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TABULATED VALUES OF SCATTERED GAMMA-RAY FLUXES IN WATER INTERPOLATED FROM MOMENTS METHOD CALCULATIONS

Description: Tables of scattered gamma-ray fluxes in water that are suitable for integration over source spectra were generated on the IBM-7090 computer by Lagrangian interpolation of moments method results for point isotropic gamma-ray sources. Values of the fluxes were obtained for a consistent set of scattered gamma-ray energies taken in steps of 0.1 Mev from 0.1 to 1.0 Mev and in steps of 0.25 Mev from 0.25 to 10 Mev. One table was obtained for each scattered energy for various values of the source energy and of the distance from the source in mean free paths. Tables were then generated in which the distance from the source was converted from mean free paths to centimeters. The latter set may be integrated over any source spectrum to obtain the flux of scattered gamma rays in water, and, with small error, in any material in which the cross section is dominated by the Compton scattering cross section. This includes the low-Z materials such as aluminum and air. (auth)
Date: July 29, 1963
Creator: Trubey, D.K.
Partner: UNT Libraries Government Documents Department

TWO AUXILIARY CODES FOR USE WITH RENUPAK

Description: Two IBM-7090 codes were written to aid the user of the neutron moments method code RENUPAK. One code computes and punches response function input cards for RENUPAK or NIOBE (another neutron transport code). The second code reads RENUPAK flux tapes and prints out a compact edit including dose rate as a function of distance. (auth)
Date: August 10, 1962
Creator: Trubey, D.K.
Partner: UNT Libraries Government Documents Department

Standard reference data for gamma-ray transport in homogeneous media. [PWR; BWR]

Description: An American Nuclear Society Standards Committee Working Group, identified as ANS-6.4.3, is developing a set of evaluated gamma-ray isotropic point-source buildup factors and attenuation coefficients for a standard reference data base. As a first step, a largely unpublished set of buildup factors calculated with the moments method is being evaluated by recalculating key values with Monte Carlo, integral transport, and discrete ordinates methods. Attention is being given to frequently neglected processes such as bremsstrahlung and the effect of introducing a tissue phantom behind the shield. The proposed standard will contain data for a source energy range from 15 keV to 15 MeV and for approximately 12 elements and 3 mixtures (water, air, and concrete).
Date: January 1, 1983
Creator: Trubey, D.K.
Partner: UNT Libraries Government Documents Department

ANS shielding standards for light-water reactors

Description: The purpose of the American Nuclear Society Standards Subcommittee, ANS-6, Radiation Protection and Shielding, is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. A total of seven published ANS-6 standards are now current. Additional projects of the subcommittee, now composed of nine working groups, include: standard reference data for multigroup cross sections, gamma-ray absorption coefficients and buildup factors, additional benchwork problems for shielding problems and energy spectrum unfolding, power plant zoning design for normal and accident conditions, process radiation monitors, and design for postaccident radiological conditions.
Date: January 1, 1982
Creator: Trubey, D.K.
Partner: UNT Libraries Government Documents Department

Radiation protection and shielding standards for the 1980s

Description: The American Nuclear Society (ANS) is a standards-writing organization member of the American National Standards Institute (ANSI). The ANS Standards Committee has a subcommittee denoted ANS-6, Radiation Protection and Shielding, whose charge is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. This paper is a progress report of this subcommittee. Significant progress has been made since the last comprehensive report to the Society.
Date: January 1, 1982
Creator: Trubey, D.K.
Partner: UNT Libraries Government Documents Department

DEPOSITION OF GAMMA-RAY HEATING IN STRATIFIED LEAD AND WATER SLABS

Description: Typical results are given from a calculation of the deposition of heat in stratified lead and water slabs caused by a monodirectional, monoenergetic beam of gamma rays incident on the slabs. A total of 512 cases were calculated for infinite slabs with finite thicknesses of 1, 2, 4, and 6 mean free paths; source energies of 1, 3, 8, and 10 Mev, and source angles of incidence which were chosen to give slant slab thicknesses of 1, 2, 3, and 4 times the normal thickness. The results were fitted to an empirical formula, which can be simplified for special cases. While for the cases examined, the fit was usually good to within 5%, it is to be emphasized that the formula has been compared only with the results from a very limited number of parameters. (auth)
Date: July 28, 1958
Creator: Bowman, L.A. & Trubey, D.K.
Partner: UNT Libraries Government Documents Department

Specific gamma-ray dose constants for nuclides important to dosimetry and radiological assessment

Description: Tables of specific gamma-ray dose constants (the unshielded gamma-ray dose equivalent rate at 1 m from a point source) have been computed for approximately 500 nuclides important to dosimetry and radiological assessment. The half life, the mean attenuation coefficient, and thickness for a lead shield providing 95% dose equivalent attenuation are also listed.
Date: May 1, 1982
Creator: Unger, L.M. & Trubey, D.K.
Partner: UNT Libraries Government Documents Department

