The amount of radiation an individual accumulates will depend on how long the individual stays in the radiation field:
Dose = Dose Rate x Time mrem = mrem/hr x hrTherefore, to limit a persons dose, one can restrict the time spent in the area. How long a person can stay in an area without exceeding a prescribed limit is called the "stay time" and is calculated from the simple relationship:
Stay Time = Limit (mrem) Dose Rate (mrem/hr)Example: How long can a radiation worker stay in a 1.5 rem/hr radiation field if we wish to limit a dose to 100 mrem?
Stay Time = 100 mrem = 0.0667 hr = 4 minutes 1500 mrem/hr
I proportional D^-2
To represent this in a more useful formula:
I1= K = 1 (d12) Therefore I2=K(12) D2 K2 I1 = d1 I2 K2 d1 I1 = d22 I2 d1 or I1d12 = I2d22 I1 I2 source* -------------|---------->| ------------>| | d1 d2And where: I1 = the radiation intensity at distance d1 from the radiation source.
I1 = 500 mrem/hr d1 = 1 foot d2 = 3 feet I2 = I1d12 =(500mR/hr)(1Foot)2 = 500 mR/hr = 55.6 mR/hr d22 (3Foot)2 9
0.33 Rad/hr at 1 meter/Curie (0.33 mrad/hr at 1 meter/mCi) I1d12 = I2d22 I1 = ? I2 = 0.33mrad/hr/mCi x 100mCi = 33rad/hr d2 = 1 meter d1 = 1 foot = 0.3048 meter I1 = I2d22 = (33mrad/hr)(1m)2=355 mrad/hr d12 (0.3048m)2Example 2: If the exposure rate from Cs-137 at one meter is 250 mrad/hr, how many Curies are present?
( = 0.33) 0.25 rad/hr/meter = 0.76 Curies 0.33 rad/hr/meter/Curie
R/hr at 1 foot = 6CEn
Where: C is the number of Curies of gamma emitter E is the gamma ray energy in MeV n is the number of gammas per disintegrationThis expression holds only for gamma emitters with energies ranging from 0.07 MeV to 4 MeV. Example: What would the exposure rate be one foot from 100 mCi of I-131?
I-131: 1 = 0.363 MeV, 81.2 % /d I-131: 2 = 0.636 MeV, 7.3 % /d rad/hr at 1 foot = 6(.1Ci)[(.364 x .812)+(.636 x .073)] = 0.21 or 210 mrad/hr at one foot
I = Ioeµx Where: I is the intensity outside of a shield of thickness x Io is the unshielded intensity µ is the linear attenuation coefficient x is the thickness of shielding materialThe linear attenuation coefficient is the sum of the probabilities of interaction per unit path length by each of the three scattering and absorption processes photoelectric effect, compton effect, and pair production. Note that µ has dimensions of inverse length. The reciprocal of µ is defined as the mean free path which is the average distance the photon travels in an absorber before an interaction takes place. Because linear attenuation coefficients are proportional to the absorber density, which usually does not have a unique value but depends somewhat on the physical state of the material, it is customary to use "mass attenuation coefficients" which removes density dependence:
Mass attenuation coefficient µm = µ = density (gm/cm3)For a given photon energy, µm does not change with the physical state of a given absorber. For example, it is the same for water whether present in liquid or vapor form. If the absorber thickness is in cm, then µm will have units of:
cm-1 which = cm2/gm gm/cm3
Values of the mass attenuation coefficient for lead are given in Appendix IV.
Example: The intensity of an unshielded Cs-137 source is 1 rad/hr. If the source is put into a lead shield two inches thick, what would be the intensity on the outside of the shield? Density of lead = 11.35 gm/cm3
I = Ioe-µX Io = 1 rad/hr µ = µm x = (0.114 cm2gm)(11.35 gm/cm3) = 1.29 cm-1 X = 2 inches x 2.54 cm/inch = 5.08 cm I= (1R/hr) x e[(1.28 cm-1)(5.08cm)] = 0.0014 R/hr = 1.4 mR/hr
I = 0.5 = e-µX Io x1/2 = 0.693 = HVL µ
Half value layers (for lead) are given for selected radioisotopes in Appendix IV.
Example: How much lead shielding must be used to reduce the exposure rate from an I-131 source from 32 mrad/hr to 2 mrad/hr? HVL of lead for I-131 is 0.178 cm.<.
