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The Nuclear Regulatory Commission's Regulatory Guide 8.39, Release of Patients Administered Radioactive Materials, permits licensees to authorize the release of patients treated with 131 I [(T 1/2 ) p = 8.04 d] if the total dose equivalent to another individual from the released individual is not likely to exceed 5 millisieverts (0.5 rem). This paper guides the reader through the derivation of an easy-to-use formula that demonstrates how to calculate the total dose equivalent, using patient-specific parameters that an individual may receive from a patient who has been administered a therapeutic dose of 131 I, as compared to the NRC's Reg. Guide 8.39 (see Appendix A). Patient-specific parameters include the individual thyroidal uptake, the patient's physical dimensions, attenuation of radiation by body tissues, buildup of secondary photons, whole body and thyroid retention functions as well as the distance and time the patient is around others. This paper also demonstrates how to calculate an inpatient's discharge time, so that the patient will not expose other individuals above the desired total dose equivalent set by the Radiation Safety Officer (RSO).

Carbon cartridge standards were prepared to assess the activity of 125 I incident on, and adsorbed in, cartridge samples during air sampling. Each cartridge standard consisted of an 125 I-spiked filter paper at a known depth, ranging from 0 to 19 mm, embedded in approximately 34 g of 20-30 mesh activated carbon contained within a 6.35 cm diameter by 2.22 cm deep metal cartridge with screened openings. The total counting efficiency values range from 17.8 to 20.8% for cartridges counted at 3.2 mm from a thin-crystal NaI(TI) detector. The standards were analyzed using a front/back counting technique, and fitting functions were developed relating the front/back net counts ratio and counting efficiency to the 125 I depth of burial. A method for determining sample activity that accounts for exponential radioiodine loading in cartridge samples is compared to a less complicated technique that assumes all the radioiodine is located at an equivalent depth of burial that is based on the sample front/back net counts ratio. In addition, methods are presented for determining airborne 125 I activity for constant and variable concentrations. Variable concentrations are assumed to occur in a fume hood duct by one or more bulk releases as a result of iodinations that are performed during a given sampling interval. The two methods are shown to have maximum relative deviations ranging from -16 to +16%.

The beta particle mass attenuation coefficient for CaSO4 in a Teflon matrix has been estimated by two methods: an experiment using collimated sources and a collimated gas flow proportional counter and simulations using the EGS4 Monte Carlo code. The attenuation coefficient has been cast in the general form of mu/rho = aE(-b), where E is the maximum beta particle energy and a and b are fitting parameters. The a and b parameters are given as a function of the atomic number of the material.

An environmental monitoring program has been developed for Harvard University, Southborough campus, to access the local environmental concentrations of radionuclides released in incinerator effluents. The campus is host to the University`s low-level radioactive waste management facility, which consists of 6,000 drum capacity decay-storage buildings; a 250 drum capacity decay-storage freezer; and a controlled-air incinerator. Developmental considerations were based on the characteristics and use of the incinerator, which has a capacity of 8 tons per day and is operated at 5% of the time for the volume reduction of Type 0 and Type 4 wastes contaminated with a variety of radionuclides used in biomedical research-some in microsphere form. Monitoring was established for air, leafy vegetation, leaf-litter, and surface soil media. Field sampling was optimized regarding location and time based on the action of atmospheric, terrestrial, and biotic transport mechanisms. Preliminary results indicate transient concentrations of {sup 3}H and {sup 125}I in vegetation directly exposed to the dispersing plume. Measurable particulate depositions have not been observed. 52 refs., 3 figs., 14 tabs.

A simple method using a large-area zinc sulfide detector to determine the total specific alpha activity of thick sources is presented. A previous paper shows how the linear absorption properties of weightless alpha sources can be applied to thick sources placed in direct contact with a varying thickness of window material. A quadratic relationship between the detector response and absorber thickness was derived for sources whose thickness exceeds the range of the alpha particle. The coefficient of the linear term in the quadratic expression is used to calculate the total specific alpha activity of a source in contact with the window of the detector. This relationship is tested by obtaining alpha absorption data from solid sources of known specific alpha activity, fitting the data to the theoretical relationship and comparing the results to the known activities.

