Chapter IV Radiation Protection and Laboratory Techniques

The three basic methods used to reduce the external radiation hazard are time, distance, and shielding. Good radiation protection practices requires optimization of these fundamental techniques.

Time

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
```

Distance

The amount of radiation an individual receives will also depend on how close the person is to the source.

The Inverse Square Law

Point sources of X and gamma radiation follow the inverse square law, which states that the intensity of the radiation decreases in proportion to the inverse of the distance squared:

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
```

Gamma Constants

Gamma radiation levels (in rem/hr) for one Curie of many radionuclides at a distance of one meter have been measured. These "gamma constants" can be used to determine the expected exposure rate at a given distance (using inverse square) for a known quantity of a radionuclide or the number of Curies of a radionuclide from a measured exposure rate. Gamma constants () 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).

```  0.33 Rad/hr at 1 meter/Curie (0.33 mrad/hr at 1 meter/mCi)

I1d12 = I2d22

I1 = ?

d2 = 1 meter
d1 = 1 foot = 0.3048 meter

d12   (0.3048m)2
```
Example 2: If the exposure rate from Cs-137 at one meter is 250 mrad/hr, how many Curies are present?
```                 ( = 0.33)

```

Gamma Exposure Rate Formula

The exposure rate from a gamma point source can be approximated from the following expression:

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(0.1Ci)[(0.364 x 0.812)+(0.636 x 0.073)] = 0.21
or

Shielding

When reducing the time or increasing the distance may not be possible, one can choose shielding material to reduce the external radiation hazard. The proper material to use depends on the type of radiation and its energy.

As discussed in Chapter I, alpha particles are easily shielded. A thin piece of paper or several cm of air is usually sufficient to stop them. Thus, alpha particles present no external radiation hazard. Beta particles are more penetrating than alpha particles. Beta shields are usually made of aluminum, brass, plastic, or other materials of low atomic number to reduce the production of bremsstrahlung radiation. Appendix IV gives the range of beta radiation for selected radionuclides in air and plastic.

Monoenergetic X or gamma rays collimated into a narrow beam are attenuated exponentially through a shield according to the following equation:

```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

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```

Half Value Layer

The half value layer (HVL) is the thickness of the shielding material required to reduce the intensity to one half of its original intensity and can be calculated from:

```                     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)