32 mrad/hr = 16 = 4 HVL (2 x 2 x 2 x 2 = 16) 2 mrad/hr 4 x 0.178 cm = 0.71 cm
The advantage of a pocket dosimeter is that it can provide an on-the-spot result of an individuals exposure to radiation. However, pocket dosimeters are susceptible to erroneous readings when exposed to excessive moisture, dust, or physical abuse. In each case, the dosimeter will read high. For this reason, two dosimeters are usually worn for periods of one day or less. The lower reading dosimeter is considered to be the more accurate. Another disadvantage is the dosimeter's limited exposure range. If the dosimeter is exposed to radiation beyond its range, then the total exposure received cannot be determined.
The badge holder contains filters that allow different radiation types (beta, X, gamma, neutron) and energies to be distinguished on the film. An "open" window (i.e., no filter) allows all radiations of sufficient energy to pass and expose the film. A plastic filter absorbs most low energy beta radiation. Other filters such as copper or lead absorb most high energy beta radiation and all but high energy gamma radiation. Fast neutrons interact with a cadmium filter to produce film blackening. Slow neutrons interact with the nitrogen atoms in the film's gelatin layer and the resulting proton tracks are counted.
Advantages of film badges are:
Caution - Radiation Area: In areas where the level of radiation could cause a major portion of an individual's body to receive an exposure from external radiation that exceeds 5 mrem/hr at 30 cm from source or container. "Radiation Area" postings should be at the point of entrance to the area.
Caution - High Radiation Area or Danger - High Radiation Area: In areas where the level of radiation could cause a major portion of an individual's body to receive an exposure from external radiation that exceeds 100 mrem/hr at 30 cm from source or container. "High Radiation Area" postings should be at the point of entrance to the area.
Grave Danger - Very High Radiation Area: In areas where the level of radiation could result in an individual receiving an absorbed dose in excess of 500 rads/hr at 1 meter from a radiation source or from any surface that the radiation penetrates.
In addition, individuals posting radiation warning signs should provide information on the sign to aid others in minimizing their exposure. Information may include:
DOT Warning Labels for Radioactive Materials Packages
The purpose of these labels is too alert individuals handling packages that special handling may be required. When the background color of the label is all white (Radioactive White-I), the external radiation level from the package is minimal and no special handling is necessary. If however, the background of the upper half of the label is yellow (Radioactive Yellow II or III), a radiation level may exist at the outside of the package, and precautions should be taken to minimize radiation exposures when handling the package. The radiation level in mrem/hr at one meter from the external surface of the package is known as the transport index, and is written in the space provided on the warning label. Furthermore, if the package bears a Radioactive Yellow III, the rail or highway vehicle in which it is carried must be placarded. The table below defines the label criteria for radioactive materials packages:
Dose Rate Limits Accessible Surface External Surface Label of Package of Package Radioactive-White I 0.5 mrem/hr 0 Radioactive-Yellow II 50 mrem/hr 1 mrem/hr Radioactive-Yellow III 200 mrem/hr 10 mrem/hr (requires vehicle placarding)
The cautionary signs and warning labels described in these sections must be removed or defaced when they are no longer needed.
The ALIs in this table are the annual intakes of a given radionuclide by "Reference Man" which would result in either (1) a committed effective dose equivalent of 5 rems (stochastic ALI) or (2) a committed dose equivalent of 50 rems to an organ or tissue (non-stochastic ALI]. The stochastic ALIs are derived to result in a risk, due to irradiation of organs and tissues, comparable to the risk associated with deep dose equivalent to the whole body of 5 rems. The derivation includes multiplying the committed dose equivalent to an organ or tissue by a weighting factor, Wt. This weighting factor is the proportion of the risk of stochastic effects resulting from irradiation of the organ or tissue, t, to the total risk of stochastic effects when the whole body is irradiated uniformly. The values of Wt are listed under the definition of weighting factor indicated below. The non-stochastic ALIs were derived to avoid non-stochastic effects, such as prompt damage to tissue or reduction in organ function.
A value of Wt=0.06 is applicable to each of the five organs or tissues in the "remainder" category receiving the highest dose equivalents, and the dose equivalents of all other remaining tissues may be disregarded. The following parts of the GI tract-stomach, small intestine, upper large intestine, and lower large intestine-are to be treated as four separate organs.
