Radiation Safety Training and Reference Manual

TABLE OF CONTENTS

1. INTRODUCTION
2. RADIATION FUNDAMENTALS
3. INTERACTION OF RADIATION WITH MATTER
4. ACTIVITY, EXPOSURE, AND DOSE
5. BIOLOGICAL EFFECTS OF IONIZING RADIATION
6. RADIATION DOSIMETRY PROGRAM
7. RADIOACTIVE MATERIAL HANDLING AND LABORATORY SAFETY
8. RADIATION SURVEY METERS
9. RADIOACTIVE WASTE DISPOSAL
10. RADIATION SAFETY FOR X-RAY UNITS
11. ACCELERATOR SAFETY

APPENDICES

1. DOSE CONCEPTS
2. RADIATION RULES OF THUMB
3. SI UNITS AND CONVERSION FACTORS
4. GLOSSARY OF TERMS
5. REFERENCES

INTRODUCTION

RADIATION SAFETY MANUAL

This manual is a companion to the Radiation Safety Manual.

TRAINING AND REFERENCE MANUAL

This Training and Reference Manual (TRM) presents the information necessary for users of radioactive materials and radiation producing machines to properly understand and follow the policies and procedures in the RHSC Rules and Regulations. Some of the topics covered are:

• The nature of radiation and its interaction with matter.
• Definitions of units and terms used to describe radiation and radioactive material.
• Methods of calculating and measuring radiation levels for a variety of sources.
• The biological effects of ionizing radiation
• Aulitional information on some of the policies and procedures in the RSM (e.g. dosimetry, waste disposal, and radionuclide handling).
• Safety precautions for the use of radiation producing machines.

ORIENTATION AND TRAINING

In aulition to receiving the two manuals, each user of radioactive materials or radiation producing machines will attend an orientation with one of the Health Physicists in the Radiation Safety Office. This is usually a one-on-one meeting and the topics covered depend on the experience and knowledge of the new user.

Those without extensive prior experience must complete a quiz on the material presented in the two manuals and the orientation.

Training for each user is also provided in the laboratory by the Radiation Safety Officer (RSO) or an experienced user designated by the RSO. Topics covered during this training include, as appropriate:

• Safe use of laboratory equipment and materials, including protective clothing.
• Experiment procedures and protocols, including operating procedures for radiation producing machines.
• Safe handling, storage, and disposal of radioactive materials.
• Methods to control and measure radiation levels and contamination.
• Proper maintenance of required records.
• Emergency procedures.

Biennial refresher training, usually presented at a meeting of each research group, keeps users up-to-date with the latest regulations and WPI policies.

RADIATION SAFETY OFFICE

The Radiation Safety staff is available for consultation and to answer questions on the safe use of radioactive materials and radiation producing machines. Radiation Safety also will keep RSO staff informed of changes in government regulations or Institute policies.

RADIATION FUNDAMENTALS

INTRODUCTION

For the purposes of this manual, we can use a simplistic model of an atom. The atom can be thought of as a system containing a positively charged nucleus and negatively charged electrons which are in orbit around the nucleus.

The nucleus is the central core of the atom and is composed of two types of particles, protons which are positively charged and neutrons which have a neutral charge. Each of these particles has a mass of approximately one atomic mass unit (AMU). (1 AMU = 1.66E-24 g)

Electrons surround the nucleus in orbitals of various energies. (In simple terms, the farther an electron is from the nucleus, the less energy is required to free it from the atom.) Electrons are very light compared to protons and neutrons. Each electron has a mass of approximately 5.5E-4 AMU.

A nuclide is an atom described by its atomic number (Z) and its mass number (A). The Z number is equal to the charge (number of protons) in the nucleus, which is a characteristic of the element. The A number is equal to the total number of protons and neutrons in the nucleus. Nuclides with the same number of protons but with different numbers of neutrons are called isotopes. For example, deuterium (2,1H) and tritium (3,1H) are isotopes of hydrogen with mass numbers two and three, respectively. There are on the order of 200 stable nuclides and over 1100 unstable (radioactive) nuclides. Radioactive nuclides can generally be described as those which have an excess or deficiency of neutrons in the nucleus.

RADIOACTIVE DECAY

Radioactive nuclides (also called radionuclides) can regain stability by nuclear transformation (radioactive decay) emitting radiation in the process. The radiation emitted can be particulate or electromagnetic or both. The various types of radiation and examples of decay are shown below.