4 x 0.178 cm = 0.71 cm
```

Personnel Monitoring

External radiation exposure is measured by personnel monitoring devices. Three major types of monitoring devices in use today are the pocked dosimeter, the film badge, and the thermoluminescent dosimeter (TLD). Personnel monitoring is required when it is likely that an individual will be exposed during any calendar year to a dose of 5.0 rems to the whole body (head and trunk, active blood forming organs, gonads); 15 rems to the lens of eye); 50 rems to the extremities (hands, forearms, feet, leg below the knee, ankles); 50 rems to the skin of the whole body; or in any work area where you can receive 100 mrems in any hour at 30 cm from the source or source container. Personnel monitoring provides a permanent, legal record of an individual's occupational exposure to radiation.

Pocket Dosimeters

Pocket dosimeters are small devices (about the size of a marking pen) one can carry in a shirt or lab coat pocket to record exposure to radiation. The dosimeter is set to zero prior to use by a separate battery or AC line operated charging device. When radiation passes through the sensitive volume of the dosimeter, the charge is dissipated in proportion to the amount of radiation received. "Self reading" dosimeters have an optical system to allow the wearer to view the amount of radiation received by looking through the dosimeter like a telescope. "Indirect reading" dosimeters require a separate readout device (that also serves as the dosimeter charger). Several exposure ranges are available, the most common being from 0 to 200 mr.

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.

A typical film badge consists of a film packet and holder. The film packet usually contains two pieces of film, one sensitive to X or gamma radiation in the energy range 15 kev to 3 MeV, and the other sensitive to beta radiation in the energy range from 200 keV to 1 MeV. Radioisotopes with energies below those values referenced above cannot be detected. This is why users of low energy beta emitters such as H-3, C-14, and S-35 are not issued film badges. Exposure to radiation causes the film to turn black upon development, the degree of film blackening is then related to the amount of radiation exposure.

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.

1. They are relatively inexpensive compared to other dosimeter types.
2. They provide a permanent record of an individual's dose (film are kept on file).
3. Films are processed and results reported by a disinterested third party.

1. Films are susceptible to extremes of heat, pressure and moisture.
2. Film processing and receipt of exposure results may take several weeks.

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.

Thermoluminescent Dosimeters (TLDs)

TLDs are small chips (1/8" x 1/8" x 1/32") of lithium fluoride or calcium fluoride. The chips absorb energy from radiation which excites atoms to higher energy levels within the crystal lattice. Heating the chip releases the excitation energy as light, proportional to the amount of radiation received. Chips are placed in badge holders containing filters to distinguish between energy and type.

1. They are small and can be used as extremity monitors.
2. They can be read on-site or through a disinterested third party.
3. They are reusable.

1. Once the chips are analyzed, the exposure information is lost and cannot be verified at a later date.
2. Chips are relatively expensive.
3. Chips are subject to physical damage such as cracking or breaking, etc.

Proper use of Personnel Dosimeters

1. Personnel dosimeters must be worn only by the person to whom it was issued. Any exposure information will then become a part of that person's exposure history record.
2. Dosimeters should be worn on the part of the body where exposure to radiation is likely. Usually, they are worn between the neck and waist. Care must be taken to prevent items like pens, buttons, lab benches, hood aprons, etc. from shielding the badge holder.
3. Store dosimeters along with the "control" dosimeter in a designated area, away from extremes in temperature and radiation. The purpose of the control is to record any non-occupational exposure while the badge is not being worn (i.e., during transit to and from the vendor).

Cautionary Signs

Cautionary signs are required to be posted under certain conditions as described below to warn other individuals in the area that radioactive material or radiation is present:

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:

1. The type of radioactive material;
2. The estimated activity;
3. Assay date;
4. The individual responsible for the material.

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:

2. The exposure rate in mr/hr or rem/hr on contact at the highest spot;
3. The name of the person posting the sign;

4. The date the sign was posted.

Department of Transportation (DOT) Warning Labels

Each package of radioactive material offered for transportation, unless exempted, must be labeled on two sides with one of the three labels shown below:

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-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.

Internal radiation exposure results when the body is contaminated internally with a radioisotope. When radioactive materials enter the body, they are metabolized and distributed to the tissues according to the chemical properties of the elements and compounds in which they are contained. For example, consider a complex molecule which can be equally satisfied with a C-12 (stable) atom or a C-14 (radioactive) atom at its regular carbon position. If the C-14 decays to nitrogen, the molecular structure is affected. If the molecule were DNA, this might be equivalent to gene mutation. Once radioactive material is in the body, little can be done to speed its removal. Thus, internal radiation protection is concerned with preventing or minimizing the deposition of radioactive substances in the body.

Methods of Entry

Radioactive substances, like other toxic agents, may gain entry into the body by four processes:

1. Inhalation - Breathing radioactive aerosols or dust.
2. Ingestion - Drinking contaminated water, or transferring radioactivity to the mouth.
3. Absorption - Entering through intact skin.
4. Injection - Entering through a puncture of the skin with an object bearing radioactive materials.

The following diagram is a summary of radionuclide entry, transfer, and exit within the body:

Guidelines

1. Minimize the amount of radioactive material being handled. Use only as much activity that is needed.
2. Contain the radioactive material. Different physical states (gas, liquid, solid) require different containment techniques. Generally, two levels of containment should be provided. For example, a vial containing a stock solution of radioactive material should be properly capped and placed or transported using a drip tray or other similar device lined with absorbent paper.
3. Follow established laboratory procedures. Proper protective clothing, designated work areas, surface contamination monitoring, etc., are required in all laboratories that use radioactive material (see Part IV 3. Radioisotope Laboratory Techniques).

Internal Exposure Limits

Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage are listed below for selected nuclides.

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.

Table 1

Note that the columns in Table 1 captioned "Ingestion ALI", "Inhalation ALI", and "DAC,"are applicable to occupational exposure to radioactive material.

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. AL. DA. 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
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.

Gray (Gy) is the SI unit of absorbed dose. One gray is equal to an absorbed dose of 1 Joule/kilogram (100 rads).

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```

Activity is expressed in the special unit of curies (Ci) or in the SI unit of becquerels (Bq), or their multiples, or disintegrations (transformations) per unit of time.

