovered the, recently reported, classical gas calculation. An ultimate upper mass limit $M = 2.43M_\\odot$ of all maximum values is found to occur at Tolman temperature $ T = 1.27mc^2$ with radius $R = 15.2km$.

Roupas, Zacharias

2014-01-01

12

The Thermal Mass limit of Neutron Cores  

E-print Network

Static thermal equilibrium of a quantum self-gravitating ideal gas in General Relativity is studied at any temperature, taking into account the Tolman-Ehrenfest effect. Thermal contribution to the gravitational stability of static neutron cores is quantified. The curve of maximum mass with respect to temperature is reported. At low temperatures is recovered the Oppenheimer-Volkoff calculation, while at high temperatures is recovered the, recently reported, classical gas calculation. An ultimate upper mass limit $M = 2.43M_\\odot$ of all maximum values is found to occur at Tolman temperature $ T = 1.27mc^2$ with radius $R = 15.2km$.

Zacharias Roupas

2014-11-19

13

Thermal mass limit of neutron cores  

NASA Astrophysics Data System (ADS)

Static thermal equilibrium of a quantum self-gravitating ideal gas in general relativity is studied at any temperature, taking into account the Tolman-Ehrenfest effect. Thermal contribution to the gravitational stability of static neutron cores is quantified. The curve of maximum mass with respect to temperature is reported. At low temperatures the Oppenheimer-Volkoff calculation is recovered, while at high temperatures the recently reported classical gas calculation is recovered. An ultimate upper mass limit M =2.43 M? of all maximum values is found to occur at Tolman temperature T =1.27 mc2 with radius R =15.2 km .

Roupas, Zacharias

2015-01-01

14

HFIR cold neutron source moderator vessel design analysis  

SciTech Connect

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.

1998-04-01

15

Annular core research reactor high flux neutron radiography facility  

Microsoft Academic Search

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

1990-01-01

16

A NOVEL MICROMEGAS DETECTOR FOR IN-CORE NUCLEAR REACTOR NEUTRON FLUX MEASUREMENTS  

E-print Network

1 A NOVEL MICROMEGAS DETECTOR FOR IN-CORE NUCLEAR REACTOR NEUTRON FLUX MEASUREMENTS S. ANDRIAMONJE Talence Cedex, France Future fast nuclear reactors designed for energy production and transmutation to neutron detection inside nuclear reactor is given. The advantage of this detector over conventional

Paris-Sud XI, Université de

17

SmAHTR-CTC Neutronic Design  

SciTech Connect

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

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

2014-01-01

18

Preliminary engineering design of sodium-cooled CANDLE core  

NASA Astrophysics Data System (ADS)

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

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

2012-06-01

19

Superfluidity in the Core of Neutron Stars  

NASA Astrophysics Data System (ADS)

The year (1958) after the publication of the BCS theory, Bohr, Mottelson & Pines showed that nuclei should also contain superfluid neutrons and superconducting protons. In 1959, A. Migdal proposed that neutron superfluidity should also occur in the interior of neutron stars. Pairing in nuclei forms Cooper pairs with zero spin, but the relevant component of the nuclear interaction becomes repulsive at densities larger than the nuclear matter density. It has been proposed that neutron-neutron interaction in the spin-triplet state, and L=1 orbital angular momentum, that is known to be attractive from laboratory experiments, may result in a new form of neutron superfluidity in the neutron star interior. I will review our present understanding of the structure of neutron stars and describe how superfluidity strongly affects their thermal evolution. I will show how a ``Minimal Model'' that excludes the presence of ``exotic'' matter (Bose condensates, quarks, etc.) is compatible with most observations of the surface temperatures of young isolated neutron stars in the case this neutron superfluid exists. Compared to the case of isotropic spin-zero Cooper pairs, the formation of anisotropic spin-one Cooper pairs results in a strong neutrino emission that leads to an enhanced cooling of neutron stars after the onset of the pairing phase transition and allows the Minimal Cooling scenario to be compatible with most observations. In the case the pairing critical temperature Tc is less than about 6 x10^8 K, the resulting rapid cooling of the neutron star may be observable. It was recently reported that 10 years of Chandra observations of the 333 year young neutron star in the Cassiopeia A supernova remnant revealed that its temperature has dropped by about 5%. This result indicates that neutrons in this star are presently becoming superfluid and, if confirmed, provides us with the first direct observational evidence for neutron superfluidity at supra-nuclear densities.

Page, Dany

2013-04-01

20

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

PubMed

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

Lashkari, A; Khalafi, H; Kazeminejad, H

2013-05-01

21

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. PMID:24976672

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

2013-01-01

22

Pyrochemical neutron multiplicity counter design  

SciTech Connect

Pyrochemical process materials are difficult to measure using conventional neutron counting methods because of significant self- multiplication and variable ({alpha},n) reaction rates. Multiplicity counters measure the first three moments of the neutron multiplicity distribution and thus make it possible to determine sample mass even when multiplication and ({alpha},n) rate are unknown. A new multiplicity counter suitable for inplant measurement of pyrochemical process materials has been designed using Monte Carlo simulations. The goals were to produce a counter that has high neutron detection efficiency, low die-away time, a flat spatial efficiency profile, and is insensitive to the neutron energy spectrum. Monte Carlo calculations were performed for several prototype models consisting of four rings of 71-cm active length {sup 3}He tubes in a polyethylene body. The cadmium-lined sample well is 25 cm in diameter to accommodate a wide variety of inplant sample containers. The counter can be free-standing or in-line without mechanical modification. The calculations were performed to determine the above design criteria for several configurations of tube spacing, cadmium liners, and sample height. Calculations were also performed for distributed sample sources to understand the integrated effects of variable neutron spectra on the counter. 5 refs., 8 figs., 1 tab.

Langner, D.G.; Ensslin, N.; Krick, M.S.

1990-01-01

23

GCFR core thermal-hydralic design  

SciTech Connect

The approach for developing the thermal-hydraulic core assembly designs for the gas-cooled fast reactor (GCFR) is reviewed, and key considerations for improving the core performance at all power and flow conditions are discussed. It is shown how the thermal-hydraulic core assembly designs evolve from evaluations of plant size, material limitations, safety criteria, and structural performance considerations.

Schleuter, G.; Baxi, C.B.; Bennett, F.O.

1980-05-01

24

Optimization of IEC grid design for maximum neutron production  

SciTech Connect

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)

1996-12-31

25

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

SciTech Connect

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

Chiang, R. T. [AREVA NP Inc., 6399 San Ignacio Ave., San Jose, CA 95119 (United States)

2012-07-01

26

One pass core design of a super fast reactor  

SciTech Connect

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

Liu, Qingjie; Oka, Yoshiaki [Cooperative Major in Nuclear Energy, Waseda University, Tokyo 169-8555 (Japan)

2013-07-01

27

Neutronic feasibility design of a small long-life HTR  

Microsoft Academic Search

Small high temperature gas-cooled reactors (HTRs) have the advantages of transportability, modular construction and flexible site selection. This paper presents the neutronic feasibility design of a 20MWth U-Battery, which is a long-life block-type HTR. Key design parameters and possible reactor core configurations of the U-Battery were investigated by SCALE 5.1. The design parameters analyzed include fuel enrichment, the packing fraction

Ming Ding; Jan Leen Kloosterman

28

Shield Design for a Space Based Vapor Core Reactor  

SciTech Connect

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

Knight, Travis; Anghaie, Samim [Innovative Nuclear Space Power and Propulsion Institute (INSPI), PO Box 116502, University of Florida, Gainesville, FL 32611-6502 (United States)

2002-07-01

29

Experiment Design and Analysis Guide - Neutronics & Physics  

SciTech Connect

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

Misti A Lillo

2014-06-01

30

A proposal of a new in-core neutron monitor using nuclear pumped laser  

Microsoft Academic Search

A new type of in-core neutron monitor using the 3He-Ne nuclear pumped laser (NPL) is proposed and the optimum design and characteristics of this detector are discussed, especially for application to the fast breeder reactor (FBR). Through this optimum design study, a highest efficiency monitor has been found to show the simple and practical relation of PR = 3.54 [atom

A. Okumura; S. Soramoto; H. Hayashida; H. Nakamura; M. Nakazawa

1991-01-01

31

Sensitivity of detecting in-core vibrations and boiling in pressurized water reactors using ex-core neutron detectors  

Microsoft Academic Search

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

1984-01-01

32

Design Study of Small Lead-Cooled Fast Reactor Cores Using SiC Cladding and Structure  

Microsoft Academic Search

Neutronics of a reactor core with SiC cladding and structure was compared with that with steel cladding and structure analytically for small lead-cooled fast reactors. Uranium nitride fuel was used for this reactor. U235 enrichment was 11% in inner core and 13% in outer core for relatively flat neutron flux distributions and power density distribution. The core design was optimized

Abu Khalid Rivai; Minoru Takahashi

2007-01-01

33

Recent Problems of Transformer Core Design  

NASA Astrophysics Data System (ADS)

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

Valkovic, Z.

1988-01-01

34

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

SciTech Connect

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.

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

1985-06-01

35

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

NASA Astrophysics Data System (ADS)

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

2011-02-01

36

NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE  

SciTech Connect

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

John D. Bess

2014-03-01

37

Design of multidirectional neutron beams for boron neutron capture synovectomy  

SciTech Connect

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

Gierga, D.P.; Yanch, J.C. [Massachusetts Institute of Technology, Cambridge, MA (United States); Shefer, R.E. [Newton Scientific, Inc., Cambridge, MA (United States)

1997-12-01

38

CHINA SPALLATION NEUTRON SOURCE DESIGN.  

SciTech Connect

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.

WEI,J.

2007-01-29

39

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

Microsoft Academic Search

A conceptual design of a lead-bismuth-eutectic (LBE)-cooled actinide burner core with innovative streaming fuel assemblies (FAs) is described. The 1800-MW(thermal) core employs metallic, fertile-free fuel where the transuranics (plutonium plus minor actinides) are dispersed in a zirconium matrix. The core contains 157 streaming FAs that enhance neutron streaming by employing gas-filled, sealed streaming tubes at the FA periphery and center.

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

2001-01-01

40

Measurement of the vortex core in sub-100 nm Fe dots using polarized neutron scattering  

E-print Network

OFFPRINT Measurement of the vortex core in sub-100 nm Fe dots using polarized neutron scattering neutron scattering Igor V. Roshchin1,2 , Chang-Peng Li2(a) , Harry Suhl2 , Xavier Batlle3 , S. Roy2(b diffraction and scattering Abstract ­ We use polarized neutron scattering to obtain quantitative information

Roshchin, Igor V.

41

Stellar encounters involving neutron stars in globular cluster cores  

NASA Technical Reports Server (NTRS)

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.

1992-01-01

42

A Monte-Carlo method for ex-core neutron response  

SciTech Connect

A Monte Carlo neutron transport kernel capability primarily for ex-core neutron response is described. The capability consists of the generation of a set of response kernels, which represent the neutron transport from the core to a specific ex-core volume. This is accomplished by tagging individual neutron histories from their initial source sites and tracking them throughout the problem geometry, tallying those that interact in the geometric regions of interest. These transport kernels can subsequently be combined with any number of core power distributions to determine detector response for a variety of reactor Thus, the transport kernels are analogous to an integrated adjoint response. Examples of pressure vessel response and ex-core neutron detector response are provided to illustrate the method.

Gamino, R.G.; Ward, J.T.; Hughes, J.C. [Lockheed Martin Corp., Schenectady, NY (United States)

1997-10-01

43

Preliminary neutronics calculations for conversion of the Tehran research reactor core from HEU to LEU fuel  

Microsoft Academic Search

The 5-MW highly enriched uranium (HEU)-fueled Tehran Research Reactor is considered for conversion to high-density, low-enriched uranium (LEU) fuel. A preliminary neutronics calculation is performed as part of the conversion goal. In this study, two cores are considered: the HEU reference core and a proposed LEU core similar to the reference core, and a proposed LEU core similar to the

Nejat; S. M. R

1993-01-01

44

Energy Efficient Engine core design and performance report  

NASA Technical Reports Server (NTRS)

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

1982-01-01

45

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

NASA Astrophysics Data System (ADS)

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

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

46

Advanced Neutron Source: Plant Design Requirements  

SciTech Connect

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.

Not Available

1990-07-01

47

Design and safety aspects of the Cornell cold neutron source  

Microsoft Academic Search

The cold neutron beam facility at the Cornell University TRIGA Mark II reactor will begin operational testing in early 1993. It is designed to provide a low background subthermal neutron beam that is as free as possible of fast neutrons and gamma rays for applied research and graduate-level instruction. The Cornell cold neutron source differs from the more conventional types

Carol G. Ouellet; David D. Clark

1992-01-01

48

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

Microsoft Academic Search

A conceptual neutronic design of the cold-neutron source (CNS) for the Egyptian second research reactor (ETRR-2) was done using the MCNP code. Parametric analysis to chose the type and geometry of the moderator, and the required CNS dimensions to maximize the cold neutron production was performed. The moderator cell has a spherical annulus structure containing liquid hydrogen. The cold neutron

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

2005-01-01

49

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

2012-01-01

50

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

Microsoft Academic Search

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

1983-01-01

51

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

SciTech Connect

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)

2012-07-01

52

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

PubMed

The fast neutron irradiation facility of the Portuguese Research Reactor was characterized after the reduction in uranium enrichment and rearrangement of the core configuration. In this work we report on the determination of the hardness parameter and the 1MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard. PMID:21071234

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

2011-08-01

53

Long-life neutron detector for instrumentation of a nuclear reactor core  

Microsoft Academic Search

The invention concerns a neutron detector adapted for use in the core ; instrumentation of nuclear reactors. It comprises an emitter electrode which, ; upon irradiation with neutrons, emits electrons by means of (n,$gamma$) ; processes, a collector electrode, and a dielectric disposed between the emitter ; and the collector electrode. Thulium 169 or terbium 159 is used as the

Klar

1975-01-01

54

The Spallation Neutron Source accelerator system design  

NASA Astrophysics Data System (ADS)

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

Henderson, S.; Abraham, W.; Aleksandrov, A.; Allen, C.; Alonso, J.; Anderson, D.; Arenius, D.; Arthur, T.; Assadi, S.; Ayers, J.; Bach, P.; Badea, V.; Battle, R.; Beebe-Wang, J.; Bergmann, B.; Bernardin, J.; Bhatia, T.; Billen, J.; Birke, T.; Bjorklund, E.; Blaskiewicz, M.; Blind, B.; Blokland, W.; Bookwalter, V.; Borovina, D.; Bowling, S.; Bradley, J.; Brantley, C.; Brennan, J.; Brodowski, J.; Brown, S.; Brown, R.; Bruce, D.; Bultman, N.; Cameron, P.; Campisi, I.; Casagrande, F.; Catalan-Lasheras, N.; Champion, M.; Champion, M.; Chen, Z.; Cheng, D.; Cho, Y.; Christensen, K.; Chu, C.; Cleaves, J.; Connolly, R.; Cote, T.; Cousineau, S.; Crandall, K.; Creel, J.; Crofford, M.; Cull, P.; Cutler, R.; Dabney, R.; Dalesio, L.; Daly, E.; Damm, R.; Danilov, V.; Davino, D.; Davis, K.; Dawson, C.; Day, L.; Deibele, C.; Delayen, J.; DeLong, J.; Demello, A.; DeVan, W.; Digennaro, R.; Dixon, K.; Dodson, G.; Doleans, M.; Doolittle, L.; Doss, J.; Drury, M.; Elliot, T.; Ellis, S.; Error, J.; Fazekas, J.; Fedotov, A.; Feng, P.; Fischer, J.; Fox, W.; Fuja, R.; Funk, W.; Galambos, J.; Ganni, V.; Garnett, R.; Geng, X.; Gentzlinger, R.; Giannella, M.; Gibson, P.; Gillis, R.; Gioia, J.; Gordon, J.; Gough, R.; Greer, J.; Gregory, W.; Gribble, R.; Grice, W.; Gurd, D.; Gurd, P.; Guthrie, A.; Hahn, H.; Hardek, T.; Hardekopf, R.; Harrison, J.; Hatfield, D.; He, P.; Hechler, M.; Heistermann, F.; Helus, S.; Hiatt, T.; Hicks, S.; Hill, J.; Hill, J.; Hoff, L.; Hoff, M.; Hogan, J.; Holding, M.; Holik, P.; Holmes, J.; Holtkamp, N.; Hovater, C.; Howell, M.; Hseuh, H.; Huhn, A.; Hunter, T.; Ilg, T.; Jackson, J.; Jain, A.; Jason, A.; Jeon, D.; Johnson, G.; Jones, A.; Joseph, S.; Justice, A.; Kang, Y.; Kasemir, K.; Keller, R.; Kersevan, R.; Kerstiens, D.; Kesselman, M.; Kim, S.; Kneisel, P.; Kravchuk, L.; Kuneli, T.; Kurennoy, S.; Kustom, R.; Kwon, S.; Ladd, P.; Lambiase, R.; Lee, Y. Y.; Leitner, M.; Leung, K.-N.; Lewis, S.; Liaw, C.; Lionberger, C.; Lo, C. C.; Long, C.; Ludewig, H.; Ludvig, J.; Luft, P.; Lynch, M.; Ma, H.; MacGill, R.; Macha, K.; Madre, B.; Mahler, G.; Mahoney, K.; Maines, J.; Mammosser, J.; Mann, T.; Marneris, I.; Marroquin, P.; Martineau, R.; Matsumoto, K.; McCarthy, M.; McChesney, C.; McGahern, W.; McGehee, P.; Meng, W.; Merz, B.; Meyer, R.; Meyer, R.; Miller, B.; Mitchell, R.; Mize, J.; Monroy, M.; Munro, J.; Murdoch, G.; Musson, J.; Nath, S.; Nelson, R.; Nelson, R.; O'Hara, J.; Olsen, D.; Oren, W.; Oshatz, D.; Owens, T.; Pai, C.; Papaphilippou, I.; Patterson, N.; Patterson, J.; Pearson, C.; Pelaia, T.; Pieck, M.; Piller, C.; Plawski, T.; Plum, M.; Pogge, J.; Power, J.; Powers, T.; Preble, J.; Prokop, M.; Pruyn, J.; Purcell, D.; Rank, J.; Raparia, D.; Ratti, A.; Reass, W.; Reece, K.; Rees, D.; Regan, A.; Regis, M.; Reijonen, J.; Rej, D.; Richards, D.; Richied, D.; Rode, C.; Rodriguez, W.; Rodriguez, M.; Rohlev, A.; Rose, C.; Roseberry, T.; Rowton, L.; Roybal, W.; Rust, K.; Salazer, G.; Sandberg, J.; Saunders, J.; Schenkel, T.; Schneider, W.; Schrage, D.; Schubert, J.; Severino, F.; Shafer, R.; Shea, T.; Shishlo, A.; Shoaee, H.; Sibley, C.; Sims, J.; Smee, S.; Smith, J.; Smith, K.; Spitz, R.; Staples, J.; Stein, P.; Stettler, M.; Stirbet, M.; Stockli, M.; Stone, W.; Stout, D.; Stovall, J.; Strelo, W.; Strong, H.; Sundelin, R.; Syversrud, D.; Szajbler, M.; Takeda, H.; Tallerico, P.; Tang, J.; Tanke, E.; Tepikian, S.; Thomae, R.; Thompson, D.; Thomson, D.; Thuot, M.; Treml, C.; Tsoupas, N.; Tuozzolo, J.; Tuzel, W.; Vassioutchenko, A.; Virostek, S.; Wallig, J.; Wanderer, P.; Wang, Y.; Wang, J. G.; Wangler, T.; Warren, D.; Wei, J.; Weiss, D.; Welton, R.; Weng, J.; Weng, W.-T.; Wezensky, M.; White, M.; Whitlatch, T.; Williams, D.; Williams, E.; Wilson, K.; Wiseman, M.; Wood, R.; Wright, P.; Wu, A.; Ybarrolaza, N.; Young, K.; Young, L.; Yourd, R.; Zachoszcz, A.; Zaltsman, A.; Zhang, S.; Zhang, W.; Zhang, Y.; Zhukov, A.

2014-11-01

55

MPACT Fast Neutron Multiplicity System Design Concepts  

SciTech Connect

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

2012-10-01

56

Advanced BWR core component designs and the implications for SFD analysis  

SciTech Connect

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.

1997-02-01

57

Design Study of Small Lead-Cooled Fast Reactor Cores Using SiC Cladding and Structure  

NASA Astrophysics Data System (ADS)

Neutronics of a reactor core with SiC cladding and structure was compared with that with steel cladding and structure analytically for small lead-cooled fast reactors. Uranium nitride fuel was used for this reactor. U235 enrichment was 11% in inner core and 13% in outer core for relatively flat neutron flux distributions and power density distribution. The core design was optimized using natural uranium blanket and nitride fuel for long life-core with reshuffling interval of 15 years. The analytical result indicated that neutron energy spectrum was slightly softer in the core with the SiC cladding and structure than that with steel cladding and structure. The SiC type reactor was designed to have criticality at the beginning of cycle (BOC), although the steel type reactor could not have criticality with the same size and geometry. In other words, the SiC type core can be designed smaller than the steel type core. The peak power densities could remain constant over the reactor operation. The consumption capability of uranium was quite high, i.e. 10% for 125 MWt reactor and 18.4% for 375 MWt reactor at the end of cycle (EOC).

Rivai, Abu Khalid; Takahashi, Minoru

58

The evolution of core and surface magnetic field in isolated neutron stars  

E-print Network

We apply the model of flux expulsion from the superfluid and superconductive core of a neutron star, developed by Konenkov & Geppert (2000), both to neutron star models based on different equations of state and to different initial magnetic field structures. When initially the core and the surface magnetic field are of the same order of magnitude, the rate of flux expulsion from the core is almost independent of the equation of state, and the evolution of the surface field decouples from the core field evolution with increasing stiffness. When the surface field is initially much stronger than the core field, the magnetic and rotational evolution resembles to those of a neutron star with a purely crustal field configuration; the only difference is the occurence of a residual field. In case of an initially submerged field significant differences from the standard evolution occur only during the early period of neutron star's life, until the field has been rediffused to the surface. The reminder of the episode of submergence is a correlation of the residual field strength with the submergence depth of the initial field. We discuss the effect of the rediffusion of the magnetic field on to the difference between the real and the active age of young pulsars and on their braking indices. Finally, we estimate the shear stresses built up by the moving fluxoids at the crust--core interface and show that preferentially in neutron stars with a soft equation of state these stresses may cause crust cracking.

D. Konenkov; U. Geppert

2001-03-04

59

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

NASA Technical Reports Server (NTRS)

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

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

1981-01-01

60

Design an Interactive Visualization System for Core Drilling Expeditions  

E-print Network

and studies have proposed participatory design [2] and user-centered design [3]. However real-world usersDesign an Interactive Visualization System for Core Drilling Expeditions Using Immersive Empathic Method Overview In this paper, we propose an immersive empathic design method and used it to create

Johnson, Andrew

61

Conceptual design for TRR-II neutron guide system  

NASA Astrophysics Data System (ADS)

The Taiwan Research Reactor modified Project (TRR-II) is going to construct the neutron guide system (NGS) in order to transport cold neutrons, with well-defined angular and energy characteristics, from cold neutron source (CNS) to the distant guide hall of 3060 m 2. The conceptual design of NGS will optimize the performance and safety of the whole system considering the scientific and engineering options. The NGS starts from a specialy designed cold neutron beam tube (BT-8) that looks onto the CNS and contains four primary neutron guides, whose cross sections are 615, 520, 615 and 617 cm 2, respectively. These neutron guides are coated with either Ni-58 mirror, or 2 ?C or 3 ?C supermirrors. In the guide tunnel of the reactor building, the primary neutron guides are connected to the curved guides in order to remove the gamma rays and fast neutrons. A curved neutron guide is subdivided into two up and down guides by inserting two partition mirrors and separated from each other by the opposite curved guides. In the guide hall, neutron beams are supplied to each instrument using benders or monochromators. The design parameters of the neutron guides are optimized using the neutron guide simulation code. The safety concepts of the NGS are also described in this paper.

Yang, T. N.; Tsai, W. F.; Chen, J. S.; Huang, Y. H.; Kawai, T.

2002-01-01

62

Advanced Neutron Source radiological design criteria  

SciTech Connect

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

Westbrook, J.L.

1995-08-01

63

Spallation neutron source target station design, development, and commissioning  

NASA Astrophysics Data System (ADS)

The spallation neutron source target station is designed to safely, reliably, and efficiently convert a 1 GeV beam of protons to a high flux of about 1 meV neutrons that are available at 24 neutron scattering instrument beam lines. Research and development findings, design requirements, design description, initial checkout testing, and results from early operation with beam are discussed for each of the primary target subsystems, including the mercury target, neutron moderators and reflector, surrounding vessels and shielding, utilities, remote handling equipment, and instrumentation and controls. Future plans for the mercury target development program are also briefly discussed.

Haines, J. R.; McManamy, T. J.; Gabriel, T. A.; Battle, R. E.; Chipley, K. K.; Crabtree, J. A.; Jacobs, L. L.; Lousteau, D. C.; Rennich, M. J.; Riemer, B. W.

2014-11-01

64

The Encapsulated Nuclear Heat Source (ENHS) Reactor Core Design  

SciTech Connect

A once-for-life, uniform composition, blanket-free and fuel-shuffling-free reference core has been designed for the Encapsulated Nuclear Heat Source (ENHS) to provide the design goals of a nearly zero burnup reactivity swing throughout {approx}20 yr of full-power operation up to the peak discharge burnup of more than 100 GWd/t HM. What limits the core life is radiation damage to the HT-9 structural material. The temperature coefficients of reactivity are all negative, except for that of the coolant expansion. However, the negative reactivity coefficient associated with the radial expansion of the core structure can compensate for the coolant thermal expansion. The void coefficient is positive but of no safety concern because the boiling temperature of lead or lead-bismuth is so high that there is no conceivable mechanism for the introduction of significant void fraction into the core. The core reactivity coefficients, reactivity worth, and power distributions are almost constant throughout the core life.It was found possible to design such once-for-life cores using different qualities of Pu and transuranics as long as U is used as the primary fertile material. It is also feasible to design ENHS cores using nitride rather than metallic fuel. Relative to the reference metallic fuel core, nitride fuel cores offer up to {approx}25% higher discharge burnup and longer life, up to {approx}38% more energy per core, a significantly more negative Doppler reactivity coefficient, and less positive coolant expansion and coolant void reactivity coefficient but a somewhat smaller negative fuel expansion reactivity coefficient. The pitch-to-diameter ratio (1.45 of the nitride fuel cores using enriched N) is larger than that (1.36) for the reference metallic fuel core, implying a reduction of the coolant friction loss, thus enabling an increase in the power level that can be removed from the core by natural circulation cooling.It is also possible to design Pu-U(10Zr) fueled ENHS-type cores using Na as the primary coolant with either Na or Pb-Bi secondary coolants. The Na-cooled cores feature a tighter lattice and are therefore more compact but have spikier power distribution, more positive coolant temperature reactivity coefficients, and smaller reactivity worth of the control elements.

Hong, Ser Gi [Korea Atomic Energy Research Institute (Korea, Republic of); Greenspan, Ehud [University of California, Berkeley (United States); Kim, Yeong Il [Korea Atomic Energy Research Institute (Korea, Republic of)

2005-01-15

65

Thermal conductivity due to phonons in the core of superfluid neutron stars  

E-print Network

We compute the contribution of phonons to the thermal conductivity in the core of superfluid neutron stars. We use effective field theory techniques to extract the phonon scattering rates, written as a function of the equation of state of the system. We also calculate the phonon dispersion law beyond linear order, which depends on the gap of superfluid neutron matter. With all these ingredients, we solve the Boltzmann equation numerically using a variational approach. We find that the thermal conductivity $\\kappa$ is dominated by combined small and large angle binary collisions. As in the color-flavor-locked superfluid, we find that our result can be well approximated by $\\kappa \\propto 1/ \\Delta^6$, where $\\Delta$ is the neutron gap, the constant of proportionality depending on the density. We further comment on the possible relevance of electron and superfluid phonon collisions in obtaining the total contribution to the thermal conductivity in the core of superfluid neutron stars.

Cristina Manuel; Sreemoyee Sarkar; Laura Tolos

2014-11-22

66

Thermal conductivity due to phonons in the core of superfluid neutron stars  

NASA Astrophysics Data System (ADS)

We compute the contribution of phonons to the thermal conductivity in the core of superfluid neutron stars. We use effective field theory techniques to extract the phonon scattering rates, written as a function of the equation of state of the system. We also calculate the phonon dispersion law beyond linear order, which depends on the gap of superfluid neutron matter. With all these ingredients, we solve the Boltzmann equation numerically using a variational approach. We find that the thermal conductivity ? is dominated by combined small- and large-angle binary collisions. As in the color-flavor-locked superfluid, we find that our result can be well approximated by ? ?1 /?6 at low temperature, where ? is the neutron gap, the constant of proportionality depending on the density. We further comment on the possible relevance of electron and superfluid phonon collisions in obtaining the total contribution to the thermal conductivity in the core of superfluid neutron stars.

Manuel, Cristina; Sarkar, Sreemoyee; Tolos, Laura

2014-11-01

67

De novo design of the hydrophobic cores of proteins.  

PubMed

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

Desjarlais, J R; Handel, T M

1995-10-01

68

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

SciTech Connect

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

Ivanov, E.; Tiberi, V.; Ecrabet, F.; Chegrani, Y.; Canuti, E.; Bisogni, D.; Sargeni, A.; Bernard, F. [Institut de Radioprotection et de Surete Nucleaire IRSN, BP 17, 92262 Fontenay-aux-roses (France)

2012-07-01

69

De novo design of the hydrophobic core of ubiquitin.  