The derived air concentration (DAC) values are derived limits intended to control chronic occupational exposures. The relationship between the DAC and the ALI is given by: DAC = ALI(in µCi)/(2000 hours per working year X 60 minutes/hour X 2E4 ml per minute) = (ALI/2.4E9) µCi/ml, where 2E4 ml is the volume of air breathed per minute at work by "Reference Man" under working conditions of "light work." The DAC values relate to one of two modes of exposure: either external submersion or the internal committed dose equivalents resulting from inhalation of radioactive materials. Derived air concentrations based upon submersion are for immersion in a semi-infinite cloud of uniform concentration and apply to each radionuclide separately.
The ALI and DAC values relate to exposure to the single radionuclide named, but also include contributions from the ingrowth of any daughter radionuclide produced in the body by the decay of the parent. However, intakes that include both the parent and daughter radionuclides should be treated by the general method appropriate for mixtures. The value of ALI and DAC do not apply directly when the individual both ingests and inhales a radionuclide, when the individual is exposed to a mixture of radionuclides by either inhalation or ingestion or both, or when the individual is exposed to both internal and external radiation. When an individual is exposed to radioactive materials which fall under several of the translocation classifications (i.e., Class D, Class W, or Class Y) of the same radionuclide, the exposure may be evaluated as if it were a mixture of different radionuclides.
It should be noted that the classification of a compound as Class D, W, or Y is based on the chemical form of the compound and does not take into account the radiological half-life of different radioisotopes. For this reason, values are given for Close D, W, and Y compounds, even for very short-lived radionuclides.Table 2 Air
The columns in Table 2 captioned "Effluents", "Air," and "Water," are applicable to the assessment and control of dose to the general public. The concentration values given in Columns 1 and 2 of Table 2 are equivalent to the radionuclide concentrations which, if inhaled or ingested continuously over the course of a year, would produce a total effective dose equivalent of 0.05 rem (50 millirem or 0.5 millisieverts).
Consideration of non-stochastic limits has not been included in deriving the air and water effluent concentration limits because non-stochastic effects are presumed not to occur at the dose levels established for individual members of the public. For radionuclides, where the non-stochastic limit is governing in deriving the occupational DAC, the stochastic ALI is used in deriving the corresponding airborne effluent limit in Table 2. For this reason, the DAC and airborne effluent limits are not always proportional.
The air concentration values listed in Table 2, Column 1, are derived by one of two methods. For those radionuclides for which the stochastic limit is governing, the occupational stochastic inhalation ALI is divided by 2.4E9 ml relating the inhalation ALI to the DAC, as explained above, and then divided by a factor of 300. The factor of 300 includes the following components: a factor of 50 to relate the 5-rem annual occupational dose limit to the 0.1-rem limit for members of the public. A factor of 3 to adjust for the difference in exposure time and the inhalation rate for a worker and that for members of the public: and a factor of 2 to adjust the occupational values (derived for adults) so that they are applicable to other age groups.
For those radionuclides for which submersion (external dose) is limiting, the occupational DAC in Table 1, Column 3, was divided by 219. The factor of 219 is composed of a factor of 50, as described above, and a factor of 4.38 relating occupational exposure for 2,000 hours per year to full-time exposure (8,760 hours per year). Note that an additional factor of 2 for age considerations is not warranted in the submersion case.
Table 2 Water
The water concentrations are derived by taking the most restrictive occupational stochastic oral ingestion ALI and dividing by 7.3E7. The factor of 7.3E7 (ml) includes the following components: the factors of 50 and 2 described above and a factor of 7.3E5 (ml) which is the annual water intake of "Reference Man."
