Browsing by Author "Wood, NR"

Now showing 1 - 6 of 6
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    DOSE - a FORTRAN program for the calculation of radiation dose from radiopharmaceuticals
    (Australian Atomic Energy Commission, 1974-09) Hetherington, ELR; Wood, NR
    A FORTRAN program DOSE has been written for the calculation of the radiation dose from diagnostic radionuclides. It uses the basic dose equation of the Medical Internal Radiation Dose Committee and published absorbed fraction and decay data. Doses are calculated for a stated initial or cumulated activity in the whole body and those organs in which significant uptake occurs.
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    IMPURE - a FORTRAN program for the analysis of the gamma spectrum of technetium-99m eluted from fission produced molybdenum-99
    (Australian Atomic Energy Commission, 1972-02) Hetherington, ELR; Wood, NR
    The FORTRAN program IMPURE is used to process the gamma spectrum of 99mTc eluted from fission produced 99Mo to determine the levels of 99Mo, 131I, 132I, 103Ru, and 140La impurities. During counting with a Ge(Li) detector and 512 channel analyser, the sample is screened in a 4.0mm lead pot to reduce the 99mTc contributions in comparison with the impurities. The program sums the counts accumulated in appropriate spectrum peaks, and calculates the standard deviation. After correction for decay, lead attenuation, detector efficiency and the interference by overlapping peaks, the activity of each nuclide and its percentage contribution are calculated. The results are displayed in an easily read format.
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    A new approach to the estimation of radiopharmaceutical radiation dose distributions
    (Australian Atomic Energy Commission, 1975-03) Hetherington, ELR; Wood, NR
    For a photon energy of 150 keV, the Monte Carlo technique of photon history simulation was used to obtain estimates of the dose distribution in a human phantom for three activity distributions relevant to diagnostic nuclear medicine. In this preliminary work, the number of photon histories considered was insufficient to produce complete dose contours and the dose distributions are presented in the form of colour-coded diagrams. The distributions obtained illustrate an important deficiency in the MIRD Schema for dose estimation. Although the Schema uses the same mathematical technique for calculating photon doses, the results are obtained as average values for the whole body and for complete organs. It is shown that the actual dose distributions, particularly those for the whole body, may differ significantly from the average value calculated using the MIRD Schema and published absorbed fractions.
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    Production of ytterbium-169 radiography sources in a high flux reactor
    (Australian Atomic Energy Commission, 1974-02) Hetherington, ELR; Wood, NR
    The irradiation of ytterbium-168 in the Hollow Fuel Element facilities of the AAEC research reactor HIFAR was studied to determine the optimum conditions for the production of ytterbium-169 for industrial radiography sources. The thermal neutron activation cross-section of ytterbium-168 is 3,470 barns; in core positions in the reactor this is significantly increased by the contribution of epithermal neutron activation. Flux depression and target burnup are such that the activation formula normally used for radioisotope production calculations is inadequate. Therefore, a modified formula which includes corrections for these effects, combined with iterative computer calculations to correct for effects resulting from the change in target mass during irradiation, was used to calculate ytterbium-169 yields. It was found that for targets containing 10 mg and 20 mg of Yb2O3, there is an optimum ytterbium-168 enrichment of approximately 20 per cent. The activation of such targets can be increased significantly by diluting the Yb2O3 with a low neutron cross-section material such as graphite. The optimum irradiation time in HIFAR in a thermal neutron flux of 1.0 x 10 14 n cm-2s-1 is approximately one operating period of 24 days.
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    SPECT - a FORTRAN program for the analysis of technetium-99m and other gamma spectra in a radioisotope quality control environment
    (Australian Atomic Energy Commission, 1974-06) Hetherington, ELR; Wood, NR
    The FORTRAN program SPECT has been written for use in a radioisotope quality control environment to process gamma ray spectra for the estimation of specific impurities in a radiopharmaceutical product. The analysis is performed using a radionuclide library containing details of the principal radioisotopes, expected impurities and known spectrum artifacts. The program provides a plot of the spectrum on which significant peaks are marked. The activities of identified radioisotopes are calculated from principal emission and decay data contained in the library. The program output lists these activities, together with artifact peaks of known origin and details of unidentified peaks. Comments to assist the user are also provided. Since the program requires < 70K of storage, it receives a high priority in the operating system used with the AAEC's IBM360/50 computer. This leads to a short 'turn around' period for the quality control of short lived radiopharmaceuticals.
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    Technetium-99m generators prepared from fission produced Molybdenum-99: quality control and performance aspects
    (Australian Atomic Energy Commission, 1971-10) Boyd, RE; Hetherington, ELR; Wood, NR
    99mTc generators distributed by the A.A.E.C. are based on 99Mo extracted from irradiated uranium. The production of these generators and the methods employed in their testing are described. The quantity of 99mTc eluted from these generators varied with pH of the eluant with maxima at pH 0.5 - 0.9, 3, 4.7 - 6.7 and > 9.5. Terminal autoclaving reduced the yield by 10 - 20 per cent.