```   One becquerel= 1 disintegration per second.
One curie = 3.7E10 disintegrations per second =
3.7E10 becquerels = 2.22E12 disintegrations per minute.
```

Internal Exposure Monitoring

Internally deposited radioactive material can be monitored by measuring the radiation emitted from the body or by measuring the amount of radioactive material contained in the urine or feces. Such monitoring techniques are called "bioassays".

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.

All laboratories authorized to use radioactive materials require special precautions to minimize the external and internal hazard from radiation and radioactive contamination. This section deals with general regulations and techniques that should be followed in the radioisotope lab.

Protective Clothing

1. Lab coats should be worn when manipulating radioactive materials to prevent contamination of street clothing.
2. Disposal plastic gloves should always be worn when using radioactive materials. Personnel with breaks in the skin should use waterproof tape to seal such breaks or not use radioactive material.
3. Care should be exercised not to transfer contamination from the hands or lab coat by reflex actions such as wiping one's brow or scratching an itch.

The Workplace

1. Areas in which radioactive material are used should be covered with plastic backed absorbent paper to contain spills and prevent contamination of the working surface.
2. Drip trays can be used to transfer beakers, test tubes, etc. from one location to another.
3. Change absorbent paper at regular intervals to prevent cross contamination.
4. Label all containers used for radioactive materials work. Keep the work areas neat and clean to prevent accidents as well as making it easier to decontaminate if accidental spills do occur.
5. Secure all radioactive materials from unauthorized removal. Close or lock the lab door when materials must be left unattended. Most refrigerator/freezers can be equipped with locks and make an ideal place for storage.
6. There must be no eating, drinking, smoking, or storage of food in areas in which radioactive materials are used.

1. No mouth pipetting of anything is allowed in a radio-isotope work area. Use a safety pipetting aid for dispensing with standard laboratory pipettes. Eppendorf or other precision pipettes can be used for smaller dispensing. Assume all pipettes and glassware in the work area are contaminated unless labelled otherwise. Contaminated glass pipettes can be placed in a pipette jar for washing. Contaminated Eppendorf tips and disposable pipettes should be placed in radioactive waste containers.
2. Containers used in vortexing, mixing, shaking, or centrifuging operations should be intact and sealed with parafilm or stoppers to prevent spillage.
3. Prepare samples carefully. Heating, drying, distilling, and other operations which could result in volatilization of the material should be performed in a fume hood or glove box.
4. Provide proper shielding to reduce exposure, but not so that you hinder the safe execution of the experiment.
5. Whenever possible, rehearse operation with non-radioactive materials to ensure that the technique will be reasonably free of incidents.
6. Accurate records of radioactive material inventory on hand should be maintained. Record withdrawals from the stock vial on inventory control forms received with the isotopes.

All spills of radioactive material must be cleaned promptly. The responsibility for cleaning up the spill rests on the individual working in the area involved and responsible for the spill. Under no circumstances should an untrained person attempt to examine or clean up a spill of radioactive material.

The following general procedures should be followed when dealing with spills of radioactive materials:

1. Major Spills - Those involving significant radiation hazards to personnel shall be decontaminated under the direct supervision of a qualified individual.
1. Notify all personnel not involved with the spill to vacate the area at once. Have an evacuee notify radiation authorities of the incident.
2. Affected persons should limit their movement to confine the spread of contamination.
3. Contain the spill from further spread.
4. Remove contaminated clothing at once: flush contaminated skin areas thoroughly.
5. Shut off ventilating equipment (if possible) that may transport contaminated air from the area to other parts of the building.
6. Vacate and post or cordon off the contaminated area.
7. Assemble in a nearby safe or clean area and begin monitoring and decontamination of affected persons. Do Not Leave the Area unless adequately decontaminated or with the permission of the responsible radiation authorities.

2. Minor Spills - Those involving little or no radiation hazard to personnel may be decontaminated by laboratory personnel under the direction of the laboratory supervisor.
1. Contain the spill: If the material is a liquid, place an absorbent material such as paper towels, tissues, cloth, etc. over the spill to prevent its spread. If the material spilled is a powdered solid, attempt to contain its spread by covering the area with a protective barrier such as a drip tray, empty beaker, section of kraft paper, etc. If appropriate, close doors and windows, turn off room ventilation fans.