ALPHA (a)

Alpha particles have a mass and charge equal to those of helium nuclei (2 protons + 2 neutrons). Alpha particles are emitted during the decay of some very heavy nuclides (Z > 83).

226,88Ra --> 222,86Rn + 4,2a

BETA (B-, B+)

Beta particles are emitted from the nucleus and have a mass equal to that of electrons. Betas can have either a negative charge or a positive charge. Negatively charged betas are equivalent to electrons and are emitted during the decay of neutron rich nuclides.

14,6C --> 14,7N + 0,-1B + neutrino

Positively charged betas (positrons) are emitted during the decay of proton rich nuclides.

22,11Na --> 22,10Ne + 0,1B + g

GAMMA (g)

Gammas (also called gamma rays) are electromagnetic radiation (photons). Gammas are emitted during energy level transitions in the nucleus. They may also be emitted during other modes of decay.

99m,43Tc --> 99,43Tc + g

ELECTRON CAPTURE

In certain neutron deficient nuclides, the nucleus will capture an orbital electron resulting in conversion of a proton into a neutron. This type of decay also involves gamma emission as well as x-ray emission as other electrons fall into the orbital vacated by the captured electrons.

125,53I + 0,-1e --> 125,52Te + g

FISSION

Fission is the splitting of an atomic nucleus into two smaller nuclei and usually two or three neutrons. This process also releases a large amount of energy in the form of gammas and kinetic energy of the fission fragments and neutrons.

235,92U + 1,0n --> 93,37Rb + 141,55Cs + 2(1,0n) + g

NEUTRONS (n)

For a few radionuclides, a neutron can be emitted during the decay process.

17,7N --> 17,8O* + 0,-1B (*excited state)

17,8O* --> 16,8O + 1,0n

X-RAYS

X-rays are photons emitted during energy level transitions of orbital electrons.

Bremsstrahlung x-rays (braking radiation) are emitted as energetic electrons (betas) are decelerated when passing close to a nucleus. Bremsstrahlung must be considered when using large activities of high energy beta emitters such as P-32 and Sr-90.

CHARACTERISTICS OF RADIOACTIVE DECAY

In aulition to the type of radiation emitted, the decay of a radionuclide can be described by the following characteristics.

HALF-LIFE

The half-life of a radionuclide is the time required for one-half of a collection of atoms of that nuclide to decay. Decay is a random process which follows an exponential curve. The number of radioactive nuclei remaining after time (t) is given by:

N(t) = N(0) x exp(-0.693t/T)

where

N(0) = original number of atoms

N(t) = number remaining at time t

t = decay time

T = half-life

ENERGY

The basic unit used to describe the energy of a radiation particle or photon is the electron volt (eV). An electron volt is equal to the amount of energy gained by an electron passing through a potential difference of one volt. The energy of the radiation emitted is a characteristic of the radionuclide. For example, the energy of the alpha emitted by Cm-238 will always be 6.52 MeV, and the gamma emitted by Ba-135m will always be 268 keV. Many radionuclides have more than one decay route. That is, there may be different possible energies that the radiation may have, but they are discreet possibilities. However, when a beta particle is emitted, the energy is divided between the beta and a neutrino. (A  neutrino is a particle with no charge and infinitesimally small mass.) Consequently, a beta particle may be emitted with an energy varying in a continuous spectrum from zero to a maximum energy (Emax) which is characteristic of the radionuclide. The average energy is generally around forty percent of the maximum.

INTERACTION OF RADIATION WITH MATTER

ENERGY ABSORPTION

The transfer of energy from the emitted particle or photon to an absorbing medium has several mechanisms. These mechanisms result in ionization and excitation of atoms or molecules in the absorber. The transferred energy is eventually dissipated as heat.

Ionization is the removal of an orbital electron from an atom or molecule, creating a positively charged ion. In order to cause an ionization, the radiation must transfer enough energy to the electron to overcome the binding force on the electron. The ejection of an electron from a molecule can cause dissociation of the molecule.

Excitation is the aulition of energy to an orbital electron, thereby transferring the atom or molecule from the ground state to an excited state.

ALPHA PARTICLES

Interactions between the electric field of an alpha and orbital electrons in the absorber cause ionization and excitation events. Because of their double charge and low velocity (due to their large mass), alpha particles lose their energy over a relatively short range. One alpha will cause tens of thousands of ionizations per centimeter in air. The range in air of the most energetic alpha particles commonly encountered is about 10 centimeters (4 inches). In denser materials, the range is much less. Alpha particles are easily stopped by a sheet of paper or the protective (dead) layers of skin.