PubMed

We have previously reported the development and evaluation of a computational program to assist in the design of hydrophobic cores of proteins. In an effort to investigate the role of core packing in protein structure, we have used this program, referred to as Repacking of Cores (ROC), to design several variants of the protein ubiquitin. Nine ubiquitin variants containing from three to eight hydrophobic core mutations were constructed, purified, and characterized in terms of their stability and their ability to adopt a uniquely folded native-like conformation. In general, designed ubiquitin variants are more stable than control variants in which the hydrophobic core was chosen randomly. However, in contrast to previous results with 434 cro, all designs are destabilized relative to the wild-type (WT) protein. This raises the possibility that beta-sheet structures have more stringent packing requirements than alpha-helical proteins. A more striking observation is that all variants, including random controls, adopt fairly well-defined conformations, regardless of their stability. This result supports conclusions from the cro studies that non-core residues contribute significantly to the conformational uniqueness of these proteins while core packing largely affects protein stability and has less impact on the nature or uniqueness of the fold. Concurrent with the above work, we used stability data on the nine ubiquitin variants to evaluate and improve the predictive ability of our core packing algorithm. Additional versions of the program were generated that differ in potential function parameters and sampling of side chain conformers. Reasonable correlations between experimental and predicted stabilities suggest the program will be useful in future studies to design variants with stabilities closer to that of the native protein. Taken together, the present study provides further clarification of the role of specific packing interactions in protein structure and stability, and demonstrates the benefit of using systematic computational methods to predict core packing arrangements for the design of proteins. PMID:9194177

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

1997-06-01

70

Effect of core structure irradiation on the RBMK neutron characteristics  

NASA Astrophysics Data System (ADS)

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

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

2014-12-01

71

Fusion Engineering and Design 38 (1997) 139158 Overview of ARIES-RS neutronics and radiation shielding: key  

E-print Network

, service lifetime, radiation damage, shielding requirements and design, and personnel protection. Major of the fusion power core (FPC). The primary function of the shield is radiation protection: protectionFusion Engineering and Design 38 (1997) 139­158 Overview of ARIES-RS neutronics and radiation

72

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

PubMed

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

2005-01-01

73

Advanced Neutron Source design: Burnout heat flux correlation development  

SciTech Connect

In the Advanced Neutron Source Reactor (ANSR) fuel element region, heat fluxes will be elevated. Early designs corresponded to average and estimated hot-spot fluxes of 11-12 and 21-22 MW/m/sup 2/, respectively. Design changes under consideration may lower these values to about 9 and 17 MW/m/sup 2/. In either event, the development of a satisfactory burnout heat flux correlation is an important element among the many thermal-hydraulic design issues, since the critical power ration will depend in part on its validity. Relatively little work in the area of subcooled-flow burnout has been published over the past 12 years. We have compared seven burnout correlations and modifications thereof with several sets of experimental data, of which the most relevant to the ANS core are presently those referenced. The best overall agreement between the correlations tested and these data is currently provided by a modification of Thorgerson's correlation. 7 refs., 1 tab.

Gambill, W.R.; Mochizuki, T.

1988-01-01

74

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

NASA Astrophysics Data System (ADS)

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

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

75

Neutron single-particle strength outside the N=50 core  

NASA Astrophysics Data System (ADS)

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.

2013-01-01

76

Gravitational waves from axisymmetric rotating stellar core collapse to a neutron star in full general relativity  

E-print Network

Gravitational waves from axisymmetric rotating stellar core collapse to a neutron star in full stars. Gravitational waves are computed using a quadrupole formula. It is found that the waveforms is among the most promis- ing sources of gravitational waves. To date, there has been no systematic work

Shibata, Masaru

77

Neutron kinetics in subcritical cores with application to the source modulation method  

E-print Network

Neutron kinetics in subcritical cores with application to the source modulation method J. Wright The subject of this paper is an investigation of the performance of the so-called source modulation technique detector, including the source modulation method, are based on the assumption of point kinetic behaviour

Pázsit, Imre

78

Simulation of reactivity transients in a miniature neutron source reactor core  

Microsoft Academic Search

Computer simulation was carried out for reactivity induced transients in a HEU core of a tank-in-pool reactor, a miniature neutron source reactor (MNSR). The reactivity transients without scram at initial power of 3 W were studied. From the low power level, the power steadily increased with time and then rose sharply to higher peak values followed by a gradual decrease

E. H. K. Akaho; B. T. Maakuu

2002-01-01

79

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

SciTech Connect

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

Schulz, K.C. [Univ. of Turabo, Gurabo, Puerto (Puerto Rico). College of Engineering; Yahr, G.T. [Oak Ridge National Lab., TN (United States)

1995-08-01

80

Neutronics Assessment of Molten Salt Breeding Blanket Design  

E-print Network

1 Neutronics Assessment of Molten Salt Breeding Blanket Design Options Mohamed Sawan Fusion adequate tritium breeding and shielding for VV and magnet Larger margins are considered to account flow channel required to cool it #12;5 Tritium Breeding Potential If neutron coverage for double null

81

HFIR cold neutron source moderator vessel design analysis  

Microsoft Academic Search

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

1998-01-01

82

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

NASA Astrophysics Data System (ADS)

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

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

1997-02-01

83

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

SciTech Connect

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)

2013-07-01

84

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

Microsoft Academic Search

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

Z. Zhong; Y. Gohar; R. Kellogg

2009-01-01

85

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

SciTech Connect

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

Varvayanni, M.; Catsaros, N.; Stakakis, E. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece); Grigoriadis, D. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece); Department of Mechanical and Manufacturing Engineering, University of Cyprus, P.O. Box 20537, Nicosia 1678 (Cyprus)

2008-07-15

86

Current advances in precious metal core-shell catalyst design  

NASA Astrophysics Data System (ADS)

Precious metal nanoparticles are commonly used as the main active components of various catalysts. Given their high cost, limited quantity, and easy loss of catalytic activity under severe conditions, precious metals should be used in catalysts at low volumes and be protected from damaging environments. Accordingly, reducing the amount of precious metals without compromising their catalytic performance is difficult, particularly under challenging conditions. As multifunctional materials, core-shell nanoparticles are highly important owing to their wide range of applications in chemistry, physics, biology, and environmental areas. Compared with their single-component counterparts and other composites, core-shell nanoparticles offer a new active interface and a potential synergistic effect between the core and shell, making these materials highly attractive in catalytic application. On one hand, when a precious metal is used as the shell material, the catalytic activity can be greatly improved because of the increased surface area and the closed interfacial interaction between the core and the shell. On the other hand, when a precious metal is applied as the core material, the catalytic stability can be remarkably improved because of the protection conferred by the shell material. Therefore, a reasonable design of the core-shell catalyst for target applications must be developed. We summarize the latest advances in the fabrications, properties, and applications of core-shell nanoparticles in this paper. The current research trends of these core-shell catalysts are also highlighted.

Wang, Xiaohong; He, Beibei; Hu, Zhiyu; Zeng, Zhigang; Han, Sheng

2014-08-01

87

A Method Using Nuclear Emulsions for Measuring the Anisotropy of Fast Neutron Fluxes due to Arrangement of Fast Core Cell  

Microsoft Academic Search

Measurements using nuclear emulsions have been made on the neutron spectra and on the fine structure of neutron fluxes in the cell of the I-4 core of the fast critical assembly at the Japan Atomic Energy Research Institute. The I-4 core is a graphite-diluted fast core with 3:1 volume ratio of 20% enriched metallic uranium and graphite. The nuclear emulsions

Shoji NOMOTO; Takehiko YASUNO; Hisashi NAKAMURA

1970-01-01

88

Maximum mass of neutron stars with quark matter core  

SciTech Connect

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)

2012-11-12

89

Design and Demonstration of a Quasi-monoenergetic Neutron Source  

SciTech Connect

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

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

2014-03-05

90

Design and demonstration of a quasi-monoenergetic neutron source  

E-print Network

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 $^{7}$Li(p,n)$^{7}$Be 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.

T. H. Joshi; S. Sangiorgio; V. Mozin; E. B. Norman; P. Sorensen; M. Foxe; G. Bench; A. Bernstein

2014-05-13

91

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

SciTech Connect

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

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

2011-01-01

92

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

SciTech Connect

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

Ehlers, G.; Podlesnyak, A. A.; Niedziela, J. L.; Iverson, E. B. [Neutron Scattering Science Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Sokol, P. E. [Department of Physics, Indiana University, Bloomington, Indiana 47405 (United States)

2011-08-15

93

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

PubMed

The design and performance of the new cold neutron chopper spectrometer (CNCS) at the Spallation Neutron Source in Oak Ridge are described. CNCS is a direct-geometry inelastic time-of-flight spectrometer, designed essentially to cover the same energy and momentum transfer ranges as IN5 at ILL, LET at ISIS, DCS at NIST, TOFTOF at FRM-II, AMATERAS at J-PARC, PHAROS at LANSCE, and NEAT at HZB, at similar energy resolution. Measured values of key figures such as neutron flux at sample position and energy resolution are compared between measurements and ray tracing Monte Carlo simulations, and good agreement (better than 20% of absolute numbers) has been achieved. The instrument performs very well in the cold and thermal neutron energy ranges, and promises to become a workhorse for the neutron scattering community for quasielastic and inelastic scattering experiments. PMID:21895276

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

2011-08-01

94

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

NASA Astrophysics Data System (ADS)

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

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

2011-08-01

95

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

NASA Astrophysics Data System (ADS)

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

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

2011-03-01

96

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

NASA Astrophysics Data System (ADS)

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

97

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

NASA Astrophysics Data System (ADS)

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.

98

Transport properties of a quark-hadron Coulomb lattice in the cores of neutron stars  

NASA Astrophysics Data System (ADS)

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

2012-12-01

99

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

SciTech Connect

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

Steiner, A. W. [Institute for Nuclear Theory, University of Washington, Seattle, WA 98195 (United States); Hempel, M. [Department of Physics, University of Basel, Klingelbergstrasse 82, 4056 Basel (Switzerland); Fischer, T. [Institute for Theoretical Physics, University of Wroclaw, pl. Maxa Borna 9, 50-204, Wroclaw (Poland)

2013-09-01

100

Massive neutron stars with hyperonic core : a case study with the IUFSU model  

E-print Network

The recent discoveries of massive neutron stars, such as PSR J$0348+0432$ and PSR J$1614-2230$, 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 (EoS's) like the GM1 and FSUGold are also questioned. We investigate the existence of hyperonic matter in the central regions of massive neutron stars using Relativistic Mean Field (RMF) theory with the recently proposed IUFSU model. The IUFSU model is extended by including hyperons to study the neutron star in $\\beta$ equilibrium. The effect of different hyperonic potentials, namely $\\Sigma$ and $\\Xi$ 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 $1.93M_{\\odot}$, which is within the 3$\\sigma$ limit of the observed mass of PSR J$0348+0432$, 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.

Bipasha Bhowmick; Madhubrata Bhattacharya; Abhijit Bhattacharyya; G. Gangopadhyay

2014-07-08

101

Massive neutron stars with a hyperonic core: A case study with the IUFSU relativistic effective interaction  

NASA Astrophysics Data System (ADS)

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.

2014-06-01

102

Optimization of a neutron detector design using adjoint transport simulation  

SciTech Connect

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)

2012-07-01

103

Observer Design for a Core Circadian Rhythm Network  

PubMed Central

The paper investigates the observer design for a core circadian rhythm network in Drosophila and Neurospora. Based on the constructed highly nonlinear differential equation model and the recently proposed graphical approach, we design a rather simple observer for the circadian rhythm oscillator, which can well track the state of the original system for various input signals. Numerical simulations show the effectiveness of the designed observer. Potential applications of the related investigations include the real-world control and experimental design of the related biological networks. PMID:25121122

Zhang, Yuhuan

2014-01-01

104

Neutronic analysis of pebble-bed cores with transuranics  

E-print Network

systems designs requires this growth in code systems to accommodate current and future research. Some of the major renovations to version 5.1 are in the following areas: square4 New ORIGEN-ARP libraries square4 New covariance libraries in TSUNAMI...

Pritchard, Megan Leigh

2009-05-15

105

Two stochastic optimization algorithms applied to nuclear reactor core design  

Microsoft Academic Search

Two stochastic optimization algorithms conceptually similar to Simulated Annealing are presented and applied to a core design optimization problem previously solved with Genetic Algorithms. The two algorithms are the novel Particle Collision Algorithm (PCA), which is introduced in detail, and Dueck's Great Deluge Algorithm (GDA). The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and

Wagner F. Sacco; Cassiano R. E. de oliveira; Cludio M. N. A. Pereira

2006-01-01

106

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

Microsoft Academic Search

A tool was developed to calculate the average cold neutron flux that could be generated for a spherically shaped cold neutron source situated in the center of a nuclear reactor core. The tool also estimates the subsequent nuclear heating of the cold source. The results were compared for three different cold source mediums; hydrogen, deuterium and helium. The tool utilizes

Malek Chatila

2003-01-01

107

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

SciTech Connect

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.

2011-09-30

108

EU Blanket Design Activities and Neutronics Support Efforts  

SciTech Connect

An overview is provided of the design activities and the related neutronics support efforts conducted in the European Union for the development of breeder blankets for future fusion power reactors. The EU fusion programme considers two blanket lines, the Helium-Cooled Pebble Bed (HCPB) blanket with Lithium ceramics pebbles (Li{sub 4}SiO{sub 4} or Li{sub 2}TiO{sub 3}) as breeder and beryllium pebbles as neutron multiplier, and the Helium-Cooled Lithium-Lead (HCLL) blanket with the Pb-Li eutectic alloy as breeder and neutron multiplier. The blanket design and the related R and D efforts are based on the use of the same coolant and the same modular blanket structure to minimise the development costs as much as possible. The neutronic support efforts include design analyses for the layout and optimization of the modular HCPB/HCLL blankets based on detailed three-dimensional Monte Carlo calculations as well as underlying neutronics activities conducted in the frame of the European Fusion and Activation File (EFF/EAF) projects to develop qualified nuclear data and computational tools for reliable neutronics design calculations.

Fischer, U. [Forschungszentrum Karlsruhe (Germany); Batistoni, P. [ENEA Fusion Division (Italy); Boccaccini, L.V. [Forschungszentrum Karlsruhe (Germany); Giancarli, L. [CEA Saclay (France); Hermsmeyer, S. [Forschungszentrum Karlsruhe (Germany); Poitevin, Y. [CEA Saclay (France)

2005-05-15

109

Advanced Core Design And Fuel Management For Pebble-Bed Reactors  

SciTech Connect

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

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

2004-10-01

110

Natural Nuclear Reactor Oklo and Variation of Fundamental Constants Part 1: Computation of Neutronics of Fresh Core  

E-print Network

Using modern methods of reactor physics we have performed full-scale calculations of the natural reactor Oklo. For reliability we have used recent version of two Monte Carlo codes: Russian code MCU REA and world wide known code MCNP (USA). Both codes produce similar results. We have constructed a computer model of the reactor Oklo 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 have estimated also the temperature and void effects for the fresh core. As would be expected, we have found for the fresh core a significant difference between reactor and Maxwell spectra, which was used before for averaging cross sections in the Oklo reactor. The averaged cross section of Sm-149 and its dependence on the shift of resonance position (due to variation of fundamental constants) are significantly different from previous results. Contrary to results of some previous papers we find no evidence for the change of the fine structure constant in the past and obtain new, most accurate limits on its variation with time: -4 10^{-17}year^{-1} < d alpha/dt/alpha < 3 10^{-17} year^{-1} A further improvement in the accuracy of the limits can be achieved by taking account of the core burnup. These calculations are in progress.

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

2005-09-15

111

Neutronic analysis of the conversion of HEU to LEU fuel for a 5MW MTR core  

Microsoft Academic Search

In recent years, due to cessation of highly enriched uranium (HEU) fuel supply, practical steps have been taken to substitute HEU fuel in almost all research reactors by medium-enriched uranium or low-enriched uranium (LEU) fuels. In this study, a neutronic calculation of a 5-MW research reactor core fueled with HEU (93% ²³⁵U) is presented. In order to assess the performance

A. Pazirandeh; G. Bartsch

1987-01-01

112

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

SciTech Connect

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

2009-02-17

113

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

NASA Astrophysics Data System (ADS)

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

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

2009-02-01

114

Design of composite flywheel rotors with soft cores  

NASA Astrophysics Data System (ADS)

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

115

The Spallation Neutron Source (SNS) conceptual design shielding analysis  

Microsoft Academic Search

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

J. O. Johnson; N. Odano; R. A. Lillie

1998-01-01

116

Weapons-Grade Plutonium-Thorium PWR Assembly Design and Core Safety Analysis  

SciTech Connect

A light water reactor (LWR) fuel assembly design consisting of a blend of weapons-grade plutonium and natural thorium oxides was examined. The design meets current thermal-hydraulic and safety criteria. Such an assembly would have enough reactivity to achieve three cycles of operation. The pin power distribution indicates a fairly level distribution across the assembly, avoiding hot spots near guide tubes, corners, and other sections where excessive power would create significant loss to thermal-hydraulic margins.This work examined a number of physics and core safety analysis parameters that impact the operation and safety of power reactors. Such parameters as moderator coefficients of reactivity, Doppler coefficients, soluble boron worth, control rod worth, prompt neutron lifetime, and delayed-neutron fractions were considered. These in turn were used to examine reactor behavior during a number of operational conditions, transients, and accidents. Such conditions as shutdown from power with one rod stuck out, steam-line break accident, feedwater line break, loss of coolant flow, locked rotor accidents, control rod ejection accidents, and anticipated transients without scram (ATWSs) were examined.The analysis of selected reactor transients demonstrated that it is feasible to license and safely operate a reactor fueled with plutonium-thorium blended fuel. In most cases analyzed, the thorium mixture had less-severe consequences than those for a core comprising low-enriched uranium fuel. In the analyzed cases where the consequences were more severe, they were still within acceptable limits. The ATWS accident condition requires more analysis.

Dziadosz, David; Ake, Timothy N.; Saglam, Mehmet; Sapyta, Joe J. [Framatome ANP, Inc. (France)

2004-07-15

117

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

NASA Astrophysics Data System (ADS)

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

118

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

SciTech Connect

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

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

1996-06-01

119

Conceptual design of a neutron camera for MAST Upgrade  

SciTech Connect

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

Weiszflog, M., E-mail: matthias.weiszflog@physics.uu.se; Sangaroon, S.; Cecconello, M.; Conroy, S.; Ericsson, G.; Klimek, I. [Department of Physics and Astronomy, Uppsala University, EURATOM-VR Association, Uppsala (Sweden); Keeling, D.; Martin, R. [CCFE, Culham Science Centre, Abingdon (United Kingdom); Turnyanskiy, M. [ITER Physics Department, EFDA CSU Garching, Boltzmannstrae 2, D-85748 Garching (Germany)

2014-11-15

120

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

NASA Astrophysics Data System (ADS)

It has been shown recently that taking into account strong, weak, electromagnetic, and gravitational interactions, and fulfilling the global charge neutrality of the system, a transition layer will happen between the core and crust of neutron stars, at the nuclear saturation density. We use relativistic mean field theory together with the Thomas-Fermi approximation to study the detailed structure of this transition layer and calculate its surface and Coulomb energy. We find that the surface tension is proportional to a power-law function of the baryon number density in the core bulk region. We also analyze the influence of the electron component and the gravitational field on the structure of the transition layer and the value of the surface tension, to compare and contrast with known phenomenological results in nuclear physics. Based on the above results we study the instability against Bohr-Wheeler surface deformations in the case of neutron stars obeying global charge neutrality. Assuming the core-crust transition at nuclear density ?core?2.71014 g cm-3, we find that the instability sets the upper limit to the crust density, ?crustcrit?1.21014 g cm-3. This result implies a nonzero lower limit to the maximum electric field of the core-crust transition surface and makes inaccessible a limit of quasilocal charge neutrality in the limit ?crust=?core. The general framework presented here can be also applied to study the stability of sharp phase transitions in hybrid stars as well as in strange stars, both bare and with outer crust. The results of this work open the way to a more general analysis of the stability of these transition surfaces, accounting for other effects such as gravitational binding, centrifugal repulsion, magnetic field induced by rotating electric field, and therefore magnetic dipole-dipole interactions.

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

2014-03-01

121

Design of the Core 2-4 GHz Betatron Equalizer  

SciTech Connect

The core betatron equalizer in the Accumulator in the Antiproton Source at Fermilab needed to be upgraded. The performance could be rated as only circa 650 MHz when the system was a 2 GHz system. The old equalizer did not correct for the strong phase mismatch for the relatively strong gain of the system slightly below 2 GHz. The design corrects this phase mismatch and is relatively well matched both in and out of band.

Deibele, C.; /Fermilab

2000-01-01

122

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

SciTech Connect

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

Dracoulis, G. D.; Lane, G. J.; Davidson, P. M.; Kibedi, T.; Nieminen, P.; Maier, K. H.; Watanabe, H. [Department of Nuclear Physics, R. S. Phys. S. E., Australian National University, Canberra ACT 0200 (Australia); Byrne, A. P.; Wilson, A. N. [Department of Nuclear Physics, R. S. Phys. S. E., Australian National University, Canberra ACT 0200 (Australia); Department of Physics, The Faculties, Australian National University, Canberra ACT 0200 (Australia)

2008-03-15

123

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

NASA Technical Reports Server (NTRS)

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.

1979-01-01

124

Beamed Core Antimatter Propulsion: Engine Design and Optimization  

E-print Network

A conceptual design for beamed core antimatter propulsion is reported, where electrically charged annihilation products directly generate thrust after being deflected and collimated by a magnetic nozzle. Simulations were carried out using the Geant4 (Geometry and tracking) software toolkit released by the CERN accelerator laboratory for Monte Carlo simulation of the interaction of particles with matter and fields. Geant permits a more sophisticated and comprehensive design and optimization of antimatter engines than the software environment for simulations reported by prior researchers. The main finding is that effective exhaust speeds Ve ~ 0.69c (where c is the speed of light) are feasible for charged pions in beamed core propulsion, a major improvement over the Ve ~ 0.33c estimate based on prior simulations. The improvement resulted from optimization of the geometry and the field configuration of the magnetic nozzle. Moreover, this improved performance is realized using a magnetic field on the order of 10 T at the location of its highest magnitude. Such a field could be produced with today's technology, whereas prior nozzle designs anticipated and required major advances in this area. The paper also briefly reviews prospects for production of the fuel needed for a beamed core engine.

Ronan Keane; Wei-Ming Zhang

2012-01-26

125

Advanced Neutronics Tools for BWR Design Calculations  

SciTech Connect

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)

2006-07-01

126

Effects of core excess reactivity and coolant average temperature on maximum operable time of NIRR-1 miniature neutron source reactor  

Microsoft Academic Search

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

2011-01-01

127

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

SciTech Connect

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

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

2006-09-15

128

Preliminary design study of advanced multistage axial flow core compressors  

NASA Technical Reports Server (NTRS)

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.

1977-01-01

129

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

NASA Astrophysics Data System (ADS)

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

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

2014-12-01

130

Shield design for next-generation, low-neutron-fluence, superconducting tokamaks  

SciTech Connect

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.

1985-01-01

131

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

SciTech Connect

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

Sun, K. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Chenu, A. [Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Mikityuk, K.; Krepel, J. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Chawla, R. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland)

2012-07-01

132

Preliminary scoping safety analyses of the limiting design basis protected accidents for the Fast Flux Test Facility tritium production core  

SciTech Connect

The SAS4A/SASSYS-l computer code is used to perform a series of analyses for the limiting protected design basis transient events given a representative tritium and medical isotope production core design proposed for the Fast Flux Test Facility. The FFTF tritium and isotope production mission will require a different core loading which features higher enrichment fuel, tritium targets, and medical isotope production assemblies. Changes in several key core parameters, such as the Doppler coefficient and delayed neutron fraction will affect the transient response of the reactor. Both reactivity insertion and reduction of heat removal events were analyzed. The analysis methods and modeling assumptions are described. Results of the analyses and comparison against fuel pin performance criteria are presented to provide quantification that the plant protection system is adequate to maintain the necessary safety margins and assure cladding integrity.

Heard, F.J.

1997-11-19

133

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

PubMed

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

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

2004-11-01

134

High Flux Isotope Reactor cold neutron source reference design concept  

SciTech Connect

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

1998-05-01

135

Core polarization for the electric quadrupole moment of neutron-rich Aluminum isotopes  

E-print Network

The core polarization effect for the electric quadrupole moment of the neutron-rich $^{31}$Al, $^{33}$Al and $^{35}$Al isotopes in the vicinity of the island of inversion are investigated by means of the microscopic particle-vibration coupling model in which the Skyrme Hartee-Fock-Bogoliubov and quasiparticle-random-phase approximation are used to calculate the single-quasiparticle wave functions and the excitation modes. It is found that the polarization charge for the proton $1d_{5/2}$ hole state in $^{33}$Al is quite sensitive to coupling to the neutrons in the $pf$-shell associated with the pairing correlations, and that the polarization charge in $^{35}$Al becomes larger due to the stronger collectivity of the low-lying quadrupole vibrational mode in the neighboring $^{36}$Si nucleus.

Kenichi Yoshida

2009-02-18

136

Electron-muon heat conduction in neutron star cores via the exchange of transverse plasmons  

E-print Network

We calculate the thermal conductivity of electrons and muons kappa_{e-mu} produced owing to electromagnetic interactions of charged particles in neutron star cores and show that these interactions are dominated by the exchange of transverse plasmons (via the Landau damping of these plasmons in nonsuperconducting matter and via a specific plasma screening in the presence of proton superconductivity). For normal protons, the Landau damping strongly reduces kappa_{e-mu} and makes it temperature independent. Proton superconductivity suppresses the reduction and restores the Fermi-liquid behavior kappa_{e-mu} ~ 1/T. Comparing with the thermal conductivity of neutrons kappa_n, we obtain kappa_{e-mu}> kappa_n for T>2 GK in normal matter and for any T in superconducting matter with proton critical temperatures T_c>3e9 K. The results are described by simple analytic formulae.

P. S. Shternin; D. G. Yakovlev

2007-05-14

137

MCNP simulation of TRR-II cold neutron source design  

Microsoft Academic Search

A cold neutron source (CNS) for the modified Taiwan Research Reactor (TRR-II) was designed using the MCNP code. Liquid hydrogen at 20K is selected as a moderator. The moderator cell has a structure of a cylindrical annulus in which the outer shell contains liquid hydrogen and the inner shell has only hydrogen vapor. The moderator cell is made of a

Tai-Cheng Guung; Wen-Fa Tsai; Weng-Sheng Kuo; Jong-Sheng Chen

2002-01-01

138

Advanced Neutron Source reactor control and plant protection systems design  

Microsoft Academic Search

This paper describes the reactor control and plant protection systems' conceptual design of the Advanced Neutron Source (ANS). The Plant Instrumentation, Control, and Data Systems and the Reactor Instrumentation and Control System of the ANS are planned as an integrated digital system with a hierarchical, distributed control structure of qualified redundant subsystems and a hybrid digital\\/analog protection system to achieve

J. L. Anderson; R. E. Battle; J. March-Leuba; M. I. Khayat

1992-01-01

139

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

SciTech Connect

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

Jean Ragusa; Karen Vierow

2011-09-01

140

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

NASA Astrophysics Data System (ADS)

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.

2014-06-01

141

Design of a boron neutron capture enhanced fast neutron therapy assembly  

SciTech Connect

The use of boron neutron capture to boost tumor dose in fast neutron therapy has been investigated at several fast neutron therapy centers worldwide. This treatment is termed boron neutron capture enhanced fast neutron therapy (BNCEFNT). It is a combination of boron neutron capture therapy (BNCT) and fast neutron therapy (FNT). It is believed that BNCEFNT may be useful in the treatment of some radioresistant brain tumors, such as glioblastoma multiform (GBM). A boron neutron capture enhanced fast neutron therapy assembly has been designed for the Fermilab Neutron Therapy Facility (NTF). This assembly uses a tungsten filter and collimator near the patient's head, with a graphite reflector surrounding the head to significantly increase the dose due to boron neutron capture reactions. The assembly was designed using Monte Carlo radiation transport code MCNP version 5 for a standard 20x20 cm{sup 2} treatment beam. The calculated boron dose enhancement at 5.7-cm depth in a water-filled head phantom in the assembly with a 5x5 cm{sup 2} collimation was 21.9% per 100-ppm {sup 10}B for a 5.0-cm tungsten filter and 29.8% for a 8.5-cm tungsten filter. The corresponding dose rate for the 5.0-cm and 8.5-cm thick filters were 0.221 and 0.127 Gy/min, respectively; about 48.5% and 27.9% of the dose rate of the standard 10x10 cm{sup 2} fast neutron treatment beam. To validate the design calculations, a simplified BNCEFNT assembly was built using four lead bricks to form a 5x5 cm{sup 2} collimator. Five 1.0-cm thick 20x20 cm{sup 2} tungsten plates were used to obtain different filter thicknesses and graphite bricks/blocks were used to form a reflector. Measurements of the dose enhancement of the simplified assembly in a water-filled head phantom were performed using a pair of tissue-equivalent ion chambers. One of the ion chambers is loaded with 1000-ppm natural boron (184-ppm {sup 10}B) to measure dose due to boron neutron capture. The measured dose enhancement at 5.0-cm depth in the head phantom for the 5.0-cm thick tungsten filter is (16.6 {+-} 1.8)%, which agrees well with the MCNP simulation of the simplified BNCEFNT assembly, (16.4 {+-} 0.5)%. The error in the calculated dose enhancement only considers the statistical uncertainties. The total dose rate measured at 5.0-cm depth using the non-borated ion chamber is (0.765 {+-} 0.076) Gy/MU, about 61% of the fast neutron standard dose rate (1.255Gy/MU) at 5.0-cm depth for the standard 10x10 cm{sup 2} treatment beam. The increased doses to other organs due to the use of the BNCEFNT assembly were calculated using MCNP5 and a MIRD phantom. The activities of the activation products produced in the BNCEFNT assembly after neutron beam delivery were computed. The photon ambient dose rate due to the radioactive activation products was also estimated.

