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1

Fissioning Plasma Core Reactor  

NASA Technical Reports Server (NTRS)

Institute for Scientific Research, Inc. (ISR) research program consist of: 1.Study core physics by adapting existing codes: MCNP4C - Monte Carlo code; COMBINE/VENTURE - diffusion theory; SCALE4 - Monte Carlo, with many utility codes. 2. Determine feasibility and study major design parameters: fuel selection, temperature and reflector sizing. 3. Study reactor kinetics: develop QCALC1 to model point kinetics; study dynamic behavior of the power release.

Albright, Dennis; Butler, Carey; West, Nicole; Cole, John W. (Technical Monitor)

2002-01-01

2

The IRIS General Plant Arrangement  

SciTech Connect

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. Bechtel, with Westinghouse consultation, has performed a layout study of the IRIS plant and this paper will discuss the results of this design effort. (authors)

Robertson, J.; Love, J.; Morgan, R. [Bechtel Power Company (United States); Conway, L.E. [Westinghouse Electric Company (United States)

2002-07-01

3

Gamma dose from activation of internal shields in IRIS reactor.  

PubMed

The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield. PMID:16381688

Agosteo, Stefano; Cammi, Antonio; Garlati, Luisella; Lombardi, Carlo; Padovani, Enrico

2005-01-01

4

Nuclear Reactor Core Clamping System.  

National Technical Information Service (NTIS)

The clamping system is for the core of a nuclear reactor having a neutron reflector control system. The clamping system is situated between the core and the reflector system and is of such a size to permit closer spacing of the reflector system to the cor...

R. W. Guenther

1965-01-01

5

Reactor core isolation cooling system  

DOEpatents

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

Cooke, F.E.

1992-12-08

6

Reactor core isolation cooling system  

DOEpatents

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

Cooke, Franklin E. (San Jose, CA)

1992-01-01

7

Lateral restraint assembly for reactor core  

DOEpatents

A restraint assembly for use in restraining lateral movement of a reactor core relative to a reactor vessel wherein a plurality of restraint assemblies are interposed between the reactor core and the reactor vessel in circumferentially spaced relation about the core. Each lateral restraint assembly includes a face plate urged against the outer periphery of the core by a plurality of compression springs which enable radial preloading of outer reflector blocks about the core and resist low-level lateral motion of the core. A fixed radial key member cooperates with each face plate in a manner enabling vertical movement of the face plate relative to the key member but restraining movement of the face plate transverse to the key member in a plane transverse to the center axis of the core. In this manner, the key members which have their axes transverse to or subtending acute angles with the direction of a high energy force tending to move the core laterally relative to the reactor vessel restrain such lateral movement.

Gorholt, Wilhelm (San Diego, CA); Luci, Raymond K. (Del Mar, CA)

1986-01-01

8

Kinetics of Coupled-Core Reactors.  

National Technical Information Service (NTIS)

A technique for obtaining the frequency response of a nuclear reactor using a pulsed neutron source has been investigated. Pulsed neutron measurements have been performed in the UTR-10 coupled-core Argonaut type reactor with neutron detectors placed in th...

R. A. Danofsky

1971-01-01

9

Gas core reactors for coal gasification  

NASA Technical Reports Server (NTRS)

The concept of using a gas core reactor to produce hydrogen directly from coal and water is presented. It is shown that the chemical equilibrium of the process is strongly in favor of the production of H2 and CO in the reactor cavity, indicating a 98% conversion of water and coal at only 1500 K. At lower temperatures in the moderator-reflector cooling channels the equilibrium strongly favors the conversion of CO and additional H2O to CO2 and H2. Furthermore, it is shown the H2 obtained per pound of carbon has 23% greater heating value than the carbon so that some nuclear energy is also fixed. Finally, a gas core reactor plant floating in the ocean is conceptualized which produces H2, fresh water and sea salts from coal.

Weinstein, H.

1976-01-01

10

Facility modernization Annular Core Research Reactor  

SciTech Connect

The Annular Core Research Reactor (ACRR) has undergone numerous modifications since its conception in response to program needs. The original reactor fuel, which was special U-ZrH TRIGA fuel designed primarily for pulsing, has been replaced with a higher pulsing capacity BeO fuel. Other advanced operating modes which use this increased capability, in addition to the pulse and steady state, have been incorporated to tailor power histories and fluences to the experiments. Various experimental facilities have been developed that range from a radiography facility to a 50 cm diameter External Fuel Ring Cavity (FREC) using 180 of the original ZrH fuel elements. Currently a digital reactor console is being produced with GA, which will give enhanced monitoring capabilities of the reactor parameters while leaving the safety-related shutdown functions with analog technology. (author)

Morris, F.M.; Luera, T.F.; McCrory, F.M.; Nelson, D.A.; Trowbridge, F.R.; Wold, S.A. [Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States)

1990-07-01

11

Gas-core reactor power transient analysis.  

NASA Technical Reports Server (NTRS)

The nuclear fuel in the gas-core reactor concept is a ball of uranium plasma radiating thermal photons. The photons are met by an inflowing hydrogen stream, which is seeded with submicron size, depleted uranium particles. A 'wall-burnout' condition exists if the thermal photons can reach the cavity liner because of insufficient absorption by the hydrogen. An analysis was conducted in order to determine the time for which the maximum steady state reactor power could be exceeded without damage to the cavity liner due to burnout. Wall-burnout time as a function of the power increase above the initial steady state condition is shown in a graph.

Kascak, A. F.

1972-01-01

12

Emergency core cooling system for a nuclear reactor  

Microsoft Academic Search

An emergency core cooling system for a nuclear reactor which preferably is supplemental to the main emergency core cooling system incorporated in the reactor at the time of construction is described. Under circumstances of a rupture in the reactor primary coolant piping and consequent drop in reactor coolant pressure, emergency supplemental coolant is supplied from tanks or accumulators through check

W. E. Desmarchais; L. R. Katz; B. L. Silverblatt

1977-01-01

13

Neutronics of a Mixed-Flow Gas-Core Reactor.  

National Technical Information Service (NTIS)

The study was made to investigate the neutronic feasibility of a mixed-flow gas-core reactor. Three reactor concepts were studied: four- and seven-cell radial reactors and a seven-cell scallop reactor. The reactors were fueled with UF sub 6 (either U-233 ...

P. D. Soran G. E. Hansen

1977-01-01

14

Multilevel transport solution of LWR reactor cores  

SciTech Connect

This work presents a multilevel approach for the solution of the transport equation in typical LWR assemblies and core configurations. It is based on the second-order, even-parity formulation of the transport equation, which is solved within the framework provided by the finite element-spherical harmonics code EVENT. The performance of the new solver has been compared with that of the standard conjugate gradient solver for diffusion and transport problems on structured and unstruc-tured grids. Numerical results demonstrate the potential of the multilevel scheme for realistic reactor calculations.

Jose Ignacio Marquez Damian; Cassiano R.E. de Oliveira; HyeonKae Park

2008-09-01

15

Conceptual Design of a Modular Island Core Fast Breeder Reactor \\  

Microsoft Academic Search

A metal fueled modular island core sodium cooled fast breeder reactor concept RAPID-M to improve reactor per- formance and proliferation resistance and to accommodate various power requirements has been demonstrated. The essential feature of the RAPID-M concept is that the reactor core consists of integrated fuel assemblies (IFAs) instead of conventional fuel subassemblies. The RAPID concept enables quick and simplified

Mitsuru KAMBE

2002-01-01

16

Study of compact fast reactor core designs  

SciTech Connect

A study is conducted to investigate conceptual liquid-metal reactor (LMR) core concepts, employing some unconventional design features for improved economics and safety. The unconventional design elements are used to supplement the conventional measures, which alone have apparently not led to an attractive LMR design for the 21st century. Better economics are obtained through simplicity and compactness of the core design. For simplicity, internal scattered blankets are omitted. Core compactness is achieved by maximum power flattening, resulting from axial and radial enrichment zones along with axial and radial (BeO) reflectors. To further enhance core compactness, the in-core compactness, the in-core control rods are replaced by reflector controls. For improved safety, the general objective is to reduce both coolant-void and burnup reactivities. However, even with the use of a wide spectrum of unconventional design features, such as burnable poisons, peripheral reflectors, and inner moderating regions, it is not possible to overcome the fact that both coolant-void and burnup reactivities cannot be reduced simultaneously to desirably low levels. The only resolution of this dilemma appears to be minimize coolant-void reactivity and to manage the burnup reactivities. However, even with the use of a wide spectrum of unconventional design features, such as burnable poisons, peripheral reflectors, and inner moderating regions, it is not possible to overcome the fact that both coolant-void and burnup reactivities cannot be reduced simultaneously to desirably low levels. The only resolution of this dilemma appears to be to minimize coolant-void reactivity and to manage the burnup reactivity losses, such that an accidental insertion of significant amounts of reactivity is mechanically not possible. A conceptual design with these characteristics is described.

Hamid, T.; Ott, K.O. (Purdue Univ., West Lafayette, IN (United States))

1993-02-01

17

Iris melanoma.  

PubMed

The iris is the least common site of primary uveal melanoma. The prognosis of iris melanoma is better than that of melanoma of the ciliary body and choroid, but the reason for this difference is unclear. One possible explanation is that iris melanoma is smaller than its posterior segment counterparts at the time of diagnosis. Most iris melanomas are spindle cell types, according to a modified Callender classification system. There is evidence that the proliferation of melanocytes of the anterior iris surface (iris plaque) and diffuse stromal invasion may be risk factors for local recurrence and metastasis, respectively. PMID:18251588

Henderson, Evita; Margo, Curtis E

2008-02-01

18

Development of Reactor Core Surveillance System for PARR,  

National Technical Information Service (NTIS)

A computer based surveillance system has been developed at PARR for an early detection and identification of any core abnormality. The system acquires various plant signals monitoring the reactor core condition and performs the frequency analysis of the f...

S. A. Ansari S. K. Ayazuddin

1984-01-01

19

Reactor pulse repeatability studies at the annular core research reactor  

SciTech Connect

The Annular Core Research Reactor (ACRR) at Sandia National Laboratories is a water-moderated pool-type reactor designed for testing many types of objects in the pulse and steady-state mode of operations. Personnel at Sandia began working to improve the repeatability of pulse operations for experimenters in the facility. The ACRR has a unique UO{sub 2}-BeO fuel that makes the task of producing repeatable pulses difficult with the current operating procedure. The ACRR produces a significant quantity of photoneutrons through the {sup 9}Be({gamma}, n){sup 8}Be reaction in the fuel elements. The photoneutrons are the result of the gammas produced during fission and in fission product decay, so their production is very much dependent on the reactor power history and changes throughout the day/week of experiments in the facility. Because the photoneutrons interfere with the delayed-critical measurements required for accurate pulse reactivity prediction, a new operating procedure was created. The photoneutron effects at delayed critical are minimized when using the modified procedure. In addition, the pulse element removal time is standardized for all pulse operations with the modified procedure, and this produces less variation in reactivity removal times. (authors)

DePriest, K.R. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Trinh, T.Q. [Nuclear Facility Operations, Sandia National Laboratories, Mail Stop 0614, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Luker, S. M. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States)

2011-07-01

20

Granular Dynamics in Pebble Bed Reactor Cores  

NASA Astrophysics Data System (ADS)

This study focused on developing a better understanding of granular dynamics in pebble bed reactor cores through experimental work and computer simulations. The work completed includes analysis of pebble motion data from three scaled experiments based on the annular core of the Pebble Bed Fluoride Salt-Cooled High- Temperature Reactor (PB-FHR). The experiments are accompanied by the development of a new discrete element simulation code, GRECO, which is designed to offer a simple user interface and simplified two-dimensional system that can be used for iterative purposes in the preliminary phases of core design. The results of this study are focused on the PB-FHR, but can easily be extended for gas-cooled reactor designs. Experimental results are presented for three Pebble Recirculation Experiments (PREX). PREX 2 and 3.0 are conventional gravity-dominated granular systems based on the annular PB-FHR core design for a 900 MWth commercial prototype plant and a 16 MWth test reactor, respectively. Detailed results are presented for the pebble velocity field, mixing at the radial zone interfaces, and pebble residence times. A new Monte Carlo algorithm was developed to study the residence time distributions of pebbles in different radial zones. These dry experiments demonstrated the basic viability of radial pebble zoning in cores with diverging geometry before pebbles reach the active core. Results are also presented from PREX 3.1, a scaled facility that uses simulant materials to evaluate the impact of coupled fluid drag forces on the granular dynamics in the PB-FHR core. PREX 3.1 was used to collect first of a kind pebble motion data in a multidimensional porous media flow field. Pebble motion data were collected for a range of axial and cross fluid flow configurations where the drag forces range from half the buoyancy force up to ten times greater than the buoyancy force. Detailed analysis is presented for the pebble velocity field, mixing behavior, and residence time distributions for each fluid flow configuration. The axial flow configurations in PREX 3.1 showed small changes in pebble motion compared to a reference case with no fluid flow and showed similar overall behavior to PREX 3.0. This suggests that dry experiments can be used for core designs with uniform one-dimensional coolant flow early in the design process at greatly reduced cost. Significant differences in pebble residence times were observed in the cross fluid flow configurations, but these were not accompanied by an overall horizontal diffusion bias. Radial zones showed only a small shift in position due to mixing in the diverging region and remained stable in the active core. The results from this study support the overall viability of the annular PB-FHR core by demonstrating consistent granular flow behavior in the presence of complex reflector geometries and multidimensional fluid flow fields. GRECO simulations were performed for each of the experiments in this study in order to develop a preliminary validation basis and to understand for which applications the code can provide useful analysis. Overall, the GRECO simulation results showed excellent agreement with the gravity-dominated PREX experiments. Local velocity errors were found to be generally within 10-15% of the experimental data. Average radial zone interface positions were predicted within two pebble diameters. GRECO simulations over predicted the amount of mixing around the average radial zone interface position and therefore can be treated as a conservative upper bound when used in neutronics analysis. Residence time distributions from the GRECO velocity data based on the Monte Carlo algorithm closely matched those derived from the experiment velocity statistics. GRECO simulation results for PREX 3.1 with coupled drag forces showed larger errors compared to the experimental data, particularly in the cases with cross fluid flow. The large discrepancies suggest that GRECO results in systems with coupled fluid drag forces cannot be used with high confidence at this point and future development

Laufer, Michael Robert

21

An efficient computational technique for light water reactor core dynamics  

Microsoft Academic Search

By combining a modified version of the so-called ''adiabatic'' method for reactor dynamic calculations with a simplified flow redistribution scheme, an efficient method for predicting three-dimensional core behavior has been developed for pressurized water reactor transients. Both the simplified core reactivity and the flow redistribution calculations are shown to yield close approximations of the results obtained by more rigorous approaches.

C. D. Wu; J. Weisman

1988-01-01

22

Relationship of observed flow patterns to gas core reactor criticality  

Microsoft Academic Search

The gas core reactor requires the establishment of stable and unique ; flow patterns. A recent series of room temperature flow tests have studied the ; hydrodynamics, particularly involving gases of differing densities. In an actual ; operating gas core reactor, the central gas of vaporized uranium will have a much ; higher density than the surrounding coolant. Testing was

P. J. Macbeth; J. F. Kunze; V. C. Rogers

1975-01-01

23

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

24

Advanced Instrumentation and Control Methods for Small and Medium Reactors with IRIS Demonstration.  

National Technical Information Service (NTIS)

Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investiga...

B. R. Upadhyaya J. Coble J. M. Doster J. W. Hines K. D. Lewis P. Turinsky R. M. Edwards

2011-01-01

25

REACTOR CORE SURROUNDED BY BERYLLIUM MODERATOR. CAMERA LOOKS DOWN AND ...  

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

REACTOR CORE SURROUNDED BY BERYLLIUM MODERATOR. CAMERA LOOKS DOWN AND TOWARD NORTH INTO LOWER GRID CASTING. HOLES OF VARIOUS SIZES ACCOMMODATE COOLANT WATER AND EXPERIMENTAL POSITIONS. INL NEGATIVE NO. 4197. Unknown Photographer, 2/11/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

26

Advanced Instrumentation and Control Methods for Small and Medium Reactors with IRIS Demonstration  

Microsoft Academic Search

Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investigate, develop, and validate advanced methods for sensing, controlling, monitoring, diagnosis, and prognosis of these reactors, and to demonstrate the methods with application to one

J. Wesley Hines; Belle R. Upadhyaya; J. Michael Doster; Robert M. Edwards; Kenneth D. Lewis; Paul Turinsky; Jamie Coble

2011-01-01

27

Thermal hydraulics model for Sandia's annular core research reactor  

Microsoft Academic Search

A thermal hydraulics model was developed for the Annular Core Research Reactor (ACRR) at Sandia National Laboratories. The coupled mass, momentum and energy equations for the core were solved simultaneously using an explicit forward marching numerical technique. The model predictions of the temperature rise across the central channel of the ACRR core were within {+-} 10 percent agreement with the

Dasari V. Rao; Mohamed S. El-Genk; Reuben A. Rubio; James W. Bryson; Fabian C. Foushee

1988-01-01

28

Lifetime embrittlement of reactor core materials  

SciTech Connect

Over a core lifetime, the reactor materials Zircaloy-2, Zircaloy-4, and hafnium may become embrittled due to the absorption of corrosion- generated hydrogen and to neutron irradiation damage. Results are presented on the effects of fast fluence on the fracture toughness of wrought Zircaloy-2, Zircaloy-4, and hafnium; Zircaloy-4 to hafnium butt welds; and hydrogen precharged beta treated and weld metal Zircaloy-4 for fluences up to a maximum of approximately 150 x 10{sup 24} n/M{sup 2} (> 1 Mev). While Zircaloy-4 did not exhibit a decrement in K{sub IC} due to irradiation, hafnium and butt welds between hafnium and Zircaloy-4 are susceptible to embrittlement with irradiation. The embrittlement can be attributed to irradiation strengthening, which promotes cleavage fracture in hafnium and hafnium-Zircaloy welds, and, in part, to the lower chemical potential of hydrogen in Zircaloy-4 compared to hafnium, which causes hydrogen, over time, to drift from the hafnium end toward the Zircaloy-4 end and to precipitate at the interface between the weld and base-metal interface. Neutron radiation apparently affects the fracture toughness of Zircaloy-2, Zircaloy-4, and hafnium in different ways. Possible explanations for these differences are suggested. It was found that Zircaloy-4 is preferred over Zircaloy-2 in hafnium-to- Zircaloy butt-weld applications due to its absence of a radiation- induced reduction in K{sub IC} plus its lower hydrogen absorption characteristics compared with Zircaloy-2.

Kreyns, P.H..; Bourgeois, W.F.; Charpentier, P.L.; Kammenzind, B.F.; Franklin, D.G. [Bettis Atomic Power Lab., West Mifflin, PA (United States); White, C.J. [Knolls Atomic Power Lab., Schenectady, NY (United States)

1994-08-01

29

Hanging core support system for a nuclear reactor. [LMFBR  

DOEpatents

For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

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

1984-04-26

30

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)  

NASA Technical Reports Server (NTRS)

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

Clement, J. D.; Rust, J. H.

1977-01-01

31

Fuel, Core Design and Subchannel Analysis of a Superfast Reactor  

Microsoft Academic Search

A compact supercritical water-cooled fast reactor (superfast reactor) core with a power of 700MWe is designed by using a three-dimensional neutronics thermal-hydraulic coupled method. The core consists of 126 seed assemblies and 73 blanket assemblies. In the seed assemblies, 251 fuel rods, consisting of MOX pellets, stainless steel (SUS304) cladding, and fission gas plenum are arranged into a tight triangle

Liangzhi CAO; Yoshiaki OKA; Yuki ISHIWATARI; Zhi SHANG

2008-01-01

32

Proposed methods for defueling the TMI2 reactor core  

Microsoft Academic Search

This report constitutes the general concensus of a Debris Defueling Working Group which was established by the US Department of Energy, through EG and G Idaho Inc., to obtain recommendations from nuclear industry representatives concerning techniques for removing fuel debris from the TMI-2 reactor vessel. The current configuration of the reactor core materials is characterized based on the best information

Henrie

1984-01-01

33

Proposed methods for defueling the TMI2 reactor core  

Microsoft Academic Search

This report constitutes the general consensus of a Debris Defueling Working Group which was established by the US Department of Energy, through EG and G Idaho Inc., to obtain recommendations from nuclear industry representatives concerning techniques for removing fuel debris from the TMI-2 reactor vessel. The current configuration of the reactor core materials is characterized based on the best information

Henrie

1983-01-01

34

Sizing an external-fueled in-core thermionic reactor.  

NASA Technical Reports Server (NTRS)

Parametric studies on sizing of external-fueled in-core thermionic reactors are presented. Reactor physics results obtained for a variety of fuel element designs are used as a basis for nuclear criticality, power distribution, and control worth design. Thermionic performance results for a single fuel element for several sets of operating conditions are presented. An algorithm combining the electrical and reactor physics results in a form amenable to preliminary systems analysis is presented.

Nakashima, A. M.; Sawyer, C. D.

1971-01-01

35

Cooling of core debris within the reactor vessel lower head  

Microsoft Academic Search

Under severe-accident conditions, the most crucial action for recovery from the accident state is to cool the core debris and prevent or terminate attack on the remaining fission product barriers. One means of preventing attack on the containment structures is to retain the core debris within the reactor vessel. The Three Mile Island Unit 2 (TMI-2) accident demonstrated that this

R. E. Henry; J. P. Burelbach; R. J. Hammersley; C. E. Henry; G. T. Klopp

1991-01-01

36

Cooling of core debris within the reactor vessel lower head  

Microsoft Academic Search

Under severe accident conditions, the most crucial action for recovery from the accident state is to cool the core debris and prevent or terminate attack on the remaining fission product barriers. One means of preventing attack on the containment structures is to retain the core debris within the reactor vessel. Some accident situations could result in the transport of molten

R. E. Henry; J. P. Burelbach; R. J. Hammersley; C. E. Henry; G. T. Klopp

1993-01-01

37

System to Reduce the Hazards of a Reactor Core Meltdown.  

National Technical Information Service (NTIS)

The core catcher is realized by the fact that at least on the inner side facing the reactor core it consists of at least one high-melting material from the group of graphite, carbides and material combinations of graphite fibers and/or filaments with carb...

L. Barleon S. Dorner O. Goetzmann G. Kussmaul

1977-01-01

38

An economic optimization of pressurized light water reactor cores  

Microsoft Academic Search

Two reactor cores (1000 MWe and 600 MWe) are optimized with respect to power cost. The power cost is minimized while retaining the thermal-hydraulic margins of the reference core. Constant thermal-hydraulic margins result in similar accident thermal-hydraulic transient behavior of the cores developed during the optimization study. The cost components impacted by the optimization are once-through fuel cycle, capital, and

Holger Pfeifer

1997-01-01

39

Core design of the upgraded TREAT reactor  

SciTech Connect

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

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

1982-01-01

40

Identification and control of a nuclear reactor core (VVER) using recurrent neural networks and fuzzy systems  

Microsoft Academic Search

Improving the methods of identification and control of nuclear power reactors core is an important area in nuclear engineering. Controlling the nuclear reactor core during load following operation encounters some difficulties in control of core thermal power while considering the core limitations in local power peaking and safety margins. In this paper, a nuclear power reactor core (VVER) is identified

Mehrdad Boroushaki; Mohammad B. Ghofrani; Caro Lucas; Mohammad J. Yazdanpanah

2003-01-01

41

Thermal barrier and support for nuclear reactor fuel core  

SciTech Connect

A nuclear reactor is described having a thermal barrier for supporting a fuel column of a nuclear reactor core within a reactor vessel having a fixed rigid metal liner. The fuel column has a refractory post extending downward. The thermal barrier comprises, in combination, a metallic core support having an interior chamber secured to the metal liner; fibrous thermal insulation material covering the metal liner and surrounding the metallic core support; means associated with the metallic core support and resting on the top for locating and supporting the full column post; and a column of ceramic material located within the interior chamber of the metallic core support, the height of the column is less than the height of the metallic core support so that the ceramic column will engage the means for locating and supporting the fuel column post only upon plastic deformation of the metallic core support; the core support comprises a metallic cylinder and the ceramic column comprises coaxially aligned ceramic pads. Each pad has a hole located within the metallic cylinder by means of a ceramic post passing through the holes in the pads.

Betts, W.S. Jr.; Pickering, J.L.; Black, W.E.

1987-06-16

42

Concept Design of Advanced Marine Reactor (1), Core Design. Design Study for Optimized Core.  

National Technical Information Service (NTIS)

We started the design study of the concept design of the Advanced Marine Reactor from FY 1987, and we researched and studied as for the optimization of the core and the components of the 100 MWt reactor plant that main research thema was minituarization a...

N. Ambo T. Yokomura

1989-01-01

43

Hanging core support system for a nuclear reactor  

DOEpatents

For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform. Motion or radiation sensing detectors can be provide at the lower ends of the tension rods for obtaining pertinent readings proximate the core.

Burelbach, James P. (Glen Ellyn, IL); Kann, William J. (Park Ridge, IL); Pan, Yen-Cheng (Naperville, IL); Saiveau, James G. (Hickory Hills, IL); Seidensticker, Ralph W. (Wheaton, IL)

1987-01-01

44

Bowing of core assemblies in advanced liquid metal fast reactors  

SciTech Connect

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

Kamal, S.A.; Orechwa, Y.

1986-01-01

45

Core length testable reactor concept neutronic analysis  

SciTech Connect

Development work on thermionic reactor systems has been ongoing in the US since the early 1950s. While significant successes were achieved, progress has been hampered by frequent changes in direction and funding instabilities (as has been true for many high technology initiatives). The recent Air Force thermionics initiative (1991) represents the latest in thermionics reactor development in the US. This Air Force initiative called for the development of thermionics reactors with the output power of about 40 kWe, and which incorporated the features of testability, fabricability, low development cost, high level of safety and reliability, and survivability. Several concepts were analyzed to define a design that would meet all the requirements set forth by the Air Force. This report describes the methodology used, the different designs analyzed and reasons for the evolution of the design, and presents the results for the different concepts.

Hanan, N.A.; Bhattacharyya, S.K.

1992-09-01

46

Unsteady Characteristics of Three-Core Molten Salt Reactor  

NASA Astrophysics Data System (ADS)

Numerical analysis has been performed for load-following capability of a 465 MWth Three-Core Molten Salt Reactor (MSR). Reactor-slaved-to-turbine control technique is adopted for reactor control. As for this control technique, a turbine is controlled by a speed regulator of a generator, and subsequently the reactor is controlled so as to follow the turbine output. In this study, the turbine power is rapidly changed in a range of 50-150% of the rated power. Then transient characteristics of fuel salt and graphite temperatures, neutron fluxes, delayed neutron precursors, and reactor output are calculated. The analysis result shows that the reactor output is capable of following the turbine power in the range of the turbine output of 50-150%.

Yamamoto, Takahisa; Mitachi, Koshi; Nishio, Masatoshi

47

Dimensional changes in elements of the BN600 reactor core  

Microsoft Academic Search

Increasing the radiation stability of the structural materials of the core is of great importance for the safe and efficient operation of fast reactors. At the beginning of the operation of the BN-600 reactor, the fuel assembly boxes and the fuel cladding tubes were made from anstenitic 08Khl6NllM3 and 08Khl6N15M3B (t~I-847) steel, respectively, the components of the rods and the

S. E. Astashov; E. A. Kozmanov; A. N. Ogorodov; G. A. Sergeev; V. V. Chuev; A. G. Sheinkman; L. M. Zabud'ko; O. S. Korostin; E. A. Rogov

1993-01-01

48

Modification of the Core Cooling System of TRIGA 2000 Reactor  

NASA Astrophysics Data System (ADS)

To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

Umar, Efrizon; Fiantini, Rosalina

2010-06-01

49

Modification of the Core Cooling System of TRIGA 2000 Reactor  

SciTech Connect

To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24 deg. C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

Umar, Efrizon; Fiantini, Rosalina [National Nuclear Energy Agency of Indonesia, Jalan Tamansari 71, Bandung, 40132 (Indonesia)

2010-06-22

50

Gas core reactors for actinide transmutation and breeder applications  

NASA Technical Reports Server (NTRS)

This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.

Clement, J. D.; Rust, J. H.

1978-01-01

51

Design of an LEU core for the MIT reactor  

SciTech Connect

A design of the MIT Reactor core using monolithic U-7Mo LEU fuel has been developed with the goal of maintaining thermal and fast neutron fluxes as well as increasing the flexibility for meeting the needs of in-core experiments. An optimum core was sought by varying the core materials, and fuel plate numbers and thicknesses, but maintaining the outside dimensions of a fuel element. A full-core model of the MITR by the Monte-Carlo transport code MCNP was used to calculate the neutron fluxes, reactivity and neutron spectrum available for experiments. The optimum reactor design consisted of the use of half-sized fuel elements made up of nine U-7Mo LEU fuel plates of 0.55 mm thickness with 0.25 mm finned aluminum cladding. This design also utilized solid beryllium fuel elements (dummies) with boron fixed absorbers or solid lead dummies, depending on the in-core experiment flux and spectrum needs. Because the new core design contains twice the amount of 235 U as does the existing HEU core, and produces much more Pu, its fuel cycle length is twice as long at the same power level. Preliminary thermal-hydraulic and neutronic safety evaluations indicate superior performance to the current HEU fuel. (authors)

Newton, T. [Massachusetts Inst. of Technology, Nuclear Reactor Laboratory, 138 Albany St., Cambridge, MA 02139 (United States); Kazimi, M.; Pilat, E. [Nuclear Science and Engineering Dept., 77 Massachusetts Ave., Cambridge, MA 02139 (United States)

2006-07-01

52

MELCOR adaptation and validation for modeling of N Reactor core phenomena  

Microsoft Academic Search

MELCOR has been adapted for use in modeling the N Reactor core as a part of the recently completed N Reactor probabilistic risk assessment. Significant adaptation of MELCOR was required because of the horizontal, water cooled, graphite-moderated nature of the N Reactor core. The generation and verification of the revised N Reactor core model are described in this paper. A

G. D. Wyss; R. M. Summers; L. A. Miller

1990-01-01

53

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

SciTech Connect

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

Camp, A.L.; Dingman, S.E.; Forester, J.A. [and others

1996-12-31

54

An economic optimization of pressurized light water reactor cores  

NASA Astrophysics Data System (ADS)

Two reactor cores (1000 MWe and 600 MWe) are optimized with respect to power cost. The power cost is minimized while retaining the thermal-hydraulic margins of the reference core. Constant thermal-hydraulic margins result in similar accident thermal-hydraulic transient behavior of the cores developed during the optimization study. The cost components impacted by the optimization are once-through fuel cycle, capital, and administrative/manpower costs. The variables in the optimization are pin diameter, moderator to fuel (H/U) ratio, core length, and the number of fuel pins in the core. A sequential quadratic programming approach is employed to solve the nonlinear optimization problem with constraints. The fuel cycle costs are evaluated by the use of the linear reactivity model, and capital costs are adjusted by suitable modifications to the nuclear energy cost database reference costs. The results of the analysis shows that for fixed assembly parameters (i.e., pin diameter, H/U ratio, and core length), the optimum core is one that operates at the thermal-hydraulic limits. Cores optimized with unconstrained assembly characteristics contain a larger number of smaller pins at a higher H/U ratio. This follows the trend in current reactor designs. While the lifetime power cost savings for the optimized core are less than 4 million dollars (versus a present day total cost of 6.9 billion dollars), the optimization analysis shows that higher thermal-hydraulic margins can be attained with minimum power cost increases. With increased emphasis on reactor safety, significantly higher safety margins may therefore be achieved without a significant power cost increase. The optimized configurations were found to be relatively insensitive to fuel cycle cost component variations.

Pfeifer, Holger

55

Photoneutron effects on pulse reactor kinetics for the Annular Core Research Reactor (ACRR)  

Microsoft Academic Search

The Annular Core Research Reactor (ACRR) is a swimming-pool type pulsed reactor that maintains an epithermal neutron flux and a nine-inch diameter central dry cavity. One of its uses is neutron and gamma-ray irradiation damage studies on electronic components under transient reactor power conditions. In analyzing the experimental results, careful attention must be paid to the kinetics associated with the

Parma; Edward J

2009-01-01

56

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype  

NASA Technical Reports Server (NTRS)

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

Bragg-Sitton, Shannon M.

2005-01-01

57

Heat transfer evaluation in a plasma core reactor  

Microsoft Academic Search

Numerical evaluations of heat transfer in a fissioning uranium plasma core reactor cavity, operating with seeded hydrogen propellant, was performed. A two-dimensional analysis is based on an assumed flow pattern and cavity wall heat exchange rate. Various iterative schemes were required by the nature of the radiative field and by the solid seed vaporization. Approximate formulations of the radiative heat

D. E. Smith; T. M. Smith; M. L. Stoenescu

1976-01-01

58

MCNP/MCNPX Model of the Annular Core Research Reactor.  

National Technical Information Service (NTIS)

Many experimenters at the Annular Core Research Reactor (ACRR) have a need to predict the neutron/gamma environment prior to testing. In some cases, the neutron/gamma environment is needed to understand the test results after the completion of an experime...

E. J. Parma K. R. DePriest P. J. Cooper

2006-01-01

59

Control Rod Reactivity Curves for the Annular Core Research Reactor  

Microsoft Academic Search

Experiments were conducted at the Annular Core Research Reactor (ACRR) to increase the fidelity of the control rod integral reactivity worth curve. This experiment series was designed to refine the integral reactivity curve used for pulse yield prediction and eliminate the need for operator compensation in the pulse setup. The experiment series consisted of delayed critical and positive period measurements

K. Russell Depriest; Karen C. Kajder; Jason N. Frye; Matthew R. Denman

2009-01-01

60

MCNP\\/MCNPX model of the annular core research reactor  

Microsoft Academic Search

Many experimenters at the Annular Core Research Reactor (ACRR) have a need to predict the neutron\\/gamma environment prior to testing. In some cases, the neutron\\/gamma environment is needed to understand the test results after the completion of an experiment. In an effort to satisfy the needs of experimenters, a model of the ACRR was developed for use with the Monte

Kendall Russell DePriest; Philip J. Cooper; Parma Edward J. Jr

2006-01-01

61

Support arrangement for core modules of nuclear reactors  

DOEpatents

A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

Bollinger, Lawrence R. (Schenectady, NY)

1987-01-01

62

Gamma thermometer based reactor core liquid level detector  

DOEpatents

A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is modified by thermocouple junctions and leads to obtain an unambiguous indication of the presence or absence of coolant liquid at the gamma thermometer location. A signal processor generates a signal based on the thermometer surface heat transfer coefficient by comparing the signals from the thermocouples at the thermometer location. The generated signal is a direct indication of loss of coolant due to the change in surface heat transfer when coolant liquid drops below the thermometer location. The loss of coolant indication is independent of reactor power at the thermometer location. Further, the same thermometer may still be used for the normal power monitoring function.

Burns, Thomas J. (Knoxville, TN)

1983-01-01

63

Gamma thermometer based reactor core liquid level detector  

SciTech Connect

A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is midified by thermocouple junctions and leads to obtain an unambiguous indication of the presence or absence of coolant liquid at the gamma thermometer location. A signal processor generates a signal based on the thermometer surface heat transfer coefficient by comparing the signals from the thermocouples at the thermometer location. The generated signal is a direct indication of loss of coolant due to the change in surface heat transfer when coolant liquid drops below the thermometer location. The loss of coolant indication is independent of reactor power at the thermometer location. Further, the same thermometer may still be used for the normal power monitoring function.

Burns, T.J.

1983-09-20

64

Support arrangements for core modules of nuclear reactors. [PWR  

DOEpatents

A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

Bollinger, L.R.