BETA PARTICLES

Normally, a beta particle loses its energy in a large number of ionization and excitation events. Due to the smaller mass, higher velocity and single charge of the beta particle, the range of a beta is considerably greater than that of an alpha of comparable energy. Since its mass is equal to that of an electron, a large deflection can occur with each interaction, resulting in many path changes in an absorbing medium.

If a beta particle passes close to a nucleus, it decreases in velocity due to interaction with the positive charge of the nucleus, emitting x-rays (Bremsstrahlung). The energy of the Bremsstrahlung x-rays has a continuous spectrum up to a maximum equal to the maximum kinetic energy of the betas. The production of Bremsstrahlung increases with the atomic number of the absorber and the energy of the beta. Therefore, low Z materials are used as beta shields.

A positron will lose its kinetic energy through ionizations and excitations in a similar fashion to a negative beta particle. However, the positron will then combine with an electron. The two particles are annihilated, producing two 511 keV photons called annihilation radiation.

PHOTONS

Gammas and x-rays differ only in their origin. Both are electromagnetic radiation, and differ only from radio waves and visible light in having much shorter wavelengths. They have zero rest mass and travel with the speed of light. They are basically distortions in the electromagnetic field of space, and interact electrically with atoms even though they have no net electrical charge. While alphas and betas have a finite maximum range and can therefore be completely stopped with a sufficient thickness of absorber, photons interact in a probabilistic manner. This means that an individual photon has no definite maximum range. However, the total fraction of photons passing through an absorber decreases exponentially with the thickness of the absorber. There are three mechanisms by which gammas and x-rays lose energy.

The photoelectric effect is one in which the photon imparts all its energy to an orbital electron. The photon simply vanishes, and the absorbing atom becomes ionized as an electron (photoelectron) is ejected. This effect has the highest probability with low energy photons ( < 50 keV) and high Z absorbers.

Compton scattering provides a means for partial absorption of photon energy by interaction with a 'free' (loosely bound) electron. The electron is ejected, and the photon continues on to lose more energy in other interactions. In this mechanism of interaction, the photons in a beam are scattered, so that radiation may appear around corners and in front of shields.

Pair production occurs only when the photon energy exceeds 1.02 MeV. In pair production the photon simply disappears in the electric field of a nucleus, and in its place two electrons, a negatron and a positron, are produced from the energy of the photon. The positron will eventually encounter a free electron in the absorbing medium. The two particles annihilate each other and their mass is converted into energy. Two photons are produced each of 0.511 MeV. The ultimate fate of these two photons is energy loss by Compton scattering or the photoelectric effect.

SECONDARY IONIZATIONS

The electrons from ionizations and pair production will themselves go on to cause more ionization and excitation events in the same way as described for betas.

ACTIVITY, EXPOSURE, AND DOSE

DEFINITIONS

Activity

Activity is the rate of decay (disintegrations/time) of a given amount of radioactive material.

Dose

Dose is a measure of energy deposited by radiation in a material, or of the relative biological damage produced by that amount of energy given the nature of the radiation.

Exposure

Exposure is a measure of the ionizations produced in air by x-ray or gamma radiation. The term exposure (with its 'normal' definition) is sometimes used to mean dose. (e.g. 'He received a radiation exposure to his hand.')

UNITS

ACTIVITY

1 Curie (Ci) = 3.7E10 disintegrations per sec (dps). The Becquerel (Bq) is also coming into use as the International System of Units (SI){XE "International System of Units (SI)"} measure of disintegration rate. 1 Bq = 1 dps, 3.7E10 Bq = 1 Ci, and 1 mCi = 37 MBq.

EXPOSURE

The unit of radiation exposure in air is the roentgen (R). It is defined as that quantity of gamma or x-radiation causing ionization in air equal to 2.58E-4 coulombs per kilogram. Exposure applies only to absorption of gammas and x-rays in air.

DOSE

The Rad is a unit of absorbed dose. One Rad is equal to an absorbed dose of 100 ergs/gram. (1 erg = 6.24E11 eV) The SI unit of absorbed dose is the Gray (Gy). 1 Gy = 1 joule/kilogram = 100 Rad. An exposure of 1 R results in an absorbed dose of 0.87 Rad.