Wang, Zhonglu; /Georgia Tech; ,

2006-08-01

142

Maximum mass of neutron stars and strange neutron-star cores  

NASA Astrophysics Data System (ADS)

Context. The recent measurement of mass of PSR J1614-2230 rules out most existing models of the equation of state (EOS) of dense matter with high-density softening due to hyperonization that were based on the recent hyperon-nucleon and hyperon-hyperon interactions, which leads to a "hyperon puzzle". Aims: We study a specific solution of this hyperon puzzle that consists of replacing a too soft hyperon core by a sufficiently stiff quark core. In terms of the quark structure of the matter, one replaces a strangeness-carrying baryon phase of confined quark triplets, some of them involving s quarks, by a quark plasma of deconfined u, d, and s quarks. Methods: We constructed an analytic approximation that fits modern EOSs of the two flavor (2SC) and the color-flavor-locked (CFL) color-superconducting phases of quark matter very well. Then, we used it to generate a continuum of EOSs of quark matter. This allowed us to simulate continua of sequences of first-order phase transitions at prescribed pressures, from hadronic matter to the 2SC and then to the CFL state of color-superconducting quark matter. Results: We obtain constraints in the parameter space of the EOS of superconducting quark cores, EOS.Q, resulting from Mmax > 2 M?. These constraints depend on the assumed EOS of baryon phase, EOS.B. We also derive constraints that would result from significantly higher measured masses. For 2.4 M? the required stiffness of the CFL quark core is close to the causality limit while the density jump at the phase transition is very small. Conclusions: The condition Mmax > 2 M? puts strong constraints on the EOSs of the 2SC and CFL phases of quark matter. Density jumps at the phase transitions have to be sufficiently small and sound speeds in quark matter sufficiently large. The condition of thermodynamic stability of the quark phase results in a maximum mass of hybrid stars similar to that of purely baryon stars. This is due to the phase transition of quark matter back to the baryon phase (reconfinement) that we find for both EOS.B. Therefore, to obtain Mmax > 2 M? for hybrid stars, both sufficiently strong additional hyperon repulsion at high-density baryon matter and a sufficiently stiff EOS of quark matter would be needed. However, we think that the high-density instability, which results in the reconfinement of quark matter, indicates actually the inadequacy of the point-particle model of baryons in dense matter at ? ? 5 8?0. We expect that reconfinement can be removed by a sufficient stiffening of the baryon phase, resulting from the repulsive finite size contribution for baryons to the EOS.

Zdunik, J. L.; Haensel, P.

2013-03-01

143

Shielding Design of the Spallation Neutron Source (SNS)  

Microsoft Academic Search

The shielding design is important for the construction of an intense high-energy accelerator facility like the;\\u000aproposed Spallation Neutron Source (SNS) due to its impact on conventional facility design, maintenance;\\u000aoperations, and since the cost for the radiation shielding shares a considerable part of the total facility costs. A;\\u000acalculational strategy utilizing coupled high energy Monte Carlo calculations and multi-dimensional

Johnson

1998-01-01

144

Neutronic design studies for an unattended, low power reactor  

SciTech Connect

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.

1986-01-01

145

Linac design study for an intense neutron-source driver  

SciTech Connect

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.

1993-06-01

146

Linac design study for an intense neutron-source driver  

SciTech Connect

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.

1993-01-01

147

Design of neutron streak camera for fusion diagnostics  

SciTech Connect

The D-T burn time for advanced laser-fusion targets is calculated to be very short, < 50 ps. We describe the design of a neutron streak camera of 16 ps resolving time that can be used to study the temporal history of fusion burn. The cathode of the neutron streak camera is sensitive to neutrons and is curved such that the difference in the neutron path lengths from a point source to various parts of the cathode is compensated by electron transit times within the streak tube. Thus the cathode can be made large for high sensitivity, without sacrificing time resolution. The cathode is coated with 1 ..mu..m UO/sub 2/. Each fission fragment leaving the cathode generates 400 secondary electrons that are all < 20 eV. These electrons are focussed to a point with an extractor and an anode, and are then purified with an electrostatic deflector. The electron beam is streaked and detected with the standard streak camera techniques. Careful shielding is needed for x-rays from the fusion target and general background. It appears that the neutron streak camera can be a viable and unique tool for studying temporal history of fusion burns in D-T plasmas of a few keV ion temperature.

Wang, C.L.; Kalibjian, R.; Singh, M.S.

1982-06-01

148

Development of Newly Designed Ultra-Light Core Structures  

NASA Astrophysics Data System (ADS)

By folding a thin flat sheet with periodically set slits or punched out portions into the third dimension, ultra-lightweight strong and functional core models are newly devised. The basic idea of this modeling arises from the application of origami technique to engineering. Based on the space filling models, fundamental flat cores and skew type sponge cores have been newly developed. By applying these models, such modified core models as curved cores and 3D honeycomb core are newly devised.

Nojima, Taketoshi; Saito, Kazuya

149

The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow  

SciTech Connect

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

2008-07-15

150

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

SciTech Connect

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: ferdman@jb.man.ac.uk [CSIRO Astronomy and Space Science, Australia Telescope National Facility, Epping, NSW 1710 (Australia)

2013-04-10

151

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

NASA Astrophysics Data System (ADS)

The paper describes the development and testing of an Inconel600 (2 mm diameter21 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 -4.410 -18 A/R/h/cm (-9.310 -24 A/ ?/cm 2-s/cm), -5.210 -18 A/R/h/cm (-1.13310 -23 A/ ?/cm 2-s/cm) and 3410 -18 A/R/h/cm (7.1410 -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 610 -23 A/nv/cm while the corresponding sensitivities for the platinum and cobalt detectors were 1.6910 -22 and 2.6410 -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.

2007-04-01

152

Advanced Neutron Source: Plant Design Requirements. Revision 4  

SciTech Connect

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.

Not Available

1990-07-01

153

3D Neutron Transport PWR Full-core Calculation with RMC code  

NASA Astrophysics Data System (ADS)

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

2014-06-01

154

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

SciTech Connect

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

2009-05-01

155

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

SciTech Connect

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)

2012-07-01

156

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

SciTech Connect

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.

1995-08-01

157

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

SciTech Connect

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)

2013-07-01

158

Neutron transport analysis for nuclear reactor design  

DOEpatents

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.

1993-11-30

159

Neutron transport analysis for nuclear reactor design  

DOEpatents

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)

1993-01-01

160

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

NASA Astrophysics Data System (ADS)

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 310 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.

2011-09-01

161

A new 122 mm electromechanical drill for deep ice-sheet coring (DISC): 1. Design concepts  

Microsoft Academic Search

The Deep Ice Sheet Coring (DISC) drill, developed by Ice Coring and Drilling Services (ICDS) under contract with the US National Science Foundation, is an electromechanical drill designed to take 122 mm diameter ice cores to depths of 4000 m. The conceptual design of the DISC drill was developed in 2002\\/03 based on science requirements written by K. Taylor and

Alexander J. Shturmakov; Donald A. Lebar; William P. Mason; Charles R. Bentley

2007-01-01

162

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

NASA Astrophysics Data System (ADS)

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

Bao, S. S.; Shen, H.

2015-01-01

163

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

E-print Network

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

Bao, S S

2015-01-01

164

Thermal neutron measurements of the Rhode Island Nuclear Science Center reactor after conversion to a compact low enriched uranium core  

NASA Astrophysics Data System (ADS)

The Rhode Island Nuclear Science Center reactor, a 2 MW light water research reactor, was converted from a highly enriched uranium core to a more compact low enriched uranium core in 1993. Thermal neutron flux measurements of the instrumented beamports indicate increased thermal flux and a large improvement in the thermal to epithermal ratio. Measurements of irradiation facilities indicate an increased thermal flux, and more potential irradiation positions are now available.

Crow, M. L.; Jeng, U.; Nunes, A. C.; Malik, S. S.; Lin, D.; Bai, S.; Tehan, T.; Jacob, N.; Johnson, D. G.; Simoneau, W. A.; DiMeglio, A. F.

1995-02-01

165

A three-dimensional neutronics-thermohydraulics simulation of core disruptive accident in sodium-cooled fast reactor  

Microsoft Academic Search

The SIMMER-III code is a two-dimensional, multi-velocity-field, multi-phase, multi-component, Eulerian, fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. Since the three-dimensional representation of the core enables realistic distribution of the materials constituting the core, including control rods, SIMMER-IV has been developed as a direct extension of SIMMER-III to three dimensions with retaining

Hidemasa Yamano; Yoshiharu Tobita; Satoshi Fujita

2009-01-01

166

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

E-print Network

removal During an accident, the reactivity feedback of the coreCore coolant temperatures stabilize at an equilibrium value as decay heat removalcore. The engineering, design and operation of systems for heat removal

Qvist, Staffan Alexander

2013-01-01

167

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

SciTech Connect

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

Aristova, E. N.; Baydin, D. F.; Gol'din, V. Y.; Pestryakova, G. A.; Stoynov, M. I. [Keldysh Inst. of Applied Mathematics RAS, Miusskaya sq., bld.4, 125047, Moscow (Russian Federation)

2012-07-01

168

Design and Demonstration of a Neutron Spin Flipper for a New Neutron Reflectometer SHARAKU at J-PARC  

NASA Astrophysics Data System (ADS)

A new neutron reflectometer SHARAKU with vertical sample-plane geometry was installed on the beam line 17 (BL17) at Materials and Life science experiment Facility (MLF) at J-PARC. Magnetism in a thin magnetic film is one of the main targets on SHARAKU and polarizing devices and neutron spin flippers are required. Since polarized neutrons of wavelength from 0.24 nm to 0.64 nm can be used on SHARAKU, a neutron spin flipper has to control white neutron beam. A two-coil neutron spin flipper (Drabkin spin flipper) is one of the powerful devices to control neutron spin with white beam. In this study, the two-coil flipper was designed and installed in SHARAKU. Demonstration of the two-coil flipper was also performed and polarization of more than 0.95 with wavelengths ranging from 0.24 nm to 0.64 nm was obtained.

Hayashida, Hirotoshi; Takeda, Masayasu; Yamazaki, Dai; Maruyama, Ryuji; Soyama, Kazuhiko; Kubota, Masato; Mizusawa, Tazuko; Yoshida, Noboru; Sakaguchi, Yoshifumi

169

Early Cosmic Evolution of Europium from Core Collapse Supernovae and/or Neutron Star Mergers  

E-print Network

The rapid neutron-capture process is known to be of fundamental importance for explaining the origin of approximately half of the A > 60 stable nuclei observed in nature. Despite important efforts, the astrophysical site of the r process remains unidentified. The two most promising astrophysical sites of the r process, namely Core Collapse SuperNovae (CCSN) and Neutron Star Mergers (NSM) are considered in the context of the early cosmic chemical evolution through the origin and evolution of a typical r process element, Eu. The Eu abundance in very low metallicity stars is used to shed light on the possible CCSN and NSM contributions in the early Universe. Predictions are made here using a hierarchical model for structure formation for which a special attention is paid to a proper description of the stellar formation rate. Eu yields from NSM are taken from recent nucleosynthesis calculations. Observations of Eu in ultra metal poor stars are considered to constrain the model. We find that the bulk of Eu observa...

Vangioni, E; Daigne, F; Francois, P; Belczynski, K

2015-01-01

170

Split-core PCB Rogowski coil designs and applications for protective relaying  

Microsoft Academic Search

Printed circuit board (PCB) Rogowski coils can be designed as non-split-core styles with different shapes such as circular, oval, or rectangular. They can also be designed in split-core style for installation without the need to disconnect a primary or secondary conductor. This paper presents split-core PCB Rogowski coil designs, characteristics, and applications for advanced protection, control, and metering systems.

Ljubomir A. Kojovic

2003-01-01

171

Static and dynamic neutronic analysis of a burst-mode multiple-cavity gas core reactor, Rankine cycle space power system  

Microsoft Academic Search

Static and dynamic neutronic analyses have been performed on an innovative burst-mode Ultrahigh-Temperature Vapor Core Reactor (UTVR) space nuclear power system. This novel reactor concept employs multiple neutronically coupled fissioning cores and operates on a direct closed Rankine cycle using a disk magnetohydrodynamic generator for energy conversion. The UTVR includes two types of fissioning core regions: (a) the central Ultrahigh-Temperature

E. T. Dugan; S. D. Kahook

1993-01-01

172

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

NASA Astrophysics Data System (ADS)

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

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

2014-10-01

173

Neutronics Assessment of Molten Salt Breeding Blanket Design Options  

SciTech Connect

Neutronics assessment has been performed for molten salt breeding blanket design options that can be utilized in fusion power plants. The concepts evaluated are a self-cooled Flinabe blanket with Be multiplier and dual-coolant blankets with He-cooled FW and structure. Three different molten salts were considered including the high melting point Flibe, a low melting point Flibe, and Flinabe. The same TBR can be achieved with a thinner self-cooled blanket compared to the dual-coolant blanket. A thicker Be zone is required in designs with Flinabe. The overall TBR will be {approx}1.07 based on 3-D calculations without breeding in the divertor region. Using Be yields higher blanket energy multiplication than obtainable with Pb. A modest amount of tritium is produced in the Be ({approx}3 kg) over the blanket lifetime of {approx}3 FPY. Using He gas in the dual-coolant blanket results in about a factor of 2 lower blanket shielding effectiveness. We show that it is possible to ensure that the shield is a lifetime component, the vacuum vessel is reweldable, and the magnets are adequately shielded. We conclude that molten salt blankets can be designed for fusion power plants with neutronics requirements such as adequate tritium breeding and shielding being satisfied.

Sawan, M.E. [University of Wisconsin-Madison (United States); Malang, S. [Fusion Nuclear Technology Consulting (United States); Wong, C.P.C. [General Atomics (United States); Youssef, M.Z. [University of California-Los Angeles (United States)

2005-04-15

174

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

SciTech Connect

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

Ganapol, Barry; Maldonado, Ivan

2014-01-23

175

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

PubMed

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; Braud, S; Oriol, L; Villard, J-F

2011-03-01

176

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

NASA Astrophysics Data System (ADS)

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.; Braud, S.; Oriol, L.; Villard, J.-F.

2011-03-01

177

Visibility evaluation of a neutron grating interferometer operated with a polychromatic thermal neutron beam  

NASA Astrophysics Data System (ADS)

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

2014-05-01

178

Core design for use with precision composite reflectors  

NASA Technical Reports Server (NTRS)

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)

1992-01-01

179

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

Microsoft Academic Search

Static and dynamic neutronic analyses are being performed, as part of an integrated series of studies, on an innovative burst power UF4 Ultrahigh Temperature Vapor Core Reactor (UTVR)\\/Disk Magnetohydrodynamic (MHD) generator for space nuclear power applications. This novel reactor concept operates on a direct, closed Rankine cycle in the burst power mode (hundreds of MWe for thousands of seconds). The

Samer D. Kahook; Edward T. Dugan

1991-01-01

180

Nuclear design of the burst power ultrahigh temperature UF sub 4 vapor core reactor system  

Microsoft Academic Search

Static and dynamic neutronic analyses are being performed, as part of an integrated series of studies, on an innovative burst power UF Ultrahigh Temperature Vapor Core Reactor (UTVR)\\/Disk Magnetohydrodynamic (MHD) generator for space nuclear power applications. This novel reactor concept operates on a direct, closed Rankine cycle in the burst power mode (hundreds of MW{sub e} for thousands of seconds).

S. D. Kahook; E. T. Dugan

1991-01-01

181

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

SciTech Connect

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)

2012-07-01

182

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

NASA Astrophysics Data System (ADS)

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

183

DESIGN AND ANALYSIS OF ELECTRIC MOTORS WITH SOFT MAGNETIC COMPOSITE CORE  

Microsoft Academic Search

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

184

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

SciTech Connect

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

2009-01-01

185

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

SciTech Connect

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

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

2011-03-01

186

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

SciTech Connect

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

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

2014-03-01

187

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

SciTech Connect

This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the codes versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research projects primary objective is to advance the state of the art for reactor analysis.

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

2013-11-29

188

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

SciTech Connect

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

Hans D. Gougar

2009-08-01

189

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

PubMed

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

erovnik, Gaper; Kaiba, Tanja; Radulovi?, Vladimir; Jazbec, Ane; Rupnik, Sebastjan; Barbot, Loc; Fourmentel, Damien; Snoj, Luka

2015-02-01

190

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

SciTech Connect

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

2012-01-01

191

Spallation neutron source cryomodule heat loads and thermal design  

SciTech Connect

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

2002-05-10

192

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

NASA Astrophysics Data System (ADS)

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

Kahook, Samer D.; Dugan, Edward T.

1991-01-01

193

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

SciTech Connect

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

Zhao, Jinkui [ORNL

2011-01-01

194

Current directions in core-shell nanoparticle design.  

PubMed

Ten years ago I wrote a review about the important field of core-shell nanoparticles, focussing mainly on our own work about tracer systems, and briefly addressing polymer-coated nanoparticles as fillers for homogeneous polymer-colloid composites. Since then, the potential use of core-shell nanoparticles as multifunctional sensors or potential smart drug-delivery vehicles in biology and medicine has gained more and more importance, affording special types of multi-functionalized and bio-compatible nanoparticles. In this new review article, I try to address the most important developments during the last ten years. This overview is mainly based on frequently cited and more specialized recent review articles from leaders in their respective field. We will consider a variety of nanoscopic core-shell architectures from highly fluorescent nanoparticles (NPs), protected magnetic NPs, multifunctional NPs, thermoresponsive NPs and biocompatible systems to, finally, smart drug-delivery systems. PMID:20644772

Schrtl, Wolfgang

2010-06-01

195

Current directions in core-shell nanoparticle design  

NASA Astrophysics Data System (ADS)

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

Schrtl, Wolfgang

2010-06-01

196

Updated pipe break analysis for Advanced Neutron Source Reactor conceptual design  

SciTech Connect

The Advanced Neutron Source Reactor (ANSR) is a research reactor to be built at the Oak Ridge National Laboratory that will supply the highest continuous neutron flux levels of any reactor in the world. It uses plate-type fuel with high-mass-flux and highly subcooled heavy water as the primary coolant. The Conceptual Safety Analysis for the ANSR was completed in June 1992. The thermal-hydraulic pipe-break safety analysis (performed with a specialized version of RELAP5/MOD3) focused primarily on double-ended guillotine breaks of the primary piping and some core-damage mitigation options for such an event. Smaller, instantaneous pipe breaks in the cold- and hot-leg piping were also analyzed to a limited extent. Since the initial analysis for the conceptual design was completed, several important changes to the RELAP5 input model have been made reflecting improvements in the fuel grading and changes in the elevation of the primary coolant pumps. Also, a new philosophy for pipe-break safety analysis (similar to that adopted for the New Production Reactor) accentuates instantaneous, limited flow area pipe-break accidents in addition to finite-opening-time, double-ended guillotine breaks of the major coolant piping. This paper discloses the results of the most recent instantaneous pipe-break calculations.

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

1994-04-01

197

Design review report for rotary mode core sample truck (RMCST) modifications for flammable gas tanks, preliminary design  

SciTech Connect

This report documents the completion of a preliminary design review for the Rotary Mode Core Sample Truck (RMCST) modifications for flammable gas tanks. The RMCST modifications are intended to support core sampling operations in waste tanks requiring flammable gas controls. The objective of this review was to validate basic design assumptions and concepts to support a path forward leading to a final design. The conclusion reached by the review committee was that the design was acceptable and efforts should continue toward a final design review.

Corbett, J.E.

1996-02-01

198

Design of Cold Neutron Imaging Facility at China Advanced Research Reactor  

NASA Astrophysics Data System (ADS)

The radiography imaging with cold neutrons is being planned at China Advanced Research Reactor (CARR). The 60MW CARR at China Institute of Atomic Energy (CIAE) has got full power in March, 2012. It is a tank-in-pool type reactor using a D2O reflector for inverse neutron trap, and the expected optimal undisturbed thermal neutron flux is 8 1014 n/cm2rad s. Cold neutron imaging facility will be built at the guide hall. At present, its conceptional and physical designs have been finished. The cold neutron imaging facilities will provide an efficient and versatile tool for basic scientific and industrial non-destructive investigation.

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

199

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

SciTech Connect

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

Reass, W. A. (William A.); Apgar, S. E. (Sean E.); Baca, D. M. (David M.); Doss, James D.; Gonzales, J. (Jacqueline); Gribble, R. F. (Robert F.); Hardek, T. W. (Thomas W.); Lynch, M. T. (Michael T.); Rees, D. E. (Daniel E.); Tallerico, P. J. (Paul J.); Trujillo, P. B. (Pete B.); Anderson, D. E. (David E.); Heidenreich, D. A. (Dale A.); Hicks, J. D. (Jim D.); Leontiev, V. N.

2003-01-01

200

Design calculations for the ANS (advanced neutron source) cold neutron source  

Microsoft Academic Search

The advanced neutron source (ANS) is a new experimental facility being planned at the Oak Ridge National Laboratory to meet the national need for an intense steady-state neutron source. A major purpose of the ANS is to provide a high flux of cold ({approx equal} 0.3-nm) neutrons for experiments. High fluxes of such cold neutrons can be obtained from a

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

1990-01-01

201

Small-angle neutron scattering on a core-shell colloidal system: a contrast-variation study.  

PubMed

Small-angle neutron scattering (SANS) measurements are reported on a sterically stabilized, core-shell colloidal system using contrast variation. Aqueous dispersions of polystyrene particles bearing grafted poly(ethylene glycol) (PEG) have been studied over a large range of particle concentrations and two different solvent conditions for the PEG polymer. SANS data are analyzed quantitatively by modeling the particles as core-shell colloids. In a good solvent and under particle contrast conditions, an effective hard-sphere interaction captures excluded-volume interactions up to high concentrations. Contrast variation, through isotopic substitution of both the core and solvent, expedite a detailed study of the PEG layer, both in the dilute limit and as a function of the particle concentration. Upon diminishing the solvent quality, subtle changes in the PEG layer translate into attractions among particles of moderate magnitude. PMID:16262360

Zackrisson, M; Stradner, A; Schurtenberger, P; Bergenholtz, J

2005-11-01

202

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

NASA Astrophysics Data System (ADS)

The main objective of this PhD research is to develop a high accuracy modeling tool using a Monte Carlo based coupled system. The presented research comprises the development of models to include the thermal-hydraulic feedback to the Monte Carlo method and speed-up mechanisms to accelerate the Monte Carlo criticality calculation. Presently, deterministic codes based on the diffusion approximation of the Boltzmann transport equation, coupled with channel-based (or sub-channel based) thermal-hydraulic codes, carry out the three-dimensional (3-D) reactor core calculations of the Light Water Reactors (LWRs). These deterministic codes utilize nuclear homogenized data (normally over large spatial zones, consisting of fuel assembly or parts of fuel assembly, and in the best case, over small spatial zones, consisting of pin cell), which is functionalized in terms of thermal-hydraulic feedback parameters (in the form of off-line pre-generated cross-section libraries). High accuracy modeling is required for advanced nuclear reactor core designs that present increased geometry complexity and material heterogeneity. Such high-fidelity methods take advantage of the recent progress in computation technology and coupled neutron transport solutions with thermal-hydraulic feedback models on pin or even on sub-pin level (in terms of spatial scale). The continuous energy Monte Carlo method is well suited for solving such core environments with the detailed representation of the complicated 3-D problem. The major advantages of the Monte Carlo method over the deterministic methods are the continuous energy treatment and the exact 3-D geometry modeling. However, the Monte Carlo method involves vast computational time. The interest in Monte Carlo methods has increased thanks to the improvements of the capabilities of high performance computers. Coupled Monte-Carlo calculations can serve as reference solutions for verifying high-fidelity coupled deterministic neutron transport methods with detailed and accurate thermal-hydraulic models. The development of such reference high-fidelity coupled multi-physics scheme is described in this dissertation on the basis of MCNP5, NEM, NJOY and COBRA-TF (CTF) computer codes. This work presents results from studies performed and implemented at the Pennsylvania State University (PSU) on both accelerating Monte Carlo criticality calculations by using hybrid nodal diffusion Monte Carlo schemes and thermal-hydraulic feedback modeling in Monte Carlo core calculations. The hybrid MCNP5/CTF/NEM/NJOY coupled code system is proposed and developed in this dissertation work. The hybrid coupled code system contains a special interface developed to update the required MCNP5 input changes to account for dimension and density changes provided by the thermal-hydraulics feedback module. The interface has also been developed to extract the flux and reaction rates calculated by MCNP5 to later transform the data into the power feedback needed by CTF (axial and radial peaking factors). The interface is contained in a master program that controls the flow of the calculations. Both feedback modules (thermal-hydraulic and power subroutines) use a common internal interface to further accelerate the data exchange. One of the most important steps to correctly include the thermal hydraulic feedback into MCNP5 calculations begins with temperature dependent cross section libraries. If the cross sections used for the calculations are not at the correct temperature, the temperature feedback cannot be included into MCNP5 (referred to the effect of temperature on cross sections: Doppler boarding of resolve and unresolved resonances, thermal scattering and elastic scattering). The only method of considering the temperature effects on cross sections is through the generation (or as introduced in this dissertation through a novel interpolation mechanism) of continuous energy temperature-dependent cross section libraries. An automated methodology for generation of continuous energy temperature-dependent cross section libraries has been developed

Espel, Federico Puente

203

Modified Anchor Shaped Post Core Design for Primary Anterior Teeth  

PubMed Central

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

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

2014-01-01

204

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

NASA Astrophysics Data System (ADS)

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

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

2005-05-01

205

Performance and design study of a directional scintillating fiber detector for 14-MeV neutrons using GEANT  

NASA Astrophysics Data System (ADS)

A directional scintillating fiber detector for 14-MeV neutrons was simulated using the GEANT4 Monte Carlo simulation tool. Detail design aspects of a prototype 14 MeV neutron fiber detector under development were used in the simulation to assess performance and design features of the detector. Saint-Gobain produced, BCF-12, plastic fiber material was used in the prototype development. The fiber consists of a core scintillating material of polystyrene with 0.48 mm 0.48 mm dimension and an acrylic outer cladding of 0.02 mm thickness. A total of 64 square fibers, each with a cross-sectional area of 0.25 mm2 and length of 100 mm were positioned parallel to each other with a spacing of 2.3 mm (fiber pitch) in the tracking of 14-MeV neutron induced recoil proton (n-p) events. Neutron induced recoil proton events, resulting energy deposition in two collinear fibers, were used in reconstructing a two dimensional (2D) direction of incident neutrons. Blurring of recoil protons signal in measurements was also considered to account uncertainty in direction reconstruction. Reconstructed direction has a limiting angular resolution of 3 due to fiber dimension. Blurring the recoil proton energy resulted in further broadening of the reconstructed direction and the angular resolution was 20. These values were determined when incident neutron beam makes an angle of 45 degree relative to the front surface of the detector. Comparable values were obtained at other angles of incidence. Results from the present simulation have demonstrated promising directional sensitivity of the scintillating fiber detector under development.

Mengesha, W.; Mascarenhas, N.; Peel, J.; Sunnarborg, D.

2005-09-01

206

Iron-core superconducting magnet design and test results for Maglev application  

Microsoft Academic Search

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

1995-01-01

207

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

SciTech Connect

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.; Truran, J. W. [Department of Astrophysics and Astronomy, University of Chicago, Chicago, IL 60637 (United States); Kratz, K.-L. [HGF Virtuelles Institut fuer Kernstruktur und Nukleare Astrophysik, Universitaet Mainz, D-55128 Mainz (Germany); Pfeiffer, B. [Institut fuer Kernchemie, Universitaet Mainz, D-55128 Mainz (Germany); Rauscher, T.; Thielemann, F.-K., E-mail: farouqi@uchicago.ed, E-mail: truran@nova.uchicago.ed, E-mail: BPfeiffe@uni-mainz.d, E-mail: k-l.Kratz@mpic.d, E-mail: Thomas.Rauscher@unibas.c, E-mail: F-K.Thielemann@unibas.c [Department of Physics, University of Basel, 4056 Basel (Switzerland)

2010-04-01

208

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

SciTech Connect

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

1987-11-01

209

DESIGN AND IMPLEMENTATION OF SYNTHESIZABLE SPACEWIRE CORES Session: SpaceWire Components  

E-print Network

a deep knowledge of the standard and the technologies involved in the development of IP cores for space applications. The SpaceWire Codec and Router here presented are intended to be on board of future missionsDESIGN AND IMPLEMENTATION OF SYNTHESIZABLE SPACEWIRE CORES Session: SpaceWire Components Short

López, Víctor

210

Design and Analysis of a High-Speed Claw Pole Motor With Soft Magnetic Composite Core  

Microsoft Academic Search

Soft magnetic composite (SMC) material is formed by surface-insulated iron powder particles, generating unique properties like magnetic and thermal isotropy, and very low eddy currents. This paper presents the design and analysis of a high-speed claw pole motor with an SMC stator core for reducing core losses and cost. The analyses of magnetic and thermal fields are conducted based on

Yunkai Huang; Jianguo Zhu; Youguang Guo; Zhiwei Lin; Qiansheng Hu

2007-01-01

211

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

NASA Technical Reports Server (NTRS)

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

1995-01-01

212

76 FR 14825 - Core Principles and Other Requirements for Designated Contact Markets  

Federal Register 2010, 2011, 2012, 2013

...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,...

2011-03-18

213

Development of optimized core design and analysis methods for high power density BWRs  

E-print Network

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 ...