1983-11-03

65

Annular core research reactor high flux neutron radiography facility  

SciTech Connect

Sandia National Laboratories (SNL) has been performing neutron radiography since 1964. The radiography facilities have evolved from an aperture in a radiation exposure room in the now retired Sandia Engineering Reactor to a divergent collimator radiography facility adjacent to the core of the Annular Core Research Reactor (ACRR). The maximum thermal neutron flux achieved in these facilities has been limited to approximately 1 {times} 10{sup 7} n-cm{sup -2}-s{sup -1}. In order to perform high-resolution, real-time neutron radiography for transient events, higher neutron fluxes are required. In response to this need, Sandia is designing a new high-flux neutron radiography facility for the ACRR. The new facility uses the central irradiation cavity of the ACRR and consists of a collimator assembly, reactor control system, an experiment support structure, and an imaging system. This new facility is described in this paper. 2 refs., 2 figs.

McCrory, F.M.; Kelly, J.G.; Vernon, M.E.; Tichenor, D.A.

1990-01-01

66

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (I)  

Microsoft Academic Search

A computer code SEFDAN is developed for one-dimensional thermal-hydraulics in a partially uncovered core of a light water reactor during a severe core damage accident. The developed models include:1. Froth level (or dry-out level) calculation2. Transition and mixing between convection flow regimes in convective heat transfer3. Radiant heat transfer between solid walls and flowing gas4. Heat generation by zirconium-water reaction5.

Fumiya TANABE; Ken MURAMATSU; Tohru SUDA

1986-01-01

67

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (III)  

Microsoft Academic Search

The SEFDAN is a computer program to analyze the one-dimensional thermal-hydraulics of a partially uncovered core of a light water reactor in a severe degraded-cooling event. In order to verify the code and to obtain better understanding of the severe core damage process, SEFDAN has been applied to analyses of the thermal response of fuel rods in the Power Burst

Ken MURAMATSU; Fumiya TANABE; Tohru SUDA

1986-01-01

68

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (II)  

Microsoft Academic Search

The SEFDAN is a computer program to analyze the one-dimensional thermal-hydraulics of a partially uncovered core of a light water reactor in a severe degraded-cooling event. In order to verify the code and to obtain better understanding of the severe core damage process, SEFDAN has been applied to analyses of the thermal response of fuel rods in the Power Burst

Fumiya TANABE; Ken MURAMATSU; Tohru SUDA

1986-01-01

69

Gamma-Thermometer-Based Reactor-Core Liquid-Level Detector (PWR).  

National Technical Information Service (NTIS)

A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is modifie...

T. J. Burns

1981-01-01

70

Development and Assessment of Advanced Reactor Core Protection System  

NASA Astrophysics Data System (ADS)

An advanced core protection system for a pressurized water reactor, Reactor Core Protection System(RCOPS), was developed by adopting a high performance hardware platform and optimal system configuration. The functional algorithms of the core protection system were also improved to enhance the plant availability by reducing unnecessary reactor trips and increasing operational margin. The RCOPS consists of four independent safety channels providing a two-out-of-four trip logic. The reliability analysis using the reliability block diagram method showed the unavailability of the RCOPS to be lower than the conventional system. The failure mode and effects analysis demonstrated that the RCOPS does not lose its intended safety functions for most failures. New algorithms for the RCOPS functional design were implemented in order to avoid unnecessary reactor trips by providing auxiliary pre-trip alarms and signal validation logic for the control rod position. The new algorithms in the RCOPS were verified by comparing the RCOPS calculations with reference results. The new thermal margin algorithm for the RCOPS was expected to increase the operational margin to the limit for Departure from Nucleate Boiling Ratio (DNBR) by approximately 1%.

in, Wang-Kee; Park, Young-Ho; Baeg, Seung-Yeob

71

Proceedings of the topical meeting on advances in reactor physics and core thermal hydraulics  

SciTech Connect

Technical papers presented at the ANS Topical Meeting on Advances in Reactor Physics and Core Thermal Hydraulics, September 22-24, 1982, at Kiamesha Lake, N.Y. are included in these Proceedings. Reactor physics, core thermal hydraulics, and the interactions between core physics and thermal hydraulics are covered both for thermal reactors and for fast breeders. There are sessions on current challenges in these areas, on measurement and analysis of fast reactor physics parameters, on coupled core physics and thermal-hydraulics analysis, on in-core fuel management, on nodal and homogenization methods in reactor physics, and on core thermal hydraulic and nuclear instrumentation.

Not Available

1982-08-01

72

Relative Iris Codes  

Microsoft Academic Search

This paper proposes a new scheme to generate iris codes based on relative measure of local iris texture. The local characteristic of iris texture is analyzed using 2D Gabor wavelets. Twelve Gabor kernels, four frequencies and three orientations, are constructed and convoluted with an iris image. To inherit relationship of local iris texture among pixels, Gabor magnitude and phase of

Peeranat Thoonsangngam; Somying Thainimit; Vutipong Areekul

2007-01-01

73

Dynamic analysis of gas-core reactor system  

NASA Technical Reports Server (NTRS)

A heat transfer analysis was incorporated into a previously developed model CODYN to obtain a model of open-cycle gaseous core reactor dynamics which can predict the heat flux at the cavity wall. The resulting model was used to study the sensitivity of the model to the value of the reactivity coefficients and to determine the system response for twenty specified perturbations. In addition, the model was used to study the effectiveness of several control systems in controlling the reactor. It was concluded that control drums located in the moderator region capable of inserting reactivity quickly provided the best control.

Turner, K. H., Jr.

1973-01-01

74

Gas core reactor concepts and technology - Issues and baseline strategy  

NASA Technical Reports Server (NTRS)

Results of a research program including phenomenological studies, conceptual design, and systems analysis of a series of gaseous/vapor fissile fuel driven engines for space power platforms and for thermal and electric propulsion are reviewed. It is noted that gas and vapor phase reactors provide the path for minimum mass in orbit and trip times, with a specific impulse from 1020 sec at the lowest technololgical risk to 5200 sec at the highest technological risk. The discussion covers various configurations of gas core reactors and critical technologies and the nuclear vapor thermal rocket engine.

Diaz, Nils J.; Dugan, Edward T.; Kahook, Samer; Maya, Isaac

1991-01-01

75

System Study: Reactor Core Isolation Cooling 19982012  

SciTech Connect

This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

T. E. Wierman

2013-10-01

76

Light Water Breeder Reactor core evaluation operations at the expended core facility (LWBR Development Program)  

Microsoft Academic Search

This report presents an overview of the processes and equipment used to receive, examine, and store the Light Water Breeder Reactor (LWBR) fuel modules once they had been partially disassembled and shipped to the Expended Core Facility (ECF) in Idaho for storage in an underwater facility. At ECF, the 39 fuel modules underwent further disassembly to provide fuel rods for

1987-01-01

77

Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors  

NASA Astrophysics Data System (ADS)

Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.

Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Subhki, Muhamad Nurul; Ismail

2010-12-01

78

Development of a core follow calculational system for research reactors  

SciTech Connect

Over the last few years a comprehensive PWR and MTR core analysis code system based on modern reactor physics methods has been under development by the Atomic Energy Corporation of South Africa. This system, known as OSCAR-3, will incorporate a customized graphical user interface and data management system to ensure user-friendliness and good quality control. The system has now reached the stage of development where it can be used for practical MTR core analyses. This paper describes the current capabilities of the components of the OSCAR-3 package, their integration within the package, and outlines future developments.

Mueller, E.Z.; Ball, G.; Joubert, W.R. [and others

1994-12-31

79

Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores  

SciTech Connect

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

Krass, A.W.

2005-12-19

80

Replacement fuel scoping studies for the Annular Core Research Reactor  

Microsoft Academic Search

Sandia National Laboratories Annular Core Research Reactor (ACRR) is undertaking a new mission for the Department of Energy: production of the radioisotope ⁹⁹Mo used in nuclear medicine applications. Isotope production is significantly different from previous programs conducted at the ACRR that typically required high intensity, short duration pulses. The current UO-BeO fuel will power the initial startup phase of the

K. Hays; L. Martin; E. Parma

1995-01-01

81

Replacement fuel scoping studies for the annular core research reactor  

Microsoft Academic Search

Sandia National Laboratories annular core research reactor (ACRR) is a candidate for a new mission for the U.S Department of Energy: production of the radioisotope ⁹⁹Mo used in nuclear medicine applications. Isotope production is significantly different from previous programs conducted at the ACRR that typically required high-intensity, short-duration pulses. The current UO-BeO fuel would power the initial startup phase of

K. Hays; L. Martin; E. J. Parma

1995-01-01

82

A reactor core on-line monitoring program - COMP  

SciTech Connect

A program named COMP is developed for on-line monitoring PWRs' in-core power distribution in this paper. Harmonics expansion method is used in COMP. The Unit 1 reactor of Daya Bay Nuclear Power Plant (Daya Bay NPP) in China is considered for verification. The numerical results show that the maximum relative error between measurement and reconstruction results from COMP is less than 5%, and the computing time is short, indicating that COMP is capable for online monitoring PWRs. (authors)

Wang, C. [State Nuclear Power Software Development Center, Beijing, 100029 (China); School of Nuclear Science and Technology, Xi'an Jiaotong Univ. (China); Building 1, Compound No.29, North Third Ring Road, Xicheng District, Beijing, 100029 (China); Wu, H.; Cao, L. [School of Nuclear Science and Technology, Xi'an Jiaotong Univ. (China)

2012-07-01

83

Gas core reactors for actinide transmutation. [uranium hexafluoride  

NASA Technical Reports Server (NTRS)

The preliminary design of a uranium hexafluoride actinide transmutation reactor to convert long-lived actinide wastes to shorter-lived fission product wastes was analyzed. It is shown that externally moderated gas core reactors are ideal radiators. They provide an abundant supply of thermal neutrons and are insensitive to composition changes in the blanket. For the present reactor, an initial load of 6 metric tons of actinides is loaded. This is equivalent to the quantity produced by 300 LWR-years of operation. At the beginning, the core produces 2000 MWt while the blanket generates only 239 MWt. After four years of irradiation, the actinide mass is reduced to 3.9 metric tonnes. During this time, the blanket is becoming more fissile and its power rapidly approaches 1600 MWt. At the end of four years, continuous refueling of actinides is carried out and the actinide mass is held constant. Equilibrium is essentially achieved at the end of eight years. At equilibrium, the core is producing 1400 MWt and the blanket 1600 MWt. At this power level, the actinide destruction rate is equal to the production rate from 32 LWRs.

Clement, J. D.; Rust, J. H.; Wan, P. T.; Chow, S.

1979-01-01

84

Photoneutron effects on pulse reactor kinetics for the Annular Core Research Reactor (ACRR).  

SciTech Connect

The Annular Core Research Reactor (ACRR) is a swimming-pool type pulsed reactor that maintains an epithermal neutron flux and a nine-inch diameter central dry cavity. One of its uses is neutron and gamma-ray irradiation damage studies on electronic components under transient reactor power conditions. In analyzing the experimental results, careful attention must be paid to the kinetics associated with the reactor to ensure that the transient behavior of the electronic device is understood. Since the ACRR fuel maintains a substantial amount of beryllium, copious quantities of photoneutrons are produced that can significantly alter the expected behavior of the reactor power, especially following a reactor pulse. In order to understand these photoneutron effects on the reactor kinetics, the KIFLE transient reactor-analysis code was modified to include the photoneutron groups associated with the beryllium. The time-dependent behavior of the reactor power was analyzed for small and large pulses, assuming several initial conditions including following several pulses during the day, and following a long steady-state power run. The results indicate that, for these types of initial conditions, the photoneutron contribution to the reactor pulse energy can have a few to tens of percent effect.

Parma, Edward J., Jr.

2009-06-01

85

Piezoelectric material for use in a nuclear reactor core  

SciTech Connect

In radiation environments ultrasonic nondestructive evaluation has great potential for improving reactor safety and furthering the understanding of radiation effects and materials. In both nuclear power plants and materials test reactors, elevated temperatures and high levels of radiation present challenges to ultrasonic NDE methodologies. The challenges are primarily due to the degradation of the ultrasonic sensors utilized. We present results from the operation of a ultrasonic piezoelectric transducer, composed of bulk single crystal AlN, in a nuclear reactor core for over 120 MWHrs. The transducer was coupled to an aluminum cylinder and operated in pulse echo mode throughout the irradiation. In addition to the pulse echo testing impedance data were obtained. Further, the piezoelectric coefficient d{sub 33} was measured prior to irradiation and found to be 5.5 pC/N which is unchanged from as-grown samples, and in fact higher than the measured d{sub 33} for many as-grown samples.

Parks, D. A.; Reinhardt, Brian; Tittmann, B. R. [EES Department, Penn State University, University Park, PA 16802 (United States)

2012-05-17

86

RMC - A Monte Carlo Code for Reactor Core Analysis  

NASA Astrophysics Data System (ADS)

A new Monte Carlo transport code RMC has been being developed by Department of Engineering Physics, Tsinghua University, Beijing as a tool for reactor core analysis on high-performance computing platforms. To meet the requirements of reactor analysis, RMC now has such functions as criticality calculation, fixed-source calculation, burnup calculation and kinetics simulations. Some techniques for geometry treatment, new burnup algorithm, source convergence acceleration, massive tally and parallel calculation, and temperature dependent cross sections processing are researched and implemented in RMC to improve the effciency. Validation results of criticality calculation, burnup calculation, source convergence acceleration, tallies performance and parallel performance shown in this paper prove the capabilities of RMC in dealing with reactor analysis problems with good performances.

Wang, Kan; Li, Zeguang; She, Ding; Liang, Jin'gang; Xu, Qi; Qiu, Yishu; Yu, Jiankai; Sun, Jialong; Fan, Xiao; Yu, Ganglin

2014-06-01

87

Development of an automated core model for nuclear reactors  

SciTech Connect

This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input.

Mosteller, R.D.

1998-12-31

88

Thermal hydraulics model for Sandia's annular core research reactor  

SciTech Connect

A thermal hydraulics model was developed for the Annular Core Research Reactor (ACRR) at Sandia National Laboratories. The coupled mass, momentum and energy equations for the core were solved simultaneously using an explicit forward marching numerical technique. The model predictions of the temperature rise across the central channel of the ACRR core were within {+-} 10 percent agreement with the in-core temperature measurements. The model was then used to estimate the coolant mass flow rate and the axial distribution of the cladding surface temperature in the central and average channels as functions of the operating power and the water inlet subcooling. Results indicated that subcooled boiling occurs at the cladding surface in the central channels of the ACRR at power levels in excess of 0.5 MW. However, the high heat transfer coefficient due to subcooled boiling causes the cladding temperature along most of the active fuel rod region to be quite uniform and to increase very little with the reactor power. (author)

Rao, Dasari V.; El-Genk, Mohamed S. [Department of Chemical and Nuclear Engineering, University of New Mexico, Albuquerque, NM 87131 (United States); Rubio, Reuben A.; Bryson, James W. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Foushee, Fabian C. [General Atomics Technologies, Inc., San Diego, CA 92138 (United States)

1988-07-01

89

GCRA review and appraisal of HTGR reactor-core-design program. [HTGR-SC, -R, -NHSDR  

SciTech Connect

The reactor-core-design program has as its principal objective and responsibility the design and resolution of major technical issues for the reactor core and core components on a schedule consistent with the plant licensing and construction program. The task covered in this review includes three major design areas: core physics, core thermal and hydraulic performance fuel element design, and in-core fuel performance evaluation.

Not Available

1980-09-01

90

Proceedings of the topical meeting on advances in reactor physics and core thermal hydraulics  

SciTech Connect

Technical papers presented at the ANS Topical Meeting on Advances in Reactor Physics and Core Thermal Hydraulics, September 22-24, 1982, at Kiamesha Lake, NY, are included in these proceedings. Reactor physics, core thermal hydraulics, and the interactions between core physics and thermal hydraulics are covered both for thermal reactors and for fast breeders. There are sessions on current challenges in these areas. On validation of fast reactor thermal hydraulic methods, on reactor theory, on measurement and analysis of thermal reactor physics parameters, on validation of thermal reactor thermal hydraulics methods, and on development and utilization of differential and integral nuclear data.

Not Available

1982-08-01

91

MCNP/MCNPX model of the annular core research reactor.  

SciTech Connect

Many experimenters at the Annular Core Research Reactor (ACRR) have a need to predict the neutron/gamma environment prior to testing. In some cases, the neutron/gamma environment is needed to understand the test results after the completion of an experiment. In an effort to satisfy the needs of experimenters, a model of the ACRR was developed for use with the Monte Carlo N-Particle transport codes MCNP [Br03] and MCNPX [Wa02]. The model contains adjustable safety, transient, and control rods, several of the available spectrum-modifying cavity inserts, and placeholders for experiment packages. The ACRR model was constructed such that experiment package models can be easily placed in the reactor after being developed as stand-alone units. An addition to the 'standard' model allows the FREC-II cavity to be included in the calculations. This report presents the MCNP/MCNPX model of the ACRR. Comparisons are made between the model and the reactor for various configurations. Reactivity worth curves for the various reactor configurations are presented. Examples of reactivity worth calculations for a few experiment packages are presented along with the measured reactivity worth from the reactor test of the experiment packages. Finally, calculated neutron/gamma spectra are presented.

DePriest, Kendall Russell; Cooper, Philip J.; Parma, Edward J., Jr. (.,; .)

2006-10-01

92

A Solid Core Heatpipe Reactor with Cylindrical Thermoelectric Converter Modules  

SciTech Connect

A nuclear space power system that consists of a solid metal nuclear reactor core with heat pipes carrying energy to a cylindrical thermoelectric converter surrounding each of the heat pipes with a heat pipe radiator surrounding the thermoelectric converter is the most simple and reliable space power system. This means no single point of failure since each heat pipe and cylindrical converter is a separate power system and if one fails it will not affect the others. The heat pipe array in the solid core is designed so that if an isolated heat pipe or even two adjacent heat pipes fail, the remaining heat pipes will still transport the core heat without undue overheating of the uranium nitride fuel. The primary emphasis in this paper is on simplicity, reliability and fabricability of such a space nuclear power source. The core and heat pipes are made of Niobium 1% Zirconium alloy (Nb1Zr), with rhenium lined fuel tubes, bonded together by hot isostatic pressure (HIPing) and with sodium as the heat pipe working fluid, can be operated up to 1250K. The cylindrical thermoelectric converter is made by depositing the constituents of the converter around a Nb1%Zr tube and encasing it in a Nb 1% Zr alloy tube and HIPing the structure to get final bonding and to produce residual compressive stresses in all brittle materials in the converter. A radiator heat pipe filled with potassium that operates at 850K is bonded to the outside of the cylindrical converter for cooling. The solid core heat pipe and cylindrical converter are mated by welding during the final assembly. A solid core reactor with 150 heat pipes with a 0.650-inch (1.65 cm) ID and a 30-inch (76.2 cm) length with an output of 8 Watts per square inch as demonstrated by the SP100 PD2 cell tests will produce about 80 KW of electrical power. An advanced solid core reactor made with molybdenum 47% rhenium alloy, with lithium heat pipes and the PD2 theoretical output of 11 watts per square inch or advanced higher temperature converter to operate at 1350K could produce a greater output of approximately 100KW.

Sayre, Edwin D. [218 Brooke Acres Drive, Los Gatos, CA 95032 (United States); Vaidyanathan, Sam [6663 Pomander Place, San Jose, CA 95120 (United States)

2006-01-20

93

Study of a new compact fast reactor core design  

SciTech Connect

A study was conducted to investigate conceptual Liquid Metal Reactor (LMR) designs, employing some unconventional design features, for improved economics and safety. The unconventional design elements were used to supplement the conventional design measures, which alone did not lead to a truly competitive LMR design. Better economics was obtained through simplicity and compactness of core design. For simplicity of core design, internal blankets were omitted. Core compactness was achieved by maximum power flattening. This was done by employing axial and radial enrichment zones along with axial and radial (BeO) reflectors. To further enhance core compactness, the in-core control rods were replaced by reflector controls. For improved safety, the objective was to reduce both coolant void and burnup reactivities. However, even with the use of a wide spectrum of unconventional design features, such as burnable poisons, peripheral reflectors and inner moderating regions, it was not possible to overcome the classical known fact that both coolant void and burnup reactivities cannot be reduced simultaneously. The only resolution of this dilemma appeared to be to minimize coolant void reactivity, and to manage the burnup reactivity losses, such that an accidental insertion of significant amounts of reactivity is mechanically not possible. A conceptual design with these characteristics is described in this thesis.

Hamid, T.

1990-01-01

94

Nuclear reactor spacer grid and ductless core component  

SciTech Connect

This invention relates to a nuclear reactor spacer grid member for use in a liquid cooled nuclear reactor and to a ductless core component employing a plurality of these spacer grid members. The spacer grid member is of the egg-shell type and is constructed so that the walls of the cell members of the grid member are formed of a single thickness of metal to avoid tolerance problems. Within each cell member is a hydraulic spring which laterally constrains the nuclear material bearing rod which passes through each cell member against a hardstop in response to coolant flow through the cell member. This hydraulic spring is also suitable for use in a water cooled nuclear reactor. A core component constructed of, among other components, a plurality of these spacer grid members, avoids the use of a full length duct by providing spacer sleeves about the sodium tubes passing through the spacer grid members at locations between the grid members, thereby maintaining a predetermined space between adjacent grid members.

Christiansen, D.W.; Karnesky, R.A.

1989-04-04

95

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (IV)  

Microsoft Academic Search

Analysis of the TMI-2 core damage behavior has been performed with the SEFDAN code. The scope of the analysis is by the time of restarting the reactor coolant pump RCP-2B at 2 h 54 min into the accident. The analysis indicates that fuel temperature would have reached the melting point of UO2 in the upper-most part of the most central

Fumiya TANABE; Tohru SUDA

1987-01-01

96

Optical iris localization approach  

Microsoft Academic Search

This paper introduces a new iris segmentation approach based on correlation filters. The iris boundaries are detected using an original technique of composite filter called indexed composite filters. For each of the iris boundary detection, an indexed composite filter is generated using a database of circular contours corresponding to all boundary radius possible values. Experimental results of pupil and iris

O. El Kheir Abra; Esmail AHOUZI; Nawfel AZAMI; Fakhita REGRAGUI

2009-01-01

97

Control Rod Reactivity Curves for the Annular Core Research Reactor  

NASA Astrophysics Data System (ADS)

Experiments were conducted at the Annular Core Research Reactor (ACRR) to increase the fidelity of the control rod integral reactivity worth curve. This experiment series was designed to refine the integral reactivity curve used for pulse yield prediction and eliminate the need for operator compensation in the pulse setup. The experiment series consisted of delayed critical and positive period measurements with various ACRR cavity configurations. An improved integral reactivity worth curve for the ACRR control rods has been constructed using the positive period measurements, the delayed critical measurements, and radiation transport modeling of the reactor. A series of prompt period measurements is used to validate that the new control rod curve more accurately predicts the energy yield of the pulse operations. The new reactivity worth curve is compared with the current curve that was developed using traditional approaches.

Depriest, K. Russell; Kajder, Karen C.; Frye, Jason N.; Denman, Matthew R.

2009-08-01

98

Closed loop dynamics of in-core thermionic reactor systems.  

NASA Technical Reports Server (NTRS)

Use of a point model of an in-core thermionic converter to investigate alternative schemes for providing closed-loop reactor control. It was found that schemes based on variable-gain power regulation buffers which use the reactor current as the control variable provide complete protection from thermionic burnout and also provide a virtually constant voltage to the user. A side benefit is that the emitter temperature transients are small; even for a complete electric load drop the emitter temperature transient is less than 100 K. The current regulation scheme was selected for further study with a distributed parameter model which was developed to account for variations in thermionic and heat transfer properties along the length of a cylindrical converter. It was found that, even though the emitter temperature distribution is about 200 K along the converter length, the dynamic properties are unchanged when using the current control scheme.

Sawyer, C. D.; Boudreau, J. E.

1972-01-01

99

Replacement fuel scoping studies for the Annular Core Research Reactor  

SciTech Connect

Sandia National Laboratories Annular Core Research Reactor (ACRR) is undertaking a new mission for the Department of Energy: production of the radioisotope {sup 99}Mo used in nuclear medicine applications. Isotope production is significantly different from previous programs conducted at the ACRR that typically required high intensity, short duration pulses. The current UO{sub 2}-BeO fuel will power the initial startup phase of the production program, and can perform exceptionally well for this mission. However, this type of fuel is no longer available, commercially or otherwise. This paper presents the results of some preliminary studies of commercially available fuels.

Hays, K.; Martin, L.; Parma, E.

1995-07-01

100

Design and analysis of a thermal core for a high performance light water reactor  

Microsoft Academic Search

The High Performance Light Water Reactor is a Generation IV light water reactor concept, operated at a supercritical pressure of 25MPa with a core outlet temperature of 500C. A thermal core design for this reactor has been worked out by a consortium of Euratom member states within the 6th European Framework Program. Aiming at peak cladding temperatures of less than

T. Schulenberg; C. Marczy; J. Heinecke; W. Bernnat

2011-01-01

101

Modular assembly for supporting, straining, and directing flow to a core in a nuclear reactor  

Microsoft Academic Search

A reactor core support arrangement for supporting, straining, and providing fluid flow to the core and periphery of a nuclear reactor during normal operation. A plurality of removable inlet modular units are contained within permanent liners in the lower supporting plate of the reactor vessel lower internals. During normal operation (1) each inlet modular unit directs main coolant flow to

Pennell; William E

1977-01-01

102

Modular assembly for supporting, straining, and directing flow to a core in a nuclear reactor  

Microsoft Academic Search

A reactor core support arrangement for supporting, straining, and providing fluid flow to the core and periphery of a nuclear reactor during normal operation is described. A plurality of removable inlet modular units are contained within permanent liners in the lower supporting plate of the reactor vessel lower internals. During normal operation (1) each inlet modular unit directs main coolant

Pennell

1977-01-01

103

Lessons learned from Sandia National Laboratories' operational readiness review of the Annular Core Research Reactor  

Microsoft Academic Search

The Sandia National Laboratories (SNL) Annular Core Research Reactor (ACRR) (a Hazard Category 2 nuclear reactor facility) was defueled in early 1997 to convert the reactor core and control system to produce medical radioisotopes for the US Department of Energy (DOE) Medical Isotope Production Program. The DOE determined that an operational readiness review (ORR) per DOE 5480.31 or DOE 420.1

A. O. Bendure; J. W. Bryson

1999-01-01

104

Axial offset control of PWR nuclear reactor core using intelligent techniques  

Microsoft Academic Search

Improved load following capability is one of the main technical performances of advanced PWR (APWR). Controlling the nuclear reactor core during load following operation encounters some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking, while the core is subject to large and sharp variation of local power density during transients. Axial offset (AO) is

Mehrdad Boroushaki; Mohammad B. Ghofrani; Caro Lucas; Mohammad J. Yazdanpanah; Nasser Sadati

2004-01-01

105

Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)  

Microsoft Academic Search

This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR.

Budd

1986-01-01

106

Optimization of the core configuration design using a hybrid artificial intelligence algorithm for research reactors  

Microsoft Academic Search

To successfully carry out material irradiation experiments and radioisotope productions, a high thermal neutron flux at irradiation box over a desired life time of a core configuration is needed. On the other hand, reactor safety and operational constraints must be preserved during core configuration selection. Two main objectives and two safety and operational constraints are suggested to optimize reactor core

Afshin Hedayat; Hadi Davilu; Ahmad Abdollahzadeh Barfrosh; Kamran Sepanloo

2009-01-01

107

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.

108

Corium retention for high power reactors by an in-vessel core catcher in combination with External Reactor Vessel Cooling  

Microsoft Academic Search

If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel lower head remains intact so that relocated core materials are retained within

J. L. Rempe; D. L. Knudson; K. G. Condie; K. Y. Suh; F.-B. Cheung; S.-B. Kim

2004-01-01

109

Finite Element Analysis of Neutron Diffusion Equations and Reactor Core Fuel Management.  

National Technical Information Service (NTIS)

The mathematical formulation for operation and reactor core fuel management was studied. Using time discrimination, the eigenvalue problem of neutron diffusion equations are governed by reactivity gain in the reactor. From the distribution of neutron flux...

A. Huang Q. Huang Y. Tang Z. Tu

1988-01-01

110

Lessons Learned from Sandia National Laboratories' Operational Readiness Review of the Annular Core Research Reactor (ACRR).  

National Technical Information Service (NTIS)

The Sandia ACRR (a Hazard Category 2 Nuclear Reactor Facility) was defueled in June 1997 to modify the reactor core and control system to produce medical radioisotopes for the Department of Energy (DOE) Isotope Production Program. The DOE determined that ...

A. O. Bendure J. W. Bryson

1999-01-01

111

Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011.  

National Technical Information Service (NTIS)

Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of m...

D. W. Nigg K. A. Steuhm

2011-01-01

112

Prosthetic iris devices.  

PubMed

Congenital iris defects may usually present either as subtotal aniridia or colobomatous iris defects. Acquired iris defects are secondary to penetrating iris injury, iatrogenic after surgical excision of iris tumours, collateral trauma after anterior segment surgery, or can be postinflammatory in nature. These iris defects can cause severe visual disability in the form of glare, loss of contrast sensitivity, and loss of best corrected visual acuity. The structural loss of iris can be reconstructed with iris suturing, use of prosthetic iris implants, or by a combination of these, depending on the relative amount of residual iris stromal tissue and health of the underlying pigment epithelium. Since the first implant of a black iris diaphragm posterior chamber intraocular lens in 1994, advances in material and design technology over the last decade have led to advances in the prosthetic material, surgical technique, and instrumentation in the field of prosthetic iris implants. In this article, we review the classification of iris defects, types of iris prosthetic devices, implantation techniques, and complications. PMID:24513351

Srinivasan, Sathish; Ting, Darren S J; Snyder, Michael E; Prasad, Somdutt; Koch, Hans-Reinhard

2014-02-01

113

In-reactor testing of the closed cycle gas core reactor: The Nuclear Light Bulb concept  

NASA Astrophysics Data System (ADS)

The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (greater than 1800 s) and thrust (greater than 445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (greater than 4000 K). Analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept are described. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRE. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRE for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept.

Gauntt, R. O.; Slutz, S. A.; Harms, G. A.; Latham, T. S.; Roman, W. C.; Rodgers, R. J.

1992-10-01

114

PRIZMA predictions of in-core detection indications in the VVER-1000 reactor  

NASA Astrophysics Data System (ADS)

The paper describes calculations which were done by the PRIZMA code(1) to predict indications of in-core rhodium detectors in the VVER-1000 reactor for some core fragments with allowance for fuel and rhodium burnout.

Kandiev, Yadgar Z.; Kashayeva, Elena A.; Malyshin, Gennady N.; Modestov, Dmitry G.; Khatuntsev, Kirill E.

2014-06-01

115

A novel iris database indexing method using the iris color  

Microsoft Academic Search

In a very large iris database, the design of an effective iris indexing method reduces the computational complexity of feature matching process. This paper presents a new iris indexing technique based on the iris color for noisy iris images. The proposed method using the chrominance components achieves high indexing performance in defocused, reflection-contained and eyelid-occluded iris images. The performance measures

N. B. Puhan; N. Sudha

2008-01-01

116

Plasma core reactor simulations using RF uranium seeded argon discharges  

NASA Technical Reports Server (NTRS)

Experimental results are described in which pure uranium hexafluoride was injected into an argon-confined, steady-state, RF-heated plasma to investigate characteristics of plasma core nuclear reactors. The 80 kW (13.56 MHz) and 1.2 MW (5.51 MHz) rf induction heater facilities were used to determine a test chamber flow scheme which offered best uranium confinement with minimum wall coating. The cylindrical fused-silica test chamber walls were 5.7-cm-ID by 10-cm-long. Test conditions included RF powers of 2-85 kW, chamber pressures of 1-12 atm, and uranium hexafluoride mass-flow rates of 0.005-0.13 g/s. Successful techniques were developed for fluid-mechanical confinement of RF-heated plasmas with pure uranium hexafluoride injection.

Roman, W. C.

1976-01-01

117

Replacement fuel scoping studies for the annular core research reactor  

SciTech Connect

Sandia National Laboratories annular core research reactor (ACRR) is a candidate for a new mission for the U.S Department of Energy: production of the radioisotope {sup 99}Mo used in nuclear medicine applications. Isotope production is significantly different from previous programs conducted at the ACRR that typically required high-intensity, short-duration pulses. The current UO{sub 2}-BeO fuel would power the initial startup phase of the production program and can perform exceptionally well for this mission. However, this type of fuel is no longer available, commercially or otherwise. This paper presents results of preliminary studies of commercially available fuels for potential use in the ACRR.

Hays, K.; Martin, L.; Parma, E.J. [Sandia National Labs., Albuquerque, NM (United States)

1995-12-31

118

Evaluation of the Start-Up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Annular Core Loadings)  

Microsoft Academic Search

The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about

John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

2010-01-01

119

Core reactivity estimation in space reactors using recurrent dynamic networks  

NASA Technical Reports Server (NTRS)

A recurrent multilayer perceptron network topology is used in the identification of nonlinear dynamic systems from only the input/output measurements. The identification is performed in the discrete time domain, with the learning algorithm being a modified form of the back propagation (BP) rule. The recurrent dynamic network (RDN) developed is applied for the total core reactivity prediction of a spacecraft reactor from only neutronic power level measurements. Results indicate that the RDN can reproduce the nonlinear response of the reactor while keeping the number of nodes roughly equal to the relative order of the system. As accuracy requirements are increased, the number of required nodes also increases, however, the order of the RDN necessary to obtain such results is still in the same order of magnitude as the order of the mathematical model of the system. It is believed that use of the recurrent MLP structure with a variety of different learning algorithms may prove useful in utilizing artificial neural networks for recognition, classification, and prediction of dynamic systems.

Parlos, Alexander G.; Tsai, Wei K.

1991-01-01

120

The Iris database system  

Microsoft Academic Search

Iris is an object-oriented database management system being developed at Hewlett-Packard Laboratories [1], [3]. This videotape provides an overview of the Iris data model and a summary of our experiences in converting a computer-integrated manufacturing application to Iris. An abstract of the videotape follows.Iris is intended to meet the needs of new and emerging database applications such as office and

Bill Kent; Peter Lyngback; Samir Mathur; Kevin Wilkinson

1990-01-01

121

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

122

Iris Recognition System  

Microsoft Academic Search

In a biometric system a person is identified automatically by processing the unique features that are posed by the individual. Iris Recognition is regarded as the most reliable and accurate biometric identification system available. In Iris Recognition a person is identified by the iris which is the part of eye using pattern matching or image processing using concepts of neural

Sanchit Mahajan Rishi Gupta Neha Kak

2010-01-01

123

Improved core design of the high temperature supercritical-pressure light water reactor  

Microsoft Academic Search

A new coolant flow scheme has been devised to raise the average coolant core outlet temperature of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530C.In previous studies, the average coolant core outlet temperature

A. Yamaji; K. Kamei; Y. Oka; S. Koshizuka

2005-01-01

124

Prosthetic iris implantation for congenital, traumatic, or functional iris deficiencies  

Microsoft Academic Search

Purpose: To determine the efficacy and safety of surgical implantation of prosthetic iris devices in patients with anatomic or functional iris deficiencies.Setting: Cincinnati Eye Institute, Cincinnati, Ohio, USA.Methods: Twenty-five patients were enrolled in an interventional prospective noncomparative case series. Twenty-eight eyes had prosthetic iris diaphragm implantation for traumatic iris defects, congenital aniridia or iris coloboma, herpetic iris atrophy, surgical iris

Scott E Burk; Andrea P Da Mata; Michael E Snyder; Robert J Cionni; John S Cohen; Robert H Osher

2001-01-01

125

Development concept for a small, split-core, heat-pipe-cooled nuclear reactor  

NASA Technical Reports Server (NTRS)

There have been two main deterrents to the development of semiportable nuclear reactors. One is the high development costs; the other is the inability to satisfy with assurance the questions of operational safety. This report shows how a split-core, heat-pipe cooled reactor could conceptually eliminate these deterrents, and examines and summarizes recent work on split-core, heat-pipe reactors. A concept for a small reactor that could be developed at a comparatively low cost is presented. The concept would extend the technology of subcritical radioisotope thermoelectric generators using 238 PuO2 to the evolution of critical space power reactors using 239 PuO2.