A quality factor (Q) is used to compare the biological damage producing potential of various types of radiation, given equal absorbed doses. The effectiveness of radiation in producing damage is related to the energy loss of the radiation per unit path length. The term used to express this is linear energy transfer (LET). Generally, the greater the LET in tissue, the more effective the radiation is in producing damage. The quality factors for radiations frequently encountered are:

Radiation	Q
Gammas and X-rays	1
Beta particles & electrons	1
Alpha particles & fission fragments	20
Neutrons	10

The rem is a unit of dose equivalent. The dose equivalent in rem is equal to the absorbed dose in Rad multiplied by the quality factor. Dose equivalent determinations for internally deposited radioactive materials also take into account other factors such as the non-uniform distribution of some radionuclides (e.g. I-125 in the thyroid). The SI unit for dose equivalent is the Sievert (Sv). 1 Sv = 100 rem.

CALCULATION OF ACTIVITIES

The half-life of a radionuclide is the time required for one-half of a collection of atoms of that nuclide to decay. This is the same as saying it is the time required for the activity of the sample to be reduced to one-half the original activity. This can be written as:

A(t) = A(0) x exp(-0.693t/T)

where

A(0) = original activity

A(t) = activity at time t

t = decay time

T = half-life

EXAMPLE

P-32 has a half-life of 14.3 days. On January 10, the activity of a P-32 sample was 10 uCi. What will the activity be on February 6? February 6 is 27 days after January 10, so

A(Feb 6) = A(Jan 10) x exp[-0.693(27/14.3)] = 2.7 uCi

A quick estimate could also have been made by noting that 27 days is about two half-lives. So the new activity would be about one-half of one-half (i.e. one-fourth) of the original activity.

CALCULATION OF EXPOSURE RATES

Gamma exposure constants (G) for some radionuclides are shown below. G is the exposure rate in R/hr at 1 cm from a 1 mCi point source.

Nuclide	G
Chromium-51	0.16
Cobalt-57	0.9
Cobalt-60	13.2
Gold-198	2.3
Iodine-125	1.5
Nickel-63	3.1
Radium-226	8.25
Tantalum-182	6.8
Zinc-65	2.7

An empirical rule which may also be used is

6 x Ci x n x E = R/hr @ 1 foot,

where

Ci = source strength in curies.

E = energy of the emitted photons in MeV.

n = fraction of decays resulting in photons with an energy of E.

It should be noted that this formula and the gamma constants are for exposure rates from gammas and x-rays only. Any dose calculations would also have to include the contribution from any particulate radiation that may be emitted.

INVERSE SQUARE LAW

Exposure rate varies inversely with the square of the distance from a point source of radiation. This is often referred to as the inverse square law.

ER2 = ER1 x (D1/D2)^2

where

ER2 = exposure rate at distance 2

ER1 = exposure rate at distance 1

D1 = distance 1

D2 = distance 2

For example, from the table above, the G for Co-60 is 13.2. Therefore, the exposure rate at 1 cm from a 1 mCi source would be 13.2 R/hr. At 30 cm from the same source, the exposure rate would be

(13.2 R/hr)(1/30)^2 = 0.0147 R/hr = 14.7 mR/hr.

BETA DOSE RATES

For a beta emitter point source, the dose rate can be calculated using the empirical equation

300 x Ci = rad/hr @ 1 foot,

where Ci = source strength in curies.

This calculation neglects any shielding provided by the air, which can be significant. For example, the maximum range in air for a beta from S-35 is less than one foot, so the dose rate at one foot is zero for any size S-35 source.

SKIN DOSE

For energies above 0.6 MeV, the dose rate to the skin from a uniform deposition of 1 mCi/cm^2 of a beta emitter on the skin is about 9 rem/hr.

INTERNAL DOSE CALCULATIONS

See Appendix 1 for methods and examples of internal dose calculations.

BIOLOGICAL EFFECTS OF IONIZING RADIATION

RADIATION HAZARDS

The hazards associated with the absorption of radiation in mammalian systems and tissue are related to both the type of radiation and the nature of the absorbing tissue or organ system.

ALPHA

Alpha particles will be stopped by the dead layers of skin, so they are not an external hazard. However, many alpha emitters or their daughters also emit gammas which are penetrating and therefore may present an external hazard. Internally, alphas can be very damaging due to their high linear energy transfer (LET). That is, they deposit all of their energy in a very small area. Based on their chemical properties, alpha emitters can be concentrated in specific tissues or organs.

BETA

Externally, beta particles can deliver a dose to the skin or the tissues of the eye. Many beta emitters also emit gammas. A large activity of a high energy beta emitter can create a significant exposure from Bremsstrahlung x-rays produced in shielding material. Internally, betas can be more damaging, especially when concentrated in specific tissues or organs.