The classic problem of alpha absorption is discussed in terms of the quantitative determination of the activity of weightless alpha sources and the specific alpha activity of extended sources accounting for absorption in the source medium and the window of a large area ZnS(Ag) scintillation detector. The relationship for the expected counting rate gamma of a monoenergetic source of active area A, specific alpha activity C, and thickness H that exceeds the effective mass density range Rs of the alpha particle in the source medium can be expressed by a quadratic equation in the window thickness x when this source is placed in direct contact with the window of the ZnS(Ag) detector. This expression also gives the expected counting rate of a finite detector of sensitive area A exposed to an infinite homogeneous source medium. Counting rates y obtained for a source separated from a ZnS(Ag) detector by different thicknesses x of window material can be used to estimate parameter values in the quadratic equation, y = a + bx + cx2. The experimental value determined for the coefficient b provides a direct estimation of the specific activity C. This coefficient, which depends on the ratio of the ranges inmore » the source medium and detector window and not the ranges themselves, is essentially independent of the energy of the alpha particle. Although certain experimental precautions must be taken, this method for estimating the specific activity C is essentially an absolute method that does not require the use of standards, special calibrations, or complicated radiochemical procedures. Applications include the quantitative determination of Rn and progeny in air, water, and charcoal, and the measurement of the alpha activity in soil and on air filter samples.« less

The chi-square statistic has many scientific applications, including the evaluation of variance in counting data and the proper functioning of a radiation counting system. This paper provides a discussion of the fundamental aspects of the chi-square test using counting data. Practical applications of the chi-square statistic are discussed, including the estimation of extra-Poisson variance and dead time for a counting system. The consequences of passing or failing the chi-square test are discussed regarding the proper estimator for the population variance of the counting data. Example scenarios are used to provide insight into the applications of the chi-square statistic and the interpretation of values obtained in hypothesis testing.

This paper presents the analytical approaches used and results obtained by the authors in the evaluation of the dose significance of maintaining a microscopic particle of beta-gamma emitting radionuclides on the skin for a finite time interval. The correlation of the response of a portable measuring instrument with radioactivity content of the particle is discussed along with the subsequent evaluation of dose to a limited skin area. Examples are given to show how two independent theoretical approaches were used to estimate the activity of 60Co from a single poor geometry measurement made with a thin window ionization chamber and how the results compared with empirical results from laboratory measurements on a prepared 60Co source. The application of Loevinger's point beta source dose distribution function and other dosimetric information is presented. A summary table and simplified equations to predict beta dose rates to 1 cm2 of tissue at a depth of 7 mg cm-2 for 3.7 X 10(4) Bq point isotropic sources of hypothetical beta emitters ranging in end point energies from 0.2 to 3 MeV and in atomic number from 10 to 90 are also included.

Ordinary sugar can be used as an emergency dosimeter for any person exposed to a nuclear or radiation accident. The number of free radicals in sugar created by radiation does not decrease at room temperature for two months after irradiation and is not changed by thermal treatment for about 18 h at even 55 degrees C. A 600 mg granulated sugar sample can detect about 0.05 Gy (5 rad) as the minimum detectable absorbed dose using electron spin resonance equipment. If sugar is present at the time of a radiation or nuclear accident, the absorbed dose can be evaluated from the sugar and will be useful for both the medical treatment and health effects of the exposed persons.

Exact and approximate kinetics equations relating to the transfer and elimination of radionuclides from the blood and various organs in the body are presented. Although they are limited to simple first order kinetics, instantaneous uniform mixing in the blood and all organ pools, and the behavior of a single metabolic species, they are not limited by the number of transfer organs. In addition, the approach used here may he extended to other less limiting cases (e.g. various metabolites or daughters, slug flow, etc.). These expressions may be used to estimate the instantaneous activity or the total number of disintegrations of a radionuclide in the blood or various organs of reference in the body, hence, also the respective dose rates and doses. The exact kinetics equations may be used to relate measurements of radionuclides in excreta to burdens in the body. The approximate expressions greatly simplify the mathematics and yet provide sufficiently accurate results with a maximum deviation of about 23% from exact mathematical expressions over most time intervals of interest. They do give better results for exposure intervals long compared to the effective mean lives of the radionuclide in the various organs of reference, and they yield the exact steady state expressions. Fortunately, this condition is often satisfied for the relatively long standard exposure interval of 50 years that is applied to occupational exposure. In addition, the steady state expressions may be used along with metabolic data of the distribution of elements in the body, diet and excreta to estimate values of the rate constants used in both the exact and approximate expressions. A comparison of the exact and approximate expressions is given for the uranium metabolic model of Wrenn et al. (Wr78), and a comparison is made with current ICRP models (ICRP68a).