2. Inform others of the spill. Adjust your response to the seriousness of the spill. Instruct those personnel present in the room at the time of the spill to remain in an evacuation area to prevent contamination spread. Evacuated personnel should not eat, drink, or smoke until they are monitored and found free of contamination.
3. Decontaminate the area: Plan ahead. Provide adequate protection and supplies for personnel involved in the cleanup. Begin at the periphery and work toward the center of the contamination. Cover cleaned area with plastic or paper to prevent its recontamination. Place all contaminated items in the proper waste containers. The degree of decontamination arranged by the supervisor should be to the limit specified in Section G.
4. Monitor the area: Using appropriate survey techniques, monitor the progress of the decontamination. Monitor all personnel and materials before releasing them to clean areas.

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 retention areas should be managed with close attention to cleanliness. Housekeeping employees should be instructed not to move or empty radioactive waste containers during the course of their duties.

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:

1. Liquid wastes containing acids should not be mixed with liquids containing bases. Aqueous liquids should not be mixed with organic liquids.
2. Pipettes and other sharp objects should be bundled together in order to prevent them from puncturing the inner liner of the dry waste container.
3. Biological waste should contain sufficient amounts of a preservative and absorbent to prevent decomposition.
4. Radioactive biohazard waste must be treated in the same manner as regular biohazard waste before being placed into radioactive waste containers.

Surveys for radiation and removable radioactive contamination must be performed after each use of radioactive materials. The purpose of this survey is to identify any contamination present and to prevent its spread. Appropriate radiation survey equipment should be available to users for the type of surveys required for the laboratory. Such survey equipment must be properly calibrated for energy and type of radionuclide in use at least every six months.

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.

Survey for Removable Contamination

Contamination not fixed to a surface can be transferred to hands, clothing, notebooks, pens, etc. leading to internal exposure or contamination of clean areas. The most common survey procedure to detect the presence of loose contamination is called a wipe test. In this procedure, a piece of filter paper (usually about 1 inch square or circular) is used to wipe over a surface suspected of being contaminated. The area which the wipe should cover is approximately 100 square centimeters. Depending on the surface being wiped, or the type of material being surveyed for, it may be necessary to wet the wipe material with alcohol or other solvent for better adhesion of contaminated particles to the wipe material. A single wipe can be used to determine contamination over a large surface such as a floor or benchtop. If radioactivity is found a series of wipes covering smaller areas should be performed to localize the contamination. After wipes have been taken, place them in numbered scintillation vials or other carriers to organize and prevent cross contamination of the samples. Wipes should be treated as potentially contaminated until analyzed. prepare the wipes for analysis as you would a regular sample. Analyze the wipe samples, using an appropriate instrument. For beta-emitting isotopes with energies above 100 kev, you can hold the wipe material one cm away from a GM tube and observe the count rate. Beta-emitting isotopes below 200 kev (H-3, C-14, S-35) should be analyzed using a liquid scintillation counting system. Wipe samples of gamma or x-ray emitters should be analyzed using a gamma counting system.

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.

Survey for Fixed Contamination

In laboratories where high energy beta (greater than 100 kev) radioisotopes are used, a survey for fixed contamination should be performed using a GM survey instrument. Where gamma emitting radioisotopes are used, a gamma survey instrument calibrated for the particular radioisotope should be used. Scan the area suspected of being contaminated with the instrument's probe. To prevent possible contamination of the probe, do not let it touch the surface being surveyed. Sites in which the results are twice the background rate could indicate radioactive contamination (Limits for fixed contamination are shown in Section J). Record the results of the survey on the laboratory diagram. Any fixed contamination should be tested for removable activity as described above.

The following removable contamination limits can be used as a guide when working with loose radioactive material. By observing good radioisotope laboratory practices, contamination levels should be kept to less than 10 % of the maximum limits.

