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The AMPX code system has been limited to processing Version 5 formats of the Evaluated Nuclear Data File (ENDF). Work is in progress to upgrade the AMPX code system to process ENDF/B-VI data. A substantial amount of code development is... more
The AMPX code system has been limited to processing Version 5 formats of the Evaluated Nuclear Data File (ENDF). Work is in progress to upgrade the AMPX code system to process ENDF/B-VI data. A substantial amount of code development is complete, and the latest version of AMPX has the capability to (a) process cross sections in the resolved resonance region (RRR) using the multipole formalism and (b) produce continuous-energy cross sections from ENDF/B-VI data. Comprehensive testing of the new modules is underway, and the purpose of this work is to establish the capabilities of the new AMPX code system for criticality safety applications.
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The development of the vector processor, which is used in the current generation of supercomputers and is beginning to be used in workstations, provides the potential for dramatic speed-up for codes that are able to process data as... more
The development of the vector processor, which is used in the current generation of supercomputers and is beginning to be used in workstations, provides the potential for dramatic speed-up for codes that are able to process data as vectors. Unfortunately, the stochastic nature of Monte Carlo codes prevents the old scalar version of these codes from taking advantage of the vector processors. New Monte Carlo algorithms that process all the histories undergoing the same event as a batch are required. Recently, new vectorized Monte Carlo codes have been developed that show significant speed-ups when compared to the scalar version of themselves or equivalent codes. This paper discusses the vectorization of an already existing and widely used criticality safety code, KENO V.a All the changes made to KENO V.a are transparent to the user making it possible to upgrade from the standard scalar version of KENO V.a to the vectorized version without learning a new code.
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Most criticality experiments and excursion models deal with single units of fissile material. In this paper, the authors report on the development of a model to predict the results of a hypothetical excursion involving an array of fissile... more
Most criticality experiments and excursion models deal with single units of fissile material. In this paper, the authors report on the development of a model to predict the results of a hypothetical excursion involving an array of fissile units. The array of interest in this work consists of a planar array of several polyethylene bottles, each containing a low-enriched solution of uranyl fluoride. The computer model calculates the power history, the total number of fissions, the total system reactivity, and the temperature of each bottle in the array. The model employs point neutronis coupled with simple lumped parameter feedback. The most important feedback effect considered is temperature. Temperature effects are treated using a single factor that describes both Doppler and thermal expansion effects.
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As the speed, available memory, and reliability of computer hardware increases and the cost decreases, the complexity and usability of computer software will increase, taking advantage of the new hardware capabilities. Computer programs... more
As the speed, available memory, and reliability of computer hardware increases and the cost decreases, the complexity and usability of computer software will increase, taking advantage of the new hardware capabilities. Computer programs today must be more flexible and user friendly than those of the past. Within available resources, the SCALE staff at Oak Ridge National Laboratory (ORNL) is committed to upgrading its computer codes to keep pace with the current level of technology. This paper examines recent additions and enhancements to the criticality safety analysis sections of the SCALE code package. These recent additions and enhancements made to SCALE can be divided into nine categories: (1) new analytical computer codes, (2) new cross-section libraries, (3) new criticality search sequences, (4) enhanced graphical capabilities, (5) additional KENO enhancements, (6) enhanced resonance processing capabilities, (7) enhanced material information processing capabilities, (8) portab...
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This paper describes the development of a computer model for predicting the excursion characteristics of a postulated, hypothetical, critically accident involving a homogeneous mixture of low-enriched UO[sub 2] powder and water contained... more
This paper describes the development of a computer model for predicting the excursion characteristics of a postulated, hypothetical, critically accident involving a homogeneous mixture of low-enriched UO[sub 2] powder and water contained in a cylindrical blender. The model uses point neutronics coupled with simple lumped-parameter thermal-hydraulic feedback. The temperature of the system is calculated using a simple time-dependent energy balance
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ABSTRACT This paper describes the Monte Carlo codes KENO V.a and KENO-VI in SCALE that are primarily used to calculate multiplication factors and flux distributions of fissile systems. Both codes allow explicit geometric representation of... more
ABSTRACT This paper describes the Monte Carlo codes KENO V.a and KENO-VI in SCALE that are primarily used to calculate multiplication factors and flux distributions of fissile systems. Both codes allow explicit geometric representation of the target systems and are used internationally for safety analyses involving fissile materials. KENO V.a has limiting geometric rules such as no intersections and no rotations. These limitations make KENO V.a execute very efficiently and run very fast. On the other hand, KENO-VI allows very complex geometric modeling. Both KENO codes can utilize either continuous-energy or multigroup cross-section data and have been thoroughly verified and validated with ENDF libraries through ENDF/B-VII.0, which has been first distributed with SCALE 6. Development of the Monte Carlo solution technique and solution methodology as applied in both KENO codes is explained in this paper. Available options and proper application of the options and techniques are also discussed. Finally, performance of the codes is demonstrated using published benchmark problems.
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A suite of heretofore unreported experiments were performed in 1969 at the Oak Ridge Critical Experiments Facility (ORCEF). These experiments are useful for validating criticality safety calculations for poorly moderated enriched uranium... more
A suite of heretofore unreported experiments were performed in 1969 at the Oak Ridge Critical Experiments Facility (ORCEF). These experiments are useful for validating criticality safety calculations for poorly moderated enriched uranium systems in that the average energy of the neutrons causing fission spans the range from 2 eV to 1 MeV - a range not typically validated, The ORCEF
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An exploration of flow regime characterization using measurements of the neutron fluctuations induced by a gasliquid flow passing through a nuclear reactor core has reached a developmental plateau that permits a quantitative evaluation of... more
An exploration of flow regime characterization using measurements of the neutron fluctuations induced by a gasliquid flow passing through a nuclear reactor core has reached a developmental plateau that permits a quantitative evaluation of the method, Albrecht et al. (1982), Crowe et al. (1977). It has been demonstrated that many significant properties of the flow can be determined by analysis of the induced neutronic fluctuations. These characteristics have been categorized as propagation, structure, and dynamics of the flow, Albrecht et al. (1982c).