Shirvan, Koroush

2013-01-01

214

Neutronic design and simulated performance of Peking University Neutron Imaging Facility (PKUNIFTY)  

NASA Astrophysics Data System (ADS)

The Peking University Neutron Imaging Facility (PKUNIFTY) is a Radio Frequency Quadruple (RFQ) accelerator based system. The fast neutrons are produced by 2 MeV deuterons bombarding beryllium target. The moderator, reflector, shielding and collimator have been optimized with Monte-Carlo simulation to improve the neutron beam quality. The neutrons are thermalized in water cylinder of ?2626 cm 2 with a polyethylene disk in front of Be target. The size of deuteron beam spot is optimized considering both the thermal neutron distribution and the demand of target cooling. The shielding is a combination of 8 cm thick lead and 42 cm thick boron doped polyethylene. The thermal neutrons are extracted through a rectangular inner collimator and a divergent outer collimator. The thermal neutron beam axis is perpendicular to the D + beam line in order to reduce the fast neutron and the ? ray components in the imaging beam. When the neutron yield is 310 12 n/s and the L/ D is 50, the thermal neutron flux is 510 5 n/cm 2/s at the imaging plane, the Cd ratio is 1.63 and the n/? ratio is 1.610 10 n/cm 2/Sv.

Wen, Weiwei; Li, Hang; Zou, Yubin; Tang, Guoyo; Mo, Dawei; Lu, Yuanrong; Guo, Zuiyu

2011-09-01

215

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

SciTech Connect

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

2010-09-01

216

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

SciTech Connect

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

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

2006-06-12

217

ATR LEU Monothlic and Dispersed with 10B Loading Minimization Design Neutronics Performance Analysis  

SciTech Connect

The Advanced Test Reactor (ATR), currently operating in the United States, is used for material testing at very high neutron fluxes. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting HEU driven reactor cores to low-enriched uranium (LEU) cores. The present work investigates the optimized LEU Monolithic and Dispersed fuel with 10B loading minimization design and evaluates the subsequent neutronics operating effects of these optimized fuel designs. The MCNP ATR 1/8th core model was used to optimize the 235U and minimize the 10B loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The fuel depletion methodology MCWO was used to calculate K eff versus effective full power days (EFPD) in this paper. The MCWO-calculated results for the optimized LEU Monolithic and Dispersed fuel cases demonstrated adequate excess reactivity such that the K-eff versus EFPD plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU Monolithic (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and 235U enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness. The proposed LEU fuel meat varies from 0.203 mm (8.0 mil) to 0.254 mm (10.0 mil) at the inner four fuel plates (1-4) and outer four fuel plates (16-19). In addition, an optimized LEU dispersed (U7Mo) case with all the fuel meat thickness of 0.635 mm (25 mil) was also proposed. Then, for both Monolithic and dispersed cases, a burnable absorber 10B, was added in the inner and outer plates to reduce the initial excess reactivity, and the higher to average ratio of the inner/outer heat flux more effectively. The final minimized 10B loading for LEU case studies will have 0.635 g in the LEU fuel meat at the inner 2 fuel plates (1-2) and outer 2 fuel plates (18-19), which can achieve peak to average ratios similar to those for the ATR reference HEU case study. The investigation of this paper shows the optimized LEU Monolithic (U-10Mo) and Dispersed (U7Mo) cases can all meet the LEU conversion objectives.

G. S. Chang

2001-10-01

218

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

NASA Astrophysics Data System (ADS)

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

Kinoshita, Takahiko

219

Strong first-order phase transition in a rotating neutron star core and the associated energy release  

NASA Astrophysics Data System (ADS)

Aims:We calculate the energy release associated with a strong first-order phase transition, from normal phase N to an exotic superdense phase S, in a rotating neutron star. Such a phase transition Nlongrightarrow S, accompanied by a density jump ?__Nlongrightarrow ?__S, is characterized by ?_S/?_N> {3over 2}(1+P_0/?_Nc^2), where P0 is the pressure at which phase transition occurs. Configurations with small S-phase cores are then unstable and collapse into stars with large S-phase cores. The energy release is equal to the difference in mass-energies between the initial (normal) configuration and the final configuration containing an S-phase core, the total stellar baryon mass and angular momentum being kept constant. Methods: The calculations of the energy release are based on precise numerical 2D calculations. Polytropic equations of state (EOSs) as well as realistic EOSs with strong first-order phase transition due to kaon condensation are used. For polytropic EOSs, a large parameter space is studied. Results: For a fixed overpressure, ?overline{P}, defined as the relative excess of central pressure of a collapsing metastable star over the pressure of the equilibrium first-order phase transition, the energy release E_rel does not depend on the stellar angular momentum. It coincides with that for nonrotating stars with the same ?overline{P}. Therefore, results of 1D calculations of E_rel(?overline{P}) for non-rotating stars can be used to predict, with very high precision, the outcome of much harder to perform 2D calculations for rotating stars with the same ?overline{P}. This result holds also for ?overline{P}_mincore. Such phase transitions could be realized in the cores of newly born, hot, pulsating neutron stars.

Zdunik, J. L.; Bejger, M.; Haensel, P.; Gourgoulhon, E.

2008-02-01

220

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

SciTech Connect

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

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

2013-03-01

221

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

SciTech Connect

The present paper is addressing the development and validation against experimental data of 3D full-core models of the BFS-2 zero-power fast-reactor using both the deterministic system code ERANOS-2.2 and the stochastic code MCNPX-2.7.0. The model configuration of BFS considered for analysis is the BFS-62-3A benchmark. To extend the - deterministic/stochastic - code-to-code comparison, neutronic parameters, i.e. reactivity, neutron spectrum and reaction rates, were also simulated at the cell level with the Monte Carlo code SERPENT-1.1.7 with two modern data libraries, ENDF-B/VII and JEFF-3.1.1. The BFS-2 critical zero-power facility at the Inst. of Physics and Power Engineering (IPPE) was designed for simulations of the core and shielding of sodium-cooled, fast reactors, for neutron data validation and comparison with experimental results. At the BFS-2 facility, the BFS-62-3A critical benchmark experiment was set-up as a mock-up of the BN-600 reactor, with hybrid MOX fuel and stainless steel reflectors. A UO{sub 2} blanket and a large non-homogeneous stainless-steel reflector surround the core. The lattice is hexagonal of pitch 5.1 cm and metallic dowels are used to keep in central position cylindrical rods made of different types of material (fissile, fertile, blanket, plenum, shielding and absorber). A typical subassembly is formed in piling up various pellets of about 1 cm in height and 4.6 cm in diameter, conferring large heterogeneity in the axial direction. The full-core model development was a complex task due to the large number of subassemblies and the axial subassembly heterogeneity. In ERANOS-2.2, it was necessary to homogenize axially per region the pellets used to form the subassembly. The self-shielded macroscopic cross-sections were calculated using the cell code ECCO in association with JEFF-3.1 and ENDF/B-VI.8 data libraries. The core calculations were performed with broad cross-sections data in 33 neutron energy groups with the solver AVNM in the diffusion approximation, mostly. In MCNPX-2.7.0, a step-by-step approach was used, starting with a model in which the fissile rods were simulated on a homogeneous level, to finally integrate the actual heterogeneous description of the subassemblies. The code-to-code cell analysis performed between ECCO, SERPENT and MCNPX with different modern nuclear data library revealed that the results for the infinite multiplication factor between Monte Carlo and deterministic analysis are in good agreement ({Delta}p < 100 pcm). The differences between the results were observed to be larger for the neutron data libraries, with reactivity differences up to 350 pcm. (authors)

Girardin, G.; Alonso, M. [Ecole Polytechnique Federale de Lausanne EPFL, CH-1015 Lausanne (Switzerland); Mikityuk, K. [Paul Scherrer Institut PSI, CH-5232 Villigen-PSI (Switzerland)

2012-07-01

222

Coded aperture Fast Neutron Analysis: Latest design advances  

NASA Astrophysics Data System (ADS)

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

Accorsi, Roberto; Lanza, Richard C.

2001-07-01

223

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

Technology Transfer Automated Retrieval System (TEKTRAN)

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

224

Cylindrical Detector and Preamplifier Design for Detecting Neutrons  

E-print Network

in the tissue-equivalent walls depend on the energy of the primary neutrons. The differences in the spectra measured by different size detectors will provide additional information on the incident neutron energy. Monte Carlo N-particle extended (MCNPX) code...

Xia, Zhenghua

2010-01-14

225

High Flux Isotope Reactor cold neutron source reference design concept  

Microsoft Academic Search

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

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

1998-01-01

226

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

PubMed

Magnetic nanoparticles are highly desirable for biomedical research and treatment of cancer especially when combined with hyperthermia. The efficacy of nanoparticle-based therapies could be improved by generating radioactive nanoparticles with a convenient decay time and which simultaneously have the capability to be used for locally confined heating. The core-shell morphology of such novel nanoparticles presented in this work involves a polysilico-tungstate molecule of the polyoxometalate family as a precursor coating material, which transforms into an amorphous tungsten oxide coating upon annealing of the FePt core-shell nanoparticles. The content of tungsten atoms in the nanoparticle shell is neutron activated using cold neutrons at the Heinz Maier-Leibnitz (FRMII) neutron facility and thereby transformed into the radioisotope W-187. The sizeable natural abundance of 28% for the W-186 precursor isotope, a radiopharmaceutically advantageous gamma-beta ratio of ???30% and a range of approximately 1mm in biological tissue for the 1.3MeV ?-radiation are promising features of the nanoparticles' potential for cancer therapy. Moreover, a high temperature annealing treatment enhances the magnetic moment of nanoparticles in such a way that a magnetic heating effect of several degrees Celsius in liquid suspension - a prerequisite for hyperthermia treatment of cancer - was observed. A rise in temperature of approximately 3C in aqueous suspension is shown for a moderate nanoparticle concentration of 0.5mg/ml after 15min in an 831kHz high-frequency alternating magnetic field of 250Gauss field strength (25mT). The biocompatibility based on a low cytotoxicity in the non-neutron-activated state in combination with the hydrophilic nature of the tungsten oxide shell makes the coated magnetic FePt nanoparticles ideal candidates for advanced radiopharmaceutical applications. PMID:25445697

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

2015-01-10

227

Preliminary Neutronics Design and Analysis of D2O Cooled High Conversion PWRs  

SciTech Connect

This report presents a neutronics analysis of tight-pitch D2O-cooled PWRs loaded with MOX fuel and focuses essentially on the Pu breeding potential of such reactors as well as on an important safety parameter, the void coefficient, which has to be negative. It is well known that fast reactors have a better neutron economy and are better suited than thermal reactors to breed fissile material from neutron capture in fertile material. Such fast reactors (e.g. sodium-cooled reactors) usually rely on technologies that are very different from those of existing water-cooled reactors and are probably more expensive. This report investigates another possibility to obtain a fast neutron reactor while still relying mostly on a PWR technology by: (1) Tightening the lattice pitch to reduce the water-to-fuel volume ratio compared to that of a standard PWR. Water-to-fuel volume ratios of between 0.45 and 1 have been considered in this study while a value of about 2 is typical of standard PWRs, (2) Using D2O instead of H2O as a coolant. Indeed, because of its different neutron physics properties, the use of D2O hardens the neutron spectrum to an extent impossible with H2O when used in a tight-pitch lattice. The neutron spectra thus obtained are not as fast as those in sodium-cooled reactor but they can still be characterized as fast compared to that of standard PWR neutron spectra. In the phase space investigated in this study we did not find any configurations that would have, at the same time, a positive Pu mass balance (more Pu at the end than at the beginning of the irradiation) and a negative void coefficient. At this stage, the use of radial blankets has only been briefly addressed whereas the impact of axial blankets has been well defined. For example, with a D2O-to-fuel volume ratio of 0.45 and a core driver height of about 60 cm, the fissile Pu mass balance between the fresh fuel and the irradiated fuel (50 GWd/t) would be about -7.5% (i.e. there are 7.5% fewer fissile Pu isotopes at the end than at the beginning of the irradiation) and the void coefficient would be negative. The addition of 1 cm of U-238 blanket at the top and bottom of the fuel would bring the fissile Pu mass balance from -7.5% to -6.5% but would also impact the void coefficient in the wrong way. In fact, it turns out that the void coefficient is so sensitive to the presence of axial blanket that it limits its size to only a few cm for driver fuel height of about 50-60 cm. For reference, the fissile Pu mass balance is about -35% in a standard PWR MOX fuel such as those used in France. In order to reduce the fissile Pu deficit and potentially reach a true breeding regime (i.e. a positive Pu mass balance), it would be necessary to make extensive use of radial blankets, both internal and external. Even though this was not addressed in detail here, it is reasonable to believe that at least as much U-238 blanket subassemblies as MOX driver fuel subassemblies would be necessary to breed enough Pu to compensate for the Pu deficit in the driver fuel. Hence, whereas a relatively simple D2O-cooled PWR core design makes it possible to obtain a near-breeder core, it may be necessary to more than double the mass of heavy metal in the core as well as the mass of heavy metal to reprocess per unit of energy produced in order to breed the few percents of Pu missing to reach a true breeding regime. It may be interesting to quantify these aspects further in the future.

Hikaru Hiruta; Gilles Youinou

2012-09-01

228

Thermal design study of a liquid hydrogen-cooled cold-neutron source  

Microsoft Academic Search

The use of both liquid hydrogen as a moderator and polycrystalline beryllium as a filter to enhance cold neutron flux at the UC Davis McClellan Nuclear Radiation Center has been studied. Although, more work is needed before an actual cold neutron source can be designed and built, the purpose of this preliminary study is to investigate the effects of liquid

D. Quach; R. C. Aldredge; H. B. Liu; W. J. Richards

2007-01-01

229

The Convex Hull of Two Core Capacitated Network Design Problems  

E-print Network

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 ...

Magnanti, Thomas L.

230

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

NASA Astrophysics Data System (ADS)

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

231

Modified Y-TZP Core Design Improves All-ceramic Crown Reliability  

PubMed Central

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.

2011-01-01

232

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

PubMed

In this work a compact system was designed for bulk sample analysis using the technique of PGNAA. The system consists of (252)Cf fission neutron source, a moderator/reflector/filter assembly, and a suitable enclosure to delimit the resulting neutron beam. The moderator/reflector/filter arrangement has been optimised to maximise the thermal neutron component useful for samples analysis with a suitably low level of beam contamination. The neutron beam delivered by this compact system is used to irradiate the sample and the prompt gamma rays produced by neutron reactions within the sample elements are detected by appropriate gamma rays detector. Neutron and gamma rays transport calculations have been performed using the Monte Carlo N-Particle transport code (MCNP5). PMID:21129990

Ghassoun, J; Mostacci, D

2011-08-01

233

Core-crust transition properties of neutron stars within systematically varied extended relativistic mean-field model  

E-print Network

The model dependence and the symmetry energy dependence of the core-crust transition properties for the neutron stars are studied using three different families of systematically varied extended relativistic mean field model. Several forces within each of the families are so considered that they yield wide variations in the values of the nuclear symmetry energy $a_{\\rm sym}$ and its slope parameter $L$ at the saturation density. The core-crust transition density is calculated using a method based on random-phase-approximation. The core-crust transition density is strongly correlated, in a model independent manner, with the symmetry energy slope parameter evaluated at the saturation density. The pressure at the transition point dose not show any meaningful correlations with the symmetry energy parameters at the saturation density. At best, pressure at the transition point is correlated with the symmetry energy parameters and their linear combination evaluated at the some sub-saturation density. Yet, such correlations might not be model independent. The correlations of core-crust transition properties with the symmetry energy parameter are also studied by varying the symmetry energy within a single model. The pressure at the transition point is correlated once again with the symmetry energy parameter at the sub-saturation density.

A. Sulaksono; Naosad Alam; B. K. Agrawal

2014-11-03

234

Core-crust transition properties of neutron stars within systematically varied extended relativistic mean-field model  

NASA Astrophysics Data System (ADS)

The model dependence and the symmetry energy dependence of the core-crust transition properties for the neutron stars (NS) are studied using three different families of systematically varied extended relativistic mean field model. Several forces within each of the families are so considered that they yield wide variations in the values of the nuclear symmetry energy asym and its slope parameter L at the saturation density. The core-crust transition density is calculated using a method based on random-phase-approximation. The core-crust transition density is strongly correlated, in a model independent manner, with the symmetry energy slope parameter evaluated at the saturation density. The pressure at the transition point does not show any meaningful correlations with the symmetry energy parameters at the saturation density. At best, pressure at the transition point is correlated with the symmetry energy parameters and their linear combination evaluated at the some sub-saturation density. Yet, such correlations might not be model independent. The correlations of core-crust transition properties with the symmetry energy parameter are also studied by varying the symmetry energy within a single model. The pressure at the transition point is correlated once again with the symmetry energy parameter at the sub-saturation density.

Sulaksono, A.; Alam, Naosad; Agrawal, B. K.

2014-11-01

235

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

PubMed

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

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

2009-07-01

236

Design/Operations review of core sampling trucks and associated equipment  

SciTech Connect

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.

1996-03-11

237

Interweaving Game Design into Core CS Curriculum Yolanda Rankin  

E-print Network

Research shows that video games are an often under- utilized learning environment that can be extended' understanding of game design principles and students' perceived learning. The results of our evaluation serve science, computer science departments are leveraging the appeal of video games to entice the next

Gooch, Bruce

238

The Design of SVPWM IP Core Based on FPGA  

Microsoft Academic Search

This paper expounds the basic principle of SVPWM from the perspective of vector analysis, and derives the necessary mathematical formula to implement this digital design. Also, this paper presents a basic structure of the digital hardware circuit based on FPGA. The proposed scheme is implemented and verified on a single Altera FPGA. The experimental results are presented in the end

Guijie Yang; Pinzhi Zhao; Zhaoyong Zhou

2008-01-01

239

A high power density radial-in-flow reactor split core design for space power systems  

NASA Astrophysics Data System (ADS)

Application of the Rankine cycle to space power systems is difficult because of the problems and complexities associated with two phase flow systems in microgravity. A direct cycle system which could provide super heated vapor to the turbine inlet would greatly enhance the development of Rankine cycle power systems for space applications. The split core radial-in-flow reactor design provides a safe reliable core design for space power systems. It makes direct Rankine cycle power systems a very competitive design, eliminating the boiler and additional pumps of an indirect cycle along with the liquid vapor separator. A continuous power Rankine cycle system using this core design would produce the least weight system of any having the same power output.

Coomes, Edmund P.

240

A fast-neutron spectrometer of advanced design  

NASA Technical Reports Server (NTRS)

Fast neutron spectrometer combines helium filled proportional counters with solid-state detectors to achieve the properties of high efficiency, good resolution, rapid response, and effective gamma ray rejection.

Moler, R. B.; Preston, C. C.

1966-01-01

241

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

Microsoft Academic Search

The China Spallation Neutron Source (CSNS) is an accelerator based multidiscipline user facility planned to be constructed in Dongguan, Guangdong, China. The CSNS complex consists of an negative hydrogen linear accelerator, a rapid cycling proton synchrotron accelerating the beam to 1.6GeV energy, a solid tungsten target station, and instruments for spallation neutron applications. The facility operates at 25Hz repetition rate

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

2009-01-01

242

Neutronics design of the low aspect ratio tokamak reactor, VECTOR  

Microsoft Academic Search

A neutronic study has been carried out for VECTOR, which is a low aspect ratio tokamak with the aspect ratio of 2.3, to achieve simultaneously sufficient shielding for super-conducting toroidal field coils (TFCs) and tritium breeding ratio (TBR) larger than unity, using the three-dimensional Monte Carlo neutron transport code MCNP-4C. The tritium production is mainly made by outboard blankets consisting

T. Nishitani; M. Yamauchi; S. Nishio; M. Wada

2006-01-01

243

Shielding design studies for a neutron irradiator system based on a 252Cf source.  

PubMed

This study aims to investigate a shielding design against neutrons and gamma rays from a source of 252Cf, using Monte Carlo simulation. The shielding materials studied were borated polyethylene, borated-lead polyethylene and stainless steel. The Monte Carlo code MCNP4B was used to design shielding for 252Cf based neutron irradiator systems. By normalising the dose equivalent rate values presented to the neutron production rate of the source, the resulting calculations are independent of the intensity of the actual 252Cf source. The results show that the total dose equivalent rates were reduced significantly by the shielding system optimisation. PMID:11707031

da Silva, A X; Crispim, V R

2001-01-01

244

Design of a compact high-power neutron sourceThe EURISOL converter target  

NASA Astrophysics Data System (ADS)

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

Samec, K.; Milenkovi?, R. .; Dementjevs, S.; Ashrafi-Nik, M.; Kalt, A.

2009-07-01

245

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

NASA Astrophysics Data System (ADS)

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

Sordo, F.; Fernandez-Alonso, F.; Gonzalez, Miguel A.; Ghiglino, A.; Magn, M.; Terrn, S.; Martnez, F.; de Vicente, J. P.; Vivanco, R.; Bermejo, F. J.; Perlado, J. M.

2014-10-01

246

Design and Analysis of a PM Wind Generator with a Soft Magnetic Composite Core  

Microsoft Academic Search

This paper presents the design and analysis of a small direct-drive axial-flux PM wind generator with a soft magnetic composite (SMC) core. A procedure is outlined for sizing the wind generator based on the wind speed requirements and the SMC core samples available for this research. A concentrated, non-overlapping winding configuration is used. Issues relating to the number of teeth

M. A. Khan; L. Dosiek; P. Pillay

2006-01-01

247

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

NASA Astrophysics Data System (ADS)

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

2014-06-01

248

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

NASA Astrophysics Data System (ADS)

The short-lived 182Hf-182W isotope system can provide powerful constraints on the timescales of planetary core formation, but its application to iron meteorites is hampered by neutron capture reactions on W isotopes resulting from exposure to galactic cosmic rays. Here we show that Pt isotopes in magmatic iron meteorites are also affected by capture of (epi)thermal neutrons and that the Pt isotope variations are correlated with variations in 182W/184W. This makes Pt isotopes a sensitive neutron dosimeter for correcting cosmic ray-induced W isotope shifts. The pre-exposure 182W/184W derived from the Pt-W isotope correlations of the IID, IVA and IVB iron meteorites are higher than most previous estimates and are more radiogenic than the initial 182W/184W of Ca-Al-rich inclusions (CAI). The Hf-W model ages for core formation range from +1.61.0 million years (Ma; for the IVA irons) to +2.71.3 Ma after CAI formation (for the IID irons), indicating that there was a time gap of at least 1 Ma between CAI formation and metal segregation in the parent bodies of some iron meteorites. From the Hf-W ages a time limit of <1.5-2 Ma after CAI formation can be inferred for the accretion of the IID, IVA and IVB iron meteorite parent bodies, consistent with earlier conclusions that the accretion of differentiated planetesimals predated that of most chondrite parent bodies.

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

2013-01-01

249

Measurements of Nuclear Heating Rate and Neutron Flux in HANARO CN Hole for Designing the Moderator Cell of Cold Neutron Source  

Microsoft Academic Search

The design of the cold neutron source(CNS) facility for HANARO, a 30 MW research reactor, is in progress. In order to measure the nuclear heating rate and thermal neutron flux in the CN hole of HANARO, a calorimeter based on the concept of the heat flow calorimeter is designed and constructed. The calorimeter sensor consists of a cylindrical Al sample,

Myong-Seop KIM; Sung-Yul HWANG; Hoan-Sung JUNG; Kye-Hong LEE

250

Design of a laboratory for experiments with a pulsed neutron source.  

PubMed

We present the results of a neutron shielding design and optimisation study performed to reduce the exposure to radiological doses arising from a 14 MeV pulsed neutron generator (PNG) having a maximum emission strength of 2.0 x 10(8) neutrons s(-1). The source was intended to be used in a new irradiation facility for the realisation of an experiment on acoustical cavitation in liquids. This paper describes in detail how the facility was designed to reduce both neutron and gamma-ray dose rates to acceptable levels, taking into account the ALARP principle in following the steps of optimisation. In particular, this work compares two different methods of optimisation to assess neutron dose rates: the use of analytical methods and the use of Monte Carlo simulations (MCNPX 2.4). The activation of the surrounding materials during operation was estimated using the neutron spectra as input to the FISPACT 3.0 code. The limitations of a first-order analytical model to determine the neutron activation levels are highlighted. The impact that activation has on the choice of the materials to be used inside the laboratory and on the waiting time before anyone can safely enter the room after the neutron source is switched off is also discussed. PMID:19454793

Memoli, G; Trusler, J P M; Ziver, A K

2009-06-01

251

Design of air-gapped magnetic-core inductors for superimposed direct and alternating currents  

NASA Technical Reports Server (NTRS)

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.

1976-01-01

252

Analysis of Stainless Steel Sandwich Panels with a Metal Foam Core for Lightweight Fan Blade Design  

NASA Technical Reports Server (NTRS)

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.

2004-01-01

253

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

SciTech Connect

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

Fouquet, Peter; Farago, Bela; Andersen, Ken H.; Bentley, Phillip M.; Pastrello, Gilles; Sutton, Iain; Thaveron, Eric; Thomas, Frederic [Institut Laue-Langevin, BP 156, F-38042 Grenoble Cedex 9 (France); Moskvin, Evgeny [Helmholtzzentrum Berlin, Glienicker Strasse 100, D-14109 Berlin (Germany); Pappas, Catherine [Helmholtzzentrum Berlin, Glienicker Strasse 100, D-14109 Berlin (Germany); Faculty of Applied Sciences, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands)

2009-09-15

254

Core Noise: Implications of Emerging N+3 Designs and Acoustic Technology Needs  

NASA Technical Reports Server (NTRS)

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.

2011-01-01

255

Boron tracedrug design for neutron dynamic therapeutics for LDL.  

PubMed

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

2013-01-01

256

Designing with advanced composites; Report on the European Core Conference, 1st, Zurich, Switzerland, Oct. 20, 21, 1988, Conference Papers  

SciTech Connect

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.

Not Available

1988-01-01

257

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

SciTech Connect

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

Livescu, V.; Clausen, B.; Sisneros, T.; Bourke, M.A.M. [Los Alamos National Laboratory, Los Alamos, New Mexico, 87545 (United States); Woodruff, T.R.; Vaidyanathan, R. [University of Central Florida, Orlando, Florida, 32816 (United States); Notardonato, W.U. [NASA Kennedy Space Center, Kennedy Space Center, Florida, 32899 (United States)

2004-06-28

258

The design and performance of IceCube DeepCore  

NASA Astrophysics Data System (ADS)

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.; Bser, 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.; Demirrs, L.; Descamps, F.; Desiati, P.; de Vries-Uiterweerd, G.; DeYoung, T.; Daz-Vlez, J. C.; Dierckxsens, M.; Dreyer, J.; Dumm, J. P.; Dunkman, M.; Eisch, J.; Ellsworth, R. W.; Engdegrd, 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.; Glsenkamp, T.; Goldschmidt, A.; Goodman, J. A.; Gra, 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.; Hl, 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.; Khne, J.-H.; Kohnen, G.; Kolanoski, H.; Kpke, 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.; Lnemann, J.; Madsen, J.; Marotta, A.; Maruyama, R.; Mase, K.; Matis, H. S.; Meagher, K.; Merck, M.; Mszros, 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.; Prez 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.; Schnwald, 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.; Stl, A.; Strahler, E. A.; Strm, R.; Ster, M.; Sullivan, G. W.; Swillens, Q.; Taavola, H.; Taboada, I.; Tamburro, A.; Tepe, A.; Ter-Antonyan, S.; Tilav, S.; Toale, P. A.; Toscano, S.; Tosi, D.; van Eijndhoven, N.; Vandenbroucke, J.; Van Overloop, A.; van Santen, J.; Vehring, M.; Voge, M.; Walck, C.; Waldenmaier, T.; Wallraff, M.; Walter, M.; Weaver, Ch.; Wendt, C.; Westerhoff, S.; Whitehorn, N.; Wiebe, K.; Wiebusch, C. H.; Williams, D. R.; Wischnewski, R.; Wissing, H.; Wolf, M.; Wood, T. R.; Woschnagg, K.; Xu, C.; Xu, D. L.; Xu, X. W.; Yanez, J. P.; Yodh, G.; Yoshida, S.; Zarzhitsky, P.; Zoll, M.