New buildup factor data for point kernel calculations

Description: An American Nuclear Society Standards Committee Working Group, identified as ANS-6.4.3, is developing a set of evaluated gamma-ray isotropic point-source buildup factors and attenuation coefficients for a standard reference data base. As a first step, a largely unpublished set of buildup factors calculated with the moments method has been evaluated by recalculating key values with Monte Carlo, integral transport, and discrete ordinates methods. Attention is being given to frequently-neglected processes such as bremsstrahlung and the effect of introducing a tissue phantom behind the shield. The proposed standard contains data for a source energy range from 15 keV to 15 MeV and for approximately 19 elements and 3 mixtures (water, air, and concrete). The data will also be represented as coefficients for the G-P fitting function. The 1985 data base was released as part of the CCC-493B/QAD-CGGP code package available from the Radiation Shielding Information Center (RSIC).
Date: January 1, 1986
Creator: Trubey, D.K. & Harima, Y.
Partner: UNT Libraries Government Documents Department

Specific gamma-ray dose constants for nuclides important to dosimetry and radiological assessment

Description: Tables of specific gamma-ray dose constants (the unshielded gamma-ray dose equivalent rate at 1 m from a point source) have been computed for approximately 500 nuclides important to dosimetry and radiological assessment. The half life, the mean attenuation coefficient, and thickness for a lead shield providing 95% dose equivalent attenuation are also listed.
Date: September 1, 1981
Creator: Unger, L.M. & Trubey, D.K.
Partner: UNT Libraries Government Documents Department

EXREM III computer code for estimating external radiation doses to populations from environmental releases

Description: EXREM III is a computer code to estimate the dose equivalent rate and the total dose equivalent from beta, positron, electron, and gamma radintion resulting from submersion in contaminated water, submersion in contaminated air, and exposure to a contaminated surface. There can be more than one environmental release, and exposure can begin at any time after the first release. EXREM III considers contributions from environmental releases and from nuclide decay chains. For a particular problem the user may choose to calculate either the dose rates, or the total doses, or both for any of the three modes of exposure. A separate solution array is printed for each mode of exposure. EXREM III is a revised version of EXREM II which was available earlier. The principal revisions include treatment of positron and electron radiations, selection of nuclear data from a data base, variable dimensioning of large data arrays, and free field input. (auth)
Date: December 1, 1973
Creator: Trubey, D.K. & Kaye, S.V.
Partner: UNT Libraries Government Documents Department

A CDC-1604 SUBROUTINE PACKAGE FOR MAKING LINEAR, LOGARITHMIC AND SEMILOGARITHMIC GRAPHS USING THE CALCOMP PLOTTER

Description: A CDC-1604 subroutine package was written to facilitate the plotting of curves and points on linear, logarithmic, and semilogarithmic graphs using the CALCOMP plotter. The subroutines accomplish the necessary computations and prepare a magnetic tape for use by the plotter. (auth)
Date: June 24, 1963
Creator: Trubey, D.K. & Emmett, M.B.
Partner: UNT Libraries Government Documents Department

AN ESTIMATE OF THE NONLEAKAGE PROBABILITY FOR BARE AQUEOUS HOMOGENEOUS U$sup 235$ REACTORS

Description: The slowing-down distribution, to thermal energy, of neutrons from a U/ sup 235/ fission source in an infinite H/sub 2/O medium up to 160 cm was determined from experimental data. The Fourier transform of this distribution and the In resonance distribution, which are the nonleakage probabilities for bare reactors, were also determined. (auth)
Date: November 1, 1957
Creator: Trubey, D.K.; Moran, H.S. & Weinberg, A.M.
Partner: UNT Libraries Government Documents Department

Available computer codes and data for radiation transport analysis

Description: The Radiation Shielding Information Center (RSIC), sponsored and supported by the Energy Research and Development Administration (ERDA) and the Defense Nuclear Agency (DNA), is a technical institute serving the radiation transport and shielding community. It acquires, selects, stores, retrieves, evaluates, analyzes, synthesizes, and disseminates information on shielding and ionizing radiation transport. The major activities include: (1) operating a computer-based information system and answering inquiries on radiation analysis, (2) collecting, checking out, packaging, and distributing large computer codes, and evaluated and processed data libraries. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results. (auth)
Date: January 1, 1975
Creator: Trubey, D.K.; Maskewitz, B.F. & Roussin, R.W.
Partner: UNT Libraries Government Documents Department

OGRE-P2, A MONTE CARLO PROGRAM FOR COMPUTING GAMMA-RAY LEAKAGE FROM LAMINATED SLABS WITH A DISTRIBUTED SOURCE

Description: A Monte Carlo program OGRE-P2 was written for the IBM-7090 to solve the problem of gamma radiation from a slab of laminated regions composed of various materials. The dose rate is calculated on one side of the slab. (J.R,D.)
Date: August 10, 1962
Creator: Trubey, D.K.; Penny, S.K. & Emmett, M.B.
Partner: UNT Libraries Government Documents Department