Lantz, E.; Breitwieser, R.; Niederauer, G. F.

1974-01-01

126

Primary disassembly of Light Water Breeder Reactor modules for core evaluation (LWBR Development Program)  

Microsoft Academic Search

After successfully operating for 29,047 effective full power hours, the Light Water Breeder Reactor (LWBR) core was defueled prior to total decommissioning of the Shippingport Atomic Power Station. All nuclear fuel and much of the reactor internal hardware was removed from the reactor vessel. Non-fuel components were prepared for shipment to disposal sites, and the fuel assemblies were partially disassembled

R. J. Greenberger; E. L. Miller

1987-01-01

127

Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR  

DOEpatents

This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

Tokarz, R.D.

1981-10-27

128

??????????????????????????????????????????????????????????????????????????????????????????????? ????????????????????????????????? Comparative Core Analysis of TRIGA Reactor Using Virgin Uranium and Reprocesssing Uranium as Nuclear Fuel  

Microsoft Academic Search

The 10 MW TRIGA Reactor is a reactor using the UZrH fuel with 45 wt % of uranium in UZrH, 19.7 % enrichment. Both of virgin uranium and reprocessed uranium can be used as nuclear fuel for TRIGA reactor. The comparative of TRIGA core using both types of uranium were analyzed. From details on dimensions and compositions of the major

Chatchai Pawong; Sunanta Patrashakorn; Teerasak Veerapasapong

129

Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power  

NASA Technical Reports Server (NTRS)

The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

1991-01-01

130

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

Microsoft Academic Search

The authors summarize the results of detailed neutronic and thermal-hydraulic evaluations of the heat pipe cooled thermionic (HPTI) reactor design, identify its key design attributes, and quantify its performance characteristics. The HPTI core uses modular, liquid-metal core heat transfer assemblies to replace the liquid-metal heat transport loop employed by in-core thermionic reactor designs of the past. The nuclear fuel, power

W. R. Determan; Brian Lewis

1991-01-01

131

A fast reactor core concept using an internal blanket  

Microsoft Academic Search

A new core concept using an internal blanket, which is one type of heterogeneous core, has been developed. The core employs a disk-shaped internal blanket at the axial central region of the core. This internal blanket extends radially all the way through the core to the external blanket and is arranged so that its thickness is greater in the radial

K. Kawashima; K. Inoue; S. Kobayashi

1981-01-01

132

Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors  

Microsoft Academic Search

Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed

G. H. Conley; G. K. Cowell; C. A. Detrick; J. Kusenko; E. G. Johnson; J. Dunyak; B. K. Flanery; M. S. Shinko; R. H. Giffen; D. S. Rampolla

1979-01-01

133

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

NASA Astrophysics Data System (ADS)

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

Mohanakrishnan, P.

1989-10-01

134

Thermal-Hydraulic Characteristics of Double Flat Core HCLWR (High Conversion Light Water Reactor).  

National Technical Information Service (NTIS)

A thermal-hydraulic characteristics of double flat core high conversion light water reactor (HCLWR) is described. The concept of flat core proposed by Ishiguro et al. is to achieve negative void reactivity coefficient in tight lattice core, and at the sam...

J. Sugimoto T. Iwamura T. Okubo Y. Murao

1989-01-01

135

Using ex-core neutron detectors to estimate fuel quantities in the reactor vessel lower head  

Microsoft Academic Search

During the accident at Three Mile Island Unit 2 (TMI-2), a significant mass of core debris slumped to the bottom head of the reactor vessel. Defueling activities caused more core debris to relocate to the lower head region. The variations in the ex-core neutron detector, or source range monitor (SRM), readings gave evidence of this effect of defueling activity. Between

R. Rainisch; V. Fricke

1988-01-01

136

Fault current limiter-predominantly resistive behavior of a BSCCO shielded-core reactor  

SciTech Connect

Tests were conducted to determine the electrical and magnetic characteristics of a superconductor shielded core reactor (SSCR). The results show that a closed-core SSCR is predominantly a resistive device and an open-core SSCR is a hybrid resistive/inductive device. The open-core SSCR appears to dissipate less than the closed-core SSCR. However, the impedance of the open-core SSCR is less than that of the closed-core SSCR. Magnetic and thermal diffusion are believed to be the mechanism that facilitates the penetration of the superconductor tube under fault conditions.

Ennis, M. G.; Tobin, T. J.; Cha, Y. S.; Hull, J. R.

2000-06-30

137

Reactor physics calculations for {sup 99}Mo production at the annular core research reactor  

SciTech Connect

The Isotope Production and Distribution Program at the U.S. Department of Energy has designated Sandia National Laboratories (SNL) as the most appropriate facility for the production of {sup 99}Mo, a radioisotope whose daughter, {sup 99m}Tc, is used in more than 36,000 medical procedures per day in the United States and is considered to be a vital medical diagnostic and treatment tool. The isotope would be produced at SNL using the annular core research reactor (ACRR) facility and collocated hot cell facility. The {sup 99}Mo would be produced using the fission process by irradiating {open_quotes}targets{close_quotes} coated with {sup 235}U in the form of highly enriched U{sub 3}O{sub 8}. After {approximately}7 days of continuous irradiation in the ACRR, a target would be re- moved from the reactor core for processing. The isotope would be extracted by chemically precipitating the molybdenum using the {open_quotes}Cintichem{close_quotes} process and would be shipped to the various pharmaceutical companies by commercial or chartered airline.

Parma, E.J. [Sandia National Labs., Albuquerque, NM (United States)

1995-12-31

138

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

NASA Technical Reports Server (NTRS)

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

Determan, W. R.; Lewis, Brian

1991-01-01

139

Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data  

SciTech Connect

The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

Jung, Y. S.; Joo, H. G. [Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of)] [Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of); Yoon, J. I. [KEPCO Nuclear Fuel, 1047 Daedukdae-ro, Yuseong-gu, Daejeon (Korea, Republic of)] [KEPCO Nuclear Fuel, 1047 Daedukdae-ro, Yuseong-gu, Daejeon (Korea, Republic of)

2013-07-01

140

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

SciTech Connect

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

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

2009-04-01

141

Iris Recognition Based on DLDA  

Microsoft Academic Search

Iris feature extraction is very important for an iris recognition system. This paper focuses on iris feature extraction. In this paper we propose direct linear discriminant analysis (DLDA) which combines with wavelet transform to extract iris feature. In our method, firstly, we apply wavelet decomposition to the normalized iris image whose size is 64times256 and just choose the coefficients of

Chengqiang Liu; Mei Xie

2006-01-01

142

Turbulence coefficients and stability studies for the coaxial flow or dissimiliar fluids. [gaseous core nuclear reactors  

NASA Technical Reports Server (NTRS)

Analytical investigations of fluid dynamics problems of relevance to the gaseous core nuclear reactor program are presented. The vortex type flow which appears in the nuclear light bulb concept is analyzed along with the fluid flow in the fuel inlet region for the coaxial flow gaseous core nuclear reactor concept. The development of numerical methods for the solution of the Navier-Stokes equations for appropriate geometries is extended to the case of rotating flows and almost completes the gas core program requirements in this area. The investigations demonstrate that the conceptual design of the coaxial flow reactor needs further development.

Weinstein, H.; Lavan, Z.

1975-01-01

143

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

144

Production of {sup 99}Mo at the annular core research reactor-recent calculative results  

SciTech Connect

Significant progress has been made over the past year in understanding the chemistry and processing challenges associated with {sup 99}Mo production using Cintichem-type targets. Targets fabricated at Los Alamos National Laboratory have been successfully irradiated in fuel element locations at the Annular Core Research Reactor (ACRR) and processed at the Sandia Hot Cell Facility. The next goal for the project is to remove the central cavity experiment tube from the reactor core, allowing for the irradiation of up to 37 targets. After the in-core work is complete, the reactor will be capable of producing significant quantities of {sup 99}Mo.

Parma, E.J. [Sandia National Labs., Albuquerque, NM (United States)

1997-12-01

145

Mo-99 production at the Annular Core Research Reactor - recent calculative results  

SciTech Connect

Significant progress has been made over the past year in understanding the chemistry and processing challenges associated with {sup 99}Mo production using Cintichem type targets. Targets fabricated at Los Alamos National Laboratory have been successfully irradiated in fuel element locations at the Annular Core Research Reactor (ACRR) and processed at the Sandia Hot Cell Facility. The next goal for the project is to remove the central cavity experiment tube from the reactor core, allowing for the irradiation of up to 37 targets. After the in-core work is complete, the reactor will be capable of producing significant quantities of {sup 99}Mo.

Parma, E.J.

1997-11-01

146

Corium Retention for High Power Reactors by An In-Vessel Core Catcher in Combination with External Reactor Vessel Cooling  

SciTech Connect

If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel lower head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe Pressurized Water Reactor (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for in-vessel retention (IVR), resulted in the United States Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing Light Water Reactors (LWRs). Accordingly, IVR of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors. However, it is not clear that currently-proposed methods to achieve ERVC will provide sufficient heat removal for higher power reactors. A USKorean International Nuclear Energy Research Initiative (INERI) project has been initiated in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) will determine if IVR is feasible for reactors up to 1500 MWe. This paper summarizes results from the first year of this 3-year project.

Joy L. Rempe; D. L. Knudson; K. G. Condie; K. Y. Suh; F. -B. Cheung; S. -B. Kim

2004-05-01

147

Burnup concept for a long-life fast reactor core using MCNPX.  

SciTech Connect

This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

2013-02-01

148

Experimental capabilities of the Annular-Core Research Reactor  

SciTech Connect

The ACRR is an extremely flexible test reactor that is employed for a variety of irradiation programs from radiation effects to reactor safety research. The BeO-UO/sub 2/ fuel element design is a significant improvement in pulse reactor technology. The fuel provides a large volumetric enthalpy which results in high energy deposition with modest temperature rise.

Reuscher, J.A.; Schmidt, T.R.; Pickard, P.S.

1981-01-01

149

The thin film microwave iris  

NASA Technical Reports Server (NTRS)

Development of waveguide iris for microwave coupling applications using thin film techniques is discussed. Production process and installation of iris are described. Iris improves power transmission properties of waveguide window.

Ramey, R. L.; Landes, H. S.; Manus, E. A.

1972-01-01

150

GNSS Measurements and IRI  

NASA Astrophysics Data System (ADS)

The International Reference Ionosphere (IRI) is an empirical model of the most important ionospheric parameters including electron and ion densities and temperatures and the Total Electron Content (TEC). IRI is a joint project of the Committee on Space Research (COSPAR) and the International Union of Radio Science (URSI) and is based on the majority of the available ground and space data. So fare Global Navigation Satellite System (GNSS) data have not been used for the model development/improvement process. Recent comparisons between GNSS measurements and IRI have highlighted some of the areas where discrepancies are observed and where GNSS data are expected to be a significant resource for IRI modeling. We will discuss ways in which GNSS results could help the IRI improvement process in critical areas. A number of studies have shown the use of GNSS data for updating IRI to near-real time conditions. We will review the different efforts and point out successes and shortcomings. A special IRI Workshop is being convened for early May to discuss in depth the topic of the Real-Time IRI and come up with recommendations. Finally we will also highlight IRI's role in improving and testing the various techniques for deducing TEC from GNSS measurements.

Bilitza, D.; Schmidt, M.

2009-04-01

151

MELCOR adaptation and validation for modeling of N Reactor core phenomena  

SciTech Connect

MELCOR has been adapted for use in modeling the N Reactor core as a part of the recently completed N Reactor probabilistic risk assessment. Significant adaptation of MELCOR was required because of the horizontal, water cooled, graphite-moderated nature of the N Reactor core. The generation and verification of the revised N Reactor core model are described in this paper. A hydrogen production and core damage benchmark calculation is presented in which all significant parameters calculated by MELCOR agreed with those in the reference calculation to within approximately 10%. The reference calculation required many CRAY CPU hours, while the MELCOR calculation was completed in less than 20 CPU minutes on a VAX 8700. 7 refs., 3 figs., 1 tab.

Wyss, G.D.; Summers, R.M.; Miller, L.A.

1990-01-01

152

Preparations to ship the TMI2 damaged reactor core  

Microsoft Academic Search

The March 1979 accident at Three Mile Island Unit 2 (TMI-2) resulted in a severely damaged core. Entries into that core using various tools and inspection devices have shown a significant void, large amounts of rubble, partially intact fuel assemblies, and some resolidified molten materials. The removal and disposition of that core has been of considerable public, regulatory, and governmental

R. C. Schmitt; G. J. Quinn

1985-01-01

153

Fast spectrum space reactor sizing code for calandria-type cores (CORSCO Code). [Li  

SciTech Connect

The CORSCO code rapidly sizes reactor cores that have calandria-type geometry. The fuel configuration modeled is a large ceramic zone that contains numerous small cylindrical coolant channels spaced apart with a triangular pitch. A minimum reactor weight is obtained for a fixed set of constraints (peak fuel temperature, peak coolant velocity, etc.) by obtaining a unique solution to a set of five thermal/hydraulic equations, as well as a required excess reactivity which is specified by a core size dependent one-group criticality expression. Typical results are shown for a W-Re/UN cermet-fueled, lithium-cooled space reactor over a power range of 25 to 100 MWt. Reactor sensitivity coefficients are also shown for changes in reactor weight and number of coolant channels due to changes in core thermal/hydraulic constraints.

Specht, E.R.; Villalobos, A. (Rockwell International, Rocketdyne Division, 6633 Canoga Avenue, HB23, Canoga Park, California (USA))

1991-01-10

154

Design and proposed utilization of the Sandia Annular Core Research Reactor (ACRR)  

Microsoft Academic Search

The Sandia ACRR became operational in 1978 and currently serves as the major in-pile fast reactor safety test facility for the US Nuclear Regulatory Commission. The ACRR is an upgrade of the Annular Core Pulse Reactor (ACPR) with the installation of a new flexible control system and a core of uniquely designed BeO-UO fuel elements for increasing the neutron fluence

J. V. Walker; J. A. Reuscher; P. S. Pickard

1979-01-01

155

A STUDY OF CORE FUEL SYSTEMS FOR A FAST BREEDER POWER REACTOR  

Microsoft Academic Search

The first phase of a program aimed toward the development of materials ;\\u000a and a core-subassembly design for the second core of the Fermi Reactor is ;\\u000a outlined. The ground rules established by APDA for the study were based upon the ;\\u000a performance requirements of the reactor plant with some modification to permit ;\\u000a hlgher power generation and upon a

Fawcett; S. L. ed

1957-01-01

156

Method of and apparatus for measuring the power distribution in nuclear reactor cores  

SciTech Connect

The invention disclosed is the method of exact calibration of gamma ray detectors called gamma thermometers prior to acceptance for installation into a nuclear reactor core. This exact calibration increases the accuracy of determining the power distribution in the nuclear reactor core. The calibration by electric resistance heating of the gamma thermometer consists of applying an electric current along the controlled heat path of the gamma thermometer and then measuring the temperature difference along this controlled heat path as a function of the amount of power generated by the electric resistance heating. Then, after the gamma thermometer is installed into the nuclear reactor core and the reactor core is operating at power producing conditions, the gamma ray heating of the detector produces a temperature difference along the controlled heat path. With the knowledge of this temperature difference, the calibration characteristic determined by the prior electric resistance heating is employed to accurately determine the local rate of gamma ray heating. The accurate measurement of the gamma heating rate at each location of a set of locations throughout the nuclear reactor core is the basis for accurately determining the power distribution within the nuclear reactor core.

Leyse, R.H.

1983-07-12

157

Core\\/concrete interaction scenarios for VVER-1000-type reactors  

Microsoft Academic Search

In continuation of the Austrian research program to investigate the safety behavior of VVER-type reactors, several accident scenarios for VVER-1000 reactors were analyzed. These reactors are equipped with four loops without isolation valves, horizontal steam generators, and hexagonal fuel assemblies. Safety features include a containment structure with spray-type steam suppression. The influence of this spray system on source-term behavior has

G. Sdouz; G. Sonneck

1994-01-01

158

Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements  

Microsoft Academic Search

The Neutron Radiography (NRAD) reactor is a 250-kW TRIGA-(Training, Research, Isotope Production, General Atomics)-conversion-type reactor at the Idaho National Laboratory; it is primarily used for neutron radiography analysis of irradiated and unirradiated fuels and materials. The NRAD reactor was converted from HEU to LEU fuel with 60 fuel elements and brought critical on March 31, 2010. This configuration of the

J. D. Bess; T. L. Maddock; M. A. Marshall

2011-01-01

159

Selecting a MAPLE research reactor core for 1-10 mW operation.  

National Technical Information Service (NTIS)

The MAPLE class of research reactors is designed so that a single reactor concept can satisfy a wide range of practical applications. This paper reports the results of physics studies performed on a number of potential core configurations fuelled with eit...

H. J. Smith M. F. Roy P. A. Carlson

1986-01-01

160

Gas Core Reactor-MHD Power System with Cascading Power Cycle  

Microsoft Academic Search

The US Department of Energy initiative Gen-IV aim is to produce an entire nuclear energy production system with next generation features for certification before 2030. A Generation 4 capable system must have superior sustainability, safety and reliability, and economic cost advantages in comparison with third generation light water reactors. A gas core reactor (GCR) with magnetohydrodynamic (MHD) power converter and

Blair M. Smith; Samim Anghaie; Travis W. Knight

2002-01-01

161

Development of Liquid-Vapor Core Reactors with MHD Generator for Space Power and Propulsion Applications  

Microsoft Academic Search

Any reactor that utilizes fuel consisting of a fissile material in a gaseous state may be referred to as a gaseous core reactor (GCR). Studies on GCRs have primarily been limited to the conceptual phase, mostly due to budget cuts and program cancellations in the early 1970's. A few scientific experiments have been conducted on candidate concepts, primarily of static

Samim Anghaie

2002-01-01

162

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

163

Performance characteristics of the annular core research reactor fuel motion detection system  

Microsoft Academic Search

Recent proof tests have shown that the annular core research reactor (ACRR) fuel motion detection system has reached its design goals of providing high temporal and spatial resolution pictures of fuel distributions in the ACRR. The coded aperture imaging system (CAIS) images the fuel by monitoring the fission gamma rays from the fuel that pass through collimators in the reactor

J. G. Kelly; K. T. Stalker

1983-01-01

164

The Annular Core Research Reactor (ACRR) postulated limiting event initial and building source terms  

Microsoft Academic Search

As part of the update of the Safety analysis Report (SAR) for the Annular Core Research Reactor (ACRR), operational limiting events under the category of inadvertent withdrawal of an experiment while at power or during a power pulse were determined to be the most limiting event(s) for this reactor. This report provides a summary of the assumptions, modeling, and results

Restrepo

1992-01-01

165

ACRR (Annular Core Research Reactor) fission product release tests: ST1 and ST2  

Microsoft Academic Search

Two experiments (ST-1 and ST-2) have been performed in the Annular Core Research Reactor (ACER) at Sandia National Laboratories (SNLA) to obtain time-resolved data on the release of fission products from irradiated fuels under light water reactor (LWR) severe accident conditions. Both experiments were conducted in a highly reducing environment at maximum fuel temperatures of greater than 2400 K. These

M. D. Allen; H. W. Stockman; K. O. Reil; A. J. Grimley; W. J. Camp

1988-01-01

166

Operational performance of the three bean salad control algorithm on the ACRR (Annular Core Research Reactor)  

Microsoft Academic Search

Experimental tests on the Annular Core Research Reactor have confirmed that the Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute. 3 refs.,

R. M. Ball; J. J. Madaras; F. R. Jr. Trowbridge; D. G. Talley; E. J. Jr. Parma

1991-01-01

167

Lessons Learned from Sandia National Laboratories' Operational Readiness Review of the Annular Core Research Reactor (ACRR)  

Microsoft Academic Search

The Sandia ACRR (a Hazard Category 2 Nuclear Reactor Facility) was defueled in June 1997 to modify the reactor core and control system to produce medical radioisotopes for the Department of Energy (DOE) Isotope Production Program. The DOE determined that an Operational Readiness Review (ORR) was required to confirm readiness to begin operations within the revised safety basis. This paper

Albert O. Bendure; James W. Bryson

1999-01-01

168

Detailed Core Design and Flow Coolant Conditions for Neutron Flux Maximization in Research Reactors  

Microsoft Academic Search

Following the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, we have already developed an asymmetric compact cylindrical core with an inner and outer reflector. The goal was to maximize the maximum thermal flux to power ratio, which is a desirable characteristic of a modern research reactor. This design, for a 10 MW power, was

F. E. Teruel; Rizwan Uddin

2006-01-01

169

Optimizing a three-element core design for the Advanced Neutron Source Reactor  

Microsoft Academic Search

Source of neutrons in the proposed Advanced Neutron Source facility is a multipurpose research reactor providing 5-10 times the flux, for neutron beams, of the best existing facilities. Baseline design for the reactor core, based on the ``no new inventions`` rule, was an assembly of two annular fuel elements similar to those used in the Oak Ridge and Grenoble high

1995-01-01

170

Conceptual Design Study of 180 MWt Small-Sized Reduced-Moderation Water Reactor Core  

Microsoft Academic Search

Conceptual design of a Small-sized Reduced-Moderation Water Reactor (S-RMWR) core, which has the thermal output of 180 MW, the conversion ratio of 1.0 and the void reactivity coefficient of negative value, has been constructed. S-RMWR is a technology demonstration reactor which also conducts material and fuel testing for commercial use of Reduced-Moderation Water Reactor (RMWR) in large-scale power plants. It

Yoshihiro NAKANO; Tsutomu OKUBO; Sadao UCHIKAWA

2006-01-01

171

FREC-II: An upgrade to SNL's annular core research reactor  

Microsoft Academic Search

The fuel-ringed external cavity, version II (FREC-II), is a recent upgrade to the annular core research reactor (ACRR) at Sandia National Laboratories (SNL). The FREC-II is neutronically coupled to the ACRR, a 2-MW steady-state\\/300-MJ pulse reactor used for a variety of simulation experiments in areas such as reactor safety and weapons effects. The FREC-II was designed to provide a large-volume

R. A. Rubio; P. J. Cooper; J. F. Schulze; J. W. Bryson; F. M. Morris; F. R. Trowbridge; T. R. Schmidt

1989-01-01

172

100-KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT  

SciTech Connect

On December 15-16, 2009, a 100-KE Reactor Core Removal Project Alternative Analysis Workshop was conducted at the Washington State University Consolidated Information Center, Room 214. Colburn Kennedy, Project Director, CH2M HILL Plateau Remediation Company (CHPRC) requested the workshop and Richard Harrington provided facilitation. The purpose of the session was to select the preferred Bio Shield Alternative, for integration with the Thermal Shield and Core Removal and develop the path forward to proceed with project delivery. Prior to this workshop, the S.A. Robotics (SAR) Obstruction Removal Alternatives Analysis (565-DLV-062) report was issued, for use prior to and throughout the session, to all the team members. The multidisciplinary team consisted ofrepresentatives from 100-KE Project Management, Engineering, Radcon, Nuclear Safety, Fire Protection, Crane/Rigging, SAR Project Engineering, the Department of Energy Richland Field Office, Environmental Protection Agency, Washington State Department of Ecology, Defense Nuclear Facility Safety Board, and Deactivation and Decommission subject matter experts from corporate CH2M HILL and Lucas. Appendix D contains the workshop agenda, guidelines and expectations, opening remarks, and attendance roster going into followed throughout the workshop. The team was successful in selecting the preferred alternative and developing an eight-point path forward action plan to proceed with conceptual design. Conventional Demolition was selected as the preferred alternative over two other alternatives: Diamond Wire with Options, and Harmonic Delamination with Conventional Demolition. The teams preferred alternative aligned with the SAR Obstruction Removal Alternative Analysis report conclusion. However, the team identified several Path Forward actions, in Appendix A, which upon completion will solidify and potentially enhance the Conventional Demolition alternative with multiple options and approaches to achieve project delivery. In brief, the Path Forward was developed to reconsider potential open air demolition areas; characterize to determine if any zircaloy exists, evaluate existing concrete data to determine additional characterization needs, size the new building to accommodate human machine interface and tooling, consider bucket thumb and use ofshape-charges in design, and finally to utilize complex-wide and industry explosive demolition lessons learned in the design approach. Appendix B documents these results from the team's use ofValue Engineering process tools entitled Weighted Analysis Alternative Matrix, Matrix Conclusions, Evaluation Criteria, and Alternative Advantages and Disadvantages. These results were further supported with the team's validation of parking-lot information sheets: memories (potential ideas to consider), issues/concerns, and assumptions, contained in Appendix C. Appendix C also includes the recorded workshop flipchart notes taken from the SAR Alternatives and Project Overview presentations. The SAR workshop presentations, including a 3-D graphic illustration demonstration video have been retained in the CHPRC project file, and were not included in this report due to size limitations. The workshop concluded with a round robin close-out where each member was engaged for any last minute items and meeting utility. In summary, the team felt the session was value added and looked forward to proceeding with the recommended actions and conceptual design.

HARRINGTON RA

2010-01-15

173

Application of advanced core process monitoring procedures in German power reactors  

Microsoft Academic Search

The nuclear reactor core design and the nuclear fuel management have been changed remarkable during the last few years. This development was initiated by increasing costs for the fuel recycling and nuclear waste storage. The fuel material, the fuel pellet fabrication, the fuel assembly structure and the core composition have been varied to get an effective fuel exploitation. Based on

J. Pohlus

2003-01-01

174

Vaporization of core materials in postulated severe light water reactor accidents  

Microsoft Academic Search

The vaporization of core materials other than fission products during a postulated severe light water reactor accident is treated by chemical thermodynamics. The core materials considered were (a) the control rod materials, silver, cadmium, and indium; (b) the structural materials, iron, chromium, nickel, and manganese; (c) cladding material, zirconium and tin; and (d) the fuel, uranium oxide. Thermodynamic data employed

D. Cubicciotti; B. R. Sehgal

1984-01-01

175

Steady State Thermal Hydraulic Analysis of LEU Cores for Pakistan Research Reactor-1.  

National Technical Information Service (NTIS)

Maximum operating power levels of the first high power and equilibrium LEU cores for PARR-1 have been assessed. The criterion followed is that nucleate boiling should not commence, at any point in the core, when reactor power approaches overpower limiting...

L. A. Khan I. H. Bokhari K. M. Akhtar S. Pervez

1991-01-01

176

Lunar Nuclear Power Plant With Solid Core Reactor, Heatpipes and Thermoelectric Conversion  

Microsoft Academic Search

This is a lunar nuclear power plant with the advantages of minimum mass, with no moving parts, no pumped liquid coolant, a solid metal rugged core, with no single point of failure. The electrical output is 100 kilowatts with a 500 kilowatt thermal reactor. The thermoelectric converters surround the potassium heatpipes from the core and water heatpipes surround the converter

Edwin D. Sayre; Peter J. Ring; Neil Brown; Norbert B. Elsner; John C. Bass

2008-01-01

177

Nuclear Criticality Calculations on a Spherically Symmetric Gaseous-Core Reactor.  

National Technical Information Service (NTIS)

The nuclear criticality calculations are performed on a two-region, spherically symmetric gaseous-core reactor. The core is fueled with U235 vapor and is externally moderated by various thicknesses of graphite at 4000K. A one-group (thermal) transport ana...

J. R. Rec

1964-01-01

178

High-performance core concept study for High Temperature Engineering Test Reactor.  

National Technical Information Service (NTIS)

A study of the high-performance core concept was carried out for the High Temperature Engineering Test Reactor (HTTR) from the viewpoint of nuclear and thermal-hydraulics design. The high-performance core should have high irradiation test capability and p...

K. Yamashita M. Nakano N. Nojiri N. Fujimoto K. Sawa

1997-01-01

179

Core design investigation for a SUPERSTAR small modular lead-cooled fast reactor demonstrator  

Microsoft Academic Search

In this paper a preconceptual neutronics design study for a SUstainable Proliferation-resistance Enhanced Refined Secure Transportable Autonomous Reactor (SUPERSTAR) demonstrator is presented. The main goal of achieving the highest realistic power level limited by natural circulation and transportability, while providing energy security and proliferation resistance thanks to a long core lifetime design has been satisfactorily attained. A preliminary core configuration

S. Bortot; A. Moisseytsev; J. J. Sienicki; Carlo Artioli

180

2ND Reactor Core of the NS Otto Hahn. Design, Operation Experience, Developments.  

National Technical Information Service (NTIS)

Details of the design of the 2nd reactor core are given, followed by a brief report summarising the operating experience gained with this 2nd core, as well as by an evaluation of measured data and statements concerning the usefulness of the knowledge gain...

H. J. Manthey H. Kracht

1979-01-01

181

Sensitivity of pressurized water reactor source term inventory and decay power to core management parameters  

Microsoft Academic Search

ORIGEN2 was used to develop a data base of pressurized water reactor isotopic concentrations at various times after discharge with core burnup, specific power, enrichment, and neutron spectrum as variables. Results were analyzed to determine source term sensitivity to core management. Fuel rod power history was found to have an important effect on the source term. Activity and decay power

J. K. Wheeler; A. Sesonske

1986-01-01

182

Preparations to load, transport, receive, and store the damaged TMI2 (Three Mile Island) reactor core  

Microsoft Academic Search

The March 1979 incident at the Three Mile Island Nuclear Power Station (TMI) which damaged the core of the Unit 2 reactor resulted in numerous scientific and technical challenges. Some of those challenges involve removing, packaging, and transporting the core debris to the Idaho National Engineering Laboratory (INEL) for storage, examination, and preparation for final disposal. This paper highlights preparations

H. W. Reno; R. C. Schmitt; G. J. Quinn; A. L. Jr. Ayers; B. J. Jr. Lilburn; D. L. Uhl

1986-01-01

183

Review of experimental results of light water reactor core melt progression  

Microsoft Academic Search

This paper reports on results from integral-effects core melt progression experiments and from the examination of the damaged core of the Three Mile Island Unit 2 (TMI-2) reactor which are reviewed to gain insight on key severe accident phenomena. The experiments and the TMI-2 accident represent a wide variety of conditions and physical scales, yet several important phenomena appear to

R. R. Hobbins; D. A. Petti; O. J. Osetek; D. L. Hagrman

1991-01-01

184

An evolutionary approach for a compact-split-core reactor.  

NASA Technical Reports Server (NTRS)

An economical approach for advanced reactor power development is presented, and systems that result from the several stages of this plan are described. The development starts with a highly modularized heat-pipe, radioisotopic design and evolves into a low-specific-weight, high-performance reactor system.

Breitwieser, R.; Lantz, E.

1973-01-01

185

Reactor Physics Parameters of Alternate Fueled FBR Core Designs.  

National Technical Information Service (NTIS)

Nuclear non-proliferation considerations have resulted in renewed interest in the thorium fuel cycle. Reactor physics parameters of a typical 1200-MW(e) Fast Breeder Reactor (FBR) design were compared when U-233 is substituted for Pu as a fissile fuel and...

D. R. Haffner, R. W. Hardie

1977-01-01

186

Preparations to ship the TMI-2 damaged reactor core  

SciTech Connect

The March 1979 accident at Three Mile Island Unit 2 (TMI-2) resulted in a severely damaged core. Entries into that core using various tools and inspection devices have shown a significant void, large amounts of rubble, partially intact fuel assemblies, and some resolidified molten materials. The removal and disposition of that core has been of considerable public, regulatory, and governmental interest for some time. In a contractual agreement between General Public Utility Nuclear (GPUN) and the US Department of Energy (DOE), DOE has agreed to accept the TMI-2 core for interim storage at the Idaho National Engineering Laboratory (INEL), conduct research on fuel and materials of the core, and eventually dispose of the core either by processing or internment at the national repository. GPUN has removed various samples of material from the core and was scheduled to begin extensive defueling operations in September 1985. EG and G Idaho, Inc. (EG and G), acting on behalf of DOE, is responsible for transporting, receiving, examining, and storing the TMI-2 core. This paper addresses the preparations to ship the core to INEL, which is scheduled to commence in March 1986.

Schmitt, R.C.; Quinn, G.J.

1985-11-01

187

Iris Recognition Using Wavelet  

Microsoft Academic Search

Biometric systems are getting more attention in the present era. Iris recognition is one of the most secure and authentic among the other biometrics and this field demands more authentic, reliable and fast algorithms to implement these biometric systems in real time. In this paper, an efficient localization technique is presented to identify pupil and iris boundaries using histogram of

K. Masood; D. M. Y. Javed; A. Basit

2007-01-01

188

In-reactor testing of the closed cycle gas core reactor-the nuclear light bulb concept  

Microsoft Academic Search

The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (>~1800 s) and thrust (>~445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (~4000

Randall O. Gauntt; Stephen A. Slutz; Gary A. Harms; Thomas S. Latham; Ward C. Roman; Richard J. Rodgers

1993-01-01

189

In-reactor testing of the closed cycle gas core reactorthe nuclear light bulb concept  

Microsoft Academic Search

The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (?1800 s) and thrust (?445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (?4000

Randall O. Gauntt; Stephen A. Slutz; Gary A. Harms; Thomas S. Latham; Ward C. Roman; Richard J. Rodgers

1993-01-01

190

Control rod worth and related nuclear characteristics of an axially heterogeneous liquidmetal fast breeder reactor core  

Microsoft Academic Search

An axially heterogeneous core (AHC) concept is applied to a 1000-MW(electric)-class tank-type liquidmetal fast breeder reactor (LMFBR). This AHC is characterized by a disk-shaped internal blanket with a radial thickness adjustment at the core midplane. The nuclear characteristics connected with control rod worth of the AHC are analyzed and compared with those of a homogeneous core (HOC) of the same

K. Kawashima; T. Inayaki; K. Inoue; K. Kaneto

1985-01-01

191

IN-CORE MEASUREMENTS OF REACTORS INTERNALS VIBRATIONS BY USE OF ACCELEROMETERS AND NEUTRON DETECTORS  

Microsoft Academic Search

A miniature biaxial accelerometer was developed for vibration measurements in radioactive environments. The sensor is small enough to be assembled in Self-powered Neutron Detector (SPND) in-core instrument strings of PWRs or to be inserted into the travelling in-core probe system of BWRs. Two accelerometers were installed inside of the core of an operating power reactor (PWR, 350 MWel). The vibrations

J. Runkel; E. Laggiard; D. Stegemann; P. Heidemann; R. Blaser; F. Schmid; H. Reinmann

192

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core  

SciTech Connect

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power.

Kotas, J.F.; Stroh, K.R.