2012-05-01

259

A multi-group Monte Carlo core analysis method and its application in SCWR design  

SciTech Connect

Complex geometry and spectrum have been the characteristics of many newly developed nuclear energy systems, so the suitability and precision of the traditional deterministic codes are doubtable while being applied to simulate these systems. On the contrary, the Monte Carlo method has the inherent advantages of dealing with complex geometry and spectrum. The main disadvantage of Monte Carlo method is that it takes long time to get reliable results, so the efficiency is too low for the ordinary core designs. A new Monte Carlo core analysis scheme is developed, aimed to increase the calculation efficiency. It is finished in two steps: Firstly, the assembly level simulation is performed by continuous energy Monte Carlo method, which is suitable for any geometry and spectrum configuration, and the assembly multi-group constants are tallied at the same time; Secondly, the core level calculation is performed by multi-group Monte Carlo method, using the assembly group constants generated in the first step. Compared with the heterogeneous Monte Carlo calculations of the whole core, this two-step scheme is more efficient, and the precision is acceptable for the preliminary analysis of novel nuclear systems. Using this core analysis scheme, a SCWR core was designed based on a new SCWR assembly design. The core output is about 1,100 MWe, and a cycle length of about 550 EFPDs can be achieved with 3-batch refueling pattern. The average and maximum discharge burn-up are about 53.5 and 60.9 MWD/kgU respectively. (authors)

Zhang, P.; Wang, K.; Yu, G. [Dept. of Engineering Physics, Tsinghua Univ., Beijing, 100084 (China)

2012-07-01

260

Use of Solid Hydride Fuel for Improved long-Life LWR Core Designs  

SciTech Connect

The primary objective of this project was to assess the feasibility of improving the performance of PWR and BWR cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the bus-bar cost of electricity (COE). Additional performance measures considered are safety, fuel bundle design simplicity in particular for BWRs, and plutonium incineration capability. It was found that hydride fuel can safely operate in PWRs and BWRs without restricting the linear heat generation rate of these reactors relative to that attainable with oxide fuel. A couple of promising applications of hydride fuel in PWRs and BWRs were identified: (1) Eliminating dedicated water moderator volumes in BWR cores thus enabling to significantly increase the cooled fuel rods surface area as well as the coolant flow cross section area in a given volume fuel bundle while significantly reducing the heterogeneity of BWR fuel bundles thus achieving flatter pin-by-pin power distribution. The net result is a possibility to significantly increase the core power density on the order of 30% and, possibly, more, while greatly simplifying the fuel bundle design. Implementation of the above modifications is, though, not straightforward; it requires a design of completely different control system that could probably be implemented only in newly designed plants. It also requires increasing the coolant pressure drop across the core. (2) Recycling plutonium in PWRs more effectively than is possible with oxide fuel by virtue of a couple of unique features of hydride fuel reduced inventory of U-238 and increased inventory of hydrogen. As a result, the hydride fuelled core achieves nearly double the average discharge burnup and the fraction of the loaded Pu it incinerates in one pass is double that of the MOX fuel. The fissile fraction of the Pu in the discharged hydride fuel is only ~2/3 that of the MOX fuel and the discharged hydride fuel is more proliferation resistant. Preliminary feasibility assessment indicates that by replacing some of the ZrH1.6 by ThH2 it will be possible to further improve the plutonium incineration capability of PWRs. Other possibly promising applications of hydride fuel were identified but not evaluated in this work. A number of promising oxide fueled PWR core designs were also found as spin-offs of this study: (1) The optimal oxide fueled PWR core design features smaller fuel rod diameter of D=6.5 mm and a larger pitch-to-diameter ratio of P/D=1.39 than presently practiced by industry 9.5mm and 1.326. This optimal design can provide a 30% increase in the power density and a 24% reduction in the cost of electricity (COE) provided the PWR could be designed to have the coolant pressure drop across the core increased from the reference 29 psia to 60 psia. (2) Using wire wrapped oxide fuel rods in hexagonal fuel assemblies it is possible to design PWR cores to operate at 54% higher power density than the reference PWR design that uses grid spacers and a square lattice, provided 60 psia coolant pressure drop across the core could be accommodated. Uprating existing PWRs to use such cores could result in 40% reduction in the COE. The optimal lattice geometry is D = 8.08 mm and P/D = 1.41. The most notable advantages of wire wraps over grid spacers are their significant lower pressure drop, higher critical heat flux and improved vibrations characteristics.

Greenspan, E

2006-04-30

261

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

NASA Astrophysics Data System (ADS)

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

2012-11-01

262

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

NASA Astrophysics Data System (ADS)

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

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

2014-11-01

263

Design of the 50 kW neutron converter for SPIRAL2 facility  

Microsoft Academic Search

SPIRAL2 is a facility for the study of fundamental nuclear physics and multidisciplinary research. SPIRAL2 represents a major advance for research on exotic nuclei. The radioactive ion beam (RIB) production system is comprised of a neutron converter, a target and an ion source. This paper is dedicated to the designing of the 50 kW neutron converter for the SPIRAL2 facility.

M. S. Avilov; L. B. Tecchio; A. T. Titov; V. S. Tsybulya; E. I. Zhmurikov

2010-01-01

264

Design of the 50 kW neutron converter for SPIRAL2 facility  

Microsoft Academic Search

SPIRAL2 is a facility for the study of fundamental nuclear physics and multidisciplinary research. SPIRAL2 represents a major advance for research on exotic nuclei. The radioactive ion beam (RIB) production system is comprised of a neutron converter, a target and an ion source. This paper is dedicated to the designing of the 50kW neutron converter for the SPIRAL2 facility.Among the

M. S. Avilov; L. B. Tecchio; A. T. Titov; V. S. Tsybulya; E. I. Zhmurikov

2010-01-01

265

Boiling water neutronic reactor incorporating a process inherent safety design  

DOEpatents

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

Forsberg, C.W.

1985-02-19

266

Boiling water neutronic reactor incorporating a process inherent safety design  

DOEpatents

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

Forsberg, Charles W. (Kingston, TN)

1987-01-01

267

Modeling and analysis of core debris recriticality during hypothetical severe accidents in the Advanced Neutron Source Reactor  

SciTech Connect

This paper discusses salient aspects of severe-accident-related recriticality modeling and analysis in the Advanced Neutron Source (ANS) reactor. The development of an analytical capability using the KENO V.A-SCALE system is described including evaluation of suitable nuclear cross-section sets to account for the effects of system geometry, mixture temperature, material dispersion and other thermal-hydraulic conditions. Benchmarking and validation efforts conducted with KENO V.A-SCALE and other neutronic codes against critical experiment data are described. Potential deviations and biases resulting from use of the 16-group Hansen-Roach library are shown. A comprehensive test matrix of calculations to evaluate the threat of a recriticality event in the ANS is described. Strong dependencies on geometry, material constituents, and thermal-hydraulic conditions are described. The introduction of designed mitigative features is described.

Taleyarkhan, R.P.; Kim, S.H.; Slater, C.O.; Moses, D.L.; Simpson, D.B.; Georgevich, V.

1993-05-01

268

Electric dipole response of He6: Halo-neutron and core excitations  

NASA Astrophysics Data System (ADS)

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.

2014-06-01

269

Narrative Plus: Designing and Implementing the Common Core State Standards with the Gift Essay  

ERIC Educational Resources Information Center

The authors of this article describe their inquiry into implementation of the writing-focused Common Core State Standards in a co-taught English 9 class in an urban school. They describe instructional moves designed to increase student success with an assignment called the Gift Essay, with particular focus on planning and other organizational

Chandler-Olcott, Kelly; Zeleznik, John

2013-01-01

270

Spring design for use in the core of a nuclear reactor  

DOEpatents

A spring design particularly suitable for use in the core of a nuclear reactor includes one surface having a first material oriented in a longitudinal direction, and another surface having a second material oriented in a transverse direction. The respective surfaces exhibit different amounts of irraditation induced strain.

Willard, Jr., H. James (Bethel Park, PA)

1993-01-01

271

Coarse-grained parallel genetic algorithm applied to a nuclear reactor core design optimization problem  

Microsoft Academic Search

This work extends the research related to genetic algorithms (GA) in core design optimization problems, which basic investigations were presented in previous work. Here we explore the use of the Island Genetic Algorithm (IGA), a coarse-grained parallel GA model, comparing its performance to that obtained by the application of a traditional non-parallel GA. The optimization problem consists on adjusting several

Cludio M. N. A. Pereira; Celso M. F. Lapa

2003-01-01

272

Post-silicon Debugging for Multi-core Designs Valeria Bertacco  

E-print Network

Post-silicon Debugging for Multi-core Designs Valeria Bertacco Dept. of Electrical Engineering in released silicon are growing in number due to the increasing complexity of modern processor de- signs. This deteriorating situation is causing a growing portion of the validation effort to shift to post-silicon, when

Bertacco, Valeria

273

Light source design using Kagome-lattice hollow core photonic crystal fibers  

NASA Astrophysics Data System (ADS)

Supercontinuum (SC) light source is designed using high pressure Xe-filled hollow core Kagome-lattice photonic crystal fiber. Using finite element method with perfectly matched layer, SC spectra in normal chromatic dispersion region have been generated using picosecond optical pulses from relatively less expensive laser sources.

Hossain, Md. Anwar; Namihira, Yoshinori

2014-09-01

274

Design and Analysis of a Claw Pole Permanent Magnet Motor With Molded Soft Magnetic Composite Core  

Microsoft Academic Search

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

2009-01-01

275

Design and analysis of a transverse flux machine with soft magnetic composite core  

Microsoft Academic Search

This paper presents the design and performance analysis of a three phase, three stack permanent magnet transverse flux motor with soft magnetic composite core. To predict and optimize the major parameters, three-dimensional finite element analysis is performed. The performance is calculated when the motor operates with a brushless DC drive.

Youguang Guo; Jian guo Zhu; P. A. Watterson; Wei Wu

2003-01-01

276

Design and Implementation of High Performance IPSec Applications with Multi-Core Processors  

Microsoft Academic Search

The rapid increasing Internet services need high performance, scalable and flexible network security devices. IPSec is a set of protocols to ensure transmission of packets in IP network. Multi-core processors are targeted to a wide range of applications with complex packet processing and high throughput requirements. Although there are several designs of IPSec system with heterogeneous hardware platforms, practical ultra-speed

Yizhen Liu; Daxiong Xu; Wuying Song; Zhixin Mu

2008-01-01

277

Designing Scalable FPGA-Based Reduction Circuits Using Pipelined Floating-Point Cores  

E-print Network

and a single pipelined floating-point unit. 1 "free FLOPS, expensive bytes" #12;1.1. Background The classic reDesigning Scalable FPGA-Based Reduction Circuits Using Pipelined Floating-Point Cores Ling Zhuo California Los Angeles, CA 90089-2562 {lzhuo,grm,prasanna}@usc.edu Abstract The use of pipelined floating

Prasanna, Viktor K.

278

Use of Solid Hydride Fuel for Improved long-Life LWR Core Designs  

Microsoft Academic Search

The primary objective of this project was to assess the feasibility of improving the performance of PWR and BWR cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the bus-bar cost of electricity (COE). Additional performance measures considered are safety, fuel bundle design simplicity in particular for BWRs,

Greenspan

2006-01-01

279

December 12-13, 2007/ARR Power Core Engineering: Design Updates  

E-print Network

#12;December 12-13, 2007/ARR 2 Engineering and Trade-Off Studies · Power cycle choice: Rankine vsDecember 12-13, 2007/ARR 1 Power Core Engineering: Design Updates and Trade-Off Studies A. René reliability assessment (in progress, UCSD/INL/Georgia Tech.) · Assessment of off-normal conditions for a power

Raffray, A. René

280

AMPLIFIER DESIGN IMPLEMENTING HOLLOW-CORE PHOTONIC BANDGAP FIBER FOR FIBER-LASER BASED  

E-print Network

to conventional solid-state system. All fiber-laser-based frequency combs consists of an octave spanning-laser based frequency combs have compared to their solid-state counterparts is that pulses directly fromAMPLIFIER DESIGN IMPLEMENTING HOLLOW-CORE PHOTONIC BANDGAP FIBER FOR FIBER-LASER BASED INFRARED

Washburn, Brian

281

AMPLIFIER DESIGN IMPLEMENTING HOLLOW-CORE PHOTONIC BANDGAP FIBER FOR FIBER-LASER BASED INFRARED  

E-print Network

to their solid- state counterparts is that pulses directly from the laser do not have enough peak powerAMPLIFIER DESIGN IMPLEMENTING HOLLOW-CORE PHOTONIC BANDGAP FIBER FOR FIBER-LASER BASED INFRARED FREQUENCY COMBS JINGANG LIM, DANIEL V. NICKEL, BRIAN R. WASHBURN Department of Physics, Kansas State

Washburn, Brian

282

Neutron-physics characteristics of the VVR core which affect the operability of the fuel elements  

Microsoft Academic Search

UDC 621.039.5 When one calculates the thermophysical features of a fuel element, one must know: the distribution of the energy liberation over the radius of the fuel core, since the distribution affects the temperature field in the fuel element [1]; the burnup at a \\

V. D. Sidorenko; A. S. Shcheglov

1993-01-01

283

Conceptual design for one megawatt spallation neutron source at Argonne  

SciTech Connect

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

Cho, Y.; Bailey, J.; Brown, B. [and others

1993-12-31

284

Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source Reactor at Oak Ridge National Laboratory  

Microsoft Academic Search

This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at the Oak Ridge National Laboratory (ORNL). Damage propagation is postulated to occur from thermal conduction between damaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur

S. H. Kim; R. P. Taleyarkhan; S. Navarro-Valenti; V. Georgevich

1995-01-01

285

Design of the 50 kW neutron converter for SPIRAL2 facility  

NASA Astrophysics Data System (ADS)

SPIRAL2 is a facility for the study of fundamental nuclear physics and multidisciplinary research. SPIRAL2 represents a major advance for research on exotic nuclei. The radioactive ion beam (RIB) production system is comprised of a neutron converter, a target and an ion source. This paper is dedicated to the designing of the 50 kW neutron converter for the SPIRAL2 facility. Among the different variants of the neutron converter, the one based on a rotating solid disk seems quite attractive due to its safety, ease in production and relatively low cost. Dense graphite used as the converter's material allows the production of high-intensity neutron flux and, at the same time, the heat removal from the converter by means of radiation cooling. Thermo-mechanical simulations performed in order to determine the basic geometry and physical characteristics of the neutron production target for SPIRAL2 facility, to define the appropriate beam power distribution, and to predict the target behaviour under the deuteron beam of nominal parameters (40 MeV, 1.2 mA, 50 kW) are presented. To study the main physical and mechanical properties and serviceability under operating conditions, several kinds of graphite have been analyzed and tested. The paper reports the results of such measurements. Radiation damage is the most important issue for the application of graphite as neutron converter. It is well known that the thermal conductivity of the neutron-irradiated graphite is reduced by a factor of 10 from the initial value after irradiation. Difference in volume expansions between the matrix and the fiber results in serious damage of neutron-irradiated C/C composites. Calculations showed that at high temperature the effect of neutron radiation is not so critical and that the change in thermal conductivity does not prevent the use of graphite as neutron converter.

Avilov, M. S.; Tecchio, L. B.; Titov, A. T.; Tsybulya, V. S.; Zhmurikov, E. I.

2010-06-01

286

Thermally Activated Post-glitch Response of the Neutron Star Inner Crust and Core. I. Theory  

NASA Astrophysics Data System (ADS)

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

2014-07-01

287

MCNP-REN: a Monte Carlo tool for neutron detector design  

NASA Astrophysics Data System (ADS)

The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo code developed at Los Alamos National Laboratory, Monte Carlo N-Particle (MCNP), was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP-Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program, predicts neutron detector response without using the point reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of mixed oxide fresh fuel were taken with the Underwater Coincidence Counter, and measurements of highly enriched uranium reactor fuel were taken with the active neutron interrogation Research Reactor Fuel Counter and compared to calculation. Simulations completed for other detector design applications are described. The method used in MCNP-REN is demonstrated to be fundamentally sound and shown to eliminate the need to use the point model for detector performance predictions.

Abhold, Mark E.; Baker, Michael C.

2002-06-01

288

Preliminary shielding analysis in support of the CSNS target station shutter neutron beam stop design  

NASA Astrophysics Data System (ADS)

The construction of China Spallation Neutron Source (CSNS) has been initiated in Dongguan, Guangdong, China. Thus a detailed radiation transport analysis of the shutter neutron beam stop is of vital importance. The analyses are performed using the coupled Monte Carlo and multi-dimensional discrete ordinates method. The target of calculations is to optimize the neutron beamline shielding design to guarantee personal safety and minimize cost. Successful elimination of the primary ray effects via the two-dimensional uncollided flux and the first collision source methodology is also illustrated. Two-dimensional dose distribution is calculated. The dose at the end of the neutron beam line is less than 2.5 ?Sv/h. The models have ensured that the doses received by the hall staff members are below the standard limit required.

Zhang, Bin; Chen, Yi-Xue; Wang, Wei-Jin; Yang, Shou-Hai; Wu, Jun; Yin, Wen; Liang, Tian-Jiao; Jia, Xue-Jun

2011-08-01

289

Effects of post-core design and ferrule on fracture resistance of endodontically treated maxillary central incisors  

Microsoft Academic Search

Statement of Problem. Studies concerning the effects of post-core design and ferrule on the fracture resistance of endodontically treated teeth remain controversial. Purpose. The purpose of this study was to investigate in vitro the effects of post-core design and ferrule on the fracture resistance of root canal treated human maxillary central incisors restored with metal ceramic crowns. Material and Methods.

Lu Zhi-Yue; Zhang Yu-Xing

2003-01-01

290

Fusion Engineering and Design 41 (1998) 371376 The ARIES-RS power core--recent development in Li/V  

E-print Network

Fusion Engineering and Design 41 (1998) 371­376 The ARIES-RS power core--recent development in Li tritium breeding, excellent high heat flux removal capability, long structural life time, low activation engineering is expected to result in a superior power plant design. This paper summarizes the power core

California at San Diego, University of

291

Insert Design and Manufacturing for Foam-Core Composite Sandwich Structures  

NASA Astrophysics Data System (ADS)

Sandwich structures have been used in the aerospace industry for many years. The high strength to weight ratios that are possible with sandwich constructions makes them desirable for airframe applications. While sandwich structures are effective at handling distributed loads such as aerodynamic forces, they are prone to damage from concentrated loads at joints or due to impact. This is due to the relatively thin face-sheets and soft core materials typically found in sandwich structures. Carleton University's Uninhabited Aerial Vehicle (UAV) Project Team has designed and manufactured a UAV (GeoSury II Prototype) which features an all composite sandwich structure fuselage structure. The purpose of the aircraft is to conduct geomagnetic surveys. The GeoSury II Prototype serves as the test bed for many areas of research in advancing UAV technologies. Those areas of research include: low cost composite materials manufacturing, geomagnetic data acquisition, obstacle detection, autonomous operations and magnetic signature control. In this thesis work a methodology for designing and manufacturing inserts for foam-core sandwich structures was developed. The results of this research work enables a designer wishing to design a foam-core sandwich airframe structure, a means of quickly manufacturing optimized inserts for the safe introduction of discrete loads into the airframe. The previous GeoSury II Prototype insert designs (v.1 & v.2) were performance tested to establish a benchmark with which to compare future insert designs. Several designs and materials were considered for the new v.3 inserts. A plug and sleeve design was selected, due to its ability to effectively transfer the required loads to the sandwich structure. The insert material was chosen to be epoxy, reinforced with chopped carbon fibre. This material was chosen for its combination of strength, low mass and also compatibility with the face-sheet material. The v.3 insert assembly is 60% lighter than the previous insert designs. A casting process for manufacturing the v.3 inserts was developed. The developed casting process, when producing more than 13 inserts, becomes more economical than machining. An exploratory study was conducted looking at the effects of dynamic loading on the v.3 insert performance. The results of this study highlighted areas for improving dynamic testing of foam-core sandwich structure inserts. Correlations were developed relating design variables such as face-sheet thickness and insert diameter to a failure load for different load cases. This was done through simulations using Computer Aided Engineering (CAE) software, and experimental testing. The resulting correlations were integrated into a computer program which outputs the required insert dimensions given a set of design parameters, and load values.

Lares, Alan

292

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

NASA Astrophysics Data System (ADS)

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

Desai, Shraddha S.

2014-07-01

293

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

NASA Astrophysics Data System (ADS)

The ultra small angle neutron scattering instrument can measure the lower limit of scattering vector to near Q 2.0x10-5 -1 while the upper limit can reach to an intermediate scattering vector Q 10-2 -1 of a typical small angle neutron scattering (SANS) depending on the contrast of sample. USANS is useful when measuring objects that are micron to submicron in size while SANS is useful when measuring objects that are micron to nano in size. When both USANS and SANS were used together, we could measure sizes from micron to nano scale, which is useful when studying the hierarchical structures in the wide scale of Q and total cross-section, d?/d?(Q). Recently, KIST has developed the USANS (so called KIST-USANS) at HANARO cold neutron guide hall of KAERI. We will present the instrument design, performance, future plan, and some examples of measurements that cover approximately 11 orders of magnitude in the d?/d?(Q) and 4 orders in the Q.

Kim, Man-Ho

2013-03-01

294

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

SciTech Connect

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))

1990-09-01

295

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

SciTech Connect

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

Grant, C. [Comision Nacional de Energia Atomica, Av del Libertador 8250, Buenos Aires 1429 (Argentina); Mollerach, R. [Nucleoelectrica Argentina S.A., Arribenos 3619, Buenos Aires 1429 (Argentina); Leszczynski, F.; Serra, O.; Marconi, J. [Comision Nacional de Energia Atomica, Av del Libertador 8250, Buenos Aires 1429 (Argentina); Fink, J. [Nucleoelectrica Argentina S.A., Arribenos 3619, Buenos Aires 1429 (Argentina)

2006-07-01

296

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

SciTech Connect

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

Gillespie, G.H. [Gillespie (G.H.) and Associates, Inc., Del Mar, CA (United States); Micklich, B.J.; McMichael, G.E. [Argonne National Lab., IL (United States)

1996-09-30

297

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

NASA Astrophysics Data System (ADS)

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

298

Design and Calibration of a High-Precision Density Gauge for Firn and Ice Cores  

NASA Astrophysics Data System (ADS)

The Maine Automated Density Gauge Experiment (MADGE) is a field deployable gamma-ray density gauging instrument designed to provide high resolution (3.3 mm) and high precision (0.004 g cm-3) density profiles of polar firn and ice cores at a typical throughput of 1.5 m h-1. The resulting density profiles are important in ice sheet mass balance and paleoclimate studies, as well as the modeling electromagnetic wave propagation in firn and ice for remote sensing and ground penetrating radar applications. This study describes the design (optimal gamma-ray energy selection, measurement uncertainty analysis, dead-time corrections) and calibration (mass-attenuation coefficient and absolute density calibrations) of the instrument, and discusses the results of additional experiments to verify the calculated measurement uncertainty. Data collected from firn cores drilled on the recent 2006-2007 U.S. Internation Trans-Antarctic Scientific Expedition are also shown and discussed.

Breton, Daniel; Hamilton, Gordon

2009-10-01

299

Large-effective-area dispersion-compensating fiber design based on dual-core microstructure.  

PubMed

We present a microstructure-based dual-core dispersion-compensating fiber (DCF) design for dispersion compensation in long-haul optical communication links. The design has been conceptualized by combining the all-solid dual-core DCF and dispersion-compensating photonic crystal fiber. The fiber design has been analyzed numerically by using a full vectorial finite difference time domain method. We propose a fiber design for narrowband as well as broadband dispersion compensation. In the narrowband DCF design, the fiber exhibits very large negative dispersion of around -42,000 ps nm(-1) km(-1) and a large mode area of 67 ?m(2). The effects of varying different structural parameters on the dispersion characteristics as well as on the trade-off between full width at half-maximum and dispersion have been investigated. For broadband DCF design, a dispersion value between -860 ps nm(-1)km(-1) and -200 ps nm(-1) km(-1) is obtained for the entire spectral range of the C band. PMID:23842244

Prabhakar, Gautam; Peer, Akshit; Rastogi, Vipul; Kumar, Ajeet

2013-07-01

300

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

Microsoft Academic Search

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

2002-01-01

301

Energy Efficient Engine integrated core/low spool design and performance report  

NASA Technical Reports Server (NTRS)

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.

Stearns, E. Marshall

1985-01-01

302

Test Planning and Design Space Exploration in a Core-Based Environment  

Microsoft Academic Search

This paper proposes a comprehensive model for testplanning in a core-based environment. The main contributionof this work is the use of several types of TAMs and theconsideration of different optimization factors (area, pinsand test time) during the global TAM and test schedule definition.This expansion of concerns makes possible an efficientyet fine-grained search in the huge design space ofa reuse-based environment.

Erika Cota; Luigi Carro; Marcelo Lubaszewski; Alex Orailoglu

2002-01-01

303

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

PubMed

The construction of the new Neutron Standards Laboratory at CIEMAT (Laboratorio de Patrones Neutrnicos) has been finalised and is ready to provide service. The facility is an ?8 m8 m8 m irradiation vault, following the International Organization for Standardization 8529 recommendations. It relies on several neutron sources: a 5-GBq (5.8 10(8) s(-1)) (252)Cf source and two (241)Am-Be neutron sources (185 and 11.1 GBq). The irradiation point is located 4 m over the ground level and in the geometrical centre of the room. Each neutron source can be moved remotely from its storage position inside a water pool to the irradiation point. Prior to this, an important task to design the neutron shielding and to choose the most appropriate materials has been developed by the Radiological Security Unit and the Ionizing Radiations Metrology Laboratory. MCNPX was chosen to simulate the irradiation facility. With this information the walls were built with a thickness of 125 cm. Special attention was put on the weak points (main door, air conditioning system, etc.) so that the ambient dose outside the facility was below the regulatory limits. Finally, the Radiation Protection Unit carried out a set of measurements in specific points around the installation with an LB6411 neutron monitor and a Reuter-Stokes high-pressure ion chamber to verify experimentally the results of the simulation. PMID:24478306

Mndez-Villafae, R; Guerrero, J E; Embid, M; Fernndez, R; Grandio, R; Prez-Cejuela, P; Mrquez, J L; Alvarez, F; Ortego, P

2014-10-01

304

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

SciTech Connect

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

Smith, K.; Forget, B. [Massachusetts Institute of Technology, 77 Massachusetts Avenue, Cambridge MA 02139 (United States)

2013-07-01

305

Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report  

SciTech Connect

This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

Parish, T.A.

1995-03-02

306

Thermal hydraulic design and decay heat removal of a solid target for a spallation neutron source  

Microsoft Academic Search

Thermal hydraulic design and thermal stress calculations were conducted for a water-cooled solid target irradiated by a MW-class proton beam for a spallation neutron source. Plate type and rod bundle type targets were examined. The thickness of the plate and the diameter of the rod were determined based on the maximum and the wall surface temperature. The thermal stress distributions

N. Takenaka; D. Nio; Y. Kiyanagi; K. Mishima; M. Kawai; M. Furusaka

2005-01-01

307

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

NASA Astrophysics Data System (ADS)

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.

2013-06-01

308

Some preliminary design considerations for the ANS (Advanced Neutron Source) reactor cold source  

Microsoft Academic Search

Two areas concerned with the design of the Advanced Neutron Source (ANS) cold source have been investigated by simple one-dimensional calculations. The gain factors computed for a possible liquid nitrogen-15 cold source moderator are considerably below those computed for the much colder liquid deuterium moderator, as is reasonable considering the difference in moderator temperature. Nevertheless, nitrogen-15 does represent a viable

1988-01-01

309

Physics Analyses in the Design of the HFIR Cold Neutron Source  

Microsoft Academic Search

Physics analyses have been performed to characterize the performance of the cold neutron source to be installed in the High Flux Isotope Reactor at the Oak Ridge National Laboratory in the near future. This paper provides a description of the physics models developed, and the resulting analyses that have been performed to support the design of the cold source. These

Bucholz

1999-01-01

310

Design calculations for the ANS cold neutron source Part II. Heating rates  

Microsoft Academic Search

Calculated results to aid in the design of a liquid deuterium cold source in the D2O reflector of a high-flux reactor are presented. The results presented include a comparison with experimental data, neutron leakage current as a function of wavelength, and heating rates in liquid deuterium and the other materials of the cold source.

R. G. Alsmiller; R. A. Lillie

1992-01-01

311

Neutronics conceptual design of a UF6-fueled gasesous laser system  

NASA Astrophysics Data System (ADS)

Using uranium hexafluoride gas to power a reactor core has been proposed previously. Considerable work was done on this concept both as basic reactor (core) physics relating to gaseous core reactors (Kunze 1969) and as an optical pumping source to drive another laser medium (Boody 1978). The first of these studies was aimed primarily at the nuclear propulsion application for space vehicles. However, the necessary core and reflector design differed substantially from the configuration required for the flashlamp driven laser application. This paper examines the specific results of the reactor physics design for the latter concept. Also briefly examined is the reactor physics of a direct UF6 laser, although the feasibility of the laser physics of this direct application has not been seriously examined.

Gu, Guo-Xiang; Kunze, Jay F.; Boody, Frederick P.; Prelas, Mark A.

312

Design and performance of a pulse transformer based on Fe-based nanocrystalline core.  

PubMed

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

Yi, Liu; Xibo, Feng; Lin, Fuchang

2011-08-01

313

A new design of fission detector for prompt fission neutron investigation  

NASA Astrophysics Data System (ADS)

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

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

2012-10-01

314

Design of bus-on-chip core for micro-satellite avionics  

NASA Astrophysics Data System (ADS)

This paper discusses a layout of bus-on-chip core referring to SoC thinking which is composed of six sections based on a physical chip of FPGA: multi-Processor cache coherence unit, external bus control module, TT&C module, Ethernet Mac interface, EDAC/DMA module, and AMBA bridges. Multi-processor cache coherence unit, as a key part of the bus core, is used to serve the rapid parallel computing by means of the breakthrough of write/read speed of EMS memory and enhances the reliability of OBC with the service of supporting the hot standby of redundancy and the reconfiguration of fault-tolerance. External bus control module is made to support the PnP of external components applying varieties of buses, which is designed by means of soft-core in order to adapt the variation of macro-design and improve the flexibility of external application. TT&C module is the interface of subsystems of telemetry, telecommand and communication, which involves the protocols of HDLC. Ethernet Mac interface based on TCP/IP acts as the access of ISL for formation flying, constellation, etc. EDAC/DMA module mainly manages the data exchange between AMBA bus and RAM, and assigns DMA for the payloads.

