WorldCat Identities

Hummel, H. H.

Overview
Works: 15 works in 19 publications in 1 language and 116 library holdings
Genres: Juvenile works  Biography 
Roles: Author
Classifications: QC787.N8, 630.92
Publication Timeline
.
Most widely held works by H. H Hummel
Stability analysis of EBR-II by H. H Hummel( )

2 editions published in 1962 in English and held by 23 WorldCat member libraries worldwide

Calculations were made for predicting the resonance of EBR-II to oscillator measurements to be made during startup of the reactor. Because of assumptions made in the calculations, which are believed to be justified by the design of the reactor, the only feedbacks are prompt negative ones. No instability in the calculated behavior is therefore possible
Calculations of the Doppler coefficient of large ceramic-fueled fast reactors by M. G Bhide( )

2 editions published in 1962 in English and held by 23 WorldCat member libraries worldwide

Calculations of the Doppler coefficient of large, ceramic-fueled fast reactors containing plutonium were made by means of the ELMOE program to provide accurate flux calculations. The temperature-dependent cross sections of P. Greebler et al., for U-238 and plutonium-239 were used. Doppler coefficients of the order of 10⁵ intermediate temperature k/ C were obtained, in agreement with Greebler's results. Coefficients of carbide fueled reactors are about 0.7 of those of oxide-fueled reactors at the same fuel enrichment. Effective coarse-group elastic-removal cross sections for light elements are tabulated. It appears that, with tabulations of this sort as a guide, coarse-group sets of cross sections can be constructed to give adequate accuracy in calculations of Doppler coefficients without use of ELMOE
Compressible analysis of inlet plenum pressure rise due to sodium boiling in fuel subassemblies during pump coast down of an LMFBR( )

2 editions published in 1980 in English and held by 19 WorldCat member libraries worldwide

The effect of sodium compressibility and steel elasticity on the rise in inlet plenum pressure occurring during boiling in a loss-of-flow accident in an LMFBR has been investigated using the require consideration in accident analysis. The pressure rise is less for pool than for loop designs. 3 refs., 1 fig., 9 tabs
ELMOE : an IBM-704 program treating elastic scattering resonances in fast reactors by A. L Rago( )

1 edition published in 1964 in English and held by 16 WorldCat member libraries worldwide

Fast fuel test reactor - : FFTR conceptual design study( )

2 editions published in 1960 in English and held by 13 WorldCat member libraries worldwide

The Fast Fuel Test Reactor (FFTR) is a nuclear facility for the purpose of irradiating samples of fuels and structural components for use in fast reactors. The core consists of a plate type element in a square configuration. Beryllium metal between the fuel elements is used to obtain a neutron energy spectrum in the hard intermediate region. Cooling of the core and test specimens is accomplished by means of liquid sodium. The design concept was carried through in sufficient degree in the following areas of preliminary concern: number and size of irradiation facilities, sample power requirements, plant layout to evaluate site requirements, plant and nuclear design parameters to evaluate essential equipment requirements. plant-capital-cost estimate, annual- operating-cost estimate, and estimate of construction time schedule
ELMOE: an IBM-704 program treating elastic scattering resonances in fast reactors by A. L Rago( Book )

1 edition published in 1964 in English and held by 8 WorldCat member libraries worldwide

A description of the operations carried out by the ELMOE code is given. This code carries out MUFT-type calculations for dealing with the elastic scattering resonances of the light elements present in fast reactors. Sources of information for the cross-section library in current use with the program, input specifications, and operating instructions are also given
Fast fuel test reactor : FFTR conceptual design study( )

1 edition published in 1960 in English and held by 6 WorldCat member libraries worldwide

The Fast Fuel Test Reactor (FFTR) is a nuclear facility for the purpose of irradiating samples of fuels and structural components for use in fast reactors. The core consisis of a plate type element in a square configuration. Beryllium metal between the fuel elements is used to obtain a neutron energy spectrum in the hard intermediate region. Cooling of the core and test specimens is accomplished by means of liquid sodium. The design concept was carried through in sufficient degree in the following areas of preliminary concern: number and size of irradiation facilities, sample power requirements, plant layout to evaluate site requirements, plant and nuclear design parameters to evaluate essential equipment requirements. plant-capital-cost estimate, annual- operating-cost estimate, and estimate of construction time schedule. (W.D.M.)
Coupled fast-thermal power breeder critical experiment by R Avery( Book )