1983-01-01

193

Fuel cycles and advanced core designs for the gas cooled fast-breeder reactor  

SciTech Connect

This report summarizes the fuel cycle and advanced core design analysis for the gas-cooled fast breeder reactor (GCFR) performed in conjunction with the Nonproliferation Alternative Systems Assessment Program (NASAP) and the International Nuclear Fuel Cycle Evaluation (INFCE) between 1976 and 1980. The report contrasts traditional fast reactor fuel cycles (plutonium/uranium) with alternative (uranium/thorium) cycles in an effort to define fuel systems which might reduce nuclear weapons proliferation risks without incurring extreme resource or economic disadvantages. It studies symbiotic reactor systems involving fast and thermal reactors and provides basic GCFR mass flow information for NASAP and INFCE. It defines an improved fuel cycle performance index (i.e., energy potential) for reactors and reactor systems. This improved index is combined with sensitivity studies of core materials and configurations to select the basic core and primary system characteristics for future core designs. This report establishes and characterizes advanced designs which may represent targets for future commercial-size fast reactors.

Hamilton, C.J.; Perkins, R.G.

1982-07-01

194

Nuclear reactor with low-level core coolant intake  

DOEpatents

A natural-circulation boiling-water reactor has skirts extending downward from control rod guide tubes to about 10 centimeters from the reactor vessel bottom. The skirts define annular channels about control rod drive housings that extend through the reactor vessel bottom. Recirculating water is forced in through the low-level entrances to these channels, sweeping bottom water into the channels in the process. The sweeping action prevents cooler water from accumulating at the bottom. This in turn minimizes thermal shock to bottom-dwelling components as would occur when accumulated cool water is swept away and suddenly replaced by warmer water.

Challberg, Roy C. (Livermore, CA); Townsend, Harold E. (Campbell, CA)

1993-01-01

195

Core subchannel thermal-hydraulic analysis methods and critical heat-flux margin in a light-water breeder reactor  

Microsoft Academic Search

Analyis methods and results are described for critical heat flux (CHF) performance margin in the core of an advanced light water moderated breeder reactor design concept. The 1000 MWe breeder reactor is basically like a large commercial pressurized water reactor (PWR); however, a number of core design features require special consideration with regard to predicting margin to CHF design limits.

R. Misiewicz; S. J. Kast; L. H. Wunderlich

1983-01-01

196

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

197

Some Nuclear Calculations of U-235-D2O Gaseous-Core Cavity Reactors  

NASA Technical Reports Server (NTRS)

The results of a multigroup, diffusion theory study of spherical gaseous-core cavity reactors are presented in this report. The reactor cavity of gaseous U235 is enclosed by a region of hydrogen gas and is separated from an external D2O moderator-reflector by a zirconium structural shell. Some cylindrical reactors are also investigated. A parametric study of spherical reactors indicates that, for the range of variables studied, critical mass increases as: (1) Fuel region is compressed within the reactor cavity, (2) moderator thickness is decreased, (3) structural shell thickness is increased, and (4) moderator temperature is increased. A buckling analogy is used to estimate the critical mass of fully reflected cylindrical reactors from spherical results without fuel compression. For a reactor cavity of a 120-centimeter radius uniformly filled with fuel, no structural shell, a moderator temperature of 70 F, and a moderator thickness of 100 centimeters, the critical mass of a spherical reactor is 3.1 kilograms while that of a cylinder with a length-to-diameter ratio of 1.0 (L/D = 1) is approximately 3.8 kilograms and, with L/D = 2, 5.9 kilograms. For the range of variables considered for U235-D2O gaseous-core cavity reactors, the systems are characterized by 95 to 99 percent thermal absorptions, with the flux reaching a maximum in the moderator about 10 to 15 centimeters from the reactor cavity.

Ragsdale, Robert G.; Hyland, Robert E.

1961-01-01

198

Annular core research reactor high flux neutron radiography facility.  

National Technical Information Service (NTIS)

Sandia National Laboratories (SNL) has been performing neutron radiography since 1964. The radiography facilities have evolved from an aperture in a radiation exposure room in the now retired Sandia Engineering Reactor to a divergent collimator radiograph...

F. M. McCrory J. G. Kelly M. E. Vernon D. A. Tichenor

1990-01-01

199

Lessons learned from Sandia National Laboratories' operational readiness review of the Annular Core Research Reactor  

SciTech Connect

The Sandia National Laboratories (SNL) Annular Core Research Reactor (ACRR) (a Hazard Category 2 nuclear reactor facility) was defueled in early 1997 to convert the reactor core and control system to produce medical radioisotopes for the US Department of Energy (DOE) Medical Isotope Production Program. The DOE determined that an operational readiness review (ORR) per DOE 5480.31 or DOE 420.1 was required to confirm the readiness of management systems, personnel, and the physical plant to refuel the reactor and begin operations within the revised safety basis. DOE stated that this was the first reactor ORR conducted within the Complex. The authors address the lessons learned from the ACRR ORR, emphasizing cost savings and the use of the ORR to confirm authorization-basis implementation.

Bendure, A.O.; Bryson, J.W.

1999-07-01

200

Heat exchanger for reactor core and the like  

DOEpatents

A compact bayonet tube type heat exchanger which finds particular application as an auxiliary heat exchanger for transfer of heat from a reactor gas coolant to a secondary fluid medium. The heat exchanger is supported within a vertical cavity in a reactor vessel intersected by a reactor coolant passage at its upper end and having a reactor coolant return duct spaced below the inlet passage. The heat exchanger includes a plurality of relatively short length bayonet type heat exchange tube assemblies adapted to pass a secondary fluid medium therethrough and supported by primary and secondary tube sheets which are releasibly supported in a manner to facilitate removal and inspection of the bayonet tube assemblies from an access area below the heat exchanger. Inner and outer shrouds extend circumferentially of the tube assemblies and cause the reactor coolant to flow downwardly internally of the shrouds over the tube bundle and exit through the lower end of the inner shroud for passage to the return duct in the reactor vessel.

Kaufman, Jay S. (Del Mar, CA) [Del Mar, CA; Kissinger, John A. (Del Mar, CA) [Del Mar, CA

1986-01-01

201

Dual iris based human identification  

Microsoft Academic Search

In this paper, a dual iris based human identification system that increases the accuracy and the performance of a typical human iris recognition system is proposed. The system detects and then isolates and extracts the iris region from eye image. It then sets the radial and angular resolution for the extracted iris region and maps the circular region into rectangular

Iftakhar Hasan; Minnatul Fatema; M. Ashraful Amin

2011-01-01

202

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor  

Microsoft Academic Search

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the

Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

2010-01-01

203

Annular core for Modular High-Temperature Gas-Cooled Reactor (MHTGR)  

SciTech Connect

The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40% greater power output over a typical cylindrical configuration. The reactor core is made up of columns of hexagonal blocks, each 793-mm high and 360-mm wide. The active core is 3.5 m in o.d., 1.65 m in i.d., and 7.93-m tall. Fuel elements contain TRISO-coated microspheres of 19.8% enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above.

Turner, R.F.; Baxter, A.M.; Stansfield, O.M.; Vollman, R.E.

1987-08-01

204

Fast reactor 3D core and burnup analysis using VESTA  

SciTech Connect

Burnup analyses using the VESTA code have been performed on a MOX-fuelled fast reactor model as specified by an IAEA computational benchmark. VESTA is a relatively new code that has been used for burnup credit calculations and thermal reactor models, but not typically for fast reactor applications. The detailed input and results of the IAEA benchmark provides an opportunity to gauge the use of VESTA in a fast reactor application. VESTA employs an ultra-fine multi-group binning approach that accelerates Monte Carlo burnup calculations. Using VESTA to compute the end of cycle (EOC) power fractions by enrichment zone showed agreement with the published values within 5%. When comparing the ultra-fine multi-group binning approach to the tally-based approach, EOC isotopic masses also agree within 5%. Using the ultra-fine multi-group binning approach, we obtain a wall-time speedup factor of 35 when compared to the tally-based approach for computing a k{sub eff} eigenvalue with burnup problem. The authors conclude the use of VESTA's ultra-fine multi-group binning approach with Monte Carlo transport performs accurate depletion calculations for this fast reactor benchmark. (authors)

Luciano, N.; Shamblin, J.; Maldonado, I. [Nuclear Engineering Dept., Univ. of Tennessee, Knoxville, TN 37996-2300 (United States)

2012-07-01

205

Internal Control Rod Drive Mechanisms, Design Options for IRIS  

SciTech Connect

IRIS (International Reactor Innovative and Secure) is a medium-power (335 MWe) PWR with an integral, primary circuit configuration, where all the reactor coolant system components are contained within the reactor vessel. This integral configuration is a key reason for the success of IRIS' 'safety-by-design' approach, whereby accident initiators are eliminated or the accident consequences and/or frequency are reduced. The most obvious example of the IRIS safety by design approach is the elimination of large LOCA's, since the integral reactor coolant system has no large loop piping. Another serious accident scenario that is being addressed in IRIS is the postulated ejection of a reactor control cluster assembly (RCCA). This accident initiator can be eliminated by locating the RCCA drive mechanisms (CRDMs) inside the reactor vessel. This eliminates the mechanical drive rod penetration between the RCCA and the external CRDM, eliminating the potential for differential pressure across the pressure boundary, and thus eliminating 'by design' the possibility for rod ejection accident. Moreover, the elimination of the 'large' drive-rod penetrations and the external CRDM pressure housings decreases the likelihood of boric acid leakage and subsequent corrosion of the reactor pressure boundary (like the Davis-Besse incident). This paper will discuss the IRIS top level design requirements and objectives for internal CRDMs, and provide examples candidate designs and their specific performance characteristics. (authors)

Conway, Lawrence E.; Petrovic, Bojan [Westinghouse Electric Company, Science and Technology Department, 1344 Beulah Rd, Pittsburgh, PA 15235 (United States)

2004-07-01

206

Core Subchannel Thermal-Hydraulic Analysis Methods and Critical Heat-Flux Margin in a Light-Water Breeder Reactor.  

National Technical Information Service (NTIS)

Analyis methods and results are described for critical heat flux (CHF) performance margin in the core of an advanced light water moderated breeder reactor design concept. The 1000 MWe breeder reactor is basically like a large commercial pressurized water ...

R. Misiewicz S. J. Kast L. H. Wunderlich

1983-01-01

207

Core and Refueling Design Studies for the Advanced High Temperature Reactor  

SciTech Connect

The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure drop across the core was restricted to no more than 1.5 atm during normal operation to minimize the upward force on the core. Also, the flow velocity in the core was restricted to 3 m/s to minimize erosion of the fuel plates. Section 3.1.1 of this report discusses the design restrictions in more detail.

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

2011-09-01

208

Explication of the Graphite Structural Design Code of core components for the High Temperature Engineering Test Reactor.  

National Technical Information Service (NTIS)

The integrity evaluation of the core graphite components for the High Temperature Engineering Test Reactor (HTTR) will be carried out based upon the Graphite Structural Design Code for core components. In the application of this design code, it is necessa...

T. Iyoku M. Ishihara J. Toyota S. Shiozawa

1991-01-01

209

Core design study on rock-like oxide fuel light water reactor and improvements of core characteristics  

NASA Astrophysics Data System (ADS)

A rock-like oxide (ROX) fuel - LWR burning system has been studied for efficient plutonium transmutation. A zirconia based ROX (Zr-ROX) core has problems such as a small negative Doppler coefficient and a large power peaking factor, which causes severe transients in accidents and high fuel temperature even under nominal condition. For the improvement of these characteristics, two approaches were considered: the additives UO 2, ThO 2 and Er 2O 3, or a heterogeneous core with Zr-ROX and UO 2 assemblies. As a result, the combination of the additives UO 2 and Er 2O 3 is found to sufficiently improve the accident behavior, while a further power peaking reduction may be necessary for the Zr-ROX + UO 2 heterogeneous core. The plutonium transmutation rate is extremely high in Zr-ROX assemblies in the heterogeneous core, to be more than 85% and 70%, respectively for weapons- and reactor-grade plutonium. The plutonium transmutation rate becomes smaller in the full-ROX core with the UO 2 or ThO 2 additive, but the annual transmutation amount of plutonium is large, in comparison with the full-MOX fuel core.

Akie, H.; Takano, H.; Anoda, Y.

1999-08-01

210

IRIS First Light Video  

NASA Video Gallery

First Interface Region Imaging Spectrograph (IRIS) movie, 21 hours after opening the telescope door. This video has been slowed forty percent and looped four times to show greater detail. Credit: N...

211

IRIS Launch Animation  

NASA Video Gallery

This animation demonstrates the launch and deployment of NASA's Interface Region Imaging Spectrograph (IRIS) mission satellite via a Pegasus rocket. The launch is scheduled for June 26, 2013 from V...

212

Sodium Based Heat Pipe Modules for Space Reactor Concepts: Stainless Steel SAFE-100 Core  

NASA Technical Reports Server (NTRS)

A heat pipe cooled reactor is one of several candidate reactor cores being considered for advanced space power and propulsion systems to support future space exploration applications. Long life heat pipe modules, with designs verified through a combination of theoretical analysis and experimental lifetime evaluations, would be necessary to establish the viability of any of these candidates, including the heat pipe reactor option. A hardware-based program was initiated to establish the infrastructure necessary to build heat pipe modules. This effort, initiated by Los Alamos National Laboratory and referred to as the Safe Affordable Fission Engine (SAFE) project, set out to fabricate and perform non-nuclear testing on a modular heat pipe reactor prototype that can provide 100 kilowatt from the core to an energy conversion system at 700 C. Prototypic heat pipe hardware was designed, fabricated, filled, closed-out and acceptance tested.

Martin, James J.; Reid, Robert S.

2004-01-01

213

Welding the Clinch River Breeder Reactor core support structure  

Microsoft Academic Search

The following accomplishments were realized in fabricating the CRBR core support structure: tools were obtained which successfully removed slag and ground starts and stops in a deep, narrow weld groove, dye penetrant examinations were successfull and routinely performed in a deep, narrow weld groove, a single weld bead per layer technique was successfully used with no requirement for interpass grinding;

W. W. Canary; E. A. Franco-Ferreira

1979-01-01

214

Gas Core Reactor with Magnetohydrodynamic Power System and Cascading Power Cycle  

Microsoft Academic Search

The U.S. Department of Energy initiative Generation IV aim is to produce an entire nuclear energy production system with next-generation features for certification before 2030. A Generation IV-capable system must have superior sustainability, safety and reliability, and economic cost advantages in comparison with third generation light water reactors (LWRs). A gas core reactor (GCR) with magnetohydrodynamic (MHD) power converter and

Blair M. Smith; Samim Anghaie

2004-01-01

215

Solid-Core, Gas-Cooled Reactor for Space and Surface Power  

SciTech Connect

The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respect0011ive.

King, Jeffrey C.; El-Genk, Mohamed S. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM 87131 (United States); Chemical and Nuclear Engineering Dept., University of New Mexico, Albuquerque, NM 87131 (United States)

2006-01-20

216

R&D program for core instrumentation improvements devoted for French Sodium fast reactors  

Microsoft Academic Search

Under the framework of French R&D studies for Generation IV reactors and more specifically for sodium-cooled fast reactors (SFR); the CEA, EDF and AREVA have launched a joint coordinated research programme. This paper deals with the R&D sets out to achieve better inspection, maintenance, availability and decommissioning. In particular the instrumentation requirements for core monitoring and detection in the case

J P. Jeannot; G. Rodriguez; C. Jammes; B. Bernardin; J L. Portier; F. Jadot; S. Maire; D. Verrier; F. Loisy; G. Prele

2011-01-01

217

Operational performance of the three bean salad control algorithm on the ACRR (Annular Core Research Reactor)  

SciTech Connect

Experimental tests on the Annular Core Research Reactor have confirmed that the Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute. 3 refs., 4 figs., 1 tab.

Ball, R.M.; Madaras, J.J. (B and W Nuclear Technologies, Lynchburg, VA (USA). Space and Defense Systems); Trowbridge, F.R. Jr.; Talley, D.G.; Parma, E.J. Jr. (Sandia National Labs., Albuquerque, NM (USA))

1991-01-01

218

Exploiting iris dynamics  

NASA Astrophysics Data System (ADS)

The human iris is a circular curtain over the light entrance pupil which is controlled directly by the intensity of blue light from photosensitive ganglions in the retina within the eye. The human iris dynamic is remarkable in that it is capable of shrinking concentrically along the radial direction by a factor 4 from 8mm to 2mm, and constantly oscillates in 1/2 second periodicity. Pupil dilation and contraction causes the iris texture to undergo nonlinear deformation with discrete components and minutia features. Thus, iris recognition must be scale invariant due to the pupil dynamics. We propose the Mandelbrot fractal dimension count of minutia iris details, at different intensity thresholds, in dilation-invariant wedge-boxes, formed at specific angular sizes, but spatially varying over 4 90 quadrants due to the cellular growth under the gravity. Despite the concentric dynamic, we have sought an invariant fractal dimensionality in the circular direction and discovered the non-isotropic effect, departed from the simple Richardson fractal law. Furthermore, we choose an optimum Rayleigh criterion ?/D matching the robust fine resolution scale for the given lens aperture D and the illumination wavelength ? for a potential application from a distant, with the help of comprehensive biometric including iris.

Hsu, Charles; Szu, Harold

2010-04-01

219

Comments on the feasibility of developing gas core nuclear reactors. [for manned interplanetary spacecraft propulsion  

NASA Technical Reports Server (NTRS)

Recent developments in the fields of gas core hydrodynamics, heat transfer, and neutronics indicate that gas core nuclear rockets may be feasible from the point of view of basic principles. Based on performance predictions using these results, mission analyses indicate that gas core nuclear rockets may have the potential for reducing the initial weight in orbit of manned interplanetary vehicles by a factor of 5 when compared to the best chemical rocket systems. In addition, there is a potential for reducing total trip times from 450 to 500 days for chemical systems to 250 to 300 days for gas core systems. The possibility of demonstrating the feasibility of gas core nuclear rocket engines by means of a logical series of experiments of increasing difficulty that ends with ground tests of full scale gas core reactors is considered.

Rom, F. E.

1969-01-01

220

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

PubMed Central

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

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

2013-01-01

221

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

222

MEKIN: MIT-EPRI nuclear reactor core kinetics code  

Microsoft Academic Search

The computer program MEKIN provides a solution in three dimensions for the time-dependent, two-group, neutron diffusion equations and corresponding thermal-hydraulic equations that model the transient behavior of a light water moderated power reactor. The code accepts initial steady state power level, inlet coolant temperature and flow rate. It first computes the initial steady state flux and power distribution throughout the

R. W. Bowring; J. W. Stewart; R. A. Shober; R. N. Sims

1975-01-01

223

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

NASA Technical Reports Server (NTRS)

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

Whitmarsh, C. L., Jr.

1972-01-01

224

Installation of state-of-the-art in-core detector system in the N reactor  

Microsoft Academic Search

A state-of-the-art extensive neutron flux monitoring system to support improved core surveillance and safety condition monitoring has been installed in the U.S. Department of Energy (DOE) Hanford site N reactor. The system uses a large number of fixed and movable self-powered detectors and a rapid data acquisition system that provides for prompt and accurate measurement of the in-core neutron flux.

H. Toffer; R. D. Crowe; T. J. Samuel; C. T. Rombough

1988-01-01

225

Method and arrangement for reducing the radiation exposure risks in the course of a nuclear reactor core melt down accident  

SciTech Connect

A method and arrangement are described for containing the core melt flowing from a nuclear reactor into a core catcher below the core wherein the core melt is permitted to gradually penetrate layers of a core catcher materials of inorganic reactor soluble oxides or salts disposed in the core catcher which core catcher materials are dissolved by the oxidic part of the core melt. The molten solution, after solidification and after being cooled down to a temperature at which hydrogen generating reactions do not take place, is leached with water and rinsed out of the core catcher without the need for humans to be present in the reactor containment and to be exposed to radiation.

Donne, M.D.; Dorner, S.; Schumacher, G.

1981-11-17

226

ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4  

NASA Astrophysics Data System (ADS)

ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14 MW. ORPHEE core is 90 cm height and has a cross section of 27x27 cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4 [1,2,3]. A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 [4]. The analysis showed the reliability of TRIPOLI-4 to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.

Damian, F.; Brun, E.

2014-06-01

227

Gamma heating in reflector heat shield of gas core reactor  

NASA Technical Reports Server (NTRS)

Heating rate measurements made in a mock-up of a BeO heat shield for a gas core nuclear rocket engine yields results nominally a factor of two greater than calculated by two different methods. The disparity is thought to be caused by errors in neutron capture cross sections and gamma spectra from the low cross-section elements, D, O, and Be.

Lofthouse, J. H.; Kunze, J. F.; Young, T. E.; Young, R. C.

1972-01-01

228

Determination of core design thermal safety limits for a two-loop pressurized water reactor  

SciTech Connect

Results are given of independent research of core thermal design limits for the Nuklearna Elektrarna Krsko (NEK) nuclear power plant; procedures for two-loop pressurized water reactor plant core design safety limit calculation are used. Emphasis is placed on researching the vessel exit boiling and the hot-channel exit quality limits and their impact on the maximum available design safety operating range and thermal operating margin of the NEK reactor core. For this purpose, the LIMITS computer code is developed. Based on the modified, well-tried COBRA-IV-I computer code, the departure of nuclear boiling ratio core safety limits are calculated. The original results complement well those of the NEK Final Safety Analysis Report. The procedures and the methods for determining the reactor core design thermal limits are successfully proven despite the unavailability of proprietary data, different models, and computer codes. In addition to the acquired capability of in-house independent checking of the vendor`s results, the bases are set for further independent analyses of the limiting safety system settings for the NEK core.

Kostadinov, V.

1996-04-01

229

Iris Recognition for Human Identification  

NASA Astrophysics Data System (ADS)

Iris recognition system is the biometric identification system. Iris has an intricate structure, uniqueness, stability, and natural protection. Due to these features of the iris it can be used for biometric identification. This system gives better performance than other biometric identification systems. A novel eyelash removal method for preprocessing of human iris images in a human iris recognition system is presented.. Discrete cosine transform (DCT) method is used for feature extraction. For matching of two-iris code Hamming distance calculation is used. EER value must be less for the optimum performance of the system.

Alandkar, Lajari; Gengaje, Sachin

2010-11-01

230

Demonstration of the reactivity constraint approach on SNL's annual core research reactor  

SciTech Connect

This paper reports on the initial demonstration of the reactivity constraint approach and its implementing algorithm, the MIT-CSDL Non-Linear Digital Controller, on the annual core research reactor (ACCR) that is operated by the Sandia National Laboratories. This demonstration constituted the first use of reactivity constraints for the closed-loop, digital control of reactor power on a facility other than the Massachusetts Institute of Technology's (MIT's) research reactor (MITR-II). Also, because the ACRR and the MITR-II are of very different design, these trials established the generic nature of the reactivity constraint approach.

Bernard, J.A.; Kwok, K.S.; Wyant, F.J.; Thome, F.V.

1989-01-01

231

FREC-II: An upgrade to SNL's annular core research reactor  

SciTech Connect

The fuel-ringed external cavity, version II (FREC-II), is a recent upgrade to the annular core research reactor (ACRR) at Sandia National Laboratories (SNL). The FREC-II is neutronically coupled to the ACRR, a 2-MW steady-state/300-MJ pulse reactor used for a variety of simulation experiments in areas such as reactor safety and weapons effects. The FREC-II was designed to provide a large-volume irradiation cavity for fissile and electronic weapon components while retaining the performance characteristics of its smaller predecessor FREC-I.

Rubio, R.A.; Cooper, P.J.; Schulze, J.F.; Bryson, J.W.; Morris, F.M.; Trowbridge, F.R.; Schmidt, T.R.

1989-01-01

232

The behavior of reactor power and flux resulting from changes in core-coolant temperature for a miniature neutron source reactor  

Microsoft Academic Search

In this work, measurements were performed to verify the theoretical predictions of reactor power and flux parameters that result from changes in core inlet temperature (Tin) and the temperature difference between the coolant inlet and outlet (?T) in the Nigeria Research Reactor-1 (NIRR-1), which is a Miniature Neutron Source Reactor (MNSR). The measured data shows that there is a strong

Y. A. Ahmed; G. I. Balogun; S. A. Jonah; I. I. Funtua

2008-01-01

233

Rugged Iris Mechanism  

NASA Technical Reports Server (NTRS)

A rugged iris mechanism has been designed to satisfy several special requirements, including a wide aperture in the "open" position, full obscuration in the "closed" position, ability to function in a cryogenic or other harsh environment, and minimization of friction through minimization of the number of components. An important element of the low-friction aspect of the design is maximization of the flatness of, and provision of small gaps between, adjacent iris blades. The tolerances of the design can be very loose, accommodating thermal expansions and contractions associated with large temperature excursions. The design is generic in that it is adaptable to a wide range of aperture sizes and can be implemented in a variety of materials to suit the thermal, optical, and mechanical requirements of various applications. The mechanism (see figure) includes an inner flat ring, an outer flat ring, and an even number of iris blades. The iris blades shown in front in the figure are denoted as "upper," and the iris blades shown partly hidden behind the front ones are denoted as "lower." Each iris blade is attached to the inner ring by a pivot assembly and to the outer ring by a roller/slider assembly. The upper and lower rings are co-centered and are kept in sliding contact. The iris is opened or closed by turning the outer ring around the center while holding the inner ring stationary. The mechanism is enclosed in a housing (not shown in the figure) that comprises an upper and a lower housing shell. The housing provides part of the sliding support for the outer ring and keeps the two rings aligned as described above. The aforementioned pivot assemblies at the inner ring also serve as spacers for the housing. The lower housing shell contains part of the lower sliding surface and features for mounting the overall mechanism and housing assembly. The upper housing shell contains part of the upper sliding surface.

Ferragut, Nelson J.

2005-01-01

234

Reactor Physics Test Program for the Light Water Breeder Reactor (LWBR) core at Shippingport. (LWBR Development Program)  

Microsoft Academic Search

This report presents and discusses the results of the reactor physics test program for the LWBR core at Shippingport during the reduced power phase of operation from 18,298 EFPH through the end of life at 28,730 EFPH. Test results covering the period from initial startup to 18,298 EFPH have previously been reported in WAPD-TM--1336 (June 1979) and WAPD-TM-1455 (December 1981).

Sarber

1983-01-01

235

Cosmic ray radiography of the damaged cores of the Fukushima reactors.  

PubMed

The passage of muons through matter is dominated by the Coulomb interaction with electrons and nuclei. The interaction with the electrons leads to continuous energy loss and stopping of the muons. The interaction with nuclei leads to angle "diffusion." Two muon-imaging methods that use flux attenuation and multiple Coulomb scattering of cosmic-ray muons are being studied as tools for diagnosing the damaged cores of the Fukushima reactors. Here, we compare these two methods. We conclude that the scattering method can provide detailed information about the core. Attenuation has low contrast and little sensitivity to the core. PMID:23102302

Borozdin, Konstantin; Greene, Steven; Luki?, Zarija; Milner, Edward; Miyadera, Haruo; Morris, Christopher; Perry, John

2012-10-12

236

Full core reactor analysis: Running Denovo on Jaguar  

SciTech Connect

Fully-consistent, full-core, 3D, deterministic neutron transport simulations using the orthogonal mesh code Denovo were run on the massively parallel computing architecture Jaguar XT5. Using energy and spatial parallelization schemes, Denovo was able to efficiently scale to more than 160 k processors. Cell-homogenized cross sections were used with step-characteristics, linear-discontinuous finite element, and trilinear-discontinuous finite element spatial methods. It was determined that using the finite element methods gave considerably more accurate eigenvalue solutions for large-aspect ratio meshes than using step-characteristics. (authors)

Jarrell, J. J.; Godfrey, A. T.; Evans, T. M.; Davidson, G. G. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States)

2012-07-01

237

Full Core Reactor Analysis: Running Denovo on Jaguar  

SciTech Connect

Fully-consistent, full-core, 3D, deterministic simulations using the orthogonal mesh code Denovo were run on the massively parallel computing architecture Jaguar XT5. Using energy and spatial parallelization schemes, Denovo was able to efficiently scale to over 160k processors. Cell-homogenized cross-sections were used with Step-Characteristics, Linear-Discontinuous Finite Element, and Tri-Linear-Discontinuous Finite Element spatial methods. It was determined that using the finite element methods gave considerably more accurate eigenvalue solutions for large aspect ratios meshes than those using Step-Characteristics.

Jarrell, Joshua J [ORNL; Godfrey, Andrew T [ORNL; Evans, Thomas M [ORNL; Davidson, Gregory G [ORNL

2012-01-01

238

In-reactor testing of the closed cycle gas core reactor---the nuclear light bulb concept  

SciTech Connect

The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse ([gt]1800 s) and thrust ([gt]445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures ([similar to]4000 K). The following paper describes analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented here include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRR. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRR for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept.

Gauntt, R.O.; Slutz, S.A.; Harms, G.A. (Sandia National Laboratories, Albuquerque, New Mexico 87185 (United States)); Latham, T.S.; Roman, W.C.; Rodgers, R.J. (United Technologies Research Center, East Hartford, Connecticut 06108 (United States))

1993-01-20

239

In-reactor testing of the closed cycle gas core reactor-the nuclear light bulb concept  

NASA Astrophysics Data System (ADS)

The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (>~1800 s) and thrust (>~445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (~4000 K). The following paper describes analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented here include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRR. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRR for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept.

Gauntt, Randall O.; Slutz, Stephen A.; Harms, Gary A.; Latham, Thomas S.; Roman, Ward C.; Rodgers, Richard J.

1993-01-01

240

Fast-Power-Reactor Core Concepts. Optimization of the Physical Characteristics of the BN-1600 Reactor.  

National Technical Information Service (NTIS)

The paper sets out the general approach to the choice of physical parameters and their optimization for the BN fast power reactors. The structural features of the reactor which are important from the point of view of breeding are discussed. The breeding c...

M. F. Troyanov V. I. Matveev A. I. Novozhilov S. B. Bobrov A. P. Ivanov

1979-01-01

241

Calculation of the Thermal and Hydraulic States in Rod Cluster Cores of Light-Water Reactors.  

National Technical Information Service (NTIS)

For calculating the three-dimensional steady distribution of the thermal and hydraulic states in rod cluster cores of light-water reactors, the subchannel analysis programs COLA 1 and COLA 2 have been developed. Both programs contain a multitude of compet...

H. Teichel

1977-01-01

242

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

Microsoft Academic Search

The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from

K. C. Schulz; G. T. Yahr

1995-01-01

243

Ex-Core CFD Analysis Results for the Prometheus Gas Reactor  

SciTech Connect

This paper presents the initial nozzle-to-nozzle (N2N) reactor vessel model scoping studies using computational fluid dynamics (CFD) analysis methods. The N2N model has been solved under a variety of different boundary conditions. This paper presents some of the basic hydraulic results from the N2N CFD analysis effort. It also demonstrates how designers were going to apply the analysis results to modify a number of the design features. The initial goals for developing a preliminary CFD N2N model were to establish baseline expectations for pressure drops and flow fields around the reactor core. Analysis results indicated that the averaged reactor vessel pressure drop for all analyzed cases was 46.9 kPa ({approx}6.8 psid). In addition, mass flow distributions to the three core fuel channel regions exhibited a nearly inverted profile to those specified for the in-core thermal/hydraulic design. During subsequent design iterations, the goal would have been to modify or add design features that would have minimized reactor vessel pressure drop and improved flow distribution to the inlet of the core.

Lorentz, Donald G. [Space Engineering, Bechtel Bettis, Inc. West Mifflin, PA 15122 (United States)

2007-01-30

244

TMI2 reactor-vessel head removal and damaged-core-removal planning  

Microsoft Academic Search

A major milestone in the cleanup and recovery effort at TMI-2 will be the removal of the reactor vessel closure head, planum, and damaged core fuel material. The data collected during these operations will provide the nuclear power industry with valuable information on the effects of high-temperature-dissociated coolant on fuel cladding, fuel materials, fuel support structural materials, neutron absorber material,

J. A. Logan; C. W. Hultman; T. J. Lewis

1982-01-01

245

Out of core testing of the North Carolina State University PULSTAR reactor positron beam  

Microsoft Academic Search

Out of core tests were performed on the North Carolina State University intense positron beam, prior to insertion into beamport number 6 of the PULSTAR nuclear reactor The beam optics were tested independently of the positron converter\\/moderator using beams of electrons and the performance of the extraction lenses was found to be in good agreement with the results of design

Jeremy Moxom; Alfred G. Hathaway; Ayman I. Hawari

2007-01-01

246

A chemical equilibrium estimate of the aerosols produced in an overheated light water reactor core  

Microsoft Academic Search

The degree of vaporization of light water reactor core materials was estimated using a highly idealized procedure involving (a) specification of the phases that are present for both structural and fuel material, (b) estimation of the vapor pressures exerted by the individual components of each phase, and (c) assuming a degree of vaporization of each phase constituent, allowing equilibration between

R. P. Wichner; R. D. Spence

1985-01-01

247

System for Incident Detection and Core Surveillance for a Fast Reactor.  

National Technical Information Service (NTIS)

The incident detection system of a Fast Reactor core covers three fields: (1) normal working (check on running order), (2) the incident situation (with alert and scram), and (3) the intermediate field in which the defect arises. The principal incidents th...

D. de Lapparent J. Gourdon A. Gouriou A. Jeannot M. Jallade

1975-01-01

248

Ex-Core CFD Analysis Results for the Prometheus Gas Reactor  

NASA Astrophysics Data System (ADS)

This paper presents the initial nozzle-to-nozzle (N2N) reactor vessel model scoping studies using computational fluid dynamics (CFD) analysis methods. The N2N model has been solved under a variety of different boundary conditions. This paper presents some of the basic hydraulic results from the N2N CFD analysis effort. It also demonstrates how designers were going to apply the analysis results to modify a number of the design features. The initial goals for developing a preliminary CFD N2N model were to establish baseline expectations for pressure drops and flow fields around the reactor core. Analysis results indicated that the averaged reactor vessel pressure drop for all analyzed cases was 46.9 kPa (~6.8 psid). In addition, mass flow distributions to the three core fuel channel regions exhibited a nearly inverted profile to those specified for the in-core thermal/hydraulic design. During subsequent design iterations, the goal would have been to modify or add design features that would have minimized reactor vessel pressure drop and improved flow distribution to the inlet of the core.

Lorentz, Donald G.

2007-01-01

249

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback  

NASA Technical Reports Server (NTRS)

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2010-01-01

250

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

National Technical Information Service (NTIS)

A new non-TRISO fuel and clad design concept is proposed for the prismatic, helium cooled Very High Temperature Reactor core. The new concept could substantially reduce the current 10-20 wt% TRISO uranium enrichments down to 4-6 wt% for both initial and r...

J. W. Sterbentz

2007-01-01

251

Critical heat flux predictions for the Sandia Annular Core Research Reactor.  

National Technical Information Service (NTIS)

This study provides best estimate predictions of the Critical Heat Flux (CHF) and the Critical Heat Flux Ratio (CHFR) to support the proposed upgrade of the Annual Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) from its present value o...

D. V. Rao M. S. El-Genk

1994-01-01

252

Design and fabrication of the instrumented fuel elements for the Annular Core Research Reactor (ACRR)  

Microsoft Academic Search

This report describes the design and fabrication techniques for the instrumented fuel elements of the Annular Core Research Reactor (ACRR). The thermocouple assemblies were designed and fabricated at Sandia Laboratories while the instrumented elements were assembled at Los Alamos Scientific Laboratory. In order to satisfy the ACRR's Technical Specifications, the thermocouples are required to measure temperature in excess of 1800°C

T. R. Schmidt; D. J. Sasmor; J. T. Martin; F. Gonzalez; D. N. Cox

1981-01-01

253

Fuel efficient hydrodynamic containment for gas core fission reactor rocket propulsion. Final report, September 30, 1992May 31, 1995  

Microsoft Academic Search

Gas core reactors can form the basis for advanced nuclear thermal propulsion (NTP) systems capable of providing specific impulse levels of more than 2,000 sec., but containment of the hot uranium plasma is a major problem. The initial phase of an experimental study of hydrodynamic confinement of the fuel cloud in a gas core fission reactor by means of an

P. M. Sforza; R. J. Cresci

1997-01-01

254

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)] [Nuclear and Radiological Engineering and Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA (United States)

2013-07-01

255

Support arrangement for core modules of nuclear reactors  

SciTech Connect

This patent describes a support arrangement in combination with a nuclear reactor, which comprises at least one fuel cell module, at least one control drive mechanism, and at least one pressure vessel head. The support arrangement is located between the fuel cell module and the control drive mechanism and is supported by the pressure vessel head, the support arrangement comprising: a module support nut, engaged with the pressure vessel head and supported therefrom, including a downwardly depending screw threaded portion, and a shroud housing for the fuel cell module including a screw threaded portion engaged with the screw threaded portion of the support nut such that the shroud housing is suspended from the support nut and thus from the pressure vessel head, the module support nut and the shroud housing including a locking means for locking the nut and housing against relative rotation.