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The commercial nuclear power industry is investing heavily in advanced fuels that can produce higher power levels with a higher safety margin and be manufactured at low cost. Although chemically stable and inexpensive to manufacture, UO2... more
The commercial nuclear power industry is investing heavily in advanced fuels that can produce higher power levels with a higher safety margin and be manufactured at low cost. Although chemically stable and inexpensive to manufacture, UO2 fuel is limited by its thermal conductivity. If the fuel thermal conductivity could be increased, existing nuclear reactors would be able to operate at higher powers and new reactors could be designed with higher power densities, thus decreasing the overall cost of electricity and the number of new electrical generating plants needed to meet demand. Movement to higher U-235 enrichments and fuel burnups is limited by UO2 thermal conductivity, which decreases with fuel burnup. Preliminary studies indicate that adding 1% to 5% by volume of long, thin fibers having a high thermal conductivity can substantially increase the bulk thermal conductivity of standard UO2 fuel with minimal decreases in its fissile inventory. The fibers need to remain intact to ...
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Research Interests: Nuclear Engineering, Materials Science, Structural Integrity, Computer Simulation, Nuclear power, and 11 moreProject Planning, Grain Growth, Nuclear reactor, Nuclear Fuel Cycle, Thermal Conductivity, Cost of Electricity, Nuclear Fuel, Burnup, Microstructures, Electricity Generation, and Power Output
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This paper presents the results of a preliminary criticality safety study of some potential effects of uranium reduction and aggregation in the Molten Salt Reactor Experiment (MSRE) fuel drain tanks (FDTs) during salt removal operations.... more
This paper presents the results of a preliminary criticality safety study of some potential effects of uranium reduction and aggregation in the Molten Salt Reactor Experiment (MSRE) fuel drain tanks (FDTs) during salt removal operations. Since the salt was transferred to the FDTs in 1969, radiological and chemical reactions have been converting the uranium and fluorine in the salt to
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KENO V.a and KENO-VI are Monte Carlo codes that solve the multigroup form of the Boltzmann transport equation. These codes are part of the SCALE system of codes and are used for performing criticality calculations of systems with... more
KENO V.a and KENO-VI are Monte Carlo codes that solve the multigroup form of the Boltzmann transport equation. These codes are part of the SCALE system of codes and are used for performing criticality calculations of systems with fissionable material. In general, continuous-energy Monte Carlo methods are preferred because such an approach avoids many of the assumptions inherent in the
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Page 1. New Capabilities to Calculate Volumes of SCALE/KENO-VI Geometry Models James E. Horwedel, Stephen M. Bowman, and Daniel F. Hollenbach Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 ...
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Herndon suggested that the inner core of the Earth consists, not of partially crystallized iron metal, but of nickel silicide. He has shown by fundamental mass ratios that i) the Earth as a whole, especially the inner 82%, has a state of... more
Herndon suggested that the inner core of the Earth consists, not of partially crystallized iron metal, but of nickel silicide. He has shown by fundamental mass ratios that i) the Earth as a whole, especially the inner 82%, has a state of oxidation like primitive enstatite chondrites, and ii) the lower mantle and core are similar in composition to the Abee enstatite chondrite. By analogy with Abee data, CaS and MgS precipitates from the core are expected to collect at the core-mantle boundary and, significantly, a major fraction of the actinides are expected to precipitate from the core and to collect at the center of the Earth. Herndon demonstrated the feasibility of a nuclear fission reactor at the center of the Earth as the energy source for the geomagnetic field and described a natural mechanism that would lead to variations in energy production and thus variations in the geomagnetic field. Hollenbach and Herndon produced numerical simulations of the operation of the geo-reactor ...
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ABSTRACT
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Geomagnetic field reversals and changes in intensity are understandable from an energy standpoint as natural consequences of intermittent and/or variable nuclear fission chain reactions deep within the Earth. Moreover, deep-Earth... more
Geomagnetic field reversals and changes in intensity are understandable from an energy standpoint as natural consequences of intermittent and/or variable nuclear fission chain reactions deep within the Earth. Moreover, deep-Earth production of helium, having 3 He/ 4 He ratios within the range observed from deep-mantle sources, is demonstrated to be a consequence of nuclear fission. Numerical simulations of a planetary-scale geo-reactor were made by using the SCALE sequence of codes. The results clearly demonstrate that such a geo-reactor ( i ) would function as a fast-neutron fuel breeder reactor; ( ii ) could, under appropriate conditions, operate over the entire period of geologic time; and ( iii ) would function in such a manner as to yield variable and/or intermittent output power.
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... Description/Abstract. In a previous paper, a parallel version of the Monte Carlo criticality safety code KENO Va was presented based on only one parallel algorithm. ... Authors: Basoglu, B.;Bentley, C.; Dunn, M.; Goluoglu, S.;... more
... Description/Abstract. In a previous paper, a parallel version of the Monte Carlo criticality safety code KENO Va was presented based on only one parallel algorithm. ... Authors: Basoglu, B.;Bentley, C.; Dunn, M.; Goluoglu, S.; Paschal, L.; Dodds, H. [Univ. ...
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... Stephen M. BOWMAN * , Daniel F. HOLLENBACH, Mark D. DeHART, Bradley T. REARDEN, Ian C. GAULD and Sedat GOLUOGLU Oak Ridge National Laboratory , PO Box 2008, Oak Ridge, Tennessee 37831-6370, USA ...