Liu, Youjun; You, Zheng; Li, Bin; Zhang, Xiangqi; Meng, Ziyang

2007-11-01

315

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

SciTech Connect

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)

2012-07-01

316

Design study of a medical proton linac for neutron therapy  

Microsoft Academic Search

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

S. Machida; D. Raparia

1988-01-01

317

Comparison of Chamfer and Deep Chamfer Preparation Designs on the Fracture Resistance of Zirconia Core Restorations  

PubMed Central

Background and aims One of the major problems of all-ceramic restorations is their probable fracture under occlusal force. The aim of the present in vitro study was to compare the effect of two marginal designs (chamfer and deep chamfer) on the fracture resistance of all-ceramic restorations, CERCON. Materials and methods This in vitro study was carried out with single-blind experimental technique. One stainless steel die with 50 chamfer finish line design (0.8 mm deep) was prepared using a milling machine. Ten epoxy resin dies were prepared. The same die was retrieved and 50' chamfer was converted into a deep chamfer design (1 mm). Again ten epoxy resin dies were prepared from the deep chamfer die. Zirconia cores with 0.4 mm thickness and 35 m cement space were fabricated on the epoxy resin dies (10 chamfer and 10 deep chamfer samples). The zirconia cores were cemented on the epoxy resin dies and underwent a fracture test with a universal testing machine and the samples were investigated from the point of view of the origin of the failure. Results The mean values of fracture resistance for deep chamfer and chamfer samples were 1426.10182.60 and 991.75112.00 N, respectively. Students t-test revealed statistically significant differences between the groups. Conclusion The results indicated a relationship between the marginal design of zirconia cores and their fracture re-sistance. A deep chamfer margin improved the biomechanical performance of posterior single zirconia crown restorations, which might be attributed to greater thickness and rounded internal angles in deep chamfer margins. PMID:23019507

Jalalian, Ezatollah; Rostami, Roghayeh; Atashkar, Berivan

2011-01-01

318

Results of a Neutronic Simulation of HTR-Proteus Core 4.2 using PEBBED and other INL Reactor Physics Tools: FY09 Report  

Microsoft Academic Search

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-PEBDAN), 1-D discrete ordinates transport (SCAMP), and nodal diffusion calculations (PEBBED) were employed to yield keff and flux profiles. Preliminary results indicate that these tools, as

Hans D. Gougar

2009-01-01

319

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

PubMed

The americium-beryllium neutron irradiation facility at the National Nuclear Research Institute (NNRI), Ghana, was re-designed with four 20 Ci sources using Monte Carlo N-Particle (MCNP) code to investigate the maximum amount of flux that is produced by the combined sources. The results were compared with a single source Am-Be irradiation facility. The main objective was to enable us to harness the maximum amount of flux for the optimization of neutron activation analysis and to enable smaller sample sized samples to be irradiated. Using MCNP for the design construction and neutronic performance calculation, it was realized that the single-source Am-Be design produced a thermal neutron flux of (1.80.0007)10(6) n/cm(2)s and the four-source Am-Be design produced a thermal neutron flux of (5.40.0007)10(6) n/cm(2)s which is a factor of 3.5 fold increase compared to the single-source Am-Be design. The criticality effective, k(eff), of the single-source and the four-source Am-Be designs were found to be 0.001150.0008 and 0.001430.0008, respectively. PMID:24792123

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

2014-08-01

320

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

SciTech Connect

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

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

2008-05-05

321

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

SciTech Connect

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

Evans, Louise G [Los Alamos National Laboratory; Schear, Melissa A [Los Alamos National Laboratory; Hendricks, John S [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory

2011-02-16

322

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

SciTech Connect

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

Evans, Louise G [Los Alamos National Laboratory; Schear, Melissa A [Los Alamos National Laboratory; Hendricks, John S [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory

2010-12-14

323

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

SciTech Connect

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

Evans, Louise G [Los Alamos National Laboratory; Schear, Melissa A [Los Alamos National Laboratory; Hendricks`, John S [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory

2011-01-13

324

Designed armadillo repeat proteins as general peptide-binding scaffolds: consensus design and computational optimization of the hydrophobic core.  

PubMed

Armadillo repeat proteins are abundant eukaryotic proteins involved in several cellular processes, including signaling, transport, and cytoskeletal regulation. They are characterized by an armadillo domain, composed of tandem armadillo repeats of approximately 42 amino acids, which mediates interactions with peptides or parts of proteins in extended conformation. The conserved binding mode of the peptide in extended form, observed for different targets, makes armadillo repeat proteins attractive candidates for the generation of modular peptide-binding scaffolds. Taking advantage of the large number of repeat sequences available, a consensus-based approach combined with a force field-based optimization of the hydrophobic core was used to derive soluble, highly expressed, stable, monomeric designed proteins with improved characteristics compared to natural armadillo proteins. These sequences constitute the starting point for the generation of designed armadillo repeat protein libraries for the selection of peptide binders, exploiting their modular structure and their conserved binding mode. PMID:18222472

Parmeggiani, Fabio; Pellarin, Riccardo; Larsen, Anders Peter; Varadamsetty, Gautham; Stumpp, Michael T; Zerbe, Oliver; Caflisch, Amedeo; Plckthun, Andreas

2008-03-01

325

Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores  

NASA Technical Reports Server (NTRS)

A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim

2007-01-01

326

Design progress of cryogenic hydrogen system for China Spallation Neutron Source  

NASA Astrophysics Data System (ADS)

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

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

2014-01-01

327

Design progress of cryogenic hydrogen system for China Spallation Neutron Source  

SciTech Connect

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

Wang, G. P.; Zhang, Y.; Xiao, J.; He, C. C.; Ding, M. Y.; Wang, Y. Q.; Li, N.; He, K. [Institute of High Energy Physics, Chinese Academy of Sciences, Beijing 100049, P.R. (China)

2014-01-29

328

Analysis of Process Parameters Affecting Spray-Dried Oily Core Nanocapsules Using Factorial Design  

Microsoft Academic Search

The purpose of this work was to optimize the process parameters required for the production of spray-dried oily core nanocapsules\\u000a (NCs) with targeted size and drug yield using a two-level four-factor fractional factorial experimental design (FFED). The\\u000a coded process parameters chosen were inlet temperature (X\\u000a 1), feed flow rate (X\\u000a 2), atomizing air flow (X\\u000a 3), and aspiration rate (X

Tao Zhang; Bi-Botti C. Youan

2010-01-01

329

Design of dispersion-compensating fibers based on a dual-concentric-core photonic crystal fiber.  

PubMed

A photonic crystal fiber based on a particular periodic arrangement of airholes and pure silica is designed for chromatic dispersion compensation. A two-concentric-core structure is obtained by introducing two different sizes of capillaries (for the airholes) and exhibits very high negative chromatic dispersion [-2200 ps/(nm km) at 1550 nm]. The variation of optogeometric parameters is also investigated to evaluate the tolerance of the fabrication. Finally, the bending influence on the modal characteristics shows that it is possible to tune the phase-matching wavelength over the C band by adjusting the diameter of the fiber. PMID:15605485

Grme, F; Auguste, J L; Blondy, J M

2004-12-01

330

PVDF core-free actuator for Braille displays: design, fabrication process, and testing  

NASA Astrophysics Data System (ADS)

Refreshable Braille displays require many, small diameter actuators to move the pins. The electrostrictive P(VDF-TrFECFE) terpolymer can provide the high strain and actuation force under modest electric fields that are required of this application. In this paper, we develop core-free tubular actuators and integrate them into a 3 2 Braille cell. The films are solution cast, stretched to 6 ?m thick, electroded, laminated into a bilayer, rolled into a 2 mm diameter tube, bonded, and provided with top and bottom contacts. Experimental testing of 17 actuators demonstrates significant strains (up to 4%). A novel Braille cell is designed and fabricated using six of these actuators.

Levard, Thomas; Diglio, Paul J.; Lu, Sheng-Guo; Gorny, Lee J.; Rahn, Christopher D.; Zhang, Q. M.

2011-04-01

331

Fuel performance models for high-temperature gas-cooled reactor core design  

SciTech Connect

Mechanistic fuel performance models are used in high-temperature gas-cooled reactor core design and licensing to predict failure and fission product release. Fuel particles manufactured with defective or missing SiC, IPyC, or fuel dispersion in the buffer fail at a level of less than 5 x 10/sup -4/ fraction. These failed particles primarily release metallic fission products because the OPyC remains intact on 90% of the particles and retains gaseous isotopes. The predicted failure of particles using performance models appears to be conservative relative to operating reactor experience.

Stansfield, O.M.; Simon, W.A.; Baxter, A.M.

1983-09-01

332

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

Microsoft Academic Search

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

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

2011-01-01

333

A single aromatic core mutation converts a designed "primitive" protein from halophile to mesophile folding.  

PubMed

The halophile environment has a number of compelling aspects with regard to the origin of structured polypeptides (i.e., proteogenesis) and, instead of a curious niche that living systems adapted into, the halophile environment is emerging as a candidate "cradle" for proteogenesis. In this viewpoint, a subsequent halophile-to-mesophile transition was a key step in early evolution. Several lines of evidence indicate that aromatic amino acids were a late addition to the codon table and not part of the original "prebiotic" set comprising the earliest polypeptides. We test the hypothesis that the availability of aromatic amino acids could facilitate a halophile-to-mesophile transition by hydrophobic core-packing enhancement. The effects of aromatic amino acid substitutions were evaluated in the core of a "primitive" designed protein enriched for the 10 prebiotic amino acids (A,D,E,G,I,L,P,S,T,V)-having an exclusively prebiotic core and requiring halophilic conditions for folding. The results indicate that a single aromatic amino acid substitution is capable of eliminating the requirement of halophile conditions for folding of a "primitive" polypeptide. Thus, the availability of aromatic amino acids could have facilitated a critical halophile-to-mesophile protein folding adaptation-identifying a selective advantage for the incorporation of aromatic amino acids into the codon table. PMID:25297559

Longo, Liam M; Tenorio, Connie A; Kumru, Ozan S; Middaugh, C Russell; Blaber, Michael

2015-01-01

334

Alternative designs for construction of the class II transfer RNA tertiary core.  

PubMed Central

The structural requirements for assembly of functional class II transfer RNA core regions have been examined by sequence analysis and tested by reconstruction of alternative folds into the tertiary domain of Escherichia coli tRNA(2)Gln. At least four distinct designs have been identified that permit stable folding and efficient synthetase recognition, as assessed by thermal melting profiles and glutaminylation kinetics. Although most large variable-arm tRNAs found in nature possess an enlarged D-loop, lack of this feature can be compensated for by insertion of nucleotides either 3' to the variable loop or within the short acceptor/D-stem connector region. Rare pyrimidines at nt 9 in the core region can be accommodated in the class II framework, but only if specific nucleotides are present either in the D-loop or 3' to the variable arm. Glutaminyl-tRNA synthetase requires one or two unpaired uridines 3' to the variable arm to efficiently aminoacylate several of the class II frameworks. Because there are no specific enzyme contacts in the tRNAGln core region, these data suggest that tRNA discrimination by GlnRS depends in part on indirect readout of RNA sequence information. PMID:11105758

Nissan, T A; Perona, J J

2000-01-01

335

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

SciTech Connect

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

Bays, Samuel E.; Anghaie, Samim; Smith, Blair; Knight, Travis [Innovative Space Power and Propulsion Institute, University of Florida, 202 Nuclear Science Building, Gainesville, FL 32611 (United States)

2004-07-01

336

On the development of computational tools for the design of beam assemblies for boron neutron capture therapy  

Microsoft Academic Search

This article is the first in a series devoted to the development of efficient and accurate computational tools for the design of beam assemblies for boron neutron capture therapy within the framework of discrete ordinates spectral nodal methods of neutron transport theory. We begin our study with a multi-layer representation of an assembly, and we derive a discrete ordinates matrix

Marcos Pimenta de Abreu

2007-01-01

337

Multi-group helium and hydrogen production cross section libraries for fusion neutronics design  

NASA Astrophysics Data System (ADS)

The helium and hydrogen production cross section libraries based on the JENDL-3 data file were compiled for use in neutronics and shielding design calculation of a fusion reactor. These libraries have the same group structures as the transport cross section sets, FUSION-J3 and FUSION-40, which are often used in fusion neutronics design and can be used as the response function libraries for the reaction rate calculation code, APPLE-3. These libraries were processed from the JENDL gas production cross section file which is one of the JENDL special purpose files. Some sample calculations using the discrete ordinate code, ANISN, with these libraries were performed and the results were compared with the existing results. Consequently it was found that the appropriate results can be obtained with these libraries. The generated multi-group cross sections for helium and hydrogen production are presented in graphs and tables.

Mori, Seiji; Zimin, S.; Takatsu, Hideyuki

1993-09-01

338

Validation of the MCNPX-PoliMi Code to Design a Fast-Neutron Multiplicity Counter  

SciTech Connect

Many safeguards measurement systems used at nuclear facilities, both domestically and internationally, rely on He-3 detectors and well established mathematical equations to interpret coincidence and multiplicity-type measurements for verifying quantities of special nuclear material. Due to resource shortages alternatives to these existing He-3 based systems are being sought. Work is also underway to broaden the capabilities of these types of measurement systems in order to improve current multiplicity analysis techniques. As a part of a Material Protection, Accounting, and Control Technology (MPACT) project within the U.S. Department of Energy's Fuel Cycle Technology Program we are designing a fast-neutron multiplicity counter with organic liquid scintillators to quantify important quantities such as plutonium mass. We are also examining the potential benefits of using fast-neutron detectors for multiplicity analysis of advanced fuels in comparison with He-3 detectors and testing the performance of such designs. The designs are being developed and optimized using the MCNPX-PoliMi transport code to study detector response. In the full paper, we will discuss validation measurements used to justify the use of the MCNPX-PoliMi code paired with the MPPost multiplicity routine to design a fast neutron multiplicity counter with liquid scintillators. This multiplicity counter will be designed with the end goal of safeguarding advanced nuclear fuels. With improved timing qualities associated with liquid scintillation detectors, we can design a system that is less limited by nuclear materials of high activities. Initial testing of the designed system with nuclear fuels will take place at Idaho National Laboratory in a later stage of this collaboration.

J. L. Dolan; A. C. Kaplan; M. Flaska; S. A. Pozzi; D. L. Chichester

2012-07-01

339

Design, assembly, and testing of the neutron imaging lens for the National Ignition Facility  

NASA Astrophysics Data System (ADS)

The National Ignition Facility will begin testing DT fuel capsules yielding greater than 1013 neutrons during 2010. Neutron imaging is an important diagnostic for understanding capsule behavior. Neutrons are imaged at a scintillator after passing through a pinhole. The pixelated, 160-mm square scintillator is made up of 1/4 mm diameter rods 50 mm long. Shielding and distance (28 m) are used to preserve the recording diagnostic hardware. Neutron imaging is light starved. We designed a large nine-element collecting lens to relay as much scintillator light as reasonable onto a 75 mm gated microchannel plate (MCP) intensifier. The image from the intensifier's phosphor passes through a fiber taper onto a CCD camera for digital storage. Alignment of the pinhole and tilting of the scintillator is performed before the relay lens and MCP can be aligned. Careful tilting of the scintillator is done so that each neutron only passes through one rod (no crosstalk allowed). The 3.2 ns decay time scintillator emits light in the deep blue, requiring special glass materials. The glass within the lens housing weighs 26 lbs, with the largest element being 7.7 inches in diameter. The distance between the scintillator and the MCP is only 27 inches. The scintillator emits light with 0.56 NA and the lens collects light at 0.15 NA. Thus, the MCP collects only 7% of the available light. Baffling the stray light is a major concern in the design of the optics. Glass cost considerations, tolerancing, and alignment of this lens system will be discussed.

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

2010-08-01

340

Design of an RFQ-Based Neutron Source for Cargo ContainerInterrogation  

SciTech Connect

An RFQ-based neutron generator system is described that produces pulsed neutrons for the active screening of sea-land cargo containers for the detection of shielded special nuclear materials (SNM).A microwave-driven deuteron source is coupled to an electrostatic LEBT that injects a 40 mA D+ beam into a 6 MeV, 5.1 meter-long 200 MHz RFQ.The RFQ has a unique beam dynamics design and is capable of operating at duty factors of 5 to 10 percent accelerating a D+ time-averaged current of up to 1.5 mA at 5 percent duty factor, including species and transmission loss. The beam is transported through a specially-designed thin window into a 2.5-atmosphere deuterium gas target. A high-frequency dipole magnet is used to scan the beam over the long dimension of the 5by 35 cm target window. The source will deliver a neutron flux of 1 cdot107 n/(cm2s) to the center of an empty cargo container. Details of the ion source, LEBT, RFQ beam dynamics and gas target design are presented.

Staples, John W.; Hoff, M.D.; Kwan, J.W.; Li, D.; Ludewigt, B.A.; Ratti, A.; Virostek, S.P.; Wells, R.P.

2006-08-01

341

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

SciTech Connect

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility based on the use of an electron accelerator driven subcritical (ADS) facility [1]. The facility uses the existing electron accelerators of KIPT in Ukraine. The neutron source of the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and the electron energy in the range of 100 to 200 MeV, [2]. The main functions of the facility are the production of medical isotopes and the support of the Ukraine nuclear power industry. Reactor physics experiments and material performance characterization will also be carried out. The subcritical assembly is driven by neutrons generated by the electron beam interactions with the target material. A fraction of these neutrons has an energy above 50 MeV generated through the photo nuclear interactions. This neutron fraction is very small and it has an insignificant contribution to the subcritical assembly performance. However, these high energy neutrons are difficult to shield and they can be slowed down only through the inelastic scattering with heavy isotopes. Therefore the shielding design of this facility is more challenging relative to fission reactors. To attenuate these high energy neutrons, heavy metals (tungsten, iron, etc.) should be used. To reduce the construction cost, heavy concrete with 4.8 g/cm{sup 3} density is selected as a shielding material. The iron weight fraction in this concrete is about 0.6. The shape and thickness of the heavy concrete shield are defined to reduce the biological dose equivalent outside the shield to an acceptable level during operation. At the same time, special attention was give to reduce the total shield mass to reduce the construction cost. The shield design is configured to maintain the biological dose equivalent during operation {le} 0.5 mrem/h inside the subcritical hall, which is five times less than the allowable dose for working forty hours per week for 50 weeks per year. This study analyzed and designed the thickness and the shape of the radial and top shields of the neutron source based on the biological dose equivalent requirements inside the subcritical hall during operation. The Monte Carlo code MCNPX is selected because of its capabilities for transporting electrons, photons, and neutrons. Mesh based weight windows variance reduction technique is utilized to estimate the biological dose outside the shield with good statistics. A significant effort dedicated to the accurate prediction of the biological dose equivalent outside the shield boundary as a function of the shield thickness without geometrical approximations or material homogenization. The building wall was designed with ordinary concrete to reduce the biological dose equivalent to the public with a safety factor in the range of 5 to 20.

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

2008-10-31

342

Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012  

SciTech Connect

Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace several obsolete components of the current analytical tool set used for ATR neutronics support. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). It will also greatly facilitate the LEU conversion effort, since the upgraded computational capabilities are now at a stage where they can be, and in fact have been, used for the required physics analysis from the beginning. In this context, extensive scoping neutronics analyses were completed for six preconceptual candidate LEU fuel element designs for the ATR (and for its companion critical facility, ATRC). Of these, four exhibited neutronics performance in what is believed to be an acceptable range. However, there are currently some concerns with regard to fabricability and mechanical performance that have emerged for one of the four latter concepts. Thus three concepts have been selected for more comprehensive conceptual design analysis during the upcoming fiscal year.

David W. Nigg; Sean R. Morrell

2012-09-01

343

Electrical Conductivity of Neutron Star Cores in the Presence of a Magnetic Field - Part Two - a Free-Particle Model of Npe - / Matter  

NASA Astrophysics Data System (ADS)

General theory of electrical conductivity of a multicomponent mixture of degenerate fermions in a magnetic fieldB, developed in the preceding article (this volume), is applied to a matter in neutron star interiors at densities ???0, where ?0 = 2.81014 g cm-3 is the standard nuclear matter density. A model of free-particle mixture ofn, p, e is used, with account for appearance of ?--hyperons at ?>? c , where ? c ?4?0. The electric resistivities along and acrossB, ?? and ??, and the Hall resistivity ? H are calculated and fitted by simple analytical formulae at ??? c and ?>? c for the cases of normal or superfluid neutrons provided other particles are normal. Charge transport alongB is produced by electrons, due to their Coulombic collisions with other charged particles; ?? is independent ofB and almost independent of the neutron superfluidity. Charge transport acrossB at largeB may be essentially determined by other charged particles. If ??? c , one has ?? = ??[1 + (B/B 0)2] for the normal neutrons, and ?? for the superfluid neutrons, while ? H = ?? B/B e for both cases. HereB e 109 T {8/2}G,B 01011 T {8/2}G, andT 8 is temperature in units of 108 K. Accordingly for the normal neutrons atB?B 0, the transverse resistivity ?? suffers an enhancement, ?1/4 ? ?1. When ??5?0 andB varies from 0 toBB p ? 1013 T {8/2}G, ?? increases by a factor of about 103 104 and ? H changes sign. WhenB?B p , ?? remains constant for the superfluid neutrons, and ? H ?B 2 for the normal neutrons, while ? H ?B for any neutron state. Strong dependence of resistivity onB, T, and ? may affect evolution of magnetic fields in neutron star cores. In particular, the enhancement of ?? at highB may noticeably speed up the Ohmic decay of those electric currents which are perpendicular toB.

Yakovlev, D. G.; Shalybkov, D. A.

1991-02-01

344

Enhancement of REBUS-3/DIF3D for whole-core neutronic analysis of prismatic very high temperature reactor (VHTR).  

SciTech Connect

Enhancements have been made to the REBUS-3/DIF3D code suite to facilitate its use for the design and analysis of prismatic Very High Temperature Reactors (VHTRs). A new cross section structure, using table-lookup, has been incorporated to account for cross section changes with burnup and fuel and moderator temperatures. For representing these cross section dependencies, three new modules have been developed using FORTRAN 90/95 object-oriented data structures and implemented within the REBUS-3 code system. These modules provide a cross section storage procedure, construct microscopic cross section data for all isotopes, and contain a single block of banded scattering data for efficient data management. Fission products other than I, Xe, Pm, and Sm, can be merged into a single lumped fission product to save storage space, memory, and computing time without sacrificing the REBUS-3 solution accuracy. A simple thermal-hydraulic (thermal-fluid) feedback model has been developed for prismatic VHTR cores and implemented in REBUS-3 for temperature feedback calculations. Axial conduction was neglected in the formulation because of its small magnitude compared to radial (planar) conduction. With the simple model, the average fuel and graphite temperatures are accurately estimated compared to reference STAR-CD results. The feedback module is currently operational for the non-equilibrium fuel cycle analysis option of REBUS-3. Future work should include the extension of this capability to the equilibrium cycle option of the code and additional verification of the feedback module. For the simulation of control rods in VHTR cores, macroscopic cross section deviations (deltas) have been defined to account for the effect of control rod insertion. The REBUS-3 code has been modified to use the appropriately revised cross sections when control rods are inserted in a calculation node. In order to represent asymmetric core blocks (e.g., fuel blocks or reflector blocks containing asymmetric absorber rods), surface-dependent discontinuity factors based on nodal equivalence theory have been introduced into the nodal diffusion theory option of the DIF3D code (DIF3D-nodal) to improve modeling accuracy. Additionally, the discontinuity factors based on the Simplified Equivalence Theory (SET) have been incorporated as an alternative and may be employed for both the DIF3D-nodal and DIF3D-VARIANT (nodal transport) solution options. Two- and three-dimensional core calculations have been performed using the routines developed and modified in this work, along with cross sections generated from single fuel block and one-dimensional or two-dimensional fuel-reflector model. Generally, REBUS-3/DIF3D results for the core multiplication factor and power distribution are found to be in good agreement with reference results (generated with MCNP continuous energy calculations) particularly when discontinuity factors are applied. The DIF3D-VARIANT option was found to provide a more accurate solution in its diffusion approximation than the DIF3D-nodal option. Control rod worths can be estimated with acceptably small errors compared to MCNP results. However, estimation of the core power tilt needs to be improved by introducing the surface-dependent discontinuity factor capability in DIF3D-VARIANT.

Lee, C. H.; Zhong, Z.; Taiwo, T.A.; Yang, W.S.; Khalil, H.S.; Smith, M.A.; Nuclear Engineering Division

2006-10-13

345

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

NASA Astrophysics Data System (ADS)

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

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

2014-08-01

346

Rotatable multifunctional load frames for neutron diffractometers at FRM IIdesign, specifications and applications  

NASA Astrophysics Data System (ADS)

Novel tensile rigs have been designed and manufactured at the research reactor Heinz Maier-Leibnitz (FRM II, Garching near Munich). Besides tensile and compressive stress, also torsion can be applied. The unique Eulerian cradle type design (?, ?, and ? axis) allows orienting the stress axis with respect to the scattering vector. Applications of these tensile rigs at our neutron diffractometers enable various investigations of structural changes under mechanical load, e.g. crystallographic texture evolution, stress-induced phase transformations or lattice expansion, and the anisotropy of mechanical response.

Hoelzel, M.; Gan, W. M.; Hofmann, M.; Randau, C.; Seidl, G.; Jttner, Ph.; Schmahl, W. W.

2013-05-01

347

Fluence-limited burnup as a function of fast reactor core parameters  

E-print Network

The limiting factor in current designs for fast reactors is not only the reactivity, but also the maximum permissible fast-neutron fluence in the cladding, especially for reduced uranium enrichment cores using high-albedo ...

Kersting, Alyssa (Alyssa Rae)

2011-01-01

348

Optimal design at inner core of the shaped pyramidal truss structure  

NASA Astrophysics Data System (ADS)

Sandwich material is a type of composite material with lightweight, high strength, good dynamic properties and high bending stiffness-to-weight ratio. This can be found well such structures in the nature (for example, internal structure of bones, plants, etc.). New trend which prefers eco-friendly products and energy efficiency is emerging in industries recently. Demand for materials with high strength and light weight is also increasing. In line with these trends, researches about manufacturing methods of sandwich material have been actively conducted. In this study, a sandwich structure named as "Shaped Pyramidal Truss Structure" is proposed to improve mechanical strength and to apply a manufacturing process suitable for massive production. The new sandwich structure was designed to enhance compressive strength by changing the cross-sectional shape at the central portion of the core. As the next step, optimization of the shape was required. Optimization technique used here was the SZGA(Successive Zooming Genetic Algorithm), which is one of GA(Genetic Algorithm) methods gradually reducing the area of design variable. The objective function was defined as moment of inertia of the cross-sectional shape of the strut. The control points of cubic Bezier curve, which was assumed to be the shape of the cross section, were used as design variables. By using FEM simulation, it was found that the structure exhibited superior mechanical properties compared to the simple design of the prior art.

Lee, Sung-Uk; Yang, Dong-Yol

2013-12-01

349

Optimal design at inner core of the shaped pyramidal truss structure  

SciTech Connect

Sandwich material is a type of composite material with lightweight, high strength, good dynamic properties and high bending stiffness-to-weight ratio. This can be found well such structures in the nature (for example, internal structure of bones, plants, etc.). New trend which prefers eco-friendly products and energy efficiency is emerging in industries recently. Demand for materials with high strength and light weight is also increasing. In line with these trends, researches about manufacturing methods of sandwich material have been actively conducted. In this study, a sandwich structure named as Shaped Pyramidal Truss Structure is proposed to improve mechanical strength and to apply a manufacturing process suitable for massive production. The new sandwich structure was designed to enhance compressive strength by changing the cross-sectional shape at the central portion of the core. As the next step, optimization of the shape was required. Optimization technique used here was the SZGA(Successive Zooming Genetic Algorithm), which is one of GA(Genetic Algorithm) methods gradually reducing the area of design variable. The objective function was defined as moment of inertia of the cross-sectional shape of the strut. The control points of cubic Bezier curve, which was assumed to be the shape of the cross section, were used as design variables. By using FEM simulation, it was found that the structure exhibited superior mechanical properties compared to the simple design of the prior art.

Lee, Sung-Uk; Yang, Dong-Yol [Department of Mechanical Engineering, KAIST 291 Daehak-ro (373-1 Guseong-dong), Yuseong-gu, Dae-jeon, 305-701 (Korea, Republic of)

2013-12-16

350

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

NASA Astrophysics Data System (ADS)

A gripping capability was designed, implemented, and tested for in situ neutron diffraction measurements during multiaxial loading and heating on the VULCAN engineering materials diffractometer at the spallation neutron source at Oak Ridge National Laboratory. The proposed capability allowed for the acquisition of neutron spectra during tension, compression, torsion, and/or complex loading paths at elevated temperatures. The design consisted of age-hardened, Inconel 718 grips with direct attachment to the existing MTS load frame having axial and torsional capacities of 100 kN and 400 N.m, respectively. Internal cooling passages were incorporated into the gripping system for fast cooling rates during high temperature experiments up to 1000 K. The specimen mounting couplers combined a threaded and hexed end-connection for ease of sample installation/removal without introducing any unwanted loads. Instrumentation of this capability is documented in this work along with various performance parameters. The gripping system was utilized to investigate deformation in NiTi shape memory alloys under various loading/control modes (e.g., isothermal, isobaric, and cyclic), and preliminary results are presented. The measurements facilitated the quantification of the texture, internal strain, and phase fraction evolution in NiTi shape memory alloys under various loading/control modes.

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

2014-10-01

351

Polar-drive designs for optimizing neutron yields on the National Ignition Facility  

SciTech Connect

Polar-drive designs are proposed for producing symmetric implosions of thin-shell, DT gas-filled targets leading to high fusion-neutron yields for neutron-diagnostic development. The designs can be implemented as soon as the National Ignition Facility (NIF) [E. M. Campbell and W. J. Hogan, Plasma Phys. Control. Fusion 41, B39 (1999)] is operational as they use indirect-drive phase plates. Two-dimensional simulations using the hydrodynamics code SAGE [R. S. Craxton and R. L. McCrory, J. Appl. Phys. 56, 108 (1984)] have shown that good low-mode uniformity can be obtained by choosing combinations of pointing and defocusing of the beams, including pointing offsets of individual beams within some of the NIF laser-beam quads. The optimizations have been carried out for total laser energies ranging from 350 kJ to 1.5 MJ, enabling the optimum pointing and defocusing parameters to be determined through interpolation for any given laser energy in this range. Neutron yields in the range of 10{sup 15}-10{sup 16} are expected.