Bollinger, L.R.

1987-02-17

256

Vaporization of core materials in postulated severe light water reactor accidents  

SciTech Connect

The vaporization of core materials other than fission products during a postulated severe light water reactor accident is treated by chemical thermodynamics. The core materials considered were (a) the control rod materials, silver, cadmium, and indium; (b) the structural materials, iron, chromium, nickel, and manganese; (c) cladding material, zirconium and tin; and (d) the fuel, uranium oxide. Thermodynamic data employed for the solid and gaseous elements and oxides were based on measurements, while the data for the gaseous hydroxides were generally based on estimates from literature. Thermodynamic criteria were derived to determine whether the metallic element or the solid oxide was the stable condensed phase for the accident environmental conditions. Equations for the partial pressures for all gaseous species were also derived. The relevant environmental conditions were provided by the pressurized water reactor and boiling water reactor heat-up thermal-hydraulic codes. The volatilities of the core materials were found to decrease roughly in the following order: cadmium, indium, tin, iron, silver, manganese, nickel, chromium, uranium, and zirconium. Cadmium and indium would provide the largest mass of core material that can be transported out of the core.

Cubicciotti, D.; Sehgal, B.R.

1984-11-01

257

Gamma-thermometer-based reactor-core liquid-level detector. [PWR  

SciTech Connect

A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is modified by thermocouple junctions and leads to obtain an unambiguous indication of the presence or absence of coolant liquid at the gamma thermometer location. A signal processor generates a signal based on the thermometer surface heat transfer coefficient by comparing the signals from the thermocouples at the thermometer location. The generated signal is a direct indication of loss of coolant due to the change in surface heat transfer when coolant liquid drops below the thermometer location. The loss of coolant indication is independent of reactor power at the thermometer location. Further, the same thermometer may still be used for the normal power monitoring function.

Burns, T.J.

1981-06-16

258

Measurements of the HEU and LEU in-core spectra at the Ford Nuclear Reactor  

SciTech Connect

The Ford Nuclear Reactor (FNR) at the University of Michigan has been serving as the test site for a low-enriched uranium (LEU) fuel whole-core demonstration. As part of the experimental program, the differential neutron spectrum has been measured in a high-enriched uranium (HEU) core and an LEU core. The HEU and LEU spectra were determined by unfolding the measured activities of foils that were irradiated in the reactor. When the HEU and LEU spectra are compared from 1 MeV to 10 MeV, significant differences between the two spectra are apparent below 10 eV. These are probably caused by the additional /sup 238/U resonance absorption in the LEU fuel. No measurable difference occurs in the shape of the spectra above 1 MeV. 7 refs., 6 figs., 2 tabs.

Wehe, D.K.; King, J.S.; Lee, J.C.; Martin, W.R.

1984-01-01

259

Contactless Autofeedback Iris Capture Design  

Microsoft Academic Search

Automated iris recognition is one of the most reliable biometrics in terms of identification and verification performance. One of the major challenges for automated iris recognition is to capture a high-quality image of the iris since system performance is greatly affected by poor-quality imaging. This paper describes the design and implementation of a high-quality imaging device for iris acquisition, which

XiaoFu He; Jingqi Yan; Guangyu Chen; Pengfei Shi

2008-01-01

260

Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor  

SciTech Connect

A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

Primm, Trent [ORNL; Gehin, Jess C [ORNL

2009-04-01

261

Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core  

SciTech Connect

Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public.

PARMA JR.,EDWARD J.

2000-01-01

262

Investigations of Alternative Steam Generator Location and Flatter Core Geometry for Lead-Cooled Fast Reactors  

SciTech Connect

This paper concerns two independent safety investigations on critical and sub-critical heavy liquid metal cooled fast reactors using simple flow paths. The first investigation applies to locating the steam generators in the risers instead of the down-comers of a simple flow path designed sub-critical reactor of 600 MW{sub th} power. This was compared to a similar design, but with the steam generators located in the downcomers. The transients investigated were Total-Loss-of-Power and unprotected Loss-Of-Flow. It is shown that this reactor peaks at 1041 K after 29 hours during a Total-Loss-Of-Power accident. The difference between locating the steam generators in the risers and the downcomers is insignificant for this accident type. During an unprotected Loss-Of-Flow accident at full power, the core outlet temperature stabilizes at 1010 K, which is 337 K above nominal outlet temperature. The second investigation concerns a 1426 MW{sub th} critical reactor where the influence of the core height versus the core outlet temperature is studied during an unprotected Loss-Of-Flow and Total-Loss-Of-Power accident. A pancake type core geometry of 1.0 m height and 5.8 m diameter, is compared to a compact core of 2 m height and 4.5 m diameter. Moderators, like BeO and hydrides, and their influence on safety coefficients and burnup swings are also presented. Both cores incinerate transuranics from spent LWR fuel with minor actinide fraction of 5%. We show that LFRs can be designed both to breed and burn transuranics from LWRs. It is shown that the hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. The computational fluid dynamics code STAR-CD was used for all thermal hydraulic calculations, and the MCNP and MCB for neutronics, and burn-up calculations. (authors)

Carlsson, Johan; Tucek, Kamil; Wider, Hartmut [Joint Research Centre, Institute for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands)

2006-07-01

263

Iris Recognition using Steerable Pyramids  

Microsoft Academic Search

This work presents a new iris recognition method based on steerable pyramid transform. This method consists of four steps: localization, normalization, features extraction and matching. After locating the iris boundaries by Hough Transform, normalization is operated by unwrapping the circular ring and isolating the noisy regions. Steerable pyramid filters are then used to capture orientation details from the iris texture.

N. Khiari; H. Mahersia; K. Hamrouni

2008-01-01

264

The Iris Architecture and Implementation  

Microsoft Academic Search

The goals of the Iris database management system are to enhance database programmer productivity and to provide generalized database support for the integration of future applications. Iris is based on an object and function model. Iris objects are typed but unlike other object systems, they contain no state. Attribute values, relationships and behavior of objects are modeled by functions. The

W. Kevin Wilkinson; Peter Lyngbk; Waqar Hasan

1990-01-01

265

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

266

Core and plant design of the power reactor cooled and moderated by supercritical light water with single tube water rods  

Microsoft Academic Search

A reactor cooled and moderated by supercritical light water with single tube water rods is designed. The plant system is a once-through direct cycle; the whole coolant which flows once through the core is fed to the turbine. This reactor is much simpler than the current light water reactors LWRs, which enhances its economy. The average outlet coolant temperature should

K. Dobashi; Y. Oka; S. Koshizuka

1997-01-01

267

A feasibility study of ferro-boron as in-core shield material in fast breeder reactors  

Microsoft Academic Search

Shields around core and blankets form major part of reactor assembly in fast reactors as the incident neutron spectrum is hard with negligible thermal component and has anisotropic angular distribution. In this paper, a study is presented on the use of ferro-boron as neutron shield material in pool type fast reactors. The reference case chosen is the Prototype Fast Breeder

D. Sunil Kumar; R. S. Keshavamurthy; P. Mohanakrishnan; S. C. Chetal

2010-01-01

268

Split-core heat-pipe reactors for out-of-pile thermionic power systems.  

NASA Technical Reports Server (NTRS)

Description of the concept of splitting a heat-pipe reactor for out-of-core thermionics into two identical halves and using the resulting center gap for reactivity control. Short Li-W reactor heat pipes penetrate the axial reflectors and form a heat exchanger with long heat pipes which wind through the shield to the thermionic diodes. With one reactor half anchored to the shield, the other is attached to a long arm with a pivot behind the shield and swings through a small arc for reactivity control. A safety shim prevents large reactivity inputs, and a fueled control arm drive shaft acts as a power stabilizer. Reactors fueled with U-235C and with U-233C have been studied.-

Niederauer, G.; Lantz, E.; Breitweiser, R.

1971-01-01

269

Development of a thermal-hydraulic design methodology for an advanced reactor core with vertical parallel channels.  

National Technical Information Service (NTIS)

A thermal-hydraulic design methodology has been developed for the analysis of an advanced reactor core and compared the results with existing analysis data. On the basis of one dimensional core design methodology for the reference core, a computer code CO...

D. H. Hwang Y. J. Yoo K. K. Kim M. Chang

1998-01-01

270

Proceedings of the Topical Meeting on Advances in Reactor Physics and Core Thermal Hydraulics Held at Kiamesha Lake, NY. on September 22-24, 1982. Volume 1.  

National Technical Information Service (NTIS)

Reactor physics, core thermal hydraulics, and the interactions between core physics and thermal hydraulics are covered both for thermal reactors and for fast breeders. There are sessions on current challenges in these areas, on measurement and analysis of...

1982-01-01

271

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

SciTech Connect

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

Casoli, P.; Authier, N.; Chapelle, A. [Commissariat a l'Energie Atomique et Aux Energies Alternatives, CEA, DAM, F-21120 Is sur Tille (France)

2012-07-01

272

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor  

SciTech Connect

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

2010-09-01

273

Using crypts as iris minutiae  

NASA Astrophysics Data System (ADS)

Iris recognition is one of the most reliable biometric technologies for identity recognition and verification, but it has not been used in a forensic context because the representation and matching of iris features are not straightforward for traditional iris recognition techniques. In this paper we concentrate on the iris crypt as a visible feature used to represent the characteristics of irises in a similar way to fingerprint minutiae. The matching of crypts is based on their appearances and locations. The number of matching crypt pairs found between two irises can be used for identity verification and the convenience of manual inspection makes iris crypts a potential candidate for forensic applications.

Shen, Feng; Flynn, Patrick J.

2013-05-01

274

Analyzing the thermionic reactor critical experiments. [thermal spectrum of uranium 235 core  

NASA Technical Reports Server (NTRS)

The Thermionic Reactor Critical Experiments (TRCE) consisted of fast spectrum highly enriched U-235 cores reflected by different thicknesses of beryllium or beryllium oxide with a transition zone of stainless steel between the core and reflector. The mixed fast-thermal spectrum at the core reflector interface region poses a difficult neutron transport calculation. Calculations of TRCE using ENDF/B fast spectrum data and GATHER library thermal spectrum data agreed within about 1 percent for the multiplication factor and within 6 to 8 percent for the power peaks. Use of GAM library fast spectrum data yielded larger deviations. The results were obtained from DOT R Theta calculations with leakage cross sections, by region and by group, extracted from DOT RZ calculations. Delineation of the power peaks required extraordinarily fine mesh size at the core reflector interface.

Niederauer, G. F.

1973-01-01

275

Using ex-core neutron detectors to estimate fuel quantities in the reactor vessel lower head  

SciTech Connect

During the accident at Three Mile Island Unit 2 (TMI-2), a significant mass of core debris slumped to the bottom head of the reactor vessel. Defueling activities caused more core debris to relocate to the lower head region. The variations in the ex-core neutron detector, or source range monitor (SRM), readings gave evidence of this effect of defueling activity. Between October 1986 and November 1987, as a result of defueling the lower core region, increases in SRM rates were noted as 50% for NI-1 and 90% for NI-2. Analysis of these increases shows that they correspond to lower head rubble bed mass increases of 30 to 50%, or between 6 and 8 tonnes of rubble. The above yields a total lower head inventory of between 21 and 28 tonnes of rubble; this equates to 15.5 to 21 tonnes of UO/sub 2/.

Rainisch, R.; Fricke, V.

1988-01-01

276

Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor  

SciTech Connect

Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve these objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.

Richard Schultz

2012-04-01

277

NonOrthogonal View Iris Recognition System  

Microsoft Academic Search

This paper proposes a non-orthogonal view iris recognition system comprising a new iris imaging module, an iris segmentation module, an iris feature extraction module and a classification module. A dual-charge-coupled device camera was developed to capture four-spectral (red, green, blue, and near-infrared) iris images which contain useful information for simplifying the iris segmentation task. An intelligent random sample consensus iris

Chia-Te Chou; Sheng-Wen Shih; Wen-Shiung Chen; Victor W. Cheng; Duan-Yu Chen

2010-01-01

278

Lunar in-core thermionic nuclear reactor power system conceptual design  

NASA Technical Reports Server (NTRS)

This paper presents a conceptual design of a lunar in-core thermionic reactor power system. The concept consists of a thermionic reactor located in a lunar excavation with surface mounted waste heat radiators. The system was integrated with a proposed lunar base concept representative of recent NASA Space Exploration Initiative studies. The reference mission is a permanently-inhabited lunar base requiring a 550 kWe, 7 year life central power station. Performance parameters and assumptions were based on the Thermionic Fuel Element (TFE) Verification Program. Five design cases were analyzed ranging from conservative to advanced. The cases were selected to provide sensitivity effects on the achievement of TFE program goals.

Mason, Lee S.; Schmitz, Paul C.; Gallup, Donald R.

1991-01-01

279

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

280

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

SciTech Connect

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

Gehin, J.C.

1995-08-01

281

Lessons Learned from Sandia National Laboratories' Operational Readiness Review of the Annular Core Research Reactor (ACRR)  

SciTech Connect

The Sandia ACRR (a Hazard Category 2 Nuclear Reactor Facility) was defueled in June 1997 to modify the reactor core and control system to produce medical radioisotopes for the Department of Energy (DOE) Isotope Production Program. The DOE determined that an Operational Readiness Review (ORR) was required to confirm readiness to begin operations within the revised safety basis. This paper addresses the ORR Process, lessons learned from the Sandia and DOE ORRS of the ACRR, and the use of the ORR to confirm authorization basis implementation.

Bendure, Albert O.; Bryson, James W.

1999-05-17

282

Iris pigment epithelium transplantation  

Microsoft Academic Search

Background: Iris pigment epithelium (IPE) cells and retinal pigment epithelium (RPE) cells possess the same embryonic origin. It is also known that the pigmented epithelial cells in the eye have a high transdifferentiation potential. In this study we transplanted IPE cells into the subretinal space of albino Royal College of Surgeons (RCS) rats and evaluated their influence on the

Kourous A. Rezai; Leon Kohen; Peter Wiedemann; Klaus Heimann

1997-01-01

283

Interaction of loading pattern and nuclear data uncertainties in reactor core calculations  

SciTech Connect

Along with best-estimate calculations for design and safety analysis, understanding uncertainties is important to determine appropriate design margins. In this framework, nuclear data uncertainties and their propagation to full core calculations are a critical issue. To deal with this task, different error propagation techniques, deterministic and stochastic are currently developed to evaluate the uncertainties in the output quantities. Among these is the sampling based uncertainty and sensitivity software XSUSA which is able to quantify the influence of nuclear data covariance on reactor core calculations. In the present work, this software is used to investigate systematically the uncertainties in the power distributions of two PWR core loadings specified in the OECD UAM-Benchmark suite. With help of a statistical sensitivity analysis, the main contributors to the uncertainty are determined. Using this information a method is studied with which loading patterns of reactor cores can be optimized with regard to minimizing power distribution uncertainties. It is shown that this technique is able to halve the calculation uncertainties of a MOX/UOX core configuration. (authors)

Klein, M.; Gallner, L.; Krzykacz-Hausmann, B.; Pautz, A.; Velkov, K.; Zwermann, W. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS MbH, Boltzmannstr. 14, D- 85748 Garching b. Muenchen (Germany)

2012-07-01

284

A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor  

SciTech Connect

This paper describes the core design and performance characteristics of a metal fuel core concept for a 1000 MWt Advanced Burner Reactor. A ternary metal fuel form of U-TRU-Zr was assumed with weapons grade plutonium feed for the startup core and TRU recovered from LWR spent fuel for the recycled equilibrium core. A compact burner core was developed by trade-off between the burnup reactivity loss and TRU conversion ratio, with a fixed cycle length of one-year. In the startup core, the average TRU enrichment is 15.5%, the TRU conversion ratio is 0.81, and the burnup reactivity loss over a cycle is 3.6% {delta}k. The heavy metal and TRU inventories are 13.1 and 2.0 metric tons, respectively. The average discharge burnup is 93 MWd/kg, and the TRU consumption rate is 55.5 kg/year. For the recycled equilibrium core, the average TRU enrichment is 22.1 %, the TRU conversion ratio is 0.73, and the burnup reactivity loss is 2.2% {delta}k. The TRU inventory and consumption rate are 2.9 metric tons and 81.6 kg/year, respectively. The evaluated reactivity coefficients provide sufficient negative feedbacks. The control systems provide shutdown margins that are more than adequate. The integral reactivity parameters for quasi-static reactivity balance analysis indicate favorable passive safety features, although detailed safety analyses are required to verify passive safety behavior. (authors)

Yang, W.S.; Kim, T.K.; Grandy, C. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne IL 60439 (United States)

2007-07-01

285

IRIS Product Recommendations  

NASA Technical Reports Server (NTRS)

This report presents the Applied Meteorology Unit's (AMU) evaluation of SIGMET Inc.'s Integrated Radar Information System (IRIS) Product Generator and recommendations for products emphasizing lightning and microburst tools. The IRIS Product Generator processes radar reflectivity data from the Weather Surveillance Radar, model 74C (WSR-74C), located on Patrick Air Force Base. The IRIS System was upgraded from version 6.12 to version 7.05 in late December 1999. A statistical analysis of atmospheric temperature variability over the Cape Canaveral Air Force Station (CCAFS) Weather Station provided guidance for the configuration of radar products that provide information on the mixed-phase (liquid and ice) region of clouds, between 0 C and -20 C. Mixed-phase processes at these temperatures are physically linked to electrification and the genesis of severe weather within convectively generated clouds. Day-to-day variations in the atmospheric temperature profile are of sufficient magnitude to warrant periodic reconfiguration of radar products intended for the interpretation of lightning and microburst potential of convectively generated clouds. The AMU also examined the radar volume-scan strategy to determine the scales of vertical gaps within the altitude range of the 0 C to -20 C isotherms over the Kennedy Space Center (KSC)/CCAFS area. This report present's two objective strategies for designing volume scans and proposes a modified scan strategy that reduces the average vertical gap by 37% as a means for improving radar observations of cloud characteristics in the critical 0 C to -20 C layer. The AMU recommends a total of 18 products, including 11 products that require use of the IRIS programming language and the IRIS User Product Insert feature. Included is a cell trends product and display, modeled after the WSR-88D cell trends display in use by the National Weather Service.

Short, David A.

2000-01-01

286

Shipment of the Light Water Breeder Reactor fuel assemblies from the Shippingport Atomic Power Station to the extended core facility (Idaho) (LWBR Development Program)  

Microsoft Academic Search

After successfully operating for 29,047 effective full power hours, the Light Water Breeder Reactor (LWBR) core was defueled prior to total decommissioning of the Shippingport Atomic Power Station. All nuclear fuel and much of the reactor internal hardware was removed from the reactor vessel and prepared for shipment to disposal sites or to the Naval Reactors Expended Core Facility in

Selsley

1987-01-01

287

Technical Safety Appraisal of the Sandia reactors, ACRR (Annular Core Research Reactor), SPR III (Sandia Pulse Reactor III), Sandia National Laboratories, Albuquerque  

SciTech Connect

This report presents findings and concerns resulting from a Technical Safety Appraisal of Sandia National Laboratories' Sandia Pulse Reactor III (SPR III) and the Annular Core Research Reactor (ACRR). It was conducted by an appraisal team for the Department of Energy's Office of Safety Appraisals during site visits July 18--22 and August 1--12, 1988. The two reactors are located in Technical Area V of the Sandia Albuquerque site. A third reactor, SPR II, is stored assembled in a vault at the SPR III facility. While the SPR II reactor is still considered operational, there are no plans to use it in the foreseeable future. It was last operated in 1984. This appraisal addresses only the operations associated with SPR III and ACRR. The principle hazards presented by operations in these facilities are routine industrial safety hazards, the beta and gamma radiation fields experienced during maintenance operations at SPR III and during the handling of experimental packages after irradiation in either of the reactors, and localized radiation fields that could result from a highly-unlikely, severe accident. The findings and concerns developed by the appraisal team were shared with senior managers of Sandia National Laboratories and the Albuquerque Operations Office in exit meetings held on August 11 and 12, 1988. The final report of the team has been validated for factual accuracy with Sandia and the Albuquerque Operations Office.

Not Available

1989-05-01

288

Medical isotope production: A new research initiative for the Annular Core Research Reactor  

SciTech Connect

An investigation has been performed to evaluate the capabilities of the Annular Core Research Reactor and its supporting Hot Cell Facility for the production of {sup 99}Mo and its separation from the fission product stream. Various target irradiation locations for a variety of core configurations were investigated, including the central cavity, fuel and reflector locations, and special target configurations outside the active fuel region. Monte Carlo techniques, in particular MCNP using ENDF B-V cross sections, were employed for the evaluation. The results indicate that the reactor, as currently configured, and with its supporting Hot Cell Facility, would be capable in meeting the current US demand if called upon. Modest modifications, such as increasing the capacity of the external heat exchangers, would permit significantly higher continuous power operation and even greater {sup 99}Mo production ensuring adequate capacity for future years.

Coats, R.L.; Parma, E.J.

1993-12-31

289

Experimental Plans for Subsystems of a Shock Wave Driven Gas Core Reactor  

NASA Technical Reports Server (NTRS)

This Contractor Report proposes a number of plans for experiments on subsystems of a shock wave driven pulsed magnetic induction gas core reactor (PMI-GCR, or PMD-GCR pulsed magnet driven gas core reactor). Computer models of shock generation and collision in a large-scale PMI-GCR shock tube have been performed. Based upon the simulation results a number of issues arose that can only be addressed adequately by capturing experimental data on high pressure (approx.1 atmosphere or greater) partial plasma shock wave effects in large bore shock tubes ( 10 cm radius). There are three main subsystems that are of immediate interest (for appraisal of the concept viability). These are (1) the shock generation in a high pressure gas using either a plasma thruster or pulsed high magnetic field, (2) collision of MHD or gas dynamic shocks, their interaction time, and collision pile-up region thickness, and (3) magnetic flux compression power generation (not included here).

Kazeminezhad, F.; Anghai, S.

2008-01-01

290

A new advanced fixed in-core instrumentation for a PWR reactor  

NASA Astrophysics Data System (ADS)

Gamma thermometer studies have been done at E.D.F. for four years. These studies started in France with a feasibility study in 1975. E.D.F.'s scope was to develop a new fixed "in-core" instrumentation for PWR based on the gamma heat measurements. The advanced gamma thermometer design has been done in such a way to be able to manufacture strings of 6 to 9 detectors each. The results of gamma thermometer make up in 1976 were encouraging and E.D.F. went on to develop a gamma thermometer assembly for a reactor application. Before being mounted on the reactor vessel, the gamma thermometer strings are calibrated in a loop test by means of an electrical current giving the ?T versus the specific power ( W/ g). The loop test simulates the thermohydraulic conditions in the reactor tube guide. Two gamma thermometer strings have been installed in the BUGEY 5 reactor since June 1979. Four gamma thermometer strings are provided for insertion in the TRICASTIN 2 reactor and four more gamma thermometer strings are manufactured to be ready for the start up of the TRICASTIN 3 reactor in 1980.

Barbet, M.; Guillery, M.

291

Results of Reactor Materials Experiments Investigating 2-D Core-Concrete Interaction and Debris Coolability  

Microsoft Academic Search

The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program is conducting reactor materials experiments and associated analysis to achieve the following objectives: 1) resolution of the ex-vessel debris coolability issue, and 2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe

M. T. Farmer; S. Lomperski; S. Basu

2004-01-01

292

Thermal-hydraulic calculations for the conversion to LEU of a research reactor core  

Microsoft Academic Search

The thermal-hydraulic analysis performed for the needs of the conversion of the open pool 5MW Greek Research Reactor (GRR-1) to a pure Low Enrichment (LEU) configuration is presented. The methodology was based on a complete set of neutronic calculations performed for the new core configuration, in compliance with pre-defined Operation Limiting Conditions. The hottest channel analysis approach was adopted, and

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

2008-01-01

293

In-core fuel management optimization of pebble-bed reactors  

Microsoft Academic Search

A reduction of the power peak in the core of High Temperature pebble-bed reactors is attractive to decrease the maximum fuel temperature and to increase fuel performance. A calculation procedure was developed, which combines fuel depletion, neutronics and thermalhydraulics to investigate the impact of a certain (re)loading scheme for the pebble-bed type HTR. The procedure has been applied to a

B. Boer; J. L. Kloosterman; D. Lathouwers; T. H. J. J. van der Hagen

2009-01-01

294

2240MW(th) high-temperature reactor core power density study  

Microsoft Academic Search

This study was done to estimate the effects of reducing the design power density of a 2240-MW(t) high-temperature gas-cooled reactor. Core history and thermal hydraulics calculations were performed for average power densities of 5.8 and 7.2 W\\/cm³ and the use of highly enriched fuel was considered. The fuel temperature conditions for the higher power density were found to be only

Vondy

1984-01-01

295

Compact Reversed-Field Pinch Reactors (CRFPR): fusion-power-core integration study  

Microsoft Academic Search

Using detailed two-dimensional neutronics studies based on the results of a previous framework study (LA-10200-MS), the fusion-power-core (FPC) integration, maintenance, and radio-activity\\/afterheat control are examined for the Compact Reversed-Field Pinch Reactor (CRFPR). While maintaining as a base case the nominal 20-MW\\/m² neutron first-wall loading design, CRFPR(20), the cost and technology impact of lower-wall-loading designs are also examined. The additional detail

C. Copenhaver; R. A. Krakowski; N. M. Schnurr; R. L. Miller; C. G. Bathke; R. L. Hagenson; C. R. Mynard; A. D. Chaffee; C. Cappiello; J. W. Davidson

1985-01-01

296

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

Microsoft Academic Search

A consistent set of designs for 1000 MWt commercial-scale sodium-cooled Advance Burner Reactors (ABR) have been developed for both metal and oxide-fueled cores with conversion ratios from breakeven (CR=1.0) to fertile-free (CR=0.0). These designs are expected to satisfy thermal and irradiation damage limits based on the currently available data. The very low conversion ratio designs require fuel that is beyond

E. A. Hoffman; W. S. Yang; R. N. Hill

2008-01-01

297

Selection and benchmarking of computer codes for research reactor core conversions  

Microsoft Academic Search

A group of computer codes has been selected and obtained from the Nuclear Energy Agency data bank in France for the core conversion study of highly enriched research reactors. The ANISN, WIMS-D4, MC², COBRA-3M, FEVER, THERMOS, GAM-2, CINDER, and EXTERMINATOR codes were selected for the study. For the final work, THERMOS, GAM-2, CINDER, and EXTERMINATOR were selected and used. A

E. Yilmaz; B. G. Jones

1984-01-01

298

Fuel and core testing plan for a target fueled isotope production reactor.  

SciTech Connect

In recent years there has been an unstable supply of the critical diagnostic medical isotope 99Tc. Several concepts and designs have been proposed to produce 99Mo the parent nuclide of 99Tc, at a commercial scale sufficient to stabilize the world supply. This work lays out a testing and experiment plan for a proposed 2 MW open pool reactor fueled by Low Enriched Uranium (LEU) 99Mo targets. The experiments and tests necessary to support licensing of the reactor design are described and how these experiments and tests will help establish the safe operating envelop for a medical isotope production reactor is discussed. The experiments and tests will facilitate a focused and efficient licensing process in order to bring on line a needed production reactor dedicated to supplying medical isotopes. The Target Fuel Isotope Reactor (TFIR) design calls for an active core region that is approximately 40 cm in diameter and 40 cm in fuel height. It contains up to 150 cylindrical, 1-cm diameter, LEU oxide fuel pins clad with Zircaloy (zirconium alloy), in an annular hexagonal array on a {approx}2.0 cm pitch surrounded, radially, by a graphite or a Be reflector. The reactor is similar to U.S. university reactors in power, hardware, and safety/control systems. Fuel/target pin fabrication is based on existing light water reactor fuel fabrication processes. However, as part of licensing process, experiments must be conducted to confirm analytical predictions of steady-state power and accident conditions. The experiment and test plan will be conducted in phases and will utilize existing facilities at the U.S. Department of Energy's Sandia National Laboratories. The first phase is to validate the predicted reactor core neutronics at delayed critical, zero power and very low power. This will be accomplished by using the Sandia Critical Experiment (CX) platform. A full scale TFIR core will be built in the CX and delayed critical measurements will be taken. For low power experiments, fuel pins can be removed after the experiment and using Sandia's metrology lab, relative power profiles (radially and axially) can be determined. In addition to validating neutronic analyses, confirming heat transfer properties of the target/fuel pins and core will be conducted. Fuel/target pin power limits can be verified with out-of-pile (electrical heating) thermal-hydraulic experiments. This will yield data on the heat flux across the Zircaloy clad and establish safety margin and operating limits. Using Sandia's Annular Core Research Reactor (ACRR) a 4 MW TRIGA type research reactor, target/fuel pins can be driven to desired fission power levels for long durations. Post experiment inspection of the pins can be conducted in the Auxiliary Hot Cell Facility to observe changes in the mechanical properties of the LEU matrix and burn-up effects. Transient tests can also be conducted at the ACRR to observe target/fuel pin performance during accident conditions. Target/fuel pins will be placed in double experiment containment and driven by pulsing the ACRR until target/fuel failure is observed. This will allow for extrapolation of analytical work to confirm safety margins.

Coats, Richard Lee; Dahl, James J.; Parma, Edward J., Jr.

2010-12-01

299

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

SciTech Connect

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

Sterbentz, James W

2007-05-01

300

Performance characteristics of the annular core research reactor fuel motion detection system  

SciTech Connect

Recent proof tests have shown that the annular core research reactor (ACRR) fuel motion detection system has reached its design goals of providing high temporal and spatial resolution pictures of fuel distributions in the ACRR. The coded aperture imaging system (CAIS) images the fuel by monitoring the fission gamma rays from the fuel that pass through collimators in the reactor core. The gamma-ray beam is modulated by coded apertures before producing a visible light coded image in thin scintillators. Each coded image is then amplified and recorded by an opticalimage-intensifier/fast-framing-camera combination. The proximity to the core and the coded aperture technique provide a high data collection rate and high resolution. Experiments of CAIS at the ACRR conducted under steady-state operation have documented the beneficial effects of changes in the radiation shielding and imaging technique. The spatial resolutions are 1.7 mm perpendicular to the axis of a single liquid-metal fast breeder reactor fuel pin and 9 mm in the axial dimension. Changes in mass of 100 mg in each resolution element can be detected each frame period, which may be from 5 to 100 ms. This diagnostic instrument may help resolve important questions in fuel motion phenomenology.

Kelly, J.G.; Stalker, K.T.

1983-12-01

301

Iris feature extraction using gabor filter  

Microsoft Academic Search

Biometric technology uses human characteristics for their reliable identification. Iris recognition is a biometric technology that utilizes iris for human identification. The human iris contains very discriminating features and hence provides the accurate authentication of persons. To extract the discriminating iris features, different methods have been used in the past. In this work, Gabor filter is applied on iris images

Saadia Minhas; Muhammad Younus Javed

2009-01-01

302

The Annular Core Research Reactor (ACRR) postulated limiting event initial and building source terms  

SciTech Connect

As part of the update of the Safety analysis Report (SAR) for the Annular Core Research Reactor (ACRR), operational limiting events under the category of inadvertent withdrawal of an experiment while at power or during a power pulse were determined to be the most limiting event(s) for this reactor. This report provides a summary of the assumptions, modeling, and results in evaluation of: Reactivity and thermal hydraulics analysis to determine the amount of fuel melt or fuel damage ratios; The reactor inventories following the limiting event; A literature review of post NUREG-0772 release fraction experiment results on severe fuel damages; Decontamination factors due to in-pool transport; and In-building transport modeling and building source term analysis.

Restrepo, L F

1992-08-01

303

R and D program for core instrumentation improvements devoted for French sodium fast reactors  

SciTech Connect

Under the framework of French R and D studies for Generation IV reactors and more specifically for sodium-cooled fast reactors (SFR); the CEA, EDF and AREVA have launched a joint coordinated research programme. This paper deals with the R and D sets out to achieve better inspection, maintenance, availability and decommissioning. In particular the instrumentation requirements for core monitoring and detection in the case of accidental events. Requirements mainly involve diversifying the means of protection and improving instrumentation performance in terms of responsiveness and sensitivity. Operation feedback from the Phenix and Superphenix prototype reactors and studies, carried out within the scope of the EFR projects, has been used to define the needs for instrumentation enhancement. (authors)

Jeannot, J. P.; Rodriguez, G.; Jammes, C.; Bernardin, B.; Portier, J. L.; Jadot, F. [Commissariat a l'Energie Atomique, Saint-Paul-lez-Durance, 13108 (France); Maire, S.; Verrier, D. [Advanced Projects and Decommissioning Div. Plant Sector AREVA NP - NEPL-FT, Lyon, 69000 (France); Loisy, F. [EDF - EDF R and D STEP Dept., 6 Quai Watier, Chatou, 78401 (France); Prele, G. [EDF, Generation/Nuclear Engineering, Basic Design Dept., Villeurbanne, 69628 (France)

2011-07-01

304

An efficient iris segmentation approach  

NASA Astrophysics Data System (ADS)

Iris recognition system became a reliable system for authentication and verification tasks. It consists of five stages: image acquisition, iris segmentation, iris normalization, feature encoding, and feature matching. Iris segmentation stage is one of the most important stages. It plays an essential role to locate the iris efficiently and accurately. In this paper, we present a new approach for iris segmentation using image processing technique. This approach is composed of four main parts. (1) Eliminating reflections of light on the eye image based on inverting the color of the grayscale image, filling holes in the intensity image, and inverting the color of the intensity image to get the original grayscale image without any reflections. (2) Pupil boundary detection based on dividing an eye image to nine sub-images and finding the minimum value of the mean intensity for each sub-image to get a suitable threshold value of pupil. (3) Enhancing the contrast of outer iris boundary using exponential operator to have sharp variation. (4) Outer iris boundary localization based on applying a gray threshold and morphological operations on the rectangular part of an eye image including the pupil and the outer boundaries of iris to find the small radius of outer iris boundary from the center of pupil. The proposed approach has been tested on CASIA v1.0 iris image database and other collected iris image database. The experimental results show that the approach is able to detect pupil and outer iris boundary with high accuracy results approximately 100% and reduce time consuming.

Gomai, Abdu; El-Zaart, A.; Mathkour, H.