```
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.

The primary concern with radioactive contamination is to prevent its spread to other areas, and prevent its uptake into the body. Once contaminated, therefore it is important to stay in the area, alert others in the area of the problem, and request assistance.

Personnel Contamination

1. Contamination of the skin or hair may be washed off with soap and water or commercially available decontamination solutions. Repeat washings with plenty of lather and water. Care must be taken not to break the skin when scrubbing. Stubborn fixed skin contamination on the hands may be "sweated" out by wearing a disposable glove sealed at the wrist.
2. Contaminated clothing should be removed in a suitable area while being properly monitored. Clothing may be discarded as waste, stored for decay, or washed. The area must be surveyed after cleanup.
3. If an uptake (inhalation, ingestion, etc.) is suspected contact authorities for appropriate bioassays.

Equipment or Area Contamination

1. Equipment or work surfaces can be cleaned using soap or commercially available decontaminating solutions. Decontaminating procedures must prevent the spread of contamination and minimize the amount of waste generated.

1. Assemble all cleaning supplies and equipment before starting, remove or cover non-contaminated items.
2. Place contaminated items in designated areas for clean-up, i.e., a fume hood if airborne activity can be generated during clean-up.
3. Wear proper protective clothing.
4. Control access to the designated work area and post appropriate cautionary signs.
5. To minimize fixation of the contaminant, perform the decontamination as soon as practical.
6. Start on the outside and work in to the most contaminated area.
7. Monitor results of clean-up -- Perform surveys regularly.

3. Fixed contamination can be sealed with paint, plastic, or covering to prevent egress.

Radiation exposure rates should be measured using an ion chamber type survey instrument (i.e., Ludlum, Victoreen, etc.). GM counters can be used to measure exposure rates in mr/hr or r/hr as long as the energy of the X or gamma radiation is known and the instrument is calibrated for this particular energy. Before using any instrument, become familiar with its proper operation. Be certain that the instrument has been properly calibrated usually indicated by a calibration sticker on the instrument (see Chapter II).

The following radiation limits should be used as a guide when planning areas for radioactive materials work, material storage, waste storage, etc.

```                       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
```

The goal of each worker should be to maintain his or her exposure to radiation as low as reasonably achievable (ALARA). When working with radioactive material yielding high radiation levels, special precautions may be necessary to limit exposure to the worker and others in the area.

