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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 hr
Therefore, 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 d2
And where:
I1 = the radiation intensity at distance d1 from the radiation source.
d1 = the shorter distance from the source where the radiation intensity is I1.
I2 = the radiation intensity at distance d2 from the radiation source.
d2 = the longer distance from the source where the radiation intensity is I2.
Therefore, by knowing the intensity at one distance, one can find the intensity at any given distance.
Example: The exposure rate one foot from a source is 500 mrem/hr. What would be the exposure rate three feet from the 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
) for selected radionuclides appear in Appendix IV.
Example 1: What is the radiation exposure rate one foot from a 100 mCi point source of Cs-137? ( = 0.33).
Example 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 disintegration
This 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 material The 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
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:
Disadvantages are:
To eliminate this latter disadvantage, pocket dosimeters can be worn along with film badges. If the pocked dosimeter indicates a possible high exposure, the film badge can be evaluated on an emergency basis, usually within twenty-four hours after the receipt by the vendor.
Advantages of TLDs are:
Disadvantages are:
Caution - Radioactive Materials: In areas or on items where radioactive material is used or stored. Each label shall provide sufficient information to permit individuals handling or using containers or working in the general vicinity to take precautions to avoid or minimize exposure. Such information should include:
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 following diagram is a summary of radionuclide entry, transfer, and exit within the body:
Introduction
For the selected radionuclides, the chemical form which is to be used for selecting the appropriate ALI or DAC value is listed. The ALIs and DACs for inhalation are given for an aerosol with an activity median aerodynamic diameter (AMAD) of 1 µm and for three classes (D,W,Y) of radioactive material, which refer to their retention (approximately days, weeks or years) in the pulmonary region of the lung. This classification applies to a range of clearance half-times of less than 10 days, for W from 10 to 100 days, and for Y greater than 100 days. The class (D,W, or Y) given in the column headed "Class" applies only to the inhalation ALIs and DACs given in Table 1 in columns 2 and 3.
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
Table 2 provides concentration limits for airborne and liquid effluents released to the general environment.
Table 3
Table 3 provides concentration limits for discharges to sanitary sewer systems.
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
*
0.30 results from 0.06 for each of 5 "remainder" organs, excluding the skin and the lens of the eye, that receive the highest doses.
**
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.
Rad is the special unit of absorbed dose. One rad is equal to an absorbed dose of 100 ergs/gram or 0.01 joule/kilogram (0.01 gray).
Rem is the special unit of any of the quantities expressed as dose equivalent. The dose equivalent in rems is equal to the absorbed dose in rads multiplied by the quality factor (1 rem= 0.01 sievert).
Sievert is the SI unit of any of the quantities expressed as dose equivalent. The dose equivalent in sieverts is equal to the absorbed dose in grays multiplied by the quality factor (l Sv=100 rems).
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.
The following general procedures should be followed when dealing with spills of radioactive materials:
Other emergencies such as fire, lost or stolen radioactive material, accidental uptake of radioactive material, radiation injury, etc. require the same basic responses as described above - Containment, Notification, Corrective Action, Monitoring. The proper authorities should be notified at once of such incidents and will act with other authorities to control emergencies of this nature.
Waste should be segregated according to its radiologic half-life. Short lived waste can be stored until decayed and then be disposed of as regular waste, however, all radioactive labels, markings, tapes, etc., must be removed or defaced before disposal.
Non-radioactive waste must not be mixed with the radioactive waste as this adds to the cost of disposal. A simple wipe survey or instrument survey of the item can determine if it still contains any radioactivity. If only a portion of an item (i.e., lab bench soaker) is contaminated, just that portion should be disposed of into radioactive waste.
Care should be exercised when disposing of waste in different physical or chemical forms:
To perform a survey, choose sites in the lab where the radioactive material has been used. Areas or equipment such as benchtops, floors near the work area, waste containers, fraction collectors, water baths, storage areas, hoods, etc. are good places to start. Each site chosen should be labeled on a diagram or floor plan of the lab for later referral in interpreting results.
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-detectable
When 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.
Plan ahead:
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
Recorded Data
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.
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