Cok, A. M.; Craxton, R. S.; McKenty, P. W. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623 (United States)

2008-08-15

352

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

SciTech Connect

A gripping capability was designed, implemented, and tested for in situ neutron diffraction measurements during multiaxial loading and heating on the VULCAN engineering materials diffractometer at the spallation neutron source at Oak Ridge National Laboratory. The proposed capability allowed for the acquisition of neutron spectra during tension, compression, torsion, and/or complex loading paths at elevated temperatures. The design consisted of age-hardened, Inconel{sup } 718 grips with direct attachment to the existing MTS load frame having axial and torsional capacities of 100 kN and 400 Nm, respectively. Internal cooling passages were incorporated into the gripping system for fast cooling rates during high temperature experiments up to ?1000 K. The specimen mounting couplers combined a threaded and hexed end-connection for ease of sample installation/removal without introducing any unwanted loads. Instrumentation of this capability is documented in this work along with various performance parameters. The gripping system was utilized to investigate deformation in NiTi shape memory alloys under various loading/control modes (e.g., isothermal, isobaric, and cyclic), and preliminary results are presented. The measurements facilitated the quantification of the texture, internal strain, and phase fraction evolution in NiTi shape memory alloys under various loading/control modes.

Benafan, O., E-mail: othmane.benafan@nasa.gov [NASA Glenn Research Center, Structures and Materials Division, Cleveland, Ohio 44135 (United States); Advanced Materials Processing and Analysis Center, Materials Science and Engineering Department, University of Central Florida, Orlando, Florida 32816 (United States); Padula, S. A. [NASA Glenn Research Center, Structures and Materials Division, Cleveland, Ohio 44135 (United States); Skorpenske, H. D.; An, K. [Spallation Neutron Source, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Vaidyanathan, R. [Advanced Materials Processing and Analysis Center, Materials Science and Engineering Department, University of Central Florida, Orlando, Florida 32816 (United States)

2014-10-15

353

Design Review Report for formal review of safety class features of exhauster system for rotary mode core sampling  

SciTech Connect

Report documenting Formal Design Review conducted on portable exhausters used to support rotary mode core sampling of Hanford underground radioactive waste tanks with focus on Safety Class design features and control requirements for flammable gas environment operation and air discharge permitting compliance.

JANICEK, G.P.

2000-06-08

354

Preliminary Neutronics Design Studies for a Molten Salt Blanket LIFE Engine  

SciTech Connect

The Laser Inertial Confinement Fusion Fission Energy (LIFE) Program being developed at Lawrence Livermore National Laboratory (LLNL) aims to design a hybrid fission-fusion subcritical nuclear engine that uses a laser-driven Inertial Confinement Fusion (ICF) system to drive a subcritical fission blanket. This combined fusion-fission hybrid system could be used for generating electricity, material transmutation or incineration, or other applications. LIFE does not require enriched fuel since it is a sub-critical system and LIFE can sustain power operation beyond the burnup levels at which typical fission reactors need to be refueled. In light of these factors, numerous options have been suggested and are being investigated. Options being investigated include fueling LIFE engines with spent nuclear fuel to aid in disposal/incineration of commercial spent nuclear fuel or using depleted uranium or thorium fueled options to enhance proliferation resistance and utilize non-fissile materials [1]. LIFE engine blanket designs using a molten salt fuel system represent one area of investigation. Possible applications of a LIFE engine with a molten salt blanket include uses as a spent nuclear fuel burner, fissile fuel breeding platform, and providing a backup alternative to other LIFE engine blanket designs using TRISO fuel particles in case the TRISO particles are found to be unable to withstand the irradiation they will be subjected to. These molten salts consist of a mixture of LiF with UF{sub 4} or ThF{sub 4} or some combination thereof. Future systems could look at using PuF{sub 3} or PuF{sub 4} as well, though no work on such system with initial plutonium loadings has been performed for studies documented in this report. The purpose of this report is to document preliminary neutronics design studies performed to support the development of a molten salt blanket LIFE engine option, as part of the LIFE Program being performed at Lawrence Livermore National laboratory. Preliminary design studies looking at fast ignition and hot spot ignition fusion options are documented, along with limited scoping studies performed to investigate other options of interest that surfaced during the main design effort. Lastly, side studies that were not part of the main design effort but may alter future work performed on LIFE engine designs are shown. The majority of all work reported in this document was performed during the Molten Salt Fast Ignition Moderator Study (MSFIMS) which sought to optimize the amount of moderator mixed into the molten salt region in order to produce the most compelling design. The studies in this report are of a limited scope and are intended to provide a preliminary neutronics analysis of the design concepts described herein to help guide decision processes and explore various options that a LIFE engine with a molten salt blanket might enable. None of the designs shown in this report, even reference cases selected for detailed description and analysis, have been fully optimized. The analyses were performed primarily as a neutronics study, though some consultation was made regarding thermal-hydraulic and structural concerns during both scoping out an initial model and subsequent to identifying a neutronics-based reference case to ensure that the design work contained no glaring mechanical or thermal issues that would preclude its feasibility. Any analyses and recommendations made in this report are either primarily or solely from the point of view of LIFE neutronics and ignore other fundamental issues related to molten salt fuel blankets such as chemical processing feasibility and political feasibility of a molten salt system.

Powers, J

2008-10-23

355

Melt spreading code assessment, modifications, and application to the EPR core catcher design.  

SciTech Connect

The Evolutionary Power Reactor (EPR) is under consideration by various utilities in the United States to provide base load electrical production, and as a result the design is undergoing a certification review by the U.S. Nuclear Regulatory Commission (NRC). The severe accident design philosophy for this reactor is based upon the fact that the projected power rating results in a narrow margin for in-vessel melt retention by external cooling of the reactor vessel. As a result, the design addresses ex-vessel core melt stabilization using a mitigation strategy that includes: (1) an external core melt retention system to temporarily hold core melt released from the vessel; (2) a layer of 'sacrificial' material that is admixed with the melt while in the core melt retention system; (3) a melt plug in the lower part of the retention system that, when failed, provides a pathway for the mixture to spread to a large core spreading chamber; and finally, (4) cooling and stabilization of the spread melt by controlled top and bottom flooding. The overall concept is illustrated in Figure 1.1. The melt spreading process relies heavily on inertial flow of a low-viscosity admixed melt to a segmented spreading chamber, and assumes that the melt mass will be distributed to a uniform height in the chamber. The spreading phenomenon thus needs to be modeled properly in order to adequately assess the EPR design. The MELTSPREAD code, developed at Argonne National Laboratory, can model segmented, and both uniform and nonuniform spreading. The NRC is thus utilizing MELTSPREAD to evaluate melt spreading in the EPR design. MELTSPREAD was originally developed to support resolution of the Mark I containment shell vulnerability issue. Following closure of this issue, development of MELTSPREAD ceased in the early 1990's, at which time the melt spreading database upon which the code had been validated was rather limited. In particular, the database that was utilized for initial validation consisted of: (1) comparison to an analytical solution for the dam break problem, (2) water spreading tests in a 1/10 linear scale model of the Mark I containment by Theofanous et al., and (3) steel spreading tests by Suzuki et al. that were also conducted in a geometry similar to the Mark I. The objective of this work was to utilize the MELTSPREAD code to check the assumption of uniform melt spreading in the EPR core catcher design. As a starting point for the project, the code was validated against the worldwide melt spreading database that emerged after the code was originally written in the very early 1990's. As part of this exercise, the code was extensively modified and upgraded to incorporate findings from these various analytical and experiment programs. In terms of expanding the ability of the code to analyze various melt simulant experiments, the options to input user-specified melt and/or substrate material properties was added. The ability to perform invisicid and/or adiabatic spreading analysis was also added so that comparisons with analytical solutions and isothermal spreading tests could be carried out. In terms of refining the capability to carry out reactor material melt spreading analyses, the code was upgraded with a new melt viscosity model; the capability was added to treat situations in which solid fraction buildup between the liquidus-solidus is non-linear; and finally, the ability to treat an interfacial heat transfer resistance between the melt and substrate was incorporated. This last set of changes substantially improved the predictive capability of the code in terms of addressing reactor material melt spreading tests. Aside from improvements and upgrades, a method was developed to fit the model to the various melt spreading tests in a manner that allowed uncertainties in the model predictions to be statistically characterized. With these results, a sensitivity study was performed to investigate the assumption of uniform spreading in the EPR core catcher that addressed parametric variations in: (1) melt pour mass, (2) melt composition, (3) me

Farmer, M. T .; Nuclear Engineering Division

2009-03-30

356

Neutron and Synchrotron Radiation Studies for Designer Materials, Sustainable Energy and Healthy Lives  

NASA Astrophysics Data System (ADS)

Probably the most prolific use of large accelerators today is in the creation of bright beams of x-ray photons or neutrons. The number of scientific users of such sources in the US alone is approaching 10,000. I will describe the some of the major applications of synchrotron and neutron radiation and their impact on society. If you have AIDS, need a better IPOD or a more efficient car, or want to clean up a superfund site, you are benefitting from these accelerators. The design of new materials is becoming more and more dependent on structural information from these sources. I will identify the trends in applications which are demanding new sources with greater capabilities.

Gibson, J. Murray

2009-05-01

357

Prospect of Spherical Tokamak towards a Power Reactor Challenge towards the Lowest Aspect Ratio Tokamak 5. Can We Obtain the Realistic Power Reactor in the ST Approach? 5.3 Neutron Shielding and Blanket Neutronics Design  

Microsoft Academic Search

Considering the geometrical characteristics of tokamak reactors with low aspect ratio, a basic neutronics strategy was derived to construct an inboard structure mainly for neutron shielding and to produce enough tritium in the outboard blanket. The designs for optimal inboard shield were surveyed and the necessary thickness was estimated to make the neutron flux sufficiently low on the super-conducting magnet.

Michinori Yamauchi; Takeo Nishitani; Satoshi Nishio

2004-01-01

358

Design of the Testing Set-up for a Nuclear Fuel Rod by Neutron Radiography at CARR  

NASA Astrophysics Data System (ADS)

In this paper, an experimental set-up dedicated to non-destructively test a 15cm-long Pressurized Water Reactor (PWR) nuclear fuel rod by neutron radiography (NR) is described. It consists of three parts: transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo Simulation by the MCNP code. The material for the shell of the transport container was chosen to be lead with the thickness of 13 cm. Also, the mechanical devices were designed to control fuel rod movement inside the container. The imaging block was designed as the exposure platform, with three openings for the neutron beam, neutron converter foil, and specimen. Development and application of this experimental set-up will help gain much experience for investigating the actual irradiated fuel rod by neutron radiography at CARR in the future.

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

359

Chemical and Colloidal Stability of Carboxylated Core-Shell Magnetite Nanoparticles Designed for Biomedical Applications  

PubMed Central

Despite the large efforts to prepare super paramagnetic iron oxide nanoparticles (MNPs) for biomedical applications, the number of FDA or EMA approved formulations is few. It is not known commonly that the approved formulations in many instances have already been withdrawn or discontinued by the producers; at present, hardly any approved formulations are produced and marketed. Literature survey reveals that there is a lack for a commonly accepted physicochemical practice in designing and qualifying formulations before they enter in vitro and in vivo biological testing. Such a standard procedure would exclude inadequate formulations from clinical trials thus improving their outcome. Here we present a straightforward route to assess eligibility of carboxylated MNPs for biomedical tests applied for a series of our core-shell products, i.e., citric acid, gallic acid, poly(acrylic acid) and poly(acrylic acid-co-maleic acid) coated MNPs. The discussion is based on physicochemical studies (carboxylate adsorption/desorption, FTIR-ATR, iron dissolution, zeta potential, particle size, coagulation kinetics and magnetization measurements) and involves in vitro and in vivo tests. Our procedure can serve as an example to construct adequate physico-chemical selection strategies for preparation of other types of core-shell nanoparticles as well. PMID:23857054

Szekeres, Mrta; Tth, Ildik Y.; Ills, Erzsbet; Hajd, Angla; Zupk, Istvn; Farkas, Katalin; Oszlnczi, Gbor; Tiszlavicz, Lszl; Tombcz, Etelka

2013-01-01

360

Core design study of a supercritical light water reactor with double row fuel rods  

SciTech Connect

An equilibrium core for supercritical light water reactor has been designed. A novel type of fuel assembly with dual rows of fuel rods between water rods is chosen and optimized to get more uniform assembly power distributions. Stainless steel is used for fuel rod cladding and structural material. Honeycomb structure filled with thermal isolation is introduced to reduce the usage of stainless steel and to keep moderator temperature below the pseudo critical temperature. Water flow scheme with ascending coolant flow in inner regions is carried out to achieve high outlet temperature. In order to enhance coolant outlet temperature, the radial power distributions needs to be as flat as possible through operation cycle. Fuel loading pattern and control rod pattern are optimized to flatten power distribution at inner regions. Axial fuel enrichment is divided into three parts to control axial power peak, which affects maximum cladding surface temperature. (authors)

Zhao, C.; Wu, H.; Cao, L.; Zheng, Y. [School of Nuclear Science and Technology, Xi'an Jiaotong Univ., No. 28, Xianning West Road, Xi'an, ShannXi, 710049 (China); Yang, J.; Zhang, Y. [China Nuclear Power Technology Research Inst., Yitian Road, ShenZhen, GuangDong, 518026 (China)

2012-07-01

361

Parameter Design and Optimal Control of an Open Core Flywheel Energy Storage System  

NASA Technical Reports Server (NTRS)

In low earth orbit (LEO) satellite applications spacecraft power is provided by photovoltaic cells and batteries. To overcome battery shortcomings the University of Maryland, working in cooperation with NASA/GSFC and NASA/LeRC, has developed a magnetically suspended flywheel for energy storage applications. The system is referred to as an Open Core Composite Flywheel (OCCF) energy storage system. Successful application of flywheel energy storage requires integration of several technologies, viz. bearings, rotor design, motor/generator, power conditioning, and system control. In this paper we present a parameter design method which has been developed for analyzing the linear SISO model of the magnetic bearing controller for the OCCF. The objective of this continued research is to principally analyze the magnetic bearing system for nonlinear effects in order to increase the region of stability, as determined by high speed and large air gap control. This is achieved by four tasks: (1) physical modeling, design, prototyping, and testing of an improved magnetically suspended flywheel energy storage system, (2) identification of problems that limit performance and their corresponding solutions, (3) development of a design methodology for magnetic bearings, and (4) design of an optimal controller for future high speed applications. Both nonlinear SISO and MIMO models of the magnetic system were built to study limit cycle oscillations and power amplifier saturation phenomenon observed in experiments. The nonlinear models include the inductance of EM coils, the power amplifier saturation, and the physical limitation of the flywheel movement as discussed earlier. The control program EASY5 is used to study the nonlinear SISO and MIMO models. Our results have shown that the characteristics and frequency responses of the magnetic bearing system obtained from modeling are comparable to those obtained experimentally. Although magnetic saturation is shown in the bearings, there are good correlations between the theoretical model and experimental data. Both simulation and experiment confirm large variations of the magnetic bearing characteristics due to air gap growth. Therefore, the gap growth effect should be considered in the magnetic bearing system design. Additionally, the magnetic bearing control system will be compared to other design methods using not only parameter design but H-infinity optimal control and mu synthesis.

Pang, D.; Anand, D. K.; Kirk, J. A.

1996-01-01

362

On the development of computational tools for the design of beam assemblies for boron neutron capture therapy  

NASA Astrophysics Data System (ADS)

This article is the first in a series devoted to the development of efficient and accurate computational tools for the design of beam assemblies for boron neutron capture therapy within the framework of discrete ordinates spectral nodal methods of neutron transport theory. We begin our study with a multi-layer representation of an assembly, and we derive a discrete ordinates matrix operator that replaces without spatial truncation error the entire multi-layer domain in neutron transmission computations. With the matrix operator derived here, we compute without further ado the angular distribution of neutrons leaving the multi-layer assembly, avoiding thus the use of general-purpose discrete ordinates codes founded in the discretization and numerical solution of the neutron transport equation over a number of spatial cells and angular directions throughout the domain. We perform numerical experiments with a four-layer model assembly, and we conclude this article with a discussion and directions for further developments.

de Abreu, Marcos Pimenta

2007-07-01

363

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

SciTech Connect

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

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

1994-12-01

364

Reliability Design for Neutron Induced Single-Event Burnout of IGBT  

NASA Astrophysics Data System (ADS)

Single-event burnout (SEB) caused by cosmic ray neutrons leads to catastrophic failures in insulated gate bipolar transistors (IGBTs). It was found experimentally that the incident neutron induced SEB failure rate increases as a function of the applied collector voltage. Moreover, the failure rate increased sharply with an increase in the applied collector voltage when the voltage exceeded a certain threshold value (SEB cutoff voltage). In this paper, transient device simulation results indicate that impact ionization at the n-drift/n+ buffer boundary is a crucially important factor in the turning-on of the parasitic pnp transistor, and eventually latch-up of the parasitic thyristor causes SEB. In addition, the device parameter dependency of the SEB cutoff voltage was analytically derived from the latch-up condition of the parasitic thyristor. As a result, it was confirmed that reducing the current gain of the parasitic transistor, such as by increasing the n-drift region thickness d was effective in increasing the SEB cutoff voltage. Furthermore, `white' neutron-irradiation experiments demonstrated that suppressing the inherent parasitic thyristor action leads to an improvement of the SEB cutoff voltage. It was confirmed that current gain optimization of the parasitic transistor is a crucial factor for establishing highly reliable design against chance failures.

Shoji, Tomoyuki; Nishida, Shuichi; Ohnishi, Toyokazu; Fujikawa, Touma; Nose, Noboru; Hamada, Kimimori; Ishiko, Masayasu

365

Design and performance of a new high accuracy combined small sample neutron/gamma detector  

SciTech Connect

This paper describes the design of an optimized combined neutron and gamma detector installed around a measurement well protruding from the floor of a glove box. The objective of this design was to achieve an overall accuracy for the plutonium element concentration in gram-sized samples of plutonium oxide powder approaching the {approximately}0.1--0.2% accuracies routinely achieved by inspectors` chemical analysis. The efficiency of the clam-shell neutron detector was increased and the flat response zone extended in axial and radial directions. The sample holder introduced from within the glove box was designed to form the upper reflector, while two graphite half-shells fitted around the thin neck of the high-resolution LEGE detector replaced the lower plug. The Institute for Reference Materials and Measurements (IRMM) in Geel prepared special plutonium oxide test samples whose plutonium concentration was determined to better than 0.05%. During a three week initial performance test in July 1992 at ITU Karlsruhe and in long term tests, it was established that the target accuracy can be achieved provided sufficient care is taken to assure the reproducibility of sample bottling and sample positioning. The paper presents and discusses the results of all test measurements.

Menlove, H. [Los Alamos National Lab., NM (United States); Davidson, D.; Verplancke, J.; Vermeulen, P. [Jomar Systems, Inc., Los Alamos, NM (United States); Wagner, H.G. [Commission of the European Communities, Luxembourg (Luxembourg); Wellum, R.; Brandelise, B. [Commission of the European Communities, Karlsruhe (Germany, F.R.). European Inst. for Transuranium Elements; Mayer, K. [Commission of the European Communities, Geel (Belgium)

1993-08-01

366

Design and performance of a new high accuracy combined small sample neutron/gamma detector  

SciTech Connect

This paper describes the design of an optimized combined neutron and gamma detector installed around a measurement well protruding from the floor of a glove box. The objective of this design was to achieve an overall accuracy for the plutonium element concentration in gram-sized samples of plutonium oxide powder approaching the {approximately}0.1--0.2% accuracies routinely achieved by inspectors` chemical analysis. The efficiency of the clam-shell neutron detector was increased and the flat response zone extended in axial and radial directions. The sample holder introduced from within the glove box was designed to form the upper reflector, while two graphite half-shells fitted around the thin neck of the high-resolution LEGe detector replaced the lower plug. The Institute for Reference Materials and Measurements (IRMM) in Geel prepared special plutonium oxide test samples whose plutonium concentration was determined to better than 0.05%. During a three week initial performance test in July 1992 at ITU Karlsruhe and in long term tests, it was established that the target accuracy can be achieved provided sufficient care is taken to assure the reproducibility of sample bottling and sample positioning. The paper presents and discusses the results of all test measurements.

Menlove, H. [Los Alamos National Lab., NM (United States); Davidson, D.; Verplancke, J.; Vermeulen, P. [Canberra Industries, Los Alamos, NM (United States). Jomar Systems Division; Wagner, H.G. [Commission of the European Communities, Luxembourg (Luxembourg). EURATOM Safeguards Directorate; Wellum, R.; Brandelise, B. [Commission of the European Communities, Karlsruhe (Germany). Institute for Transuranium Elements; Mayer, K. [Commission the European Communities, Geel (Germany). Institute for Reference Materials and Measurements

1993-12-31

367

MIC-SVM: Designing A Highly Efficient Support Vector Machine For Advanced Modern Multi-Core and Many-Core Architectures  

SciTech Connect

Support Vector Machine (SVM) has been widely used in data-mining and Big Data applications as modern commercial databases start to attach an increasing importance to the analytic capabilities. In recent years, SVM was adapted to the ?eld of High Performance Computing for power/performance prediction, auto-tuning, and runtime scheduling. However, even at the risk of losing prediction accuracy due to insuf?cient runtime information, researchers can only afford to apply of?ine model training to avoid signi?cant runtime training overhead. To address the challenges above, we designed and implemented MICSVM, a highly efficient parallel SVM for x86 based multi-core and many core architectures, such as the Intel Ivy Bridge CPUs and Intel Xeon Phi coprocessor (MIC).

You, Yang; Song, Shuaiwen; Fu, Haohuan; Marquez, Andres; Mehri Dehanavi, Maryam; Barker, Kevin J.; Cameron, Kirk; Randles, Amanda; Yang, Guangwen

2014-08-16

368

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

PubMed

A radiation shielding has been designed to reduce scattered neutrons and background gamma-rays for the new double-ring Time Of Flight Enhanced Diagnostics (TOFED). The shielding was designed based on simulation with the Monte Carlo code MCNP5. Dedicated model of the EAST tokamak has been developed together with the emission neutron source profile and spectrum; the latter were simulated with the Nubeam and GENESIS codes. Significant reduction of background radiation at the detector can be achieved and this satisfies the requirement of TOFED. The intensities of the scattered and direct neutrons in the line of sight of the TOFED neutron spectrometer at EAST are studied for future data interpretation. PMID:25430294

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

2014-11-01

369

Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2006  

SciTech Connect

Neutronics and thermal-hydraulics studies show that, for equivalent operating power [85 MW(t)], a low-enriched uranium (LEU) fuel cycle based on uranium-10 wt % molybdenum (U-10Mo) metal foil with radially, continuously graded fuel meat thickness results in a 15% reduction in peak thermal flux in the beryllium reflector of the High Flux Isotope Reactor (HFIR) as compared to the current highly enriched uranium (HEU) cycle. The uranium-235 content of the LEU core is almost twice the amount of the HEU core when the length of the fuel cycle is kept the same for both fuels. Because the uranium-238 content of an LEU core is a factor of 4 greater than the uranium-235 content, the LEU HFIR core would weigh 30% more than the HEU core. A minimum U-10Mo foil thickness of 84 ?m is required to compensate for power peaking in the LEU core although this value could be increased significantly without much penalty. The maximum U-10Mo foil thickness is 457?m. Annual plutonium production from fueling the HFIR with LEU is predicted to be 2 kg. For dispersion fuels, the operating power for HFIR would be reduced considerably below 85 MW due to thermal considerations and due to the requirement of a 26-d fuel cycle. If an acceptable fuel can be developed, it is estimated that $140 M would be required to implement the conversion of the HFIR site at Oak Ridge National Laboratory from an HEU fuel cycle to an LEU fuel cycle. To complete the conversion by fiscal year 2014 would require that all fuel development and qualification be completed by the end of fiscal year 2009. Technological development areas that could increase the operating power of HFIR are identified as areas for study in the future.

Primm, R. T. [ORNL] [ORNL; Ellis, R. J. [ORNL] [ORNL; Gehin, J. C. [ORNL] [ORNL; Clarno, K. T. [ORNL] [ORNL; Williams, K. A. [ORNL] [ORNL; Moses, D. L. [ORNL] [ORNL

2006-11-01

370

The Design and Construction of a Cold Neutron Source for Use in the Cornell University Triga Reactor  

NASA Astrophysics Data System (ADS)

A cold neutron source has been designed and constructed for insertion into the 6"-radial beam port of the Cornell University TRIGA reactor for use with a neutron guide tube system. The main differences between this cold source and other existing sources are the use of heat conduction as the method of cooling and the use of mesitylene (1,3,5 -trimethylbenzene; melting point, 228(DEGREES)K; boiling point, 437(DEGREES)K) as the moderating material. This thesis describes the design and construction details of the cold neutron source, discusses its safety aspects, and presents its cryogenic performance curves and also the results of a test of its neutron moderating ability. A closed-cycle helium gas refrigerator, located outside the reactor shielding, cools the 500 cm('3) moderator chamber and its surrounding heat shield by heat conduction through two meters of copper and rod tubing. Moderator temperatures of 23 (+OR-) 3(DEGREES)K have been achieved. Mesitylene, a hydrocarbon, is an effective cold moderator because even at low temperatures the weakly hindered rotational motions of its methyl groups enable the absorption of small amounts of energy ((LESSTHEQ) 0.005 eV) from neutrons. The use of mesitylene simplifies the cold source design because it is a liquid at room temperature and thus, the usual design safeguards required for sources using gaseous moderators are not necessary. Moreover, the flammability of mesitylene is much smaller than that of hydrogen and methane, which are the commonly used cold moderators. A method of transferring and handling the mesitylene, a carcinogen, was devised to ensure minimal contact with this substance. To test the neutron moderating ability of the cold neutron source, an out-of-reactor neutron transmission experiment was performed with the moderator chamber first at room temperature and then at about 23(DEGREES)K. The results indicate that the neutron energy spectrum is strongly shifted to lower energies when the chamber is cold. The neutrons have an apparent temperature of 98(DEGREES)K, if the moderated flux is assumed to be Maxwellian. This, however, is unlikely to be the actual case since the geometry of the chamber, neutron beam, and detector and the complexities of the neutron scattering and absorption processes in mesitylene doubtlessly distort the leakage energy spectrum.

Young, Lydia Jane

371

Mixed enrichment core design for the NC State University PULSTAR Reactor  

SciTech Connect

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

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

1997-12-01

372

On the development of computational tools for the design of beam assemblies for boron neutron capture therapy  

Microsoft Academic Search

This article is the first in a series devoted to the development of efficient and accurate computational tools for the design\\u000a of beam assemblies for boron neutron capture therapy within the framework of discrete ordinates spectral nodal methods of\\u000a neutron transport theory. We begin our study with a multi-layer representation of an assembly, and we derive a discrete ordinates\\u000a matrix

Marcos Pimenta de Abreu

2007-01-01

373

Thermal hydraulic design and decay heat removal of a solid target for a spallation neutron source  

NASA Astrophysics Data System (ADS)

Thermal hydraulic design and thermal stress calculations were conducted for a water-cooled solid target irradiated by a MW-class proton beam for a spallation neutron source. Plate type and rod bundle type targets were examined. The thickness of the plate and the diameter of the rod were determined based on the maximum and the wall surface temperature. The thermal stress distributions were calculated by a finite element method (FEM). The neutronics performance of the target is roughly proportional to its average density. The averaged densities of the designed targets were calculated for tungsten plates, tantalum clad tungsten plates, tungsten rods sheathed by tantalum and Zircaloy and they were compared with mercury density. It was shown that the averaged density was highest for the tungsten plates and was high for the tantalum cladding tungsten plates, the tungsten rods sheathed by tantalum and Zircaloy in order. They were higher than or equal to that of mercury for the 1 2 MW proton beams. Tungsten target without the cladding or the sheath is not practical due to corrosion by water under irradiation condition. Therefore, the tantalum cladding tungsten plate already made successfully by HIP and the sheathed tungsten rod are the candidate of high performance solid targets. The decay heat of each target was calculated. It was low enough low compared to that of ISIS for the target without tantalum but was about four times as high as that of ISIS when the thickness of the tantalum cladding was 0.5 mm. Heat removal methods of the decay heat with tantalum were examined. It was shown that a special cooling system was required for the target exchange when tantalum was used for the target. It was concluded that the tungsten rod target sheathed with stainless steel or Zircaloy was the most reliable from the safety considerations and had similar neutronics performance to that of mercury.

Takenaka, N.; Nio, D.; Kiyanagi, Y.; Mishima, K.; Kawai, M.; Furusaka, M.

2005-08-01

374

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

NASA Astrophysics Data System (ADS)

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

LiCausi, Nicholas

375

Computer simulations for rf design of a Spallation Neutron Source external antenna H ion source  

SciTech Connect

Electromagnetic modeling of the multicusp external antenna H ion source for the Spallation Neutron Source SNS has been performed in order to optimize high-power performance. During development of the SNS external antenna ion source, antenna failures due to high voltage and multicusp magnet holder rf heating concerns under stressful operating conditions led to rf characteristics analysis. In rf simulations, the plasma was modeled as an equivalent lossy metal by defining conductivity as . Insulation designs along with material selections such as ferrite and Teflon could be included in the computer simulations to compare antenna gap potentials, surface power dissipations, and input impedance at the operating frequencies, 2 and 13.56 MHz. Further modeling and design improvements are outlined in the conclusion.