1 edition published in 1958 in English and held by 3 WorldCat member libraries worldwide

Physics calculations for the Clinch River Breeder Reactor by Kalimullah( Book )

1 edition published in 1977 in English and held by 1 WorldCat member library worldwide

Calculations of distributions of power and sodium void reactivity, unvoided and voided Doppler coefficients and steel and fuel worths have been performed using diffusion theory and first-order perturbation theory for the LWR discharge plutonium-fueled CRBR at BOL, the FFTF-grade plutonium-fueled CRBR at BOL and for the beginning and end of equilibrium cycle of the LWR-plutonium-fueled CRBR. The results of the burnup and breeding ratio calculations performed for obtaining the reactor compositions during the equilibrium cycle are also reported. Effects of sodium and steel contents on the distributions of sodium void reactivity and steel worth have also been studied. Errors and uncertainties in the reactivity coefficients due to cross-sections and the two-dimensional geometric representations of the reactor used in the calculations have also been estimated. Comparisons of the results with those in the CRBR PSAR are also discussed
Parametric studies of the reactivity coefficients for large U²³³-Th-fueled fast reactors by R. S Singh( Book )

1 edition published in 1966 in English and held by 1 WorldCat member library worldwide

The feasibility of using uranium-233-thorium as a fuel in a large fast breeder reactor has been studied from the point of view of the Doppler and sodium-void coefficients. The composition and temperature-dependent cross sections for thorium have been evaluated by using the latest resonance parameters available. The sodium-void coefficient has been studied in detail over a range of compositions and sizes of core for metal and oxide fuels, and the effect of the addition of uranium-234 and fission products on the coefficient is also investigated. The reactivity coefficients due to the Doppler effect are calculated for metal- and oxide-fueled systems of interest. The results indicate that, even with present cross-section uncertainties, the reactivity changes due to both the sodium loss and the Doppler effect are encouraging for the development of fast breeder reactors with very large single cores if uranium-233 are used as the fuel
<> by H. H Hummel( Book )

1 edition published in 1971 in English and held by 1 WorldCat member library worldwide

Studies of unprotected loss-of-flow accidents for the Clinch River breeder reactor by H. H Hummel( Book )

1 edition published in 1976 in English and held by 1 WorldCat member library worldwide

Studies of unprotected loss-of-flow accidents in the CRBR for various rates of flow coast-down and with various options in the SAS 3A code did not lead to conditions for a violent disassembly. Maximum fuel temperatures using the SLUMPY module for disassembly were in the range 4000-4500 deg C. An approximate treatment of the LOF-driven TOP accident, not properly modeled by SAS 3A, indicates the possibility of some increase in accident severity. The effect of fission gas in dispersing fuel was not taken into account in these calculations. Parameter variations included the presence or absence of axial fuel expansion and of clad motion and use of the moving coolant film model versus the static film model. Study of severe pipe rupture accidents with scram indicated that pin power density and fuel-clad conductance were important parameters in determining what coolant flow rate was needed to prevent boiling after the rupture. It appears that for the CRBR when engineering hot channel factors are considered, this fraction would have to exceed 25 percent
Shrine waltz-song by E Riley( )

1 edition published in 1922 in English and held by 1 WorldCat member library worldwide

Grandpa's chain : companion to Grandma's chain by M. Edna Hummel-Townsend( Book )

1 edition published in 1982 in English and held by 0 WorldCat member libraries worldwide

True stories from the early-twentieth-century life of a farm family in Darke County, Ohio. Accompanying sheet contains a glossary and educational activities related to each story
Uncle Mike : a link in Grandpa's chain by M. Edna Hummel-Townsend( Book )

1 edition published in 1984 in English and held by 0 WorldCat member libraries worldwide

The author's grandfather recalls raising vegetables for sale with his Uncle Mike on the family farm in Ohio during the early 1900s. Includes a brief list of enrichment activities
 
moreShow More Titles
fewerShow Fewer Titles
Audience Level
0
Audience Level
1
  Kids General Special  
Audience level: 0.76 (from 0.52 for <> ... to 1.00 for Uncle Mike ...)

Languages
English (19)