Table 1 Table 2 Table 3 Occup.Values Effluent Releases Conc. To Sewers ------------- --------- ------- Column--> 1 2 3 1 2 Oral Inh. Monthly Inj. --------- Average ALI ALI DAC Air water Conc. µCi/ µCi/ µCi/ µCi/ Radionuclide Compound/ µCi µCi ml ml ml ml 1 Hydrogen-3 Water, DAC includes skin absorption 8E+4 8E+4 2E-5 1E-7 1E-3 1E-2 6 Carbon-14 Monoxide - 2E+6 7E-4 2E-6 - - Dioxide - 2E+5 9E-5 3E-7 - - Compounds 2E+3 2E+3 1E-6 3E-9 3E-5 3E-4 20 Calcium-45 W, all compounds 2E+3 8E+2 4E-7 1E-9 2E-5 2E-4 24 Chromium-51 D, all compounds except those given for W and Y 4E+4 5E+4 2E-5 6E-8 5E-4 5E-3 W, halides and nitrates - 2E+4 1E-5 3E-8 - - Y, oxides and hydroxides - 2E+4 8E-6 3E-8 - - 27 Cobalt-60 W, all compounds except those given for Y 5E+2 2E+2 7E-8 2E-10 3E-6 3E-5 Y, oxides, hydroxides, halides and nitrates 2E+2 3E+1 1E-8 5E-11 - - 29 Copper-64 D, all compounds except 1E+4 3E+4 1E-5 4E-8 2E-4 2E-3 those given for W and Y W, halides and nitrates - 2E+4 1E-5 3E-8 - - and sulfides Y, oxides, hydroxides - 2E+4 9E-6 3E-8 - - 55 Cesium-137 D, all compounds 1E+2 2E+2 6E-8 2E-10 1E-6 1E-5 53 Iodine-125 D, all compounds 4E+1 6E+1 3E-8 - - - Thyr Thyr (1E+2) (2E+2) - 3E-10 2E-6 2E-5 15 Phosphorus-32 D, all compounds except phosphates given for W 6E+2 9E+2 4E-7 1E-9 9E-6 9E-5 W, phosphates of Zn(2+), S(3+), Mg(2+), Fe(3+), Bi(3+) and lanthanides - 4E+2 2E-7 5E-10 - - 15 Phosphorus-33 D, see P-32 6E+3 8E+3 4E-6 1E-8 8E-5 8E-4 W, see P-32 - 3E+3 1E-6 4E-9 - - 16 Sulfur-35 Vapor - 1E+4 6E-6 2E-8 - - D, sulfides and sulfates except those given for W 1E+4 2E+4 7E-6 2E-8 - - LLI Wall (8E+3) - - - 1E+4 1E-3 W, elemental sulfur, 6E+3 - - - - - sulfides of Sr, Ba, Ge, Sn, Pb, As, Sb, Bi, Cu, Ag, Au, Zn, Cd, Hg, W, and Mo. Sulfates of Ca, Sr, Ba, Ra, As, Sb, and Bi - 2E+3 9E-7 3E-9 - -
Organ Dose Weighting Factors Organ or Tissue Wt Gonads 0.25 Breast 0.15 Red bone marrow 0.12 Lung 0.12 Thyroid 0.03 Bone surfaces 0.03 * Remainder 0.30 ** Whole Body 1.00*
For the purposes of weighting the external whole body dose, (for adding it to the internal dose) a single weighting factor, Wt = 1.0 is specified.
Qualtity Factors and Absorbed Dose Equivalencies Quality Absorbed Type of radiation factor dose (Q) equivalent X-, gamma, or beta radiation............ 1 1 Alpha particles, multiple-charged particles, fission fragments and heavy particles of unknown charge..... 20 0.05 Neutrons of unknown energy ............. 10 0.1 High-energy protons..................... 10 0.1 Absorbed dose in rad is equal to 1 rem or the absorbed dose in gray is equal to 1 sievert.If it is more convenient to measure the neutron fluence rate than to determine the neutron dose equivalent rate in rems per hour or sieverts per hour, 1 rem (0.01 Sv) of neutron radiation of unknown energies may be assumed to result from a total fluence of 25 million neutrons per square centimeter incident upon the body. If sufficient information exists to estimate the approximate energy distribution of the neutrons, the fluence rate per unit dose equivalent or the appropriate Q value from table above may be used to convert a measured tissue dose in rads to dose equivalent in rems.
Mean Quality Factors, Q, And Fluence Per Unit Dose Equivalent for Monoenergetic Neutrons Neutron Quality unit dose energy factor equivalent (MeV) (Q) (neutrons/cm2/rem) (thermal) 2.5E-8 2 980E6 1E-7 2 980E6 1E-6 2 810E6 1E-5 2 810E6 1E-4 2 840E6 1E-3 2 980E6 1E-2 2.5 1010E6 1E-1 7.5 170E6 5E-1 11 39E6 1 11 27E6 2.5 9 29E6 5 8 23E6 7 7 24E6 10 6.5 24E6 14 7.5 17E6 20 8 16E6 40 7 14E6 60 5.5 16E6 1E2 4 20E6 2E2 3.5 19E6 3E2 3.5 16E6 4E2 3.5 14E6
One becquerel= 1 disintegration per second. One curie = 3.7E10 disintegrations per second = 3.7E10 becquerels = 2.22E12 disintegrations per minute.