2011-10-01

305

Effects of Iris Surface Curvature on Iris Recognition  

SciTech Connect

To focus on objects at various distances, the lens of the eye must change shape to adjust its refractive power. This change in lens shape causes a change in the shape of the iris surface which can be measured by examining the curvature of the iris. This work isolates the variable of iris curvature in the recognition process and shows that differences in iris curvature degrade matching ability. To our knowledge, no other work has examined the effects of varying iris curvature on matching ability. To examine this degradation, we conduct a matching experiment across pairs of images with varying degrees of iris curvature differences. The results show a statistically signi cant degradation in matching ability. Finally, the real world impact of these ndings is discussed

Thompson, Joseph T [ORNL; Flynn, Patrick J [ORNL; Bowyer, Kevin W [University of Notre Dame, IN; Santos-Villalobos, Hector J [ORNL

2013-01-01

306

Searching for 'Fragile Bits' in Iris Codes Generated with Gabor Analytic Iris Texture Binary Encoder  

Microsoft Academic Search

This paper presents a new methodology for generating iris binary codes using Circular Fuzzy Iris Segmentation and Gabor Analytic Iris Texture Binary Encoder. Iris images from Bath University Iris Database are encoded as iris codes at three difierent lengths (192, 512, 768 Bytes) and used to test two hypostases of the concept of fragile bits in both single-enrollment and multi-enrollment

Nicolaie Popescu-Bodorin

2009-01-01

307

Results of Reactor Materials Experiments Investigating 2-D Core-Concrete Interaction and Debris Coolability  

SciTech Connect

The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program is conducting reactor materials experiments and associated analysis to achieve the following objectives: 1) resolution of the ex-vessel debris coolability issue, and 2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants and provide the technical basis for better containment designs of future plants. With respect to the second objective, there remain uncertainties in the models that evaluate the lateral vs. axial power split during core-concrete interaction because of a lack of truly two-dimensional experiment data. As a first step in bridging this gap, a large scale Core-Concrete Interaction experiment (CCI-1) has been conducted as part of the MCCI program. This test investigated the interaction of a 400 kg core-oxide melt with a crucible made of siliceous concrete along two walls and the base. The two remaining walls were made of non-ablative magnesium oxide. The initial phase of the test was conducted under dry conditions. After a predefined ablation depth was achieved, the cavity was flooded to obtain data on the coolability of a core melt after core-concrete interaction has progressed for some time. This paper provides a description of the test facility and an overview of results from this test. (authors)

Farmer, M. T.; Lomperski, S. [Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Basu, S. [U.S. Nuclear Regulatory Commission, MS-T10K8, 11545 Rockville Pike, Rockville, MD 20852 (United States)

2004-07-01

308

Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model''  

SciTech Connect

OAK (B204) Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model'' The present generation of reactor analysis methods uses few-group nodal diffusion approximations to calculate full-core eigenvalues and power distributions. The cross sections, diffusion coefficients, and discontinuity factors (collectively called ''group constants'') in the nodal diffusion equations are parameterized as functions of many variables, ranging from the obvious (temperature, boron concentration, etc.) to the more obscure (spectral index, moderator temperature history, etc.). These group constants, and their variations as functions of the many variables, are calculated by assembly-level transport codes. The current methodology has two main weaknesses that this project addressed. The first weakness is the diffusion approximation in the full-core calculation; this can be significantly inaccurate at interfaces between different assemblies. This project used the nodal diffusion framework to implement nodal quasidiffusion equations, which can capture transport effects to an arbitrary degree of accuracy. The second weakness is in the parameterization of the group constants; current models do not always perform well, especially at interfaces between unlike assemblies. The project developed a theoretical foundation for parameterization and homogenization models and used that theory to devise improved models. The new models were extended to tabulate information that the nodal quasidiffusion equations can use to capture transport effects in full-core calculations.

Dmitriy Y. Anistratov; Marvin L. Adams; Todd S. Palmer; Kord S. Smith; Kevin Clarno; Hikaru Hiruta; Razvan Nes

2003-08-04

309

A particle-bed gas cooled fast reactor core design for waste minimization.  

SciTech Connect

The issue of waste minimization in advanced reactor systems has been investigated using the Particle-Bed Gas-Cooled Fast Reactor (PB-GCFR) design being developed and funded under the U.S. Department of Energy Nuclear Energy Research Initiative (USDOE NERI) Program. Results indicate that for the given core power density and constraint on the maximum TRU enrichment allowable, the lowest amount of radiotoxic transuranics to be processed and hence sent to the repository is obtained for long-life core designs. Calculations were additionally done to investigate long-life core designs using LWR spent fuel TRU and recycle TRU, and different feed, matrix and reflector materials. The recycled TRU and LWR spent TRU fuels give similar core behaviors, because of the fast spectrum environment which does not significantly degrade the TRU composition. Using light elements as reflector material was found to be unattractive because of power peaking problems and large reactivity swings. The application of a lead reflector gave the longest cycle length and lowest TRU processing requirement. Materials compatibility and performance issues require additional investigation.

Hoffman, E. A.; Taiwo, T. A.; Yang, W. S.; Fatone, M.

2002-10-11

310

Investigation on the Core Bypass Flow in a Very High Temperature Reactor  

SciTech Connect

Uncertainties associated with the core bypass flow are some of the key issues that directly influence the coolant mass flow distribution and magnitude, and thus the operational core temperature profiles, in the very high-temperature reactor (VHTR). Designers will attempt to configure the core geometry so the core cooling flow rate magnitude and distribution conform to the design values. The objective of this project is to study the bypass flow both experimentally and computationally. Researchers will develop experimental data using state-of-the-art particle image velocimetry in a small test facility. The team will attempt to obtain full field temperature distribution using racks of thermocouples. The experimental data are intended to benchmark computational fluid dynamics (CFD) codes by providing detailed information. These experimental data are urgently needed for validation of the CFD codes. The following are the project tasks: Construct a small-scale bench-top experiment to resemble the bypass flow between the graphite blocks, varying parameters to address their impact on bypass flow. Wall roughness of the graphite block walls, spacing between the blocks, and temperature of the blocks are some of the parameters to be tested. Perform CFD to evaluate pre- and post-test calculations and turbulence models, including sensitivity studies to achieve high accuracy. Develop the state-of-the art large eddy simulation (LES) using appropriate subgrid modeling. Develop models to be used in systems thermal hydraulics codes to account and estimate the bypass flows. These computer programs include, among others, RELAP3D, MELCOR, GAMMA, and GAS-NET. Actual core bypass flow rate may vary considerably from the design value. Although the uncertainty of the bypass flow rate is not known, some sources have stated that the bypass flow rates in the Fort St. Vrain reactor were between 8 and 25 percent of the total reactor mass flow rate. If bypass flow rates are on the high side, the quantity of cooling flow through the core may be considerably less than the nominal design value, causing some regions of the core to operate at temperatures in excess of the design values. These effects are postulated to lead to localized hot regions in the core that must be considered when evaluating the VHTR operational and accident scenarios.

Hassan, Yassin

2013-10-22

311

Thermohydraulics in a high-temperature gas-cooled reactor prestressed-concrete reactor vessel during unrestricted core-heatup accidents  

Microsoft Academic Search

The hypothetical accident considered for siting considerations in High Temperature Gas-Cooled Reactors (HTGR) is the so called Unrestricted Core Heatup Accident (UCHA), in which all forced circulation is lost at initiation, and none of the auxillary cooling loops can be started. The result is a gradual slow core heatup, extending over days. Whether the liner cooling system (LCS) operates during

P. G. Kroeger; J. Colman; K. Araj

1983-01-01

312

Non-Proliferative, Thorium-Based, Core and Fuel Cycle for Pressurized Water Reactors  

SciTech Connect

Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a 'conventional' assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical 'blanket') and the 'driving' part of the core (a supercritical 'seed'). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER's/PWR's. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the amount of discharged spent fuel, for a given energy production, compared with standard VVER/PWR. The total Pu production rate of RTF cycles is only 30 % of standard reactor. In addition, the isotopic compositions of the RTF's and standard reactor grade Pu are markedly different due to the very high burnup accumulated by the RTF spent fuel.

Todosow M.; Todosow M.; Raitses, G. (BNL) Galperin, A. (Ben Gurion University)

2009-07-12

313

Explication of design data of the graphite structural design code for core support components of High Temperature Engineering Test Reactor.  

National Technical Information Service (NTIS)

The graphite and carbon materials used for the core support graphite components of High Temperature Engineering Test Reactor (HTTR) are nuclear grade fine-grained isotropic graphite (IG-110), nuclear grade medium-grained near-isotropic graphite (PGX) and ...

M. Ishihara T. Iyoku S. Sato S. Shiozawa J. Toyota

1991-01-01

314

A point kernel model for the energy deposited on samples from gamma radiation in a research reactor core  

Microsoft Academic Search

A basic safety requirement for a research reactor is the reliable estimation of the gamma heating of samples irradiated in the reactor core. A three-dimensional numerical code of gamma heating using a point kernel parameterization is developed. The heating due to ?-rays, produced from U235 fission and from (n,?) reactions with the core materials is considered. The dose build-up due

M. Varvayanni; N. Catsaros; M. Antonopoulos-Domis

2008-01-01

315

THERMIT: a computer program for three-dimensional thermal-hydraulic analysis of light-water-reactor cores. Final report  

Microsoft Academic Search

THERMIT is a computer code that solves the time dependent two-fluid thermal-hydraulic equations in three space dimensions along with the fuel pin and clad temperatures. Its purpose is to predict the thermal-hydraulic response of a modeling of a reactor core to situations which do not lead to reconfiguration of the core geometry. Both boiling and pressurized water reactors may be

J. Loomis; W. H. Reed; A. Schor; H. B. Stewart; L. Wolf

1981-01-01

316

Proliferation resistance potential and burnup characteristics of an equilibrium core of novel natural uranium fueled nuclear research reactor  

Microsoft Academic Search

Standard reactor simulation codes WIMS-D\\/4 and CITATION were employed to analyze the proliferation resistance potential and burnup characteristics of a novel natural uranium fueled nuclear research reactor [Annals of Nuclear Energy 31(12), 13311356]. It was found that the proposed core, which provides twice the flux per unit core power compared to similar natural uranium fueled, light water cooled, heavy water

Mohammad Javed Khan; Aslam; Nasir Ahmad

2005-01-01

317

Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris  

DOEpatents

The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

Gabor, John D. (Western Springs, IL); Cassulo, John C. (Stickney, IL); Pedersen, Dean R. (Naperville, IL); Baker, Jr., Louis (Downers Grove, IL)

1986-01-01

318

Simulated Verification of Fuel Element Inventory in a Small Reactor Core Using the Nuclear Materials Identification System (NMIS)  

SciTech Connect

The International Panel on Climate Change projects that by 2050 the world energy demand may double. Although the primary focus for new nuclear power plants in industrialized nations is on large plants in the 1000-1600 MWe range, there is an increasing demand for small and medium reactors (SMRs). About half of the innovative SMR concepts are for small (<300 MWe) reactors with a 5-30 year life without on-site refueling. This type of reactor is also known as a battery-type reactor. These reactors are particularly attractive to countries with small power grids and for non-electrical purposes such as heating, hydrogen production, and seawater desalination. Traditionally, this type of reactor has been used in a nautical propulsion role. This type of reactor is designed as a permanently sealed unit to prevent the diversion of the uranium in the core by the user. However, after initial fabrication it will be necessary to verify that the newly fabricated reactor core contains the quantity of uranium that initially entered the fuel fabrication plant. In most instances, traditional inspection techniques can be used to perform this verification, but in certain situations the core design will be considered sensitive. Non-intrusive verification techniques must be utilized in these situations. The Nuclear Materials Identification System (NMIS) with imaging uses active interrogation and a fast time correlation processor to characterize fissile material. The MCNP-PoliMi computer code was used to simulate NMIS measurements of a small, sealed reactor core. Because most battery-type reactor designs are still in the early design phase, a more traditional design based on a Russian icebreaker core was used in the simulations. These simulations show how the radiography capabilities of the NMIS could be used to detect the diversion of fissile material by detecting void areas in the assembled core where fuel elements have been removed.

Grogan, Brandon R [ORNL; Mihalczo, John T [ORNL

2009-01-01

319

IRI: An international Rawer initiative  

NASA Technical Reports Server (NTRS)

This paper was presented during the special session that was held at the 1993 International Reference Ionosphere (IRI) Workshop in honor of Karl Rawer's 80th birthday. It retraces the steps that led from the start of the IRI project to the present edition of the model highlighting the important role that the honoree played in guiding IRI from infancy to maturity. All summary view graphs are reproduced at the end of the article.

Bilitza, D.

1995-01-01

320

Complications of cosmetic iris implants.  

PubMed

Cosmetic intraocular iris implants for the purpose of changing iris color have recently been developed; however, little is known about their safety. We report a patient who had bilateral implantation of colored silicone iris implants solely for cosmetic reasons. The rapid development of uveitis, corneal decompensation, and ocular hypertension resulted in the need for explantation of the implants. Placement of these devices should require specific medical indications and meticulous surgery with early and long-term evaluation. PMID:18571095

Thiagalingam, Sureka; Tarongoy, Pamela; Hamrah, Pedram; Lobo, Ann-Marie; Nagao, Karina; Barsam, Charles; Bellows, Robert; Pineda, Roberto

2008-07-01

321

Ordinal measures for iris recognition.  

PubMed

Images of a human iris contain rich texture information useful for identity authentication. A key and still open issue in iris recognition is how best to represent such textural information using a compact set of features (iris features). In this paper, we propose using ordinal measures for iris feature representation with the objective of characterizing qualitative relationships between iris regions rather than precise measurements of iris image structures. Such a representation may lose some image-specific information, but it achieves a good trade-off between distinctiveness and robustness. We show that ordinal measures are intrinsic features of iris patterns and largely invariant to illumination changes. Moreover, compactness and low computational complexity of ordinal measures enable highly efficient iris recognition. Ordinal measures are a general concept useful for image analysis and many variants can be derived for ordinal feature extraction. In this paper, we develop multilobe differential filters to compute ordinal measures with flexible intralobe and interlobe parameters such as location, scale, orientation, and distance. Experimental results on three public iris image databases demonstrate the effectiveness of the proposed ordinal feature models. PMID:19834142

Sun, Zhenan; Tan, Tieniu

2009-12-01

322

Improved Topside Model For Iri  

NASA Astrophysics Data System (ADS)

The International Reference Ionosphere (IRI) is a data-based standard model for iono- spheric densities and temperatures. Shortcomings of the IRI topside electron density model have been noted in comparisons with insitu measurements and topside sounder data. We will discuss improvements of the IRI topside model based on a large data base of electron density profiles deduced from Alouette and ISIS topside ionograms. The data were used to obtain correction factors for the IRI model and to establish a better functional description of the transition region from oxygen ions to light ions.

Bilitza, D.

323

Burnup analysis and in-core fuel management study of the 3 MW TRIGA MARK II research reactor  

Microsoft Academic Search

The principal objective of this study is to formulate an effective optimal fuel management strategy for the TRIGA MARK II research reactor at AERE, Savar. The core management study has been performed by utilizing four basic types of information calculated for the reactor: criticality, power peaking, neutron flux and burnup calculation. This paper presents the results of the burnup calculations

M. Q. Huda; S. I. Bhuiyan; T. Obara

2008-01-01

324

Fission-chamber-compensated self-powered detector for in-core flux measurement and reactor control  

Microsoft Academic Search

An apparatus is described for in-core flux measurement and nuclear reactor control consisting of: a self-powered rhodium neutron detector for producing an output signal corresponding to reactor power level; first amplifier means having an input for receiving the neutron detector output signal, and the first amplifier means producing a corresponding first amplifier output signal (V); a fission chamber for producing

Neissel

1986-01-01

325

Gas Core Reactor-MHD Power System with Cascading Power Cycle  

SciTech Connect

The US Department of Energy initiative Gen-IV aim is to produce an entire nuclear energy production system with next generation features for certification before 2030. A Generation 4 capable system must have superior sustainability, safety and reliability, and economic cost advantages in comparison with third generation light water reactors. A gas core reactor (GCR) with magnetohydrodynamic (MHD) power converter and cascading power cycle forms the basis for a Generation IV concept that is expected to set the upper performance limits in sustainability and power conversion efficiency among all existing and proposed fission powered systems. A gaseous core reactor delivering 1000's MW fission power acts as the heat source for a high temperature magnetohydrodynamic power converter. A uranium tetrafluoride fuel mix, with {approx}95% mole fraction helium gas, provides a stable working fluid for the primary MHD-Brayton cycle. A helium Brayton cycle extracts waste heat from the MHD generator with about 20% energy efficiency, but the low temperature side is still hot enough ({approx}1600 K) to drive a second conventional helium Brayton cycle with about 35% efficiency. There is enough heat at the low temperature side of the He-Brayton cycle to generate steam, and so another heat recovery cycle can be added, this time a Rankine steam cycle with up to 40% efficiency. The proof of concept does not require a tremendously efficient (first law) MHD cycle, the high temperature direct energy conversion capability of an MHD dynamo, combined with already sophisticated steam powered turbine industry knowledge base allows the cascading cycle design to achieve break-through first law energy efficiencies previously unheard of in the nuclear power industry. Although simple in concept, the gas core reactor design has not achieved the state of technological maturity that, say, molten salt or high-temperature gas-cooled reactors have pioneered. However, even on paper the GCR-MHD concept holds considerable promise, for example, like molten salt reactors the fuel is continuously cycled, allowing high-burnup, and continuous burning of actinides, and hence greatly improved fuel utilization. The fuel inventory is two orders of magnitude lower than LWR's of comparable power output and fissile plutonium production is likewise lower than in spent LWR fuel. Besides these features this paper discusses specific GCR-MHD design challenges such as fission enhanced gas conductivity in the MHD channel, GCR safety issues and related engineering problems. (authors)

Smith, Blair M.; Anghaie, Samim; Knight, Travis W. [Innovative Nuclear Space Power and Propulsion Institute, University of Florida, PO Box 116502, Gainesville, FL, 32611 (United States)

2002-07-01

326

Thermal-Hydraulic Evaluation Study of the Effectiveness of Emergency Core Cooling System for Light Water Reactors.  

National Technical Information Service (NTIS)

In order to evaluate the core cooling capability of the emergeny core cooling system, which is a safety guard system of light water reactors for a loss-of-coolant accident, a variety of large scale test were performed. Through the results, many phenomena ...

M. Sobajima

1985-01-01

327

Disassembly and defueling of the TMI2 (Three Mile Island Unit 2) reactor vessel lower core support assembly  

Microsoft Academic Search

Planning for the disassembly and defueling of the Three Mile Island Unit 2 (TMI-2) reactor lower core support assembly (LCSA) began early in 1985. Evaluations of methods of defueling were performed based on various assumed LCSA conditions. Tooling was conceptualized and various defueling sequences were evaluated. As defueling of the core region progressed, information and data were obtained that clearly

W. E. Austin; L. H. Porter

1988-01-01

328

Feasibility of Monitoring the Strength of HTGR (High-Temperature Gas-Cooled Reactor) Core Support Graphite. Part III.  

National Technical Information Service (NTIS)

Methods are being developed to monitor, in-situ, the strength changes of graphite core-support components in a High-Temperature Gas-Cooled Reactor (HTGR). The results reported herein pertain to the development of techniques for monitoring the core-support...

W. C. Morgan T. J. Davis M. T. Thomas

1983-01-01

329

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code  

SciTech Connect

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

Vondy, D.R.

1984-07-01

330

A 100 MWe advanced sodium-cooled fast reactor core concept  

SciTech Connect

An Advanced sodium-cooled Fast Reactor core concept (AFR-100) was developed targeting a small electrical grid to be transportable to the plant site and operable for a long time without frequent refueling. The reactor power rating was strategically decided to be 100 MWe, and the core barrel diameter was limited to 3.0 m for transportability. The design parameters were determined by relaxing the peak fast fluence limit and bulk coolant outlet temperature to beyond irradiation experience assuming that advanced cladding and structural materials developed under US-DOE programs would be available when the AFR-100 is deployed. With a de-rated power density and U-Zr binary metallic fuel, the AFR-100 can maintain criticality for 30 years without refueling. The average discharge burnup of 101 MWd/kg is comparable to conventional design values, but the peak discharge fast fluence of {approx}6x10{sup 23} neutrons/cm{sup 2} is beyond the current irradiation experiences with HT-9 cladding. The evaluated reactivity coefficients provide sufficient negative feedbacks and the reactivity control systems provide sufficient shutdown margins. The integral reactivity parameters obtained from quasi-static reactivity balance analysis indicate that the AFR-100 meets the sufficient conditions for acceptable asymptotic core outlet temperature following postulated unprotected accidents. Additionally, the AFR-100 has sufficient thermal margins by grouping the fuel assemblies into eight orifice zones. (authors)

Kim, T. K.; Grandy, C.; Hill, R. N. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

2012-07-01

331

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

332

An improved resonance self-shielding method for heterogeneous fast reactor assembly and core calculations  

SciTech Connect

An improved resonance self-shielding method has been developed to accurately estimate the effective multigroup cross sections for heterogeneous fast reactor assembly and core calculations. In the method, the heterogeneity effect is considered by the use of isotopic escape cross sections while the resonance interference effect is accounted for through the narrow resonance approximation or slowing-down calculations for specific compositions. The isotopic escape cross sections are calculated by solving fixed-source transport equations with the method of characteristics for the whole problem domain. This method requires no pre-calculated resonance integral tables or parameters that are typically necessary in the subgroup method. Preliminary results for multi pin-cell fast reactor problems show that the escape cross sections estimated from the explicit-geometry fixed source calculations produce more accurate eigenvalue and self-shielded effective cross sections than those from conventional one-dimensional geometry models. (authors)

Lee, C. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439-4842 (United States)] [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439-4842 (United States); Yang, W. S. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907-2017 (United States)] [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907-2017 (United States)

2013-07-01

333

Optical phenomena in KU-1 silica core fiber waveguides under pulsed reactor irradiation  

NASA Astrophysics Data System (ADS)

The light emission intensity and transient optical absorption (TOA) have been measured in the wavelength range 400-750 nm in KU-1 silica core fibers under irradiation of the BARS-6 pulsed fission reactor (pulse duration 80 ?s, fluence per pulse up to 510 12 n/ cm2 ( E>0.5 eV), dose rate up to 10 5 Gy/s). The fast light emission component attributed to Cherenkov radiation is followed by a weak emission tail with the characteristic time (15050) ?s. The transient absorption reaches 2.510 -4 cm-1 (relaxation time 600-1200 ?s). The sub-linear dependence of Cherenkov radiation on the dose rate and the occurrence of both the tail of light emission and transient absorption are ascribed to the appearance of optical inhomogeneities of the silica glass under intense pulsed reactor irradiation.

Demenkov, P. V.; Plaksin, O. A.; Stepanov, V. A.; Stepanov, P. A.; Chernov, V. M.; Golant, K. M.; Tomashuk, A. L.

334

Investigation of aerosols released at high temperature from nuclear reactor core models  

NASA Astrophysics Data System (ADS)

Two experiments were performed to simulate severe reactor accident with air ingress into the hot reactor core. The model bundles contained nine PWR type fuel rods. Their cladding was pre-oxidised by argon-oxygen (test 1) and steam (test 2). The released aerosol was measured continuously by laser particle counters. Morphology and elemental composition of the aerosol particles were studied on samples collected by impactors and quartz filters. The highest aerosol release was detected at the steepest rise of the bundle temperature. A second increase of the aerosol release appeared at the cooling down period. Because of the higher maximum temperature at test 2 about two orders of magnitude more uranium was released than in test 1. The highest emission was found for tin at test 1 and for zirconium and iron at test 2.

Pintr Csords, A.; Matus, L.; Czitrovszky, A.; Jani, P.; Marti, L.; Hzer, Z.; Windberg, P.; Hummel, R.

2000-12-01

335

Analysis of cracked core spray piping from the Quad Cities Unit 2 boiling water reactor  

SciTech Connect

The results of a metallurgical analysis of leaking cracks detected in the core spray injection piping of Commonwealth Edison Company's Quad Cities Unit 2 Boiling Water Reactor are described. The cracks were present in a welded 105/sup 0/ elbow assembly in the line, and were found to be caused by intergranular stress corrosion cracking associated with the probable presence of dissolved oxygen in the reactor cooling water and the presence of grain boundary sensitization and local residual stresses induced by welding. The failure is unusual in several respects, including the very large number of cracks (approximately 40) present in the failed component, the axial orientation of the cracks, and the fact that at least one crack completely penetrated a circumferential weld. Virtually all of the cracking occurred in forged material, and the microstructural evidence presented suggests that the orientation of the cracks was influenced by the presence of axially banded delta ferrite in the microstructure of the forged components.

Diercks, D.R.; Gaitonde, S.M.

1982-09-01

336

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

337

BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III  

SciTech Connect

This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

1981-06-01

338

An analysis of the flow field near the fuel injection location in a gas core reactor.  

NASA Technical Reports Server (NTRS)

An analytical study is presented which shows the effects of large energy release and the concurrent high acceleration of inner stream fluid on the coaxial flow field in a gas core reactor. The governing equations include the assumptions of only radial radiative transport of energy represented as an energy diffusion term in the Euler equations. The method of integral relations is used to obtain the numerical solution. Results show that the rapidly accelerating, heat generating inner stream actually shrinks in radius as it expands axially.

Weinstein, H.; Murty, B. G. K.; Porter, R. W.

1971-01-01

339

Design and fabrication of the instrumented fuel elements for the Annular Core Research Reactor (ACRR)  

SciTech Connect

This report describes the design and fabrication techniques for the instrumented fuel elements of the Annular Core Research Reactor (ACRR). The thermocouple assemblies were designed and fabricated at Sandia Laboratories while the instrumented elements were assembled at Los Alamos Scientific Laboratory. In order to satisfy the ACRR's Technical Specifications, the thermocouples are required to measure temperature in excess of 1800/sup 0/C under rapid heating conditions. Because of the potentially high failure rates for thermocouples in such environments, the instrumented fuel elements are designed so that the thermocouples can be replaced easily.

Schmidt, T.R.; Sasmor, D.J.; Martin, J.T.; Gonzalez, F.; Cox, D.N.

1981-04-01

340

The scalability of OTR (out-of-core thermionic reactor) space nuclear power systems  

SciTech Connect

In this document, masses of the STAR-C power system and an optimized out-of-core thermionic reactor (OTR) power system versus power level are investigated. The impacts of key system parameters on system performance are also addressed. The STAR-C is mass competitive below about 15 kWe, but at higher power levels the scalability is relatively poor. An optimized OR is the least massive space nuclear power system below 25 kWe, and scales well to 50 kWe. The system parameters that have a significant impact on the scalability of the STAR-C are core thermal flux, thermionic converter efficiency, and core length to diameter ratio. The emissivity of the core surface is shown to be a relatively unimportant parameter. For an optimized OR power system, the most significant system parameter is the maximum allowable fuel temperature. It is also shown that if advanced radiation-hardened electronics are used in the satellite payload, a very large mass savings is realized. 10 refs., 23 figs., 7 tabs.

Gallup, D.R.

1990-03-01

341

Does Iris Change Over Time?  

PubMed Central

Iris as a biometric identifier is assumed to be stable over a period of time. However, some researchers have observed that for long time lapse, the genuine match score distribution shifts towards the impostor score distribution and the performance of iris recognition reduces. The main purpose of this study is to determine if the shift in genuine scores can be attributed to aging or not. The experiments are performed on the two publicly available iris aging databases namely, ND-Iris-Template-Aging-20082010 and ND-TimeLapseIris-2012 using a commercial matcher, VeriEye. While existing results are correct about increase in false rejection over time, we observe that it is primarily due to the presence of other covariates such as blur, noise, occlusion, and pupil dilation. This claim is substantiated with quality score comparison of the gallery and probe pairs.

Mehrotra, Hunny; Vatsa, Mayank; Singh, Richa; Majhi, Banshidhar

2013-01-01

342

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

SciTech Connect

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

Fourmentel, D.; Villard, J. F.; Lyoussi, A. [DEN Reactor Studies Dept., French Nuclear Energy and Alternative Energies Commission, CEA Cadarache, 13108 Saint Paul-Lez-Durance (France); Reynard-Carette, C. [Laboratoire Chimie Provence LCP UMR 6264, Univ. of Provence, Centre St. Jerome, 13397 Marseille Cedex 20 (France); Bignan, G.; Chauvin, J. P.; Gonnier, C.; Guimbal, P.; Malo, J. Y. [DEN Reactor Studies Dept., French Nuclear Energy and Alternative Energies Commission, CEA Cadarache, 13108 Saint Paul-Lez-Durance (France); Carette, M.; Janulyte, A.; Merroun, O.; Brun, J.; Zerega, Y.; Andre, J. [Laboratoire Chimie Provence LCP UMR 6264, Univ. of Provence, Centre St. Jerome, 13397 Marseille Cedex 20 (France)

2011-07-01

343

Hurricane Iris Hits Belize  

NASA Technical Reports Server (NTRS)

Hurricane Iris hit the small Central American country of Belize around midnight on October 8, 2001. At the time, Iris was the strongest Atlantic hurricane of the season, with sustained winds up to 225 kilometers per hour (140 mph). The hurricane caused severe damage-destroying homes, flooding streets, and leveling trees-in coastal towns south of Belize City. In addition, a boat of American recreational scuba divers docked along the coast was capsized by the storm, leaving 20 of the 28 passengers missing. Within hours the winds had subsided to only 56 kph (35 mph), a modest tropical depression, but Mexico, Guatemala, El Salvador, and Honduras were still expecting heavy rains. The above image is a combination of visible and thermal infrared data (for clouds) acquired by a NOAA Geostationary Operational Environmental Satellite (GOES-8) on October 8, 2001, at 2:45 p.m., and the Moderate-resolution Imaging Spectroradiometer (MODIS) (for the color of the ground). The three-dimensional view is from the south-southeast (north is towards the upper left). Belize is off the image to the left. Image courtesy Marit Jentoft-Nilsen, NASA GSFC Visualization Analysis Lab

2002-01-01

344

Hybrid parallel code acceleration methods in full-core reactor physics calculations  

SciTech Connect

When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadrature required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)

Courau, T.; Plagne, L.; Ponicot, A. [EDF R and D, 1, Avenue du General de Gaulle, 92141 Clamart Cedex (France); Sjoden, G. [Nuclear and Radiological Engineering, Georgia Inst. of Technology, Atlanta, GA 30332 (United States)

2012-07-01

345

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

346

High temperature UF6 RF plasma experiments applicable to uranium plasma core reactors  

NASA Technical Reports Server (NTRS)

An investigation was conducted using a 1.2 MW RF induction heater facility to aid in developing the technology necessary for designing a self critical fissioning uranium plasma core reactor. Pure, high temperature uranium hexafluoride (UF6) was injected into an argon fluid mechanically confined, steady state, RF heated plasma while employing different exhaust systems and diagnostic techniques to simulate and investigate some potential characteristics of uranium plasma core nuclear reactors. The development of techniques and equipment for fluid mechanical confinement of RF heated uranium plasmas with a high density of uranium vapor within the plasma, while simultaneously minimizing deposition of uranium and uranium compounds on the test chamber peripheral wall, endwall surfaces, and primary exhaust ducts, is discussed. The material tests and handling techniques suitable for use with high temperature, high pressure, gaseous UF6 are described and the development of complementary diagnostic instrumentation and measurement techniques to characterize the uranium plasma, effluent exhaust gases, and residue deposited on the test chamber and exhaust system components is reported.

Roman, W. C.

1979-01-01

347

Recriticality in a BWR (boiling water reactor) following a core damage event  

SciTech Connect

This report describes the results of a study conducted by Pacific Northwest Laboratory to assist the US Nuclear Regulatory Commission in evaluating the potential for recriticality in boiling water reactors (BWRs) during certain low probability severe accidents. Based on a conservative bounding analysis, this report concludes that there is a potential for recriticality in BWRs if core reflood occurs after control blade melting has begun but prior to significant fuel rod melting. However, a recriticality event will most likely not generate a pressure pulse significant enough to fail the vessel. Instead, a quasi-steady power level would result and the containment pressure and temperature would increase until the containment failure pressure is reached, unless actions are taken to terminate the event. Two strategies are identified that would aid in regaining control of the reactor and terminate the recriticality event before containment failure pressures are reached. The first strategy involves initiating boration injection at or before the time of core reflood if the potential for control blade melting exists. The second strategy involves initiating residual heat removal suppression pool cooling to remove the heat load generated by the recriticality event and thus extend the time available for boration. 31 figs., 17 tabs.

Scott, W.B.; Harrison, D.G.; Libby, R.A.; Tokarz, R.D. (Pacific Northwest Lab., Richland, WA (USA)); Wooton, R.D.; Denning, R.S.; Tayloe, R.W. Jr. (Battelle Memorial Inst., Columbus, OH (USA))

1990-12-01

348

Learning Based Resolution Enhancement of Iris Images  

Microsoft Academic Search

Iris recognition is one of the most reliable personal identification meth- ods. The potential requirement of obtaining high accuracy is that users sup- ply iris images with good quality. It is thus necessary for an iris recognition system to operate the possibly blurred iris images due to less cooperation of users and camera with low resolution. This paper proposes a

Junzhou Huang; Li Ma; Tieniu Tan; Yunhong Wang

349

Improved Masek approach for iris localization  

Microsoft Academic Search

Iris recognition technology has become famous in security applications because of its accuracy, safety and noninvasive biometric technologies. It demonstrates its efficiency as biometric-based authentication. This technology take advantages of random variations in the visible features of iris which is the colored part surrounding the pupil. Iris segmentation is the first and the key step at any iris recognition system.

Walid Aydi; Nouri Masmoudi; Lotfi Kamoun

2011-01-01

350

Exploring New Directions in Iris Recognition  

Microsoft Academic Search

A new approach in iris recognition based on Circular Fuzzy Iris Segmentation (CFIS) and Gabor Analytic Iris Texture Binary Encoder (GAITBE) is proposed and tested here. CFIS procedure is designed to guarantee that similar iris segments will be obtained for similar eye images, despite the fact that the degree of occlusion may vary from one image to another. Its result

Nicolaie Popescu-Bodorin

2011-01-01

351

Exploring New Directions in Iris Recognition  

Microsoft Academic Search

A new approach in iris recognition based on Circular Fuzzy Iris Segmentation (CFIS) and Gabor Analytic Iris Texture Binary Encoder (GAITBE) is proposed and tested here. CFIS procedure is designed to guarantee that similar iris segments will be obtained for similar eye images, despite the fact that the degree of occlusion may vary from one image to another. Its result

N. Popescu-Bodorin

2009-01-01

352

Learning Based Enhancement Model of Iris  

Microsoft Academic Search

Iris recognition is one of the most reliable personal identification meth- ods. The potential requirement of obtaining high accuracy is that users sup- ply iris images with good quality. It is thus necessary for an iris recognition system to operate the possibly blurred iris images due to less cooperation of users and camera with low resolution. This paper proposes a

Junzhou Huang; Li Ma; Tieniu Tan; Yunhong Wang

353

The DF4 fuel damage experiment in ACRR (Annual Core Research Reactor) with a BWR (Boiling Water Reactor) control blade and channel box  

Microsoft Academic Search

The DF-4 test was an experimental investigation into the melt progression behavior of boiling water reactor (BWR) core components under high temperature severe core damage conditions. In this study 14 zircaloy clad UO fuel rods, and representations of the zircaloy fuel canister and stainless steel\\/BC control blade were assembled into a 0.5 m long test bundle. The test bundle was

R. O. Gauntt; R. D. Gasser; L. J. Ott

1989-01-01

354

Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)  

NASA Technical Reports Server (NTRS)

The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

Clement, J. D.; Rust, J. H.; Schneider, A.; Hohl, F.

1976-01-01

355

Liquid-metal fast breeder reactor core transient modeling for faster than real-time analysis  

SciTech Connect

A model was developed for faster than real-time liquid-metal fast breeder reactor core transient analysis for purposes of continuous on-line data validation, plant state verification, and fault identification. The basic feature of this model is the use of a nodal approximation for the coolant, cladding, and fuel temperatures that gives adequately accurate power and temperature predictions with very few axial nodes. In applications of this methodology to fast loss-of-flow and overpower transients, computation times of about one-thirtieth of the real transient time per thermal-hydraulic channel were obtained. The predicted coolant and cladding temperature distributions were practically identical to those resulting from detailed finite difference computations. The predicted fuel temperatures differed by -- 1% or less from those obtained from the same finite difference computations. The analysis of the Transient Reactor Test Facility experiment TS-1C and the Experimental Breeder Reactor II experiment SHRT-17 showed very good agreement between model predictions and measurements.