1. Perform a radiation survey of the material to determine what kind of radiation levels exist. If appropriate, a survey for removable contamination should be performed.
2. Determine if the material can be safely handled without creating a hazard to yourself or others in the area. A general rule of thumb is that the guide for radiation workers is 100 mrem per week (5 Rem/yr - 50 weeks/work year). Therefore, base how long you should work in an area without exceeding this or any other specified limit by calculating your stay time:
3. ```      Stay Time =    Limit (100 mrem)
. Dose Rate (mrem/hr)
```

4. Use appropriate handling tools and shielding to reduce overall radiation exposure.
5. When finished, perform an appropriate contamination or radiation survey.

Problem Set 4

Multiple choice questions may have more than one correct response.
Refer to Appendix IV for reference data.

1. Film badge results are reported in units of:
2. mrem/hr
3. rems
4. mCi

2. Film badges cannot detect H-3, C-14, or S-35 because:
1. They are pure beta minus emitters
2. They have beta energies below the sensitivity of the film
3. They have beta energies above the sensitivity of the film
4. The specific ionization of the beta particles is to low

3. The purpose of filters in a film badge holder is to:
1. Help in identifying the type of energy of radiation
2. Determine the amount of radiation exposure
3. Shield the film from radiation exposure

4. Film badges and other personnel dosimeters should be worn:
1. Generally, between the neck and waist
2. On the area of the body where exposure to radiation is most likely
3. On only the person to whom it was issued
4. For extremity monitors, on the inside of protective gloves

5. A radioactive package displaying a DOT "Radioactive Yellow II" warning label with a Transport Index of 0.2 means that:
1. The transport vehicle requires placarding
2. The radiation level at the surface of the package is 0.1 mrem/hr
3. The radiation level at 3 feet from the package is 0.1 mrem/hr

6. If you have a source of radiation which emits high energy beta particles only, what is the most appropriate shielding material to use?
2. A container of plastic.
3. A container of plastic inside a container of lead.
4. A container of lead inside a container of plastic.

7. If you have a source of radiation which emits both high energy beta particles and gamma rays, what is the most appropriate shielding material to use?
2. A container of plastic.
3. A container of plastic inside a container of lead.
4. A container of lead inside a container of plastic.

8. What are the three basic methods for reducing exposure to radiation?
9. Name four factors which are considered in evaluating the potential internal hazards from a radioisotope?

10. If the intensity of a gamma source is 5 mrem/hr at a distance of one meter, what is the intensity one foot away from the source?
11. The intensity of a source measured outside a 2 cm thick lead pig is 2 mrem/hr. If the pig is known to contain Cr-51, what would be the intensity of the source without the pig? (Density of lead = 11.35 gm/cm3
12. A 20 Mci Cs-137 source (calibrated on 9/10/79) is to be used for the calibration of pocket dosimeters. The source is to be place at the center of a board with the dosimeters distributed around a circle of a radius of 22 inches. What dose rate will the chambers receive? Where, how and at what distance should such a facility be posted?
13. What would be the approximate dose to an operator using this calibration facility, assuming the operator spent a total of 30 minutes near or adjacent to the dosimeters?
14. A radiation worker begins work six feet away from a source determined to be 5 mrem/hr at that point. The worker's daily dose limit is 50 mrem. If this person worked in the first area for five hours and then moved to an area three feet from the source, what would be the new exposure rate in this area and how long could the person remain there until the daily limit has been expended?
15. If commercially available lead blocks are 1 inch thick, how many blocks are needed to reduce the unshielding exposure rate from a vial of Co-60 to 1/8 its original value?
16. >

17. You have completed a survey for radiation levels and removable contamination for a radioisotope laboratory that uses H-3 and P-32. From the laboratory floor plan, raw data sheet, and quench curves supplied, answer the following questions:
1. For each of the areas surveyed, what is the level of removable contamination in dpm? (fill in the bottom section labelled "Removable Contamination Survey" on laboratory floor diagram).
2. Which of the contaminated areas are due to H-3? Which are due to P-32?
3. Which of the gross counts can be ignored due to statistical fluctuations? (Hint: = N1/2)
4. Which of the areas exceed the allowable removable contamination limits?
5. What areas require posting with a "Caution-Radiation Area" sign. A "Caution-High Radiation Area" sign?
6. What is the most probable cause for contamination at site #6? What methods of decontamination would you recommend?
7. Which of the users is responsible for the contamination on the floor.
8. What recommendations would you give to the supervisor of this laboratory in order to improve radiation safety?

Example: Rm 1245 Teaching Lab

Recorded Data

```
??? (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.