Lee, Sung-Woo [ORNL] [ORNL; Goulding, Richard Howell [ORNL] [ORNL; Kang, Yoon W [ORNL] [ORNL; Shin, Ki [ORNL] [ORNL; Welton, Robert F [ORNL] [ORNL

2010-01-01

376

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

Microsoft Academic Search

The Large Core Code Evaluation Working Group Benchmark Problem Four was specified to provide a stringent test of the current methods which are used in the nuclear design and analyses process. The benchmark specifications provided a base for performing detailed burnup calculations over the first two irradiation cycles for a large heterogeneous fast reactor. Particular emphasis was placed on the

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

1984-01-01

377

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

NASA Astrophysics Data System (ADS)

Advances in microelectronics result in sub-micrometer electronic technologies as predicted by Moore's Law, 1965, which states the number of transistors in a given space would double every two years. The most available memory architectures today have submicrometer transistor dimensions. The International Technology Roadmap for Semiconductors (ITRS), a continuation of Moore's Law, predicts that Dynamic Random Access Memory (DRAM) will have an average half pitch size of 50 nm and Microprocessor Units (MPU) will have an average gate length of 30 nm over the period of 2008-2012. Decreases in the dimensions satisfy the producer and consumer requirements of low power consumption, more data storage for a given space, faster clock speed, and portability of integrated circuits (IC), particularly memories. On the other hand, these properties also lead to a higher susceptibility of IC designs to temperature, magnetic interference, power supply, and environmental noise, and radiation. Radiation can directly or indirectly affect device operation. When a single energetic particle strikes a sensitive node in the micro-electronic device, it can cause a permanent or transient malfunction in the device. This behavior is called a Single Event Effect (SEE). SEEs are mostly transient errors that generate an electric pulse which alters the state of a logic node in the memory device without having a permanent effect on the functionality of the device. This is called a Single Event Upset (SEU) or Soft Error . Contrary to SEU, Single Event Latchup (SEL), Single Event Gate Rapture (SEGR), or Single Event Burnout (SEB) they have permanent effects on the device operation and a system reset or recovery is needed to return to proper operations. The rate at which a device or system encounters soft errors is defined as Soft Error Rate (SER). The semiconductor industry has been struggling with SEEs and is taking necessary measures in order to continue to improve system designs in nano-scale technologies. Prevention of SEEs has been studied and applied in the semiconductor industry by including radiation protection precautions in the system architecture or by using corrective algorithms in the system operation. Decreasing 10B content (20%of natural boron) in the natural boron of Borophosphosilicate glass (BPSG) layers that are conventionally used in the fabrication of semiconductor devices was one of the major radiation protection approaches for the system architecture. Neutron interaction in the BPSG layer was the origin of the SEEs because of the 10B (n,alpha) 7Li reaction products. Both of the particles produced have the capability of ionization in the silicon substrate region, whose thickness is comparable to the ranges of these particles. Using the soft error phenomenon in exactly the opposite manner of the semiconductor industry can provide a new neutron detection system based on the SERs in the semiconductor memories. By investigating the soft error mechanisms in the available semiconductor memories and enhancing the soft error occurrences in these devices, one can convert all memory using intelligent systems into portable, power efficient, directiondependent neutron detectors. The Neutron Intercepting Silicon Chip (NISC) project aims to achieve this goal by introducing 10B-enriched BPSG layers to the semiconductor memory architectures. This research addresses the development of a simulation tool, the NISC Soft Error Analysis Tool (NISCSAT), for soft error modeling and analysis in the semiconductor memories to provide basic design considerations for the NISC. NISCSAT performs particle transport and calculates the soft error probabilities, or SER, depending on energy depositions of the particles in a given memory node model of the NISC. Soft error measurements were performed with commercially available, off-the-shelf semiconductor memories and microprocessors to observe soft error variations with the neutron flux and memory supply voltage. Measurement results show that soft errors in the memories increase proportionally with the neutron flux, wherea

Celik, Cihangir

378

MCNP5 and GEANT4 comparisons for preliminary Fast Neutron Pencil Beam design at the University of Utah TRIGA system  

NASA Astrophysics Data System (ADS)

The main objective of this thesis is twofold. The starting objective was to develop a model for meaningful benchmarking of different versions of GEANT4 against an experimental set-up and MCNP5 pertaining to photon transport and interactions. The following objective was to develop a preliminary design of a Fast Neutron Pencil Beam (FNPB) Facility to be applicable for the University of Utah research reactor (UUTR) using MCNP5 and GEANT4. The three various GEANT4 code versions, GEANT4.9.4, GEANT4.9.3, and GEANT4.9.2, were compared to MCNP5 and the experimental measurements of gamma attenuation in air. The average gamma dose rate was measured in the laboratory experiment at various distances from a shielded cesium source using a Ludlum model 19 portable NaI detector. As it was expected, the gamma dose rate decreased with distance. All three GEANT4 code versions agreed well with both the experimental data and the MCNP5 simulation. Additionally, a simple GEANT4 and MCNP5 model was developed to compare the code agreements for neutron interactions in various materials. Preliminary FNPB design was developed using MCNP5; a semi-accurate model was developed using GEANT4 (because GEANT4 does not support the reactor physics modeling, the reactor was represented as a surface neutron source, thus a semi-accurate model). Based on the MCNP5 model, the fast neutron flux in a sample holder of the FNPB is obtained to be 6.52107 n/cm2s, which is one order of magnitude lower than gigantic fast neutron pencil beam facilities existing elsewhere. The MCNP5 model-based neutron spectrum indicates that the maximum expected fast neutron flux is at a neutron energy of ~1 MeV. In addition, the MCNP5 model provided information on gamma flux to be expected in this preliminary FNPB design; specifically, in the sample holder, the gamma flux is to be expected to be around 108 ?/cm 2s, delivering a gamma dose of 4.54103 rem/hr. This value is one to two orders of magnitudes below the gamma exposure as exists in the currently used fast neutron irradiation facility at the UUTR. The GEANT4.9.4 semi-accurate model of the FNPB design provided higher values for neutron and gamma fluxes, indicating the importance of transfering the data from MCNP5 rather than using the GEANT4 default neutron spectra.

Adjei, Christian Amevi

379

The fuzzy clearing approach for a niching genetic algorithm applied to a nuclear reactor core design optimization problem  

Microsoft Academic Search

This article extends previous efforts on genetic algorithms (GAs) applied to a core design optimization problem. We introduce the application of a new Niching Genetic Algorithm (NGA) to this problem and compare its performance to these previous works. The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average

Wagner F. Sacco; Marcelo D. Machado; Cludio M. N. A. Pereira; Roberto Schirru

2004-01-01

380

Advanced Neutron Source (ANS) Project progress report  

SciTech Connect

This report discusses the following topics on the advanced neutron source: quality assurance (QA) program; reactor core development; fuel element specification; corrosion loop tests and analyses; thermal-hydraulic loop tests; reactor control concepts; critical and subcritical experiments; material data, structural tests, and analysis; cold source development; beam tube, guide, and instrument development; hot source development; neutron transport and shielding; I C research and development; facility concepts; design; and safety.

McBee, M.R.; Chance, C.M. (eds.) (Oak Ridge National Lab., TN (USA)); Selby, D.L.; Harrington, R.M.; Peretz, F.J. (Oak Ridge National Lab., TN (USA))

1990-04-01

381

Evaluating secondary neutron doses of a refined shielded design for a medical cyclotron using the TLD approach  

NASA Astrophysics Data System (ADS)

An increasing number of cyclotrons at medical centers in Taiwan have been installed to generate radiopharmaceutical products. An operating cyclotron generates immense amounts of secondary neutrons from reactions such the 18O(p, n)18F, used in the production of FDG. This intense radiation can be hazardous to public health, particularly to medical personnel. To increase the yield of 18F-FDG from 4200 GBq in 2005 to 48,600 GBq in 2011, Chung Shan Medical University Hospital (CSMUH) has prolonged irradiation time without changing the target or target current to meet requirements regarding the production 18F. The CSMUH has redesigned the CTI Radioisotope Delivery System shield. The lack of data for a possible secondary neutron doses has increased due to newly designed cyclotron rooms. This work aims to evaluate secondary neutron doses at a CTI cyclotron center using a thermoluminescent dosimeter (TLD-600). Two-dimensional neutron doses were mapped and indicated that neutron doses were high as neutrons leaked through self-shielded blocks and through the L-shaped concrete shield in vault rooms. These neutron doses varied markedly among locations close to the H218O target. The Monte Carlo simulation and minimum detectable dose are also discussed and demonstrated the reliability of using the TLD-600 approach. Findings can be adopted by medical centers to identify radioactive hot spots and develop radiation protection.

Lin, Jye-Bin; Tseng, Hsien-Chun; Liu, Wen-Shan; Lin, Ding-Bang; Hsieh, Teng-San; Chen, Chien-Yi

2013-11-01

382

Design and Performance Improvements of the Prototype Open Core Flywheel Energy Storage System  

NASA Technical Reports Server (NTRS)

A prototype magnetically suspended composite flywheel energy storage (FES) system is operating at the University of Maryland. This system, designed for spacecraft applications, incorporates recent advances in the technologies of composite materials, magnetic suspension, and permanent magnet brushless motor/generator. The current system is referred to as an Open Core Composite Flywheel (OCCF) energy storage system. This paper will present design improvements for enhanced and robust performance. Initially, when the OCCF prototype was spun above its first critical frequency of 4,500 RPM, the rotor movement would exceed the space available in the magnetic suspension gap and touchdown on the backup mechanical bearings would occur. On some occasions it was observed that, after touchdown, the rotor was unable to re-suspend as the speed decreased. Additionally, it was observed that the rotor would exhibit unstable oscillations when the control system was initially turned on. Our analysis suggested that the following problems existed: (1) The linear operating range of the magnetic bearings was limited due to electrical and magnetic saturation; (2) The inductance of the magnetic bearings was affecting the transient response of the system; (3) The flywheel was confined to a small movement because mechanical components could not be held to a tight tolerance; and (4) The location of the touchdown bearing magnifies the motion at the pole faces of the magnetic bearings when the linear range is crucial. In order to correct these problems an improved design of the flywheel energy storage system was undertaken. The magnetic bearings were re-designed to achieve a large linear operating range and to withstand load disturbances of at least 1 g. The external position transducers were replaced by a unique design which were resistant to magnetic field noise and allowed cancellation of the radial growth of the flywheel at high speeds. A central rod was utilized to ensure the concentricity of the magnetic bearings, the motor/generator, and the mechanical touchdown bearings. In addition, the mechanical touchdown bearings were placed at two ends of the magnetic bearing stack to restrict the motion at pole faces. A composite flywheel was made using a multi-ring interference assembled design for a high specific energy density. To achieve a higher speed and better efficiency, a permanent magnet DC brushless motor was specially designed and fabricated. A vacuum enclosure was constructed to eliminate windage losses for testing at high speeds. With the new improvements the OCCF system was tested to 20,000 RPM with a total stored energy of 15.9 WH and an angular momentum of 54.8 N-m-s (40.4 lb-ft-s). Motor current limitation, caused by power loss in the magnetic bearings, was identified as causing the limit in upper operating speed.

Pang, D.; Anand, D. K. (Editor); Kirk, J. A. (Editor)

1996-01-01

383

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

PubMed

Neutron supermirrors and supermirror polarizers are thin film multilayer based devices which are used for reflecting and polarizing neutrons in various neutron based experiments. In the present communication, the in-house development of a 9 m long in-line dc sputtering system has been described which is suitable for deposition of neutron supermirrors on large size (1500 mm 150 mm) substrates and in large numbers. The optimisation process of deposition of Co and Ti thin film, Co/Ti periodic multilayers, and a-periodic supermirrors have also been described. The system has been used to deposit thin film multilayer supermirror polarizers which show high reflectivity up to a reasonably large critical wavevector transfer of ?0.06 (-1) (corresponding to m = 2.5, i.e., 2.5 times critical wavevector transfer of natural Ni). The computer code for designing these supermirrors has also been developed in-house. PMID:25554268

Biswas, A; Sampathkumar, R; Kumar, Ajaya; Bhattacharyya, D; Sahoo, N K; Lagoo, K D; Veerapur, R D; Padmanabhan, M; Puri, R K; Bhattacharya, Debarati; Singh, Surendra; Basu, S

2014-12-01

384

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

NASA Astrophysics Data System (ADS)

Neutron supermirrors and supermirror polarizers are thin film multilayer based devices which are used for reflecting and polarizing neutrons in various neutron based experiments. In the present communication, the in-house development of a 9 m long in-line dc sputtering system has been described which is suitable for deposition of neutron supermirrors on large size (1500 mm 150 mm) substrates and in large numbers. The optimisation process of deposition of Co and Ti thin film, Co/Ti periodic multilayers, and a-periodic supermirrors have also been described. The system has been used to deposit thin film multilayer supermirror polarizers which show high reflectivity up to a reasonably large critical wavevector transfer of 0.06 -1 (corresponding to m = 2.5, i.e., 2.5 times critical wavevector transfer of natural Ni). The computer code for designing these supermirrors has also been developed in-house.

Biswas, A.; Sampathkumar, R.; Kumar, Ajaya; Bhattacharyya, D.; Sahoo, N. K.; Lagoo, K. D.; Veerapur, R. D.; Padmanabhan, M.; Puri, R. K.; Bhattacharya, Debarati; Singh, Surendra; Basu, S.

2014-12-01

385

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

SciTech Connect

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

Mueller, W.F.; Croft, S.; McElroy, R.D.; Venkataraman, R.; Zhu, H. [Canberra Industries, Inc., 800 Research Parkway, Meriden, Connecticut 06450 (United States)

2006-07-01

386

Lead-Cooled Fast Reactor (LFR) Design: Safety, Neutronics, Thermal Hydraulics, Structural Mechanics, Fuel, Core, and Plant Design  

SciTech Connect

The idea of developing fast spectrum reactors with molten lead (or lead alloy) as a coolant is not a new one. Although initially considered in the West in the 1950s, such technology was not pursued to completion because of anticipated difficulties associated with the corrosive nature of these coolant materials. However, in the Soviet Union, such technology was actively pursued during the same time frame (1950s through the 1980s) for the specialized role of submarine propulsion. More recently, there has been a renewal of interest in the West for such technology, both for critical systems as well as for Accelerator Driven Subcritical (ADS) systems. Meanwhile, interest in the former Soviet Union, primarily Russia, has remained strong and has expanded well beyond the original limited mission of submarine propulsion. This section reviews the past and current status of LFR development.

Smith, C

2010-02-22

387

Neutron Stars  

NASA Technical Reports Server (NTRS)

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

Cottam, J.

2007-01-01

388

The influence of various core designs on stress distribution in the veneered zirconia crown: a finite element analysis study  

PubMed Central

PURPOSE The purpose of this study was to evaluate various core designs on stress distribution within zirconia crowns. MATERIALS AND METHODS Three-dimensional finite element models, representing mandibular molars, comprising a prepared tooth, cement layer, zirconia core, and veneer porcelain were designed by computer software. The shoulder (1 mm in width) variations in core were incremental increases of 1 mm, 2 mm and 3 mm in proximal and lingual height, and buccal height respectively. To simulate masticatory force, loads of 280 N were applied from three directions (vertical, at a 45 angle, and horizontal). To simulate maximum bite force, a load of 700 N was applied vertically to the crowns. Maximum principal stress (MPS) was determined for each model, loading condition, and position. RESULTS In the maximum bite force simulation test, the MPSs on all crowns observed around the shoulder region and loading points. The compressive stresses were located in the shoulder region of the veneer-zirconia interface and at the occlusal region. In the test simulating masticatory force, the MPS was concentrated around the loading points, and the compressive stresses were located at the 3 mm height lingual shoulder region, when the load was applied horizontally. MPS increased in the shoulder region as the shoulder height increased. CONCLUSION This study suggested that reinforced shoulder play an essential role in the success of the zirconia restoration, and veneer fracture due to occlusal loading can be prevented by proper core design, such as shoulder. PMID:23755346

Ha, Seung-Ryong; Kim, Sung-Hun; Yoo, Seung-Hyun; Jeong, Se-Chul; Lee, Jai-Bong; Yeo, In-Sung

2013-01-01

389

Design study of an air pump and integral lift engine ALF-504 using the Lycoming 502 core  

NASA Technical Reports Server (NTRS)

Design studies were conducted for an integral lift fan engine utilizing the Lycoming 502 fan core with the final MQT power turbine. The fan is designed for a 12.5 bypass ratio and 1.25:1 pressure ratio, and provides supercharging for the core. Maximum sea level static thrust is 8370 pounds with a specific fuel consumption of 0.302 lb/hr-lb. The dry engine weight without starter is 1419 pounds including full-length duct and sound-attenuating rings. The engine envelope including duct treatment but not localized accessory protrusion is 53.25 inches in diameter and 59.2 inches long from exhaust nozzle exit to fan inlet flange. Detailed analyses include fan aerodynamics, fan and reduction gear mechanical design, fan dynamic analysis, engine noise analysis, engine performance, and weight analysis.

Rauch, D.

1972-01-01

390

The Practical Turn in Teacher Education: Designing a Preparation Sequence for Core Practice Frames  

ERIC Educational Resources Information Center

Amid calls for more practice-based teacher education, this article presents a concrete illustration of a practice-based bridging strategy for preparing high school biology teachers to enact open-inquiry labs. Open-inquiry labs were considered a core practice frame that served as a context for identifying core practices and for giving coherence to

Janssen, Fred; Westbroek, Hanna; Doyle, Walter

2014-01-01

391

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

SciTech Connect

This study investigates the neutronics design aspects of a hybrid fusion-fission energy system called the Laser Fusion-Fission Hybrid (LFFH). A LFFH combines current Laser Inertial Confinement fusion technology with that of advanced fission reactor technology to produce a system that eliminates many of the negative aspects of pure fusion or pure fission systems. When examining the LFFH energy mission, a significant portion of the United States and world energy production could be supplied by LFFH plants. The LFFH engine described utilizes a central fusion chamber surrounded by multiple layers of multiplying and moderating media. These layers, or blankets, include coolant plenums, a beryllium (Be) multiplier layer, a fertile fission blanket and a graphite-pebble reflector. Each layer is separated by perforated oxide dispersion strengthened (ODS) ferritic steel walls. The central fusion chamber is surrounded by an ODS ferritic steel first wall. The first wall is coated with 250-500 {micro}m of tungsten to mitigate x-ray damage. The first wall is cooled by Li{sub 17}Pb{sub 83} eutectic, chosen for its neutron multiplication and good heat transfer properties. The {sub 17}Pb{sub 83} flows in a jacket around the first wall to an extraction plenum. The main coolant injection plenum is immediately behind the Li{sub 17}Pb{sub 83}, separated from the Li{sub 17}Pb{sub 83} by a solid ODS wall. This main system coolant is the molten salt flibe (2LiF-BeF{sub 2}), chosen for beneficial neutronics and heat transfer properties. The use of flibe enables both fusion fuel production (tritium) and neutron moderation and multiplication for the fission blanket. A Be pebble (1 cm diameter) multiplier layer surrounds the coolant injection plenum and the coolant flows radially through perforated walls across the bed. Outside the Be layer, a fission fuel layer comprised of depleted uranium contained in Tristructural-isotropic (TRISO) fuel particles having a packing fraction of 20% in 2 cm diameter fuel pebbles. The fission blanket is cooled by the same radial flibe flow that travels through perforated ODS walls to the reflector blanket. This reflector blanket is 75 cm thick comprised of 2 cm diameter graphite pebbles cooled by flibe. The flibe extraction plenum surrounds the reflector bed. Detailed neutronics designs studies are performed to arrive at the described design. The LFFH engine thermal power is controlled using a technique of adjusting the {sup 6}Li/{sup 7}Li enrichment in the primary and secondary coolants. The enrichment adjusts system thermal power in the design by increasing tritium production while reducing fission. To perform the simulations and design of the LFFH engine, a new software program named LFFH Nuclear Control (LNC) was developed in C++ to extend the functionality of existing neutron transport and depletion software programs. Neutron transport calculations are performed with MCNP5. Depletion calculations are performed using Monteburns 2.0, which utilizes ORIGEN 2.0 and MCNP5 to perform a burnup calculation. LNC supports many design parameters and is capable of performing a full 3D system simulation from initial startup to full burnup. It is able to iteratively search for coolant {sup 6}Li enrichments and resulting material compositions that meet user defined performance criteria. LNC is utilized throughout this study for time dependent simulation of the LFFH engine. Two additional methods were developed to improve the computation efficiency of LNC calculations. These methods, termed adaptive time stepping and adaptive mesh refinement were incorporated into a separate stand alone C++ library name the Adaptive Burnup Library (ABL). The ABL allows for other client codes to call and utilize its functionality. Adaptive time stepping is useful for automatically maximizing the size of the depletion time step while maintaining a desired level of accuracy. Adaptive meshing allows for analysis of fixed fuel configurations that would normally require a computationally burdensome number of depletion zones. Alternatively, Adaptive M

Kramer, K

2010-04-08

392

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

PubMed

A novel capability was designed, implemented, and tested for in situ neutron diffraction measurements during loading at cryogenic temperatures on the spectrometer for materials research at temperature and stress at Los Alamos National Laboratory. This capability allowed for the application of dynamic compressive forces of up to 250 kN on standard samples controlled at temperatures between 300 and 90 K. The approach comprised of cooling thermally isolated compression platens that in turn conductively cooled the sample in an aluminum vacuum chamber which was nominally transparent to the incident and diffracted neutrons. The cooling/heat rate and final temperature were controlled by regulating the flow of liquid nitrogen in channels inside the platens that were connected through bellows to the mechanical actuator of the load frame and by heaters placed on the platens. Various performance parameters of this system are reported here. The system was used to investigate deformation in Ni-Ti-Fe shape memory alloys at cryogenic temperatures and preliminary results are presented. PMID:20590248

Woodruff, T R; Krishnan, V B; Clausen, B; Sisneros, T; Livescu, V; Brown, D W; Bourke, M A M; Vaidyanathan, R

2010-06-01

393

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

SciTech Connect

A novel capability was designed, implemented, and tested for in situ neutron diffraction measurements during loading at cryogenic temperatures on the spectrometer for materials research at temperature and stress at Los Alamos National Laboratory. This capability allowed for the application of dynamic compressive forces of up to 250 kN on standard samples controlled at temperatures between 300 and 90 K. The approach comprised of cooling thermally isolated compression platens that in turn conductively cooled the sample in an aluminum vacuum chamber which was nominally transparent to the incident and diffracted neutrons. The cooling/heat rate and final temperature were controlled by regulating the flow of liquid nitrogen in channels inside the platens that were connected through bellows to the mechanical actuator of the load frame and by heaters placed on the platens. Various performance parameters of this system are reported here. The system was used to investigate deformation in Ni-Ti-Fe shape memory alloys at cryogenic temperatures and preliminary results are presented.

Woodruff, T. R.; Krishnan, V. B.; Vaidyanathan, R. [Department of Mechanical, Materials, and Aerospace Engineering, Advanced Materials Processing and Analysis Center (AMPAC), University of Central Florida, Orlando, Florida 32816 (United States); Clausen, B.; Sisneros, T.; Livescu, V.; Brown, D. W.; Bourke, M. A. M. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

2010-06-15

394

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

NASA Astrophysics Data System (ADS)

A novel capability was designed, implemented, and tested for in situ neutron diffraction measurements during loading at cryogenic temperatures on the spectrometer for materials research at temperature and stress at Los Alamos National Laboratory. This capability allowed for the application of dynamic compressive forces of up to 250 kN on standard samples controlled at temperatures between 300 and 90 K. The approach comprised of cooling thermally isolated compression platens that in turn conductively cooled the sample in an aluminum vacuum chamber which was nominally transparent to the incident and diffracted neutrons. The cooling/heat rate and final temperature were controlled by regulating the flow of liquid nitrogen in channels inside the platens that were connected through bellows to the mechanical actuator of the load frame and by heaters placed on the platens. Various performance parameters of this system are reported here. The system was used to investigate deformation in Ni-Ti-Fe shape memory alloys at cryogenic temperatures and preliminary results are presented.

Woodruff, T. R.; Krishnan, V. B.; Clausen, B.; Sisneros, T.; Livescu, V.; Brown, D. W.; Bourke, M. A. M.; Vaidyanathan, R.

2010-06-01

395

Space nuclear-power reactor design based on combined neutronic and thermal-fluid analyses  

SciTech Connect

The design and performance analysis of a space nuclear-power system requires sophisticated analytical capabilities such as those developed during the nuclear rocket propulsion (Rover) program. In particular, optimizing the size of a space nuclear reactor for a given power level requires satisfying the conflicting requirements of nuclear criticality and heat removal. The optimization involves the determination of the coolant void (volume) fraction for which the reactor diameter is a minimum and temperature and structural limits are satisfied. A minimum exists because the critical diameter increases with increasing void fraction, whereas the reactor diameter needed to remove a specified power decreases with void fraction. The purpose of this presentation is to describe and demonstrate our analytical capability for the determination of minimum reactor size. The analysis is based on combining neutronic criticality calculations with OPTION-code thermal-fluid calculations.

Koenig, D.R.; Gido, R.G.; Brandon, D.I.

1985-01-01

396

Strategies for nanoplasmonic core-satellite biomolecular sensors: Theory-based Design  

PubMed Central

We present a systematic theoretical study of core-satellite gold nanoparticle assemblies using the Generalized Multiparticle Mie formalism. We consider the importance of satellite number, satellite radius, the core radius, and the satellite distance, and we present approaches to optimize spectral shift due to satellite attachment or release. This provides clear strategies for improving the sensitivity and signal-to-noise ratio for molecular detection, enabling simple colorimetric assays. We quantify the performance of these strategies by introducing a figure of merit. In addition, we provide an improved understanding of the nanoplasmonic interactions that govern the optical response of core-satellite nanoassemblies. PMID:19997582

Ross, Benjamin M.; Waldeisen, John R.; Wang, Tim; Lee, Luke P.

2009-01-01

397

Strategies for nanoplasmonic core-satellite biomolecular sensors: Theory-based Design  

NASA Astrophysics Data System (ADS)

We present a systematic theoretical study of core-satellite gold nanoparticle assemblies using the Generalized Multiparticle Mie formalism. We consider the importance of satellite number, satellite radius, the core radius, and the satellite distance, and we present approaches to optimize spectral shift due to satellite attachment or release. This provides clear strategies for improving the sensitivity and signal-to-noise ratio for molecular detection, enabling simple colorimetric assays. We quantify the performance of these strategies by introducing a figure of merit. In addition, we provide an improved understanding of the nanoplasmonic interactions that govern the optical response of core-satellite nanoassemblies.

Ross, Benjamin M.; Waldeisen, John R.; Wang, Tim; Lee, Luke P.

2009-11-01

398

Design of single-mode large-mode area bandgap fibre with microstructured-core  

NASA Astrophysics Data System (ADS)

Tailoring the modal characteristics of all-solid photonic bandgap fibre by the inclusion of low-index rods in the fibre core is investigated. By lowering the core index of the fibre, the high-order modes in the fibre can be shifted to the border of the bandgap, as a result, single-mode operation in large-mode area bandgap fibre can be realized without bending the fibre. In addition, large-mode area and single-mode operation in the all-solid microstructured-core bandgap fibre can be achieved by bending the fibre at a wide bending radius range.

Chen, Ming-Yang; Gong, Tian-Yi; Gao, Yong-Feng; Zhou, Jun

2014-11-01

399

Design of a horizontal neutron reflectometer for the European Spallation Source  

NASA Astrophysics Data System (ADS)

A design study of a horizontal neutron reflectometer adapted to the general baseline of the long pulse European Spallation Source (ESS) is presented. The instrument layout comprises advanced solutions for the neutron guide, high-resolution pulse shaping and beam bending onto a sample surface being thoroughly adjusted to the properties of the ESS. The length of this instrument is roughly 55 m, enabling ??/? resolutions from 0.5% to 10%. The incident beam is focused in horizontal plane to boost measurements of sample sizes of 11 cm2 and smaller with potential beam deflection in both downward and upward directions. The primary range of neutron wavelengths utilized by the instrument is 2-7.1 . If the wavelength range needs to be extended, then this is possible by utilizing only every second (third, fourth) pulse by suppressing all other pulses by the chopper system and thus increase the longest usable wavelength to 12.2 (17.3, 22.4) . Angles of incidence can be set between 0 and 9 with a total accessible q-range from 410-3 -1 up to 1 -1, while the ??/? resolution can be freely set. The instrument operates in both ?/? (free liquid surfaces) and ?/2? (solid-liquid, air-solid interfaces) geometries. The experimental setup will in particular enable direct studies on ultrathin films (d ?10 ) and buried monolayers to multilayered structures of up to 3000 total thickness. The horizontal reflectometer will further foster investigations of hierarchical systems from nanometer to micrometer length scale (the latter by off-specular scattering), as well as their kinetics and dynamical properties, in particular under load (shear, pressure, external fields). Polarization and polarization analysis as well as the GISANS option are designed as potential modules to be implemented in the generic instrument layout. The instrument is highly flexible and offers a variety of different measurement modes. With respect to its mechanical components the instrument is exclusively based on current technology. Risks of failure for the chosen setup are minimum.

Nekrassov, D.; Trapp, M.; Lieutenant, K.; Moulin, J.-F.; Strobl, M.; Steitz, R.

2014-08-01

400

Thermal Hydraulics Design and Analysis Methodology for a Solid-Core Nuclear Thermal Rocket Engine Thrust Chamber  

NASA Technical Reports Server (NTRS)

Nuclear thermal propulsion is a leading candidate for in-space propulsion for human Mars missions. This chapter describes a thermal hydraulics design and analysis methodology developed at the NASA Marshall Space Flight Center, in support of the nuclear thermal propulsion development effort. The objective of this campaign is to bridge the design methods in the Rover/NERVA era, with a modern computational fluid dynamics and heat transfer methodology, to predict thermal, fluid, and hydrogen environments of a hypothetical solid-core, nuclear thermal engine