Tzanos, C.P.

1987-06-01

356

A chemical equilibrium estimate of the aerosols produced in an overheated light water reactor core  

SciTech Connect

The degree of vaporization of light water reactor core materials was estimated using a highly idealized procedure involving (a) specification of the phases that are present for both structural and fuel material, (b) estimation of the vapor pressures exerted by the individual components of each phase, and (c) assuming a degree of vaporization of each phase constituent, allowing equilibration between gaseous and condensed species within the assumed pressure vessel volume. Using this procedure, the aerosol was estimated to consist mainly of silver, indium oxide, cesium hydroxide, and cadmium for pressurized water reactors and cesium hydroxide, cesium iodide, and tellurium for boiling water reactors. If boron is included in the thermodynamic estimate, then boron will significantly alter or dominate the composition of the aerosol in the form of boron oxide and cesium borate. The structural materials make up < 9% of the aerosol at 36 to 57 kg, but this figure is in good agreement with estimates from severe accident sequence analysis studies (17 kg) and from Parker (10.7 kg). The SASCHA data are used in NUREG-0772 and give much higher estimates at 295 and 250 kg.

Wichner, R.P.; Spence, R.D.

1985-09-01

357

International Referecne Ionosphere (IRI) - 2006  

NASA Astrophysics Data System (ADS)

With this presentation the newest version of IRI IRI-2006 will be officially released The new version includes a number of critical improvements and long sought-over additions For the electron density in the topside two new options were added a correction term based on Alouette ISIS topside sounder data and the new NeQuick model In the D-region new models are included for high-latitudes based on rocket and incoherent scatter data The occurrence probability of spread-F is added as a new parameter to IRI although currently only in the form of a regional model for the South-American sector IRI-2006 also includes new models for the description of topside ion composition and equatorial disturbance ion drift In addition the newest version of the International Geomagnetic Reference Field IGRF Version 10 is now implemented in IRI and used for all internal magnetic coordinate computations This presentation will describe and discuss the newest IRI version in great detail and give examples on how they improvements and new additions will benefit specific applications of the IRI model

Bilitza, D.

358

Coarse Iris Classification by Learned Visual Dictionary  

Microsoft Academic Search

In state-of-the-art iris recognition systems, the input iris image has to be compared with a large number of templates in\\u000a database. When the scale of iris database increases, they are much less efficient and accurate. In this paper, we propose\\u000a a novel iris classification method to attack this problem in iris recognition systems. Firstly, we learned a small finite\\u000a dictionary

Xianchao Qiu; Zhenan Sun; Tieniu Tan

2007-01-01

359

Evaluation of the Calculated Results of an Unprotected Transient Undercooling Accident in a Large, Heterogeneous-Core, Liquid-Metal-Cooled Fast Breeder Reactor.  

National Technical Information Service (NTIS)

Disrupted-core (transition-phase) behavior has been evaluated for a hypothetical, unprotected transient undercooling accident in an early version of the heterogeneous-core liquid-metal-cooled fast breeder reactor (LMFBR) developed for the Conceptual Desig...

L. B. Luck G. P. DeVault M. W. Asprey C. R. Bell

1982-01-01

360

Acute endothelial failure after cosmetic iris implants (NewIris)  

PubMed Central

We report a case of an acute endothelial failure after the implantation of a new cosmetic, colored, artificial iris diaphragm implant called NewIris. A 21-year-old woman came to us complaining of progressive loss of vision and pain after NewIris lenses had been implanted. Decreased visual acuity, corneal edema, and increased intraocular pressure in both eyes appeared only 3 weeks after the surgery. The lenses were removed as soon as possible but had already severely affected the endothelial cell count. NewIris implants are an alternative to cosmetic contact lenses, but they are not as safe as other phakic anterior chamber intraocular lenses, nor are they a good option for the patient.

Garcia-Pous, Maria; Udaondo, Patricia; Garcia-Delpech, Salvador; Salom, David; Diaz-Llopis, Manuel

2011-01-01

361

Core subchannel thermal-hydraulic analysis methods and critical heat-flux margin in a light-water breeder reactor  

SciTech Connect

Analyis methods and results are described for critical heat flux (CHF) performance margin in the core of an advanced light water moderated breeder reactor design concept. The 1000 MWe breeder reactor is basically like a large commercial pressurized water reactor (PWR); however, a number of core design features require special consideration with regard to predicting margin to CHF design limits. Three notable features are: (1) fuel rods closely spaced in a triangular pitch lattice; (2) high power seed fuel regions adjacent to low power blanket regions in an open lattice; and (3) power producing thoria shim rods enclosed in individual guide tubes. CHF performance of the breeder core was analyzed with the HOTROD and COBRA computer codes.

Misiewicz, R.; Kast, S.J.; Wunderlich, L.H.

1983-05-01

362

Neutron transport with the method of characteristics for 3-D full core boiling water reactor applications  

NASA Astrophysics Data System (ADS)

The Numerical Nuclear Reactor (NNR) is a code suite that is being developed to provide high-fidelity multi-physics capability for the analysis of light water nuclear reactors. The focus of the work here is to extend the capability of the NNR by incorporation of the neutronics module, DeCART, for Boiling Water Reactor (BWR) applications. The DeCART code has been coupled to the NNR fluid mechanics and heat transfer module STAR-CD for light water reactor applications. The coupling has been accomplished via an interface program, which is responsible for mapping the STAR-CD and DeCART meshes, managing communication, and monitoring convergence. DeCART obtains the solution of the 3-D Boltzmann transport equation by performing a series of 2-D modular ray tracing-based method of characteristics problems that are coupled within the framework of 3-D coarse-mesh finite difference. The relatively complex geometry and increased axial heterogeneity found in BWRs are beyond the modeling capability of the original version of DeCART. In this work, DeCART is extended in three primary areas. First, the geometric capability is generalized by extending the modular ray tracing scheme and permitting an unstructured mesh in the global finite difference kernel. Second, numerical instabilities, which arose as a result of the severe axial heterogeneity found in BWR cores, have been resolved. Third, an advanced nodal method has been implemented to improve the accuracy of the axial flux distribution. In this semi-analytic nodal method, the analytic solution to the transverse-integrated neutron diffusion equation is obtained, where the nonhomogeneous neutron source was first approximated by a quartic polynomial. The successful completion of these three tasks has allowed the application of the coupled DeCART/STAR-CD code to practical BWR problems.

Thomas, Justin W.

363

Modification of the Penn State Reactor to allow transverse and rotational core motion to increase operational versatility  

SciTech Connect

At Penn State the Nuclear Engineering students have the opportunity to perform experiments in reactor physics, work with reactor and radiation instrumentation, and operate a nuclear reactor. These activities are done at the Penn State Breazeale Reactor (PSBR), a General Atomics Mark III TRIGA reactor. Unfortunately this activity alone can not fully support the facility. The PSBR is mandated by Penn State to provide a portion of its operating budget by selling services to users outside as well as inside Penn State. In order to increase the marketability of PSBR an upgrade program was started to increase the quality and versatility of operation. The PSBR is the longest operating university reactor in the United States. The first phase of the upgrade program began in 1992. The quality of operation was increased by replacing a 1965 vintage console with a more reliable digital control and monitoring system. The present phase of the upgrade program is to increase the versatility of operation by modifying the reactor to allow transverse and rotational core motion. Adding two more degrees of motion to the reactor core increases the capability of the facility to meet the needs of present and future users. This upgrade is being financed by a grant from the Department of Energy and matching funds from Penn State. (author)

Hughes, Daniel E. [Penn State University (United States)

1994-07-01

364

Shape adaptive, robust iris feature extraction from noisy iris images.  

PubMed

In the current iris recognition systems, noise removing step is only used to detect noisy parts of the iris region and features extracted from there will be excluded in matching step. Whereas depending on the filter structure used in feature extraction, the noisy parts may influence relevant features. To the best of our knowledge, the effect of noise factors on feature extraction has not been considered in the previous works. This paper investigates the effect of shape adaptive wavelet transform and shape adaptive Gabor-wavelet for feature extraction on the iris recognition performance. In addition, an effective noise-removing approach is proposed in this paper. The contribution is to detect eyelashes and reflections by calculating appropriate thresholds by a procedure called statistical decision making. The eyelids are segmented by parabolic Hough transform in normalized iris image to decrease computational burden through omitting rotation term. The iris is localized by an accurate and fast algorithm based on coarse-to-fine strategy. The principle of mask code generation is to assign the noisy bits in an iris code in order to exclude them in matching step is presented in details. An experimental result shows that by using the shape adaptive Gabor-wavelet technique there is an improvement on the accuracy of recognition rate. PMID:24696801

Ghodrati, Hamed; Dehghani, Mohammad Javad; Danyali, Habibolah

2013-10-01

365

Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012  

SciTech Connect

Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. 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). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009, Cycle 145A through Cycle 151B, was successfully completed during 2012. This major effort supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR Core Safety Analysis Package (CSAP) preparation process, in parallel with the established PDQ-based methodology, beginning late in Fiscal Year 2012. Acquisition of the advanced SERPENT (VTT-Finland) and MC21 (DOE-NR) Monte Carlo stochastic neutronics simulation codes was also initiated during the year and some initial applications of SERPENT to ATRC experiment analysis were demonstrated. These two new codes will offer significant additional capability, including the possibility of full-3D Monte Carlo fuel management support capabilities for the ATR at some point in the future. Finally, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system has been implemented and initial computational results have been obtained. This capability will have many applications as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation.

David W. Nigg, Principal Investigator; Kevin A. Steuhm, Project Manager

2012-09-01

366

End-of-life irradiation performance of core structural components in the Shippingport Light Water Breeder Reactor  

Microsoft Academic Search

Nondestructive and destructive end-of-life examinations of Light Water Breeder Reactor (LWBR) core structural components were performed following operation in the Shippingport Atomic Power Station for 29,047 effective full power hours. The Shippingport LWBR demonstrated that breeding can be achieved in a light water reactor with thorium and uranium-233 oxide fuel pellets contained in Zircaloy-4 tubes. The purpose of this presentation

J. C. Clayton; B. C. Smith

1991-01-01

367

Simultaneous measurement of neutron and gamma-ray radiation levels from a TRIGA reactor core using silicon carbide semiconductor detectors  

Microsoft Academic Search

The ability of a SiC detector to measure neutron and gamma radiation levels in a TRIGA reactor's mixed neutron\\/gamma field was demonstrated. Linear responses to an epicadmium neutron fluence rate (up to 3107 cm-2 s-1) and to a gamma dose rate (0.6-234 krad-Si h-1) were obtained with the detector. Axial profiles of the reactor core's neutron and gamma-ray radiation levels

A. R. Dulloo; F. H. Ruddy; J. G. Seidel; C. Davison; T. Flinchbaugh; T. Daubenspeck

1998-01-01

368

Simultaneous measurement of neutron and gamma-ray radiation levels from a TRIGA reactor core using silicon carbide semiconductor detectors  

Microsoft Academic Search

The ability of a silicon carbide radiation detector to measure neutron and gamma radiation levels in a TRIGA reactor's mixed neutron\\/gamma field was demonstrated. Linear responses to epicadmium neutron fluence rate (up to 3107 cm-2 s-1) and to gamma dose rate (0.6-234 krad-Si h-1) were obtained with the detector. Axial profiles of the reactor core's neutron and gamma-ray radiation levels

A. R. Dulloo; F. H. Ruddy; J. G. Seidel; C. Davison; T. Flinchbaugh; T. Daubenspeck

1999-01-01

369

Reactor moderator, pressure vessel, and heat rejection system of an open-cycle gas core nuclear rocket concept  

NASA Technical Reports Server (NTRS)

A preliminary design study of a conceptual 6000-megawatt open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 196,600 newtons (44,200 lb) and a specific impulse of 4400 seconds. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel, and waste heat rejection system) were considered conceptually and were sized.

Taylor, M. F.; Whitmarsh, C. L., Jr.; Sirocky, P. J., Jr.; Iwanczyke, L. C.

1973-01-01

370

Microstructure and nano-hardness analyses of stress corrosion cracking, utilizing 316L core shroud of BWR power reactors  

Microsoft Academic Search

The water cooled shield blanket made of Type 316L SS for the international thermonuclear experimental reactor (ITER) has potential issues related to stress corrosion cracking (SCC). Shroud mock-ups and boat samples taken from the core shroud of the boiling water reactor (BWR) with SCC were investigated from the viewpoint of microstructures and nano-hardness. Fine grains and deformation bands were observed

Y. Sueishi; A. Kohyama; H. Kinoshita; M. Narui; K. Fukumoto

2006-01-01

371

Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core  

Microsoft Academic Search

The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool type Research Reactor [IAEA-TECDOC-233, 1980. IAEA Research Reactor Core Conversion from the use of high-enriched uranium to the use of low enriched uranium fuels Guidebook] under flow blockage of a single Fuel Assembly (FA) conditions. Such event was rarely investigated

Martina Adorni; Anis Bousbia-Salah; Tewfik Hamidouche; Beniamino Di Maro; Franco Pierro; Francesco DAuria

2005-01-01

372

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY2011 Activities  

Microsoft Academic Search

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the

Michael A. Pope

2011-01-01

373

SAS4A: A computer model for the analysis of hypothetical core disruptive accidents in liquid metal reactors  

SciTech Connect

To ensure that the public health and safety are protected under any accident conditions in a Liquid Metal Fast Breeder Reactor (LMFBR), many accidents are analyzed for their potential consequences. The SAS4A code system, described in this paper, provides such an analysis capability, including the ability to analyze low probability events such as the Hypothetical Core Disruptive Accidents (HCDAs). The SAS4A code system has been designed to simulate all the events that occur in a LMFBR core during the initiating phase of a Hypothetical Core Disruptive Accident. During such postulated accident scenarios as the Loss-of-Flow and Transient Overpower events, a large number of interrelated physical phenomena occur during a relatively short time. These phenomena include transient heat transfer and hydrodynamic events, coolant boiling and fuel and cladding melting and relocation. During to the strong neutronic feedback present in a nuclear reactor, these events can significantly influence the reactor power. The SAS4A code system is used in the safety analysis of nuclear reactors, in order to estimate the energetic potential of very low probability accidents. The results of SAS4A simulations are also used by reactor designers in order to build safer reactors and eliminate the possibility of any accident which could endanger the public safety.

Tentner, A.M.; Birgersson, G.; Cahalan, J.E.; Dunn, F.E.; Kalimullah; Miles, K.J.

1987-01-01

374

PERFLUOROOCTANE SULFONATE (PFOS) - IRIS ASSESSMENT  

EPA Science Inventory

An assessment of perfluorooctane sulfonate (PFOS) is underway that will establish the RfD/RfC as appropriate which will be made available to the public through the Integrated Risk Information System (IRIS)....

375

International Referecne Ionosphere (IRI) - 2006  

Microsoft Academic Search

With this presentation the newest version of IRI IRI-2006 will be officially released The new version includes a number of critical improvements and long sought-over additions For the electron density in the topside two new options were added a correction term based on Alouette ISIS topside sounder data and the new NeQuick model In the D-region new models are included

D. Bilitza

2006-01-01

376

Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model  

NASA Technical Reports Server (NTRS)

Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.

Kazeminezhad, F.; Anghaie, S.

2008-01-01

377

Optical phenomena in pure-silica-core fiber under pulsed reactor irradiation  

NASA Astrophysics Data System (ADS)

The light emission intensity at the wavelength of 400-750 nm in the KU-1 silica core (OH content 1000 ppm) fiber waveguide under irradiation at BARS-6 pulsed fission reactor (pulse duration 80 microsecond(s) , dose per pulse <5.5x1012 n/cm2 (9 Gy), dose rate <7x1016 n/cm2s (1.1x105 Gy/s) have been measured. The intensity of radiation-induced light emission has been found to depend on intensity of probing light. Lower intensity of the light emission has been observed for higher intensity of probing light (lasers, wavelength 532 and 632 nm). The light emission quenching occurs at the wavelengths shorter and longer than the wavelength of the probing light, and also at the equal wavelengths.

Demenkov, Pavel V.; Plaksin, Oleg A.; Stepanov, Vladimir A.; Stepanov, Peter A.; Zabezhailov, Maxim O.; Golant, Konstantin M.; Tomashuk, Alexander L.

2002-01-01

378

Flowing gas, non-nuclear experiments on the gas core reactor  

NASA Technical Reports Server (NTRS)

Variations in cavity wall and injection configurations of the gas core reactor were aimed at establishing flow patterns that give a maximum of the nuclear criticality eigenvalue. Correlation with the nuclear effect was made using multigroup diffusion theory normalized by previous benchmark critical experiments. Air was used to simulate the hydrogen propellant in the flow tests, and smoked air, argon, or Freon to simulate the central nuclear fuel gas. Tests were run both in the down-firing and upfiring directions. Results showed that acceptable flow patterns with volume fraction for the simulated nuclear fuel gas and high flow rate ratios of propellant to fuel can be obtained. Using a point injector for the fuel, good flow patterns are obtained by directing the outer gas at high velocity long the cavity wall, using louvered injection schemes. Recirculation patterns were needed to stabilize the heavy central gas when different gases are used.

Kunze, J. F.; Cooper, C. G.; Macbeth, P. J.

1973-01-01

379

Evaluation of instrumentation for detection of inadequate core cooling in boiling water reactors  

SciTech Connect

This report is a review of the Approach to Inadequate Core Cooling issue in Boiling Water Reactors (BWR). The report consists of seven sections. The principal conclusion is that the condition of the reference leg, and operator awareness of that condition are of primary importance in level indication reliability for safety. An indication of reference leg level and temperature displayed to the operators would be a useful enhancement of reliability and a guide to further operator action in all circumstances. We conclude that the BWR practice of multiple, redundant coolant level measurements, with overlapping ranges, can be a reliable basis for indication of approach to an ICC condition, and, in correlation with the other control and safety systems of modern BWRs, will prevent unsafe conditions.

Lewin, J.

1984-04-01

380

Isotope production target irradiation experience at the annular core research reactor  

SciTech Connect

As a result of an Environmental Impact Statement (EIS) recently issued by the Department of Energy, Sandia National Laboratories (SNL) has been selected as the {open_quotes}most appropriate facility{close_quotes} for the production of {sup 99}Mo. The daughter product of {sup 99}Mo is {sup 99m}Tc, a radioisotope used in 36,000 medical procedures per day in the U.S.{close_quote} At SNL, the {sup 99}Mo would be created by the fission process in UO{sub 2} coated {open_quotes}targets{close_quotes} and chemically separated in the SNL Hot Cell Facility (HCF). SNL has recently completed the irradiation of five production targets at its Annular Core Research Reactor (ACRR). Following irradiation, four of the targets were chemically processed in the HCF using the Cintichem process.

Talley, D.G.

1997-02-01

381

Non-Invasive Imaging of Reactor Cores Using Cosmic Ray Muons  

NASA Astrophysics Data System (ADS)

Cosmic ray muons penetrate deeply in material, with some passing completely through very thick objects. This penetrating quality is the basis of two distinct, but related imaging techniques. The first measures the number of cosmic ray muons transmitted through parts of an object. Relatively fewer muons are absorbed along paths in which they encounter less material, compared to higher density paths, so the relative density of material is measured. This technique is called muon transmission imaging, and has been used to infer the density and structure of a variety of large masses, including mine overburden, volcanoes, pyramids, and buildings. In a second, more recently developed technique, the angular deflection of muons is measured by trajectory-tracking detectors placed on two opposing sides of an object. Muons are deflected more strongly by heavy nuclei, since multiple Coulomb scattering angle is approximately proportional to the nuclear charge. Therefore, a map showing regions of large deflection will identify the location of uranium in contrast to lighter nuclei. This technique is termed muon scattering tomography (MST) and has been developed to screen shipping containers for the presence of concealed nuclear material. Both techniques are a good way of non-invasively inspecting objects. A previously unexplored topic was applying MST to imaging large objects. Here we demonstrate extending the MST technique to the task of identifying relatively thick objects inside very thick shielding. We measured cosmic ray muons passing through a physical arrangement of material similar to a nuclear reactor, with thick concrete shielding and a heavy metal core. Newly developed algorithms were used to reconstruct an image of the ``mock reactor core,'' with resolution of approximately 30 cm.

Milner, Edward

2011-10-01

382

Operating experience of natural circulation core cooling in boiling water reactors  

SciTech Connect

General Electric (GE) has proposed an advanced boiling water reactor, the Simplified Boiling Water Reactor (SBWR), which will utilize passive, gravity-driven safety systems for emergency core coolant injection. The SBWR design includes no recirculation loops or recirculation pumps. Therefore the SBWR will operate in a natural circulation (NC) mode at full power conditions. This design poses some concerns relative to stability during startup, shutdown, and at power conditions. As a consequence, the NRC has directed personnel at several national labs to help investigate SBWR stability issues. This paper will focus on some of the preliminary findings made at the INEL. Because of the broad range of stability issues this paper will mainly focus on potential geysering instabilities during startup. The two NC designs examined in detail are the US Humboldt Bay Unit 3 BWR-1 plant and Dodewaard plant in the Netherlands. The objective of this paper will be to review operating experience of these two plants and evaluate their relevance to planned SBWR operational procedures. For completeness, experimental work with early natural circulation GE test facilities will also be briefly discussed.

Kullberg, C.; Jones, K.; Heath, C.

1993-08-01

383

Core cooling under accident conditions at the high-flux beam reactor  

SciTech Connect

The High-Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is cooled and moderated by heavy water and contains {sup 235}U in the form of narrow-channel, parallel-plate-type fuel elements. During normal operation, the flow direction is downward through the core. This flow direction is maintained at a reduced flow rate during routine shutdown and on loss of commercial power by means of redundant pumps and power supplies. However, in certain accident scenarios, e.g. loss-of-coolant accidents (LOCAs), all forced-flow cooling is lost. Although there was experimental evidence during the reactor design period (1958-1963) that the heat removal capacity in the fully developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. Accordingly, a test program was initiated using an electrically heated section to simulate the fuel channel and a cooling loop to simulate the balance of the primary cooling system.

Tichler, P.; Cheng, L. (Brookhaven National Lab., Upton, NY (United States)); Fauske, H. (Fauske and Associates, In., Burr Ridge, IL (United States))

1991-01-01

384

The design and installation of a core discharge monitor for CANDU-type reactors  

SciTech Connect

A new type of surveillance systems that monitors neutron and gamma radiation in a reactor containment is being installed at the Ontario Hydro Darlington Nuclear Generating Station A, Unit 2. Unlike video or film surveillance that monitors mechanical motion, this system measures fuel-specific radiation emanating from irradiated fuel as it is pushed from the core of CANDU-type reactors. Proof-of-principle measurements have been carried out at Bruce Nuclear Generating Station A, Unit 3. The system uses ({gamma},n) threshold detectors and ionization detectors. A microprocessor-based electronics package, GRAND-II (Gamma Ray and Neutron Detector electronics package), provides detector bias, preamplifier power, and signal processing. Firmware in the GRAND-2 controls the surveillance activities, including data acquisition and a level of detector authentication, and it handles authenticated communication with a central data logging computer. Data from the GRAND-II are transferred to an MS-DOS-compatible computer and stored. These data are collected and reviewed for fuel-specific radiation signatures from the primary detector and proper ratios of signals from secondary detectors. 5 figs.

Halbig, J.K. (Los Alamos National Lab., NM (USA)); Monticone, A.C.; Ksiezak, L. (International Atomic Energy Agency, Vienna (Austria)); Smiltnieks, V. (International Atomic Energy Agency, Toronto, ON (Canada). Regional Office)

1990-01-01

385

ACRR (Annular Core Research Reactor) fission product release tests: ST-1 and ST-2  

SciTech Connect

Two experiments (ST-1 and ST-2) have been performed in the Annular Core Research Reactor (ACER) at Sandia National Laboratories (SNLA) to obtain time-resolved data on the release of fission products from irradiated fuels under light water reactor (LWR) severe accident conditions. Both experiments were conducted in a highly reducing environment at maximum fuel temperatures of greater than 2400 K. These experiments were designed specifically to investigate the effect of increased total pressure on fission product release; ST-1 was performed at approximately 0.16 MPa and ST-2 was run at 1.9 MPa, whereas other parameters were matched as closely as possible. Release rate data were measured for Cs, I, Ba, Sr, Eu, Te, and U. The release rates were higher than predicted by existing codes for Ba, Sr, Eu, and U. Te release was very low, but Te did not appear to be sequestered by the zircaloy cladding; it was evenly distributed in the fuel. In addition, in posttest analysis a unique fuel morphology (fuel swelling) was observed which may have enhanced fission product release, especially in the high pressure test (ST-2). These data are compared with analytical results from the CORSOR correlation and the VICTORIA computer model. 8 refs., 8 figs., 2 tabs.

Allen, M.D.; Stockman, H.W.; Reil, K.O.; Grimley, A.J.; Camp, W.J.

1988-01-01

386

Generation IV nuclear energy system initiative. Pin core subassembly designfor the Gas-Cooled Fast Reactor.  

SciTech Connect

The Gas-Cooled Fast Reactor (GFR) is one of six systems selected for viability assessment in the Generation IV program. It features a closed nuclear fuel cycle, consisting of a high-temperature helium-cooled fast spectrum reactor, coupled to a direct-cycle helium turbine for electricity production. The GFR combines the advances of fast spectrum systems with those of high-temperature systems. It was clear from the very beginning that GFR design should be driven by the objective to offer a complementary approach to liquid metal cooling. On this basis, CEA and the US DOE decided to collaborate on the pre-conceptual design of a GFR. This reactor design will provide a high level of safety and full recycling of the actinides, and will also be highly proliferation resistant and economically attractive. The status of this collaborative project is that two unit sizes, 600 MWt and 2400 MWt were selected as the focus of the design and safety studies. Researchers studied fuel forms, fuel assembly/element designs, core configurations, primary and balance-of-plant layouts, and safety approaches for both of these unit sizes. Results regarding the feasibility of this GFR design are encouraging. For example, sustainability and non-proliferation goals can be met and the proposed concept has attractive safety features. These features take advantage of the helium in terms of its neutronic quasi-transparency as well as the enhanced Doppler effect in connection with candidate fuel and structural materials. The current design trend is to consider high unit power for the GFR (2400 MWt), an attractive level for the power density (100 MW/m{sup 3}), and the implementation of an innovative plate type fuel or pin type sub-assembly with carbide-based actinide compounds and SiC-based structural materials. Work is still needed to refine the safety approach, to select the main system options, and to more definitively establish economic parameters.

Farmer, M. T.; Hoffman, E. A.; Pfeiffer, P. F.; Therios, I. U.

2006-07-31

387

Surgical technique for congenital iris coloboma repair.  

PubMed

We describe a surgical technique for managing congenital iris coloboma. After phacoemulsification with placement of an intraocular lens in the capsular bag, coloboma repair is begun by bisecting the iris sphincter on both sides of its attachment near the chamber angle. The iris leaflets central to the sphincterectomies are approximated using a 10-0 polypropylene suture (Prolene, Ethicon, Inc.) and a modified Siepser pupilloplasty technique. The remaining peripheral iris defect is closed in a similar fashion. In patients with congenital iris coloboma, phacoemulsification with in-the-bag IOL implantation followed by this pupilloplasty technique was effective in providing functional and cosmetic repair of a congenital iris coloboma. PMID:17081895

Cionni, Robert J; Karatza, Ekaterini C; Osher, Robert H; Shah, Manan

2006-11-01

388

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

389

Thermal-hydraulic modeling of the primary coolant system of light water reactors during severely degraded core accidents  

Microsoft Academic Search

A computer code has been developed to predict the time history of the primary coolant system (PCS) thermalhydraulics during severely degraded core accidents. The developed PCS code is modular and simulates both pressurized and boiling water reactor primary systems. The developed model assumes a one dimensional flow of steam and hydrogen along the PCS. The heat transfer between the gas

A. T. Wassel; M. S. Hoseyni; J. L. Jr. Farr; S. M. Ghiaasiaan

1984-01-01

390

Stress Relaxation and Creep of High-Temperature Gas-Cooled Reactor Core Support Ceramic Materials: A Literature Search.  

National Technical Information Service (NTIS)

Creep and stress relaxation in structural ceramics are important properties to the high-temperature design and safety analysis of the core support structure of the HTGR. The ability of the support structure to function for the lifetime of the reactor is d...

J. E. Selle V. J. Tennery

1980-01-01

391

Analytical study on the effects of rolling on the flow in the core of the marine reactor.  

National Technical Information Service (NTIS)

The effect of rolling on the flow in reactor core have been investigated theoretically. The flow was analyzed utilizing the subchannel analysis code THERMIT-2 developed by MIT modified by us to compute the acceleration fields due to rolling. The analysis ...

T. Kusunoki Y. Itoh O. Suzuki S. Hosoda

1990-01-01

392

Conceptual design of an integrated information system for safety related analysis of nuclear power plants (IRIS Phase 1).  

National Technical Information Service (NTIS)

This report deals with a conceptual design of an integrated information management system, called PSI-IRIS, as needed to assist the analysts for reactor safety related investigations on Swiss nuclear power plants within the project STARS. Performing compl...

K. Hofer P. Zehnder A. Galperin

1994-01-01

393

GPU Based General-Purpose Parallel computing to Solve Nuclear Reactor In-Core fuel Management Design and Operation Problem  

SciTech Connect

In-core fuel management study is a crucial activity in nuclear power plant design and operation. Its common problem is to find an optimum arrangement of fuel assemblies inside the reactor core. Main objective for this activity is to reduce the cost of generating electricity, which can be done by altering several physical properties of the nuclear reactor without violating any of the constraints imposed by operational and safety considerations. This research try to address the problem of nuclear fuel arrangement problem, which is, leads to the multi-objective optimization problem. However, the calculation of the reactor core physical properties itself is a heavy computation, which became obstacle in solving the optimization problem by using genetic algorithm optimization.This research tends to address that problem by using the emerging General Purpose Computation on Graphics Processing Units (GPGPU) techniques implemented by C language for CUDA (Compute Unified Device Architecture) parallel programming. By using this parallel programming technique, we develop parallelized nuclear reactor fitness calculation, which is involving numerical finite difference computation. This paper describes current prototype of the parallel algorithm code we have developed on CUDA, that performs one hundreds finite difference calculation for nuclear reactor fitness evaluation in parallel by using GPU G9 Hardware Series developed by NVIDIA.

Prayudhatama, D.; Waris, A.; Kurniasih, N.; Kurniadi, R. [Bosscha Laboratory, Department of Physics, Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

2010-06-22

394

A Computational Fluid Dynamic and Heat Transfer Model for Gaseous Core and Gas Cooled Space Power and Propulsion Reactors  

NASA Technical Reports Server (NTRS)

A computational model based on the axisymmetric, thin-layer Navier-Stokes equations is developed to predict the convective, radiation and conductive heat transfer in high temperature space nuclear reactors. An implicit-explicit, finite volume, MacCormack method in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve the thermal and fluid governing equations. Simulation of coolant and propellant flows in these reactors involves the subsonic and supersonic flows of hydrogen, helium and uranium tetrafluoride under variable boundary conditions. An enthalpy-rebalancing scheme is developed and implemented to enhance and accelerate the rate of convergence when a wall heat flux boundary condition is used. The model also incorporated the Baldwin and Lomax two-layer algebraic turbulence scheme for the calculation of the turbulent kinetic energy and eddy diffusivity of energy. The Rosseland diffusion approximation is used to simulate the radiative energy transfer in the optically thick environment of gas core reactors. The computational model is benchmarked with experimental data on flow separation angle and drag force acting on a suspended sphere in a cylindrical tube. The heat transfer is validated by comparing the computed results with the standard heat transfer correlations predictions. The model is used to simulate flow and heat transfer under a variety of design conditions. The effect of internal heat generation on the heat transfer in the gas core reactors is examined for a variety of power densities, 100 W/cc, 500 W/cc and 1000 W/cc. The maximum temperature, corresponding with the heat generation rates, are 2150 K, 2750 K and 3550 K, respectively. This analysis shows that the maximum temperature is strongly dependent on the value of heat generation rate. It also indicates that a heat generation rate higher than 1000 W/cc is necessary to maintain the gas temperature at about 3500 K, which is typical design temperature required to achieve high efficiency in the gas core reactors. The model is also used to predict the convective and radiation heat fluxes for the gas core reactors. The maximum value of heat flux occurs at the exit of the reactor core. Radiation heat flux increases with higher wall temperature. This behavior is due to the fact that the radiative heat flux is strongly dependent on wall temperature. This study also found that at temperature close to 3500 K the radiative heat flux is comparable with the convective heat flux in a uranium fluoride failed gas core reactor.

Anghaie, S.; Chen, G.

1996-01-01

395

An efficient color and texture based iris image retrieval technique  

Microsoft Academic Search

This paper proposes a hierarchical approach to retrieve an iris image which is based on a new indexing scheme for a large iris database. The hierarchical approach is a combination of both iris color and texture. Iris color is used for indexing and texture is used for retrieval of iris images from the indexed iris database. An index which is

Umarani Jayaraman; Surya Prakash; Phalguni Gupta

396

Iris Localization Scheme Based on Morphology and Gaussian Filtering  

Microsoft Academic Search

One of the basic techniques in iris recognition system is iris localization. To find a fast, effective and exact iris localization algorithm is the key step of iris recognition. After analyzing the principle, strong points and short points of some common used iris localization methods, a morphological theory based iris localization algorithm was proposed in this paper. Based on the

Feng Gui; Lin Qiwei

2007-01-01

397

Estimation of Specific Mass for Multimegawatt NEP Systems Based on Vapor Core Reactors with MHD Power Conversion  

NASA Astrophysics Data System (ADS)

Very low specific-mass power generation in space is possible using Vapor Core Reactors with Magnetohydrodynamic (VCR/MHD) generator. These advanced reactors at the conceptual design level have potential for the generation of tens to hundreds of megawatts of power in space with specific mass of about 1 kg/kWe. Power for nuclear electric propulsion (NEP) is possible with almost direct power conditioning and coupling of the VCR/MHD power output to the VASIMR engine, MPD, and a whole host of electric thrusters. The VCR/MHD based NEP system is designed to power space transportation systems that dramatically reduce the mission time for human exploration of the entire solar system or for aggressive long-term robotic missions. There are more than 40 years of experience in the evaluation of the scientific and technical feasibility of gas and vapor core reactor concepts. The proposed VCR is based on the concept of a cavity reactor made critical through the use of a reflector such as beryllium or beryllium oxide. Vapor fueled cavity reactors that are considered for NEP applications operate at maximum core center and wall temperatures of 4000 K and 1500K, respectively. A recent investigation has resulted in the conceptual design of a uranium tetrafluoride fueled vapor core reactor coupled to a MHD generator. Detailed neutronic design and cycle analyses have been performed to establish the operating design parameters for 10 to 200 MWe NEP systems. An integral system engineering-simulation code is developed to perform parametric analysis and design optimization studies for the VCR/MHD power system. Total system weight and size calculated based on existing technology has proven the feasibility of achieving exceptionally low specific mass (? ~1 kg/kWe) with a VCR/MHD powered system.