Bioassays are required whenever surveys or calculations indicate that an individual has been exposed to concentrations of radioactive material in excess of established limits or when required by State or Federal regulations.
In most labs, a liquid scintillation counter (or gamma counter) is already optimized for the particular nuclides in use. Liquid scintillation and gamma counters should have a calibrated reference standard readily available. These standards should be counted each time samples are counted to verify efficiencies and machine settings. A control vial consisting of a "clean" wipe and counting solution must be counted with your samples to determine machine background. A set of quenched standards should be used to determine the counter's efficiency for various degree's of quench (see Chapter II - Liquid Scintillation Counting).Report results on your diagram in terms of disintegrations per minute (dpm) per area surveyed according to the following formula:
Sample dpm = Sample Gross cpm - Background cpm Counter Efficiency (counts/disintegration)Any wipe indicating a gross cpm greater than twice background cpm should be cleaned up. Record on the survey diagrams the maximum contamination levels found as well as the final levels. Limits for removable contamination are shown in Section G.
Limits Removable Contamination Limits (dpm/100 cm2) Application Alpha Beta/Gamma Control Point Maximum Control Point Maximum Basic Guide for equipment or surface 250 500 2500 5000 in a controlled area Clean Areas, release of 50 100 500 1000 materials Skin, personal clothing non-detectable non-detectableWhen contamination levels approach the Control Point values, appropriate control measures as described below should be taken. Contamination results greater than the maximum limits in any laboratory area should be reported to the laboratory supervisor and cleaned up right away. Corrective steps should be taken to prevent reoccurrences.
Limits Radiation Limits Application From Fixed From Other Contamination Sources Basic Guide for equipment or surface in a controlled area facilities 1.0 2.5 Clean Areas 0.5 2.0 Skin, personal clothing 0.1 --- Release of materials or 0.2 --- facilities
Stay Time = Limit (100 mrem) Dose Rate (mrem/hr)
Multiple choice questions may have more than one correct response.
Refer to Appendix IV for reference data.
Routine Radiation and Contamination Survey
Example: Rm 1245 Teaching Lab
Radiation Survey ??? (mR/hr) Remarks a. 1 1 ft above floor side of waste container b. 5 On contact with side of waste container c. 125 Unshielded containers in freezer d. 0.1 On contact with freezer door closed Instrument used: Victoreen Ion Chamber Beta Shield off, neglecting air absorption. Background = 0.1 mrem/hr. Removable Contamination Survey Net cpm ÷ Eff = dpm 1. ______ ___ ___ 2. ______ ___ ___ 3. ______ ___ ___ 4. ______ ___ ___ 5. ______ ___ ___ 6. ______ ___ ___ 7. ______ ___ ___ 8. ______ ___ ___ 9. ______ ___ ___ 10. ______ ___ ___Raw Counting Data Sheet Example
Room 1245: Teaching Lab
Counter used: Packard Tricarb Liquid Scintillation Counter
Counting Time: 1 Minute
Sampling Technique: Filter Paper Smear Covering an area of 100 cm2
Survey Red Channel Green Channel External Site Identification Counts Counts Std. Ratio 0 H-3 Standard 154,000 77,000 0.99 0 C-14 Standard 17,000 87,000 0.99 0 P-32 Standard 7,000 70,000 0.99 0 Background 50 25 0.98 1 Absorbent Paper 54 27 0.95 2 Laboratory Note 101 65 0.95 3 Floor 71 237 0.80 4 Pipettor Bulb 350 75 0.98 5 Isotope Storage Inside Freezer 120 60 0.45 6 Freezer Handle 450 4,025 0.80 7 Liquid Waste Cover 150 925 0.90 8 Floor 110 525 0.85 9 Floor 70 225 0.80 10 Hood Apron 47 24 0.75
This graph is to be used only with Problem Set 4. It represents the theoretical quench for a typical liquid scintillation counter. You must generate your own quench curves for the particular counter you are using to determine what efficiencies apply to your data.