Knight, Travis; Anghaie, Samim

2004-02-01

398

Comparing Haar-Hilbert and Log-Gabor Based Iris Encoders on Bath Iris Image Database  

Microsoft Academic Search

This papers introduces a new family of iris encoders which use 2-dimensional Haar Wavelet Transform for noise attenuation, and Hilbert Transform to encode the iris texture. In order to prove the usefulness of the newly proposed iris encoding approach, the recognition results obtained by using these new encoders are compared to those obtained using the classical Log- Gabor iris encoder.

Nicolaie Popescu-Bodorin; Valentina E. Balas

2011-01-01

399

Comparing Haar-Hilbert and Log-Gabor based iris encoders on Bath Iris Image Database  

Microsoft Academic Search

This papers introduces a new family of iris encoders which use 2-dimensional Haar Wavelet Transform for noise attenuation, and Hilbert Transform to encode the iris texture. In order to prove the usefulness of the newly proposed iris encoding approach, the recognition results obtained by using these new encoders are compared to those obtained using the classical Log-Gabor iris encoder. Twelve

N. Popescu-Bodorin; V. E. Balas

2010-01-01

400

Reliability comparison of computer based core temperature monitoring system with two and three thermocouples per sub-assembly for Fast Breeder Reactors  

Microsoft Academic Search

Prototype Fast Breeder Reactor (PFBR) is a mixed oxide fuelled, sodium cooled, 500 MWe, pool type fast breeder reactor under construction at Kalpakkam, India. The reactor core consists of fuel pins assembled in a number of hexagonal shaped, vertically stacked SubAssemblies (SA). Sodium flows from the bottom of the SAs, takes heat from the fission reaction, comes out through the

R. Dheenadhayalan; M. Sakthivel; A. J. Arul; K. Madhusoodanan; P. Mohanakrishnan

2010-01-01

401

United States Department of Energy`s reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1  

SciTech Connect

This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage.

NONE

1997-06-01

402

IRIS Toxicological Review of Hexachloroethane (2011 Final)  

EPA Science Inventory

EPA is announcing the release of the final report Toxicological Review of Hexachloroethane: in support of the Integrated Risk Information System (IRIS). The updated Summary for Acrylamide and accompanying Quickview have also been added to the IRIS database. ...

403

'Fragile Bits' vs. Multi-Enrollment - A Case Study of Iris Recognition on Bath University Iris Database  

Microsoft Academic Search

This paper explores the use of fragile bits in the context of a re- cently proposed iris recognition methodology based on Circular Fuzzy Iris Segmentation and Gabor Analytic Iris Texture Binary Encoder. Iris images from Bath University Iris Database are encoded as iris codes at three difierent lengths (192, 512, 768 Bytes) and used to test the concept of fragile

Nicolaie Popescu-Bodorin

404

Effects of mascara on iris recognition  

NASA Astrophysics Data System (ADS)

Iris biometrics systems rely on analysis of a visual presentation of the human iris, which must be extracted from the periocular region. Topical cosmetics can greatly alter the appearance of the periocular region, and can occlude portions of the iris texture. In this paper, the presence of topical cosmetics is shown to negatively impact the authentic distribution of iris match scores, causing an increase in the false non-match rate at a fixed false match rate.

Doyle, James S.; Flynn, Patrick J.; Bowyer, Kevin W.

2013-05-01

405

MHTGR [modular high-temperature gas-cooled reactor] core physics validation plan  

SciTech Connect

This document contains the verification and validation (V&V) plan for analytical methods utilized in the nuclear design for normal and off-normal conditions within the Modular High-Temperature Gas-Cooled Reactor (MHTGR). Regulations, regulatory guides, and industry standards have been reviewed and the approach for V&V has been developed. MHTGR core physics methods are described and the status of previous V&V is summarized within this document. Additional work required to verify and validate these methods is identified. The additional validation work includes comparison of calculations with available experimental data, benchmark comparison of calculations with available experimental data, benchmark comparisons with other validated codes, results from a cooperative program now underway at the Arbeitsgemeinschaft Versuchs-Reaktor GmbH (AVR) facility in Germany, results from a planned series of experiments on the Compact Nuclear Power Source (CNPS) facility at Los Alamos, and detailed documentation of all V&V studies. In addition, information will be obtained from planned international cooperative agreements to provide supplemental data for V&V. The regulatory technology development plan will be revised to include these additional experiments. A work schedule and cost estimate for completing this plan is also provided. This work schedule indicates the timeframe in which major milestones must be performed in order to complete V&V tasks prior to the issuance of preliminary design approval from the NRC. The cost to complete V&V tasks for core physics computational methods is estimated to be $2.2M. 41 refs., 13 figs., 8 tabs.

Baxter, A.; Hackney, R.

1988-01-01

406

Iris Recognition System Using Combined Colour Statistics  

Microsoft Academic Search

This paper proposes a high performance iris recognition system based on the probability distribution functions (PDF) of pixels in different colour channels. The PDFs of the segmented iris images are used as statistical feature vectors for the recognition of irises by minimizing the Kullback-Leibler distance (KLD) between the PDF of a given iris and the PDFs of irises in the

Hasan Demirel; G. Anbarjafari

2008-01-01

407

Person authentication technique using human iris recognition  

Microsoft Academic Search

Research topic covered was the identification of a characteristic method of authentication based on biometric iris reading to achieve a solution to secure communications. Biometric identification solution based on iris reading was combined with conventional authentication methods to achieve more secure communications and computers better protected. The paper presents three iris classification techniques: Euclidean classifier, MLP classifier and the Hybrid

DAVID MARIUS DANIEL; BORDA MONICA

2010-01-01

408

Gabor Analytic Iris Texture Binary Encoder  

Microsoft Academic Search

The present paper proposes a new method for generating binary iris codes using Hilbert Transform. The strong analytic signal as- sociated with the chromatic iris sequence is used to recover the phase information which it contains. A meaningful synthetic example is given in order to illustrate the iris binary code extraction. The reasons for canceling the discovery of the radial

Nicolaie Popescu-Bodorin

409

Iris segmentation using geodesic active contours  

Microsoft Academic Search

The richness and apparent stability of the iris texture make it a robust biometric trait for personal authentication. The performance of an automated iris recognition system is affected by the accuracy of the segmentation process used to localize the iris structure. Most segmentation models in the literature assume that the pupillary, limbic, and eyelid boundaries are circular or elliptical in

Samir Shah; Arun Ross

2009-01-01

410

Iris recognition using rapid Haar wavelet decomposition  

Microsoft Academic Search

In this paper, we propose an iris recognition system using a basic and fast Haar wavelet decomposition method to analyze the pattern of a human iris. This system has two main modules, which are the feature encoding and iris code matching modules. Among all feature extraction methods, Haar wavelet decomposition is chosen for its computational simplicity and speed in filtering

Tze Weng Ng; Thien Lang Tay; Siak Wang Khor

2010-01-01

411

Iris Recognition Using Circular Symmetric Filters  

Microsoft Academic Search

This paper proposes a new method for personal identification based on iris recognition. The method consists of three major components: image preprocessing, feature extraction and classifier design. A bank of circular symmetric filters is used to capture local iris characteristics to form a fixed length feature vector. In iris matching, an efficient approach called nearest feature line (NFL) is used.

Li Ma; Yunhong Wang; Tieniu Tan

2002-01-01

412

SLIC: Short-length iris codes  

Microsoft Academic Search

The texture in a human iris has been shown to have good individual distinctiveness and thus is suitable for use in reliable identification. A conventional iris recognition system unwraps the iris image and generates a binary feature vector by quantizing the response of selected filters applied to the rows of this image. Typically there are 360 angular sectors, 64 radial

James E. Gentile; Nalini Ratha; Jonathan Connell

2009-01-01

413

Biometric Personal Identification Based on Iris Patterns  

Microsoft Academic Search

A new system for personal identification based on iris patterns is presented in this paper. It is composed of iris image acquisition, image preprocessing, feature extraction and classifier design. The algorithm for iris feature extraction is based on texture analysis using multi-channel Gabor filtering and wavelet transform. Compared with existing methods, our method employs the rich 2-D information of the

Yong Zhu; Tieniu Tan; Yunhong Wang

2000-01-01

414

Image understanding for iris biometrics: A survey  

Microsoft Academic Search

This survey covers the historical development and current state of the art in image understanding for iris biometrics. Most research publications can be categorized as making their primary contribution to one of the four major modules in iris biometrics: image acquisition, iris segmentation, texture analysis and matching of texture representations. Other important research includes experimental evaluations, image databases, applications and

Kevin W. Bowyer; Karen Hollingsworth; Patrick J. Flynn

2008-01-01

415

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

SciTech Connect

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

Stratton, W.R. (GPU Nuclear Corp., Middletown, PA (USA))

1987-04-15

416

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

SciTech Connect

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

Petrov, Yu. V.; Nazarov, A. I.; Onegin, M. S.; Petrov, V. Yu.; Sakhnovsky, E. G. [St. Petersburg Nuclear Physics Institute, Gatchina, RU-188-300, St. Petersburg (Russian Federation)

2006-12-15

417

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

National Technical Information Service (NTIS)

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 fue...

A. M. Ougouag B. Boer

2010-01-01

418

A Fully Implicit, Second Order in Time, Simulation of a Nuclear Reactor Core  

SciTech Connect

This paper will present a high fidelity solution algorithm for a model of a nuclear reactor core barrel. This model consists of a system of nine nonlinearly coupled partial differential equations. The coolant is modeled with the 1-D six-equation two-phase flow model of RELAP5. Nonlinear heat conduction is modeled with a single 2-D equation. The fission power comes from two 2-D equations for neutron diffusion and precursor concentration. The solution algorithm presented will be the physics-based preconditioned Jacobian-free Newton-Krylov (JFNK) method. In this approach all nine equations are discretized and then solved in a single nonlinear system. Newton's method is used to iterate the nonlinear system to convergence. The Krylov linear solution method is used to solve the matrices in the linear steps of the Newton iterations. The physics-based pre-conditioner provides an approximation to the solution of the linear system that accelerates the Krylov iterations. Results will be presented for two algorithms. The first algorithm will be the traditional approach used by RELAP5. Here the two-phase flow equations are solved separately from the nonlinear conduction and neutron diffusion. Because of this splitting of the physics, and the linearizations employed this method is first order accurate in time. A second algorithm will be the JFNK method solved second order in time accurate. Results will be presented which compare these two algorithms in terms of accuracy and efficiency. (author)

Mousseau, Vincent A. [Los Alamos National Laboratory, P.O. Box 1663 Los Alamos, NM 87545 (United States)

2006-07-01

419

Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building  

SciTech Connect

This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

Lata

1996-09-26

420

Extending Iris: The VAO SED Analysis Tool  

NASA Astrophysics Data System (ADS)

Iris is a tool developed by the Virtual Astronomical Observatory (VAO) for building and analyzing Spectral Energy Distributions (SEDs). Iris was designed to be extensible, so that new components and models can be developed by third parties and then included at runtime. Iris can be extended in different ways: new file readers allow users to integrate data in custom formats into Iris SEDs; new models can be fitted to the data, in the form of template libraries for template fitting, data tables, and arbitrary Python functions. The interoperability-centered design of Iris and the Virtual Observatory standards and protocols can enable new science functionalities involving SED data.

Laurino, O.; Busko, I.; Cresitello-Dittmar, M.; D'Abrusco, R.; Doe, S.; Evans, J.; Pevunova, O.

2013-10-01

421

A new approach for cancelable iris recognition  

NASA Astrophysics Data System (ADS)

The iris is a stable and reliable biometric for positive human identification. However, the traditional iris recognition scheme raises several privacy concerns. One's iris pattern is permanently bound with him and cannot be changed. Hence, once it is stolen, this biometric is lost forever as well as all the applications where this biometric is used. Thus, new methods are desirable to secure the original pattern and ensure its revocability and alternatives when compromised. In this paper, we propose a novel scheme which incorporates iris features, non-invertible transformation and data encryption to achieve "cancelability" and at the same time increases iris recognition accuracy.

Yang, Kai; Sui, Yan; Zhou, Zhi; Du, Yingzi; Zou, Xukai

2010-04-01

422

Reactor flooding system for a retaining molten core materials in a reactor vessel by the improved external vessel cooling capability  

US Patent & Trademark Office Database

A reactor cavity flooding system, which is used to immerse the hemispherical lower head of a nuclear reactor vessel by flooding the reactor cavity, is connected to both coolant injection nozzles located at the annulus gap between the lower head and the thermal insulator of a reactor and the discharge loops which are used to drain the hot water of the annulus gap into either the cavity floor or a liquid eductor. The subcooled water at a fire protection system can be directly injected into the annulus gap through twenty-five (25) nozzles at the lowest, middle, and top injection headers by a pump. The hot water heated at the lower head will be drained into either the cavity floor and/or the liquid eductor via two discharge loops that consist of both a suction header in the annulus gap at the equator level of the lower head and four (4) leakage collectors at the outside of four (4) shear keys of a reactor vessel. Drainage and recirculation of the hot water can be achieved in two ways. The first way uses the pump for injecting the subcooled water and for recirculating the drained water in a reactor cavity. The second way uses a liquid eductor for draining the hot water, instead of discharging it into the reactor cavity floor, and a pump for recirculating the drained water blended with subcooled water through the liquid eductor.

1998-10-20

423

Automated personal identification system based on human iris analysis  

Microsoft Academic Search

In general, a typical iris recognition system includes iris imaging, iris liveness detection, iris image quality assessment,\\u000a and iris recognition. This paper presents an algorithm focusing on the last two steps. The novelty of this algorithm includes\\u000a improving the speed and accuracy of the iris segmentation process, assessing the iris image quality such that only the clear\\u000a images are accepted

R. T. Al-zubi; D. I. Abu-al-nadi

2007-01-01

424

Thermodynamics of vaporization of fission products and materials under severe reactor accident conditions: Analysis of molten core/concrete chemistry  

NASA Astrophysics Data System (ADS)

Vaporization-condensation processes can generate radioactive aerosols in the event of a core dryout and meltdown accident at a nuclear power station. The time sequence of fission produce vaporization and aerosol formation in relation to processes that can transport them out of the reactor containment is important for assessing their potential biohazard. Thermodynamics of vaporization of fission products and other materials are evaluated for the extreme environmental conditions projected by computer models if a molten core penetrates the reactor vessel and melts into the concrete base. A free energy minimization treatment was used to estimate partial pressures of gases in this many-component, multiphase system. The amounts of fission products and condensable materials vaporized were calculated for a test case involving basalt-aggregate concrete.

Cubicciotti, Daniel

1985-02-01

425

End-of-life irradiation performance of core structural components in the Shippingport Light Water Breeder Reactor  

SciTech Connect

Nondestructive and destructive end-of-life examinations of Light Water Breeder Reactor (LWBR) core structural components were performed following operation in the Shippingport Atomic Power Station for 29,047 effective full power hours. The Shippingport LWBR demonstrated that breeding can be achieved in a light water reactor with thorium and uranium-233 oxide fuel pellets contained in Zircaloy-4 tubes. The purpose of this presentation is to report results of LWBR core structural component examinations that were carried out to assess the effects of irradiation on support structure and to provide a data base for the evaluation of design procedures. The postirradiation nondestructive examinations included visual inspection and, in some cases, dye penetrant testing to assess structural integrity and surface conditions of the components. Destructive metallography was performed to assess cracking, corrosion buildup, and microstructural condition.

Clayton, J.C.; Smith, B.C.

1991-12-31

426

Three-dimensional Core Design of High Temperature Supercritical-Pressure Light Water Reactor with Neutronic and Thermal-Hydraulic Coupling  

Microsoft Academic Search

The equilibrium core of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H) is designed by three-dimensional neutronic and thermal-hydraulic coupled core calculations. The average coolant core outlet temperature of 500C is accurately evaluated for the first time in the development of the SCLWR-H.The average coolant core outlet temperature is one of the key parameters, which must be accurately determined in

Akifumi YAMAJI; Yoshiaki OKA; Seiichi KOSHIZUKA

2005-01-01

427

Tuberculosis IRIS: a mediastinal problem  

PubMed Central

We present a case of a 39 year old male patient with Acquired Immune Deficiency Syndrome (AIDS) who developed Mycobacterium tuberculosis related Immune Reconstitution Inflammatory Syndrome (IRIS) after initiation of Highly Active Antiretroviral Therapy (HAART) treatment. The inflammatory response resulted in mediastinal necrotic lymphadenopathy and subsequent perforation of the esophageal wall.

Valentin, Leonardo

2013-01-01

428

EXPOSURE SUMMARIES FOR IRIS CHEMICALS.  

EPA Science Inventory

The Integrated Risk Information System (IRIS), prepared and maintained by the National Center for Environmental Assessment (NCEA) of the U.S. Environmental Protection Agency (U.S. EPA), is an electronic database containing information on human health effects that may result from ...

429

Characterization of an energy source for modeling hypothetical core disruptive accidents in nuclear reactors. First interim report. [LMFBR  

Microsoft Academic Search

The expansion characteristics of the detonation products of a high-explosive energy source used to simulate the pressure-volume change relationships for sodium-vapor expansions during hypothetical core disruptive accidents in a Fast Test Reactor were determined experimentally. Rigid cylinder-piston experiments performed at two scales (ratio 1:3) were undertaken to determine a pressure-volume relationship as a function of source mass and expansion environment.

D. J. Cagliostro; A. L. Florence

1972-01-01

430

Large-Scale Water-Vapor Two-Phase Flow Simulations in Advanced Light Water Reactor Cores  

Microsoft Academic Search

Fluid flow characteristics in a fuel bundle of a reduced-moderation light water reactor (RMWR) with a tight-lattice core were analyzed numerically using a newly developed two-phase flow analysis code under the full bundle size condition. Conventional analysis methods such as subchannel codes need composition equations based on the experimental data. In case that there are no experimental data regarding to

Yoshida Hiroyuki; Takase Kazuyuki; Tamai Hidesada; Akimoto Hajime; Ose Yasuo

2004-01-01

431

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

SciTech Connect

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

Allen, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Shaw, S.E. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom); Huggon, A.P.; Steadman, R.J.; Thornton, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Whiley, G.S. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom)

2011-07-01

432

Performance of prompt- and delayed-responding self-powered in-core neutron detectors in a pressurized water reactor  

Microsoft Academic Search

An assembly of self-powered in-core neutron detectors has been tested for 6 yr over four fuel cycles in the Oconee 2 pressurized water reactor. The assembly contained both prompt-responding ytterbium and delayed-responding rhodium detectors. Two ytterbium detectors were paired with two rhodium detectors in the assembly. The experiment was conducted to define the long-term performance characteristics of the ytterbium detectors.

H. D. Warren; M. F. Sulcoski

1984-01-01

433

Core conversion analyses of the Syrian MNSR reactor from HEU to LEU and MEU fuel with homogeneously mixed burnable poisons.  

PubMed

A comprehensive analysis has been performed to investigate the conversion of the Syrian MNSR (miniature neutron source reactor) from current HEU fuel to selected alternatives LEU and MEU fuels. For this purposes the core design calculations related to design and engineering of LEU and MEU fuels have been carried out using the codes WIMSD/4 and BORGES-part of the MTR-PC and the code CITATION. Aiming at reducing the fuel enrichment by maintaining reactor power, thermal neutron flux and excess reactivity in the same range of the current MNSR design, two fuel alternatives of LEU (UO(2)-Mg) and MEU (U(3)Si(x)-Al) have been investigated. The results indicate that the first type (UO(2)-Mg) realizes the criticality conditions with low enrichment of 20% using the similar overall design of the present HEU fuel pins, whereas the second type (U(3)Si-Al) requires increasing the enrichment up to 33%. For the purpose of reactor core lifetime extension the possibility of mixing the burnable poisons Gd(157) and Cd(113) in the fresh core has been also explored. Thus, the calculation results indicate that the long-term control effect of Cd(113) on the excess reactivity is more homogeneous over the time due to the lower burn up rate of this burnable poison. PMID:19628402

Ghazi, N; Haj Hassan, H; Hainoun, A

2009-10-01

434

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

Microsoft Academic Search

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

Boyd D. Christensen

2009-01-01

435

Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual report for FY 2009  

SciTech Connect

This report documents progress made during FY 2009 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Studies are reported of the application of a silicon coating to surrogates for spheres of uranium-molybdenum alloy. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. A description of the progress in developing a finite element thermal hydraulics model of the LEU core is provided.

Chandler, David [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Sease, John D [ORNL; Guida, Tracey [University of Pittsburgh; Jolly, Brian C [ORNL

2010-02-01

436

Core burnup calculation and accidents analyses of a pressurized water reactor partially loaded with rock-like oxide fuel  

NASA Astrophysics Data System (ADS)

A rock-like oxide (ROX) fuel - light water reactor (LWR) burning system has been studied for efficient plutonium transmutation. For the improvement of small negative reactivity coefficients and severe transient behaviors of ROX fueled LWRs, a partial loading core of ROX fuel assemblies with conventional UO 2 assemblies was considered. As a result, although the reactivity coefficients could be improved, the power peaking tends to be large in this heterogeneous core configuration. The reactivity initiated accident (RIA) and loss of coolant accident (LOCA) behaviors were not sufficiently improved. In order to reduce the power peaking, the fuel composition and the assembly design of the ROX fuel were modified. Firstly, erbium burnable poison was added as Er 2O 3 in the ROX fuel to reduce the burnup reactivity swing. Then pin-by-pin Pu enrichment and Er content distributions within the ROX fuel assembly were considered. In addition, the Er content distribution was also considered in the axial direction of the ROX fuel pin. With these modifications, a power peaking factor even lower than the one in a conventional UO 2 fueled core can be obtained. The RIA and LOCA analyses of the modified core have also shown the comparable transient behaviors of ROX partial loading core to those of the UO 2 core.

Akie, H.; Sugo, Y.; Okawa, R.

2003-06-01

437

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities  

SciTech Connect

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

Michael A. Pope

2011-10-01

438

Disassembly and defueling of the TMI-2 (Three Mile Island Unit 2) reactor vessel lower core support assembly  

SciTech Connect

Planning for the disassembly and defueling of the Three Mile Island Unit 2 (TMI-2) reactor lower core support assembly (LCSA) began early in 1985. Evaluations of methods of defueling were performed based on various assumed LCSA conditions. Tooling was conceptualized and various defueling sequences were evaluated. As defueling of the core region progressed, information and data were obtained that clearly showed that a large amount of fuel was located within the LCSA and an even larger amount was located below the LSCA in the lower head of the vessel. It became apparent that it would be impractical to defuel this area of the reactor without massive cutting of the structural steel internal components. A plan was developed to defuel the LCSA by using two techniques: (a) core-boring LCSA components that had a round configuration (support posts and in-core instrument guide tubes) and (b) plasma arc cutting of components, which required linear cutting (plates and forgings). This paper describes the approach to LCSA disassembly, including the design basis of the equipment and sequence of in-vessel operations.

Austin, W.E.; Porter, L.H.

1988-01-01

439

Assessing the IRIS Professional Development Model: Impact Beyond the Workshops  

NASA Astrophysics Data System (ADS)

The IRIS Education and Outreach (E&O) Program has developed a highly effective, one-day professional development experience for formal educators. Leveraging the expertise of its consortium, IRIS delivers content including: plate tectonics, propagation of seismic waves, seismographs, Earth's interior structure. At the core of the IRIS professional development model is the philosophy that changes in teacher behavior can be affected by increasing teacher comfort in the classroom. Science and research organizations such as IRIS are able to increase teachers' comfort in the classroom by providing professional development which: increases an educator's knowledge of scientific content, provides educators with a variety of high-quality, scientifically accurate activities to deliver content to students, and provides educators with experiences involving both the content and the educational activities as the primary means of knowledge transfer. As reflected in a 2002-2003 academic year assessment program, this model has proven to be effective at reaching beyond participants and extending into the educators' classrooms. 76% of respondents report increasing the amount of time they spend teaching seismology or related topics in their classroom as a result of participating in IRIS professional development experience. This increase can be directly attributed to the workshop as 90% of participants report using at least one activity modeled during the workshop upon returning to their classrooms. The reported mean activity usage by teachers upon was 4.5 activities per teacher. Since the inception of the professional development model in 1999, IRIS E&O has been committed to evaluation. Data derived from assessment is utilized as a key decision making tool, driving a continuous improvement process. As a result, both the model and the assessment methods have become increasingly refined and sophisticated. The alignment of the professional development model within the IRIS E&O Program framework has resulted in a clarified a definition of success and an increased demand for the collection of new data. Currently, the assessment program is testing tools to examine participant learning, measure the transfer of knowledge and resources from professional development into in classrooms, and measure the use of individual activities.

Hubenthal, M.; Braile, L. W.; Taber, J. J.

2003-12-01

440

MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents  

SciTech Connect

The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR.

Ball, S.J. (Oak Ridge National Lab., TN (United States))

1991-10-01

441

Study of the Porosity Fluctuations in the Pebble Bed Core of a HTGR Reactor.  

National Technical Information Service (NTIS)

Experimentally studied is the structure of irregular filling of pebble bed fuel elements and estimated is the effect of spatial porosity fluctuations in the temperature field in the HTGR core with core filling. The porosity fluctuations of irregular filli...

K. I. Pelagejchenko M. D. Segal' N. A. Strebnev

1979-01-01

442

Iris recognition based on bant-limited phase correlation  

Microsoft Academic Search

In this paper, an iris recognition method based on phase correlation works on infrared iris images is explained. Pupil region is obtained using an edge detection based method and iris pattern is extracted. Iris pattern is normalized utilizing polar coordinate transform and enhanced using local histogram equalization. Iris matching is performed on database using band-limited phase correlation. The method is

S. Durmus?; Aysun Tayapi elebi; M. Kemal Gll

2010-01-01

443

An accurate and easy method towards iris localization  

Microsoft Academic Search

Iris segmentation is an essential module in iris recognition because it defines the effective image region used for subsequent processing such as feature extraction and matching of the iris. Traditional iris segmentation methods often involve in searching large region and the arithmetic is complex. Some suppose that pupil and iris are concentric circles, but most of time, the two circles

Peng Yu; Mei Xie

2010-01-01

444

Results of a Neutronic Simulation of HTR-Proteus Core 4.2 Using PEBBED and Other INL Reactor Physics Tools: FY09 Report.  

National Technical Information Service (NTIS)

The Idaho National Laboratorys deterministic neutronics analysis codes and methods were applied to the computation of the core multiplication factor of the HTR-Proteus pebble bed reactor critical facility. A combination of unit cell calculations (COMBINE-...

H. D. Gougar

2009-01-01

445

The DF-4 fuel damage experiment in ACRR (Annual Core Research Reactor) with a BWR (Boiling Water Reactor) control blade and channel box  

SciTech Connect

The DF-4 test was an experimental investigation into the melt progression behavior of boiling water reactor (BWR) core components under high temperature severe core damage conditions. In this study 14 zircaloy clad UO{sub 2} fuel rods, and representations of the zircaloy fuel canister and stainless steel/B{sub 4}C control blade were assembled into a 0.5 m long test bundle. The test bundle was fission heated in a flowing steam environment, using the Annular Core Research Reactor at Sandia Laboratories, simulating the environmental conditions of an uncovered BWR core experiencing high temperature damage as a result residual fission product decay heating. The experimental results provide information on the thermal response of the test bundle components, the rapid exothermic oxidation of the zircaloy fuel cladding and canister, the production of hydrogen from metal-steam oxidation, and the failure behavior of the progressively melting bundle components. This information is provided in the form of thermocouple data, steam and hydrogen flow rate data, test bundle fission power data and visual observation of the damage progression. In addition to BWR background information, this document contains a description of the experimental hardware with details on how the experiment was instrumented and diagnosed, a description of the test progression, and a presentation of the on-line measurements. Also in this report are the results of a thermal analysis of the fueled test section of the fueled test section of the experiment demonstrating an overall consistency in the measurable quantities from the test. A discussion of the results is provided. 38 refs., 72 figs., 7 tabs.

Gauntt, R.O.; Gasser, R.D.; Ott, L.J. (Sandia National Labs., Albuquerque, NM (USA))

1989-11-01

446

IRIS: Animations of Plate Tectonics  

NSDL National Science Digital Library

This is a collection of animations on dynamic earth processes: plate tectonics, earthquakes, volcanoes, and seismic waves. Users can explore the interaction of Earth's tectonic plates, view models of P and S wave propagation, study how seismographs work, monitor earthquakes and volcanoes, and get instructions for modeling earthquakes in the classroom. This resource is part of IRIS, the Incorporated Research Institutions for Seismology, a consortium of international laboratories and data collection centers.

2011-03-18

447

DCT-based iris recognition.  

PubMed

This paper presents a novel iris coding method based on differences of discrete cosine transform (DCT) coefficients of overlapped angular patches from normalized iris images. The feature extraction capabilities of the DCT are optimized on the two largest publicly available iris image data sets, 2,156 images of 308 eyes from the CASIA database and 2,955 images of 150 eyes from the Bath database. On this data, we achieve 100 percent Correct Recognition Rate (CRR) and perfect Receiver-Operating Characteristic (ROC) Curves with no registered false accepts or rejects. Individual feature bit and patch position parameters are optimized for matching through a product-of-sum approach to Hamming distance calculation. For verification, a variable threshold is applied to the distance metric and the False Acceptance Rate (FAR) and False Rejection Rate (FRR) are recorded. A new worst-case metric is proposed for predicting practical system performance in the absence of matching failures, and the worst case theoretical Equal Error Rate (EER) is predicted to be as low as 2.59 x 10(-4) on the available data sets. PMID:17299216

Monro, Donald M; Rakshit, Soumyadip; Zhang, Dexin

2007-04-01

448

IRIS thermal balance test within ESTEC LSS  

NASA Technical Reports Server (NTRS)

The Italian Research Interim Stage (IRIS) thermal balance test was successfully performed in the ESTEC Large Space Simulator (LSS) to qualify the thermal design and to validate the thermal mathematical model. Characteristics of the test were the complexity of the set-up required to simulate the Shuttle cargo bay and allowing IRIS mechanism actioning and operation for the first time in the new LSS facility. Details of the test are presented, and test results for IRIS and the LSS facility are described.

Messidoro, Piero; Ballesio, Marino; Vessaz, J. P.

1988-01-01

449

Robust Iris Recognition Using Advanced Correlation Techniques  

Microsoft Academic Search

\\u000a The iris is considered one of the most reliable and stable biometrics as it is believed to not change significantly during\\u000a a persons lifetime. Standard techniques for iris recognition, popularized by Daugman, apply Gabor wavelet analysis for feature\\u000a extraction. In this paper, we consider an alternative method for iris recognition, the use of advanced distortion-tolerant\\u000a correlation filters for robust pattern

Jason Thornton; Marios Savvides; B. V. K. Vijaya Kumar

2005-01-01

450

A Robust Iris Segmentation with Fuzzy Supports  

Microsoft Academic Search

\\u000a Today, iris recognition is reported as one of the most reliable biometric approaches. With the strength of contactless, the\\u000a hygienic issue is therefore minimized and the possibility of disease infection through the device as a medium is low. In this\\u000a paper, a MMU2 iris database with such consideration is created for this study. Moreover, the proposed iris segmentation method\\u000a has

Chuan Chin Teo; Han Foon Neo; Michael Goh Kah Ong; Connie Tee; K. S. Sim

2010-01-01

451

Iris segmentation using variational level set method  

NASA Astrophysics Data System (ADS)

Continuous efforts have been made to process degraded iris images for enhancement of the iris recognition performance in unconstrained situations. Recently, many researchers have focused on developing the iris segmentation techniques, which can deal with iris images in a non-cooperative environment where the probability of acquiring unideal iris images is very high due to gaze deviation, noise, blurring, and occlusion by eyelashes, eyelids, glasses, and hair. Although there have been many iris segmentation methods, most focus primarily on the accurate detection of iris images captured in a closely controlled environment. The novelty of this research effort is that we propose to apply a variational level set-based curve evolution scheme that uses a significantly larger time step to numerically solve the evolution partial differential equation (PDE) for segmentation of an unideal iris image accurately, and thereby, speeding up the curve evolution process drastically. The iris boundary represented by the variational level set may break and merge naturally during evolution, and thus, the topological changes are handled automatically. The proposed variational model is also robust against poor localization and weak iris/sclera boundaries. In order to solve the size irregularities occurring due to arbitrary shapes of the extracted iris/pupil regions, a simple method is applied based on connection of adjacent contour points. Furthermore, to reduce the noise effect, we apply a pixel-wise adaptive 2D Wiener filter. The verification and identification performance of the proposed scheme is validated on three challenging iris image datasets, namely, the ICE 2005, the WVU Unideal, and the UBIRIS Version 1.

Roy, Kaushik; Bhattacharya, Prabir; Suen, Ching Y.

2011-04-01

452

Securing iris recognition systems against masquerade attacks  

NASA Astrophysics Data System (ADS)

A novel two-stage protection scheme for automatic iris recognition systems against masquerade attacks carried out with synthetically reconstructed iris images is presented. The method uses different characteristics of real iris images to differentiate them from the synthetic ones, thereby addressing important security flaws detected in state-of-the-art commercial systems. Experiments are carried out on the publicly available Biosecure Database and demonstrate the efficacy of the proposed security enhancing approach.

Galbally, Javier; Gomez-Barrero, Marta; Ross, Arun; Fierrez, Julian; Ortega-Garcia, Javier

2013-05-01

453

Monte Carlo modeling of eye iris color  

NASA Astrophysics Data System (ADS)

Based on the presented two-layer eye iris model, the iris diffuse reflectance has been calculated by Monte Carlo technique in the spectral range 400-800 nm. The diffuse reflectance spectra have been recalculated in L*a*b* color coordinate system. Obtained results demonstrated that the iris color coordinates (hue and chroma) can be used for estimation of melanin content in the range of small melanin concentrations, i.e. for estimation of melanin content in blue and green eyes.

Koblova, Ekaterina V.; Bashkatov, Alexey N.; Dolotov, Leonid E.; Sinichkin, Yuri P.; Kamenskikh, Tatyana G.; Genina, Elina A.; Tuchin, Valery V.

2007-06-01

454

Junction device between the delivery duct of a primary pump and a duct joined to the core support of a fast neutron nuclear reactor  

Microsoft Academic Search

A junction device between the delivery duct of a primary pump and a duct joined to the core support of a liquid metal cooled fast neutron nuclear reactor comprises a frusto-conical sleeve widening out towards the inlet of the duct joined to the core support, a connecting member joined to the sleeve, a sealing device interposed between the outlet end

M. Thevenin; G. Jullien

1985-01-01

455

Radiation transport in a liquid-metal fast breeder reactor during a loss of sodium coolant reactor core disassembly  

Microsoft Academic Search

The time-dependent radiation transport for a demonstration scale liquid-metal-cooled fast breeder reactor that has undergone a severe loss of sodium coolant is calculated with both a discrete ordinates and a diffusion theory solution for the real neutron flux shape. It is found that diffusion theory underpredicts reactivity levels by about $6 when compared to discrete ordinates. It is also found

Rzepecki

1985-01-01

456

Safety Analysis of Fast Reactor Core with Uranium-Free Fuel for Actinide Transmutation  

Microsoft Academic Search

The solving of ecological problems of future nuclear power is connected with the solving of long-lived radioactive waste utilization problems. This concerns primarily plutonium and minor actinides (neptunium, americium, and curium), accumulated in the spent fuel of nuclear reactors. One of the ways this can be solved is to use a fast reactor with uranium-free fuel. The physics of this

Igor Krivitski; Mikhail Vorotyntsev; Valentin Pyshin; Ludmila Korobeinikova

2003-01-01

457

AVR-Pebble-Bed Reactor: Core-Physics Control Without Incore-Instrumentation.  

National Technical Information Service (NTIS)

The AVR 15 MWe test power station is equipped with the first high temperature pebble bed reactor. It is He-cooled and Graphite-moderated. Recent developments in the German HTR-area towards