PHOTONS

Externally, the hazard from low energy (< 30 keV) gammas and x-rays is primarily to the skin or the tissues of the eye. Higher energies are more penetrating and therefore a whole body hazard. Internally, gamma emitters can effect not only the tissues or organs in which they are deposited, but also surrounding tissues.

MECHANISMS OF DAMAGE

As discussed earlier, radiation causes atoms and molecules to become ionized or excited. These ionizations and excitations can result in:

• Production of free radicals.
• Breakage of chemical bonds.
• Production of new chemical bonds and cross-linkage between macromolecules.
• Damage to molecules which regulate vital cell processes (e.g. DNA, RNA, proteins).

TISSUE SENSITIVITY

In general, the radiation sensitivity of a tissue varies directly with the rate of proliferation of its cells and inversely with the degree of differentiation.

EFFECTS OF ACUTE HIGH RADIATION DOSES

A whole body radiation dose of greater than 25 to 50 rem received in a short time results in the clinical 'acute radiation syndrome.' This syndrome, which is dose related, can result in disruption of the functions of the bone marrow system (>25 rem), the gastro-intestinal system (>500 rem), and the central nervous system (>2000 rem). An acute dose over 300 rem can be lethal.

EFFECTS OF LOW RADIATION DOSES

There is no disease uniquely associated with low radiation doses.

Immediate effects are not seen below doses of 25 rem. Latent effects may appear years after a dose is received. The effect of greatest concern is the development of some form of cancer.

The National Academy of Sciences Committee on Biological Effects of Ionizing Radiation (BEIR) issued a report in 1990 entitled Health Effects of Exposure to Low Levels of Ionizing Radiation , also known as BEIR V . The following is an excerpt from the Executive Summary of the report:

On the basis of the available evidence, the population-weighted average lifetime risk of death from cancer following an acute dose equivalent to all body organs of 0.1 Sv (0.1 Gy of low-LET radiation) is estimated to be 0.8%, although the lifetime risk varies considerably with age at the time of exposure. For low LET radiation, accumulation of the same dose over weeks or months, however, is expected to reduce the lifetime risk appreciably, possibly by a factor of 2 or more. The Committee's estimated risks for males and females are similar. The risk from exposure during childhood is estimated to be about twice as large as the risk for adults, but such estimates of lifetime risk are still highly uncertain due to the limited follow-up of this age group.

. . . . .

The Committee examined in some detail the sources of uncertainty in its risk estimates and concluded that uncertainties due to chance sampling variation in the available epidemiological data are large and more important than potential biases such as those due to differences between various exposed ethnic groups. Due to sampling variation alone, the 90% confidence limits for the Committee's preferred risk models, of increased cancer mortality due to an acute whole body dose of 0.1 Sv to 100,000 males of all ages range from about 500 to 1200 (mean 760); for 100,000 females of all ages, from about 600 to 1200 (mean 810). This increase in lifetime risk is about 4% of the current baseline risk of death due to cancer in the United States. The Committee also estimated lifetime risks with a number of other plausible linear models which were consistent with the mortality data. The estimated lifetime risks projected by these models were within the range of uncertainty given above. The committee recognizes that its risk estimates become more uncertain when applied to very low doses. Departures from a linear model at low doses, however, could either increase or decrease the risk per unit dose.

WPI's whole body dose limit for planned exposures is 5 Rem/year(50 mSv/yr), with a 1 Rem/year (10 mSv/yr) lifetime average for the maximally exposed worker under normal conditions. If a WPI worker were to receive the maximum allowable planned dose each year for twenty years, the total dose received would be 20 rem (0.2 Sv). According to the BEIR V report, the worker's chance of death from cancer would increase by approximately 0.6%. This is fairly small compared to the normal chance of death from cancer in the U. S. of about 20% .

RADIATION DOSIMETRY PROGRAM

EXTERNAL DOSIMETRY

WPI currently uses thermoluminescent dosimeters (TLDs) supplied and processed by an independent outside company.

TLD Badge

The TLD is a small crystal which absorbs the energy from radiation. When heated, it releases the stored energy in the form of visible light. The TLD badge is used to measure whole body dose and shallow dose. It consists of a TLD crystal and a holder. The holder has several filters which help in determining the type and energy of radiation. The badge will detect gamma and x-rays, high energy beta particles, and in certain special cases, neutrons. It does not register radiation from low energy beta emitters such as H-3, C-14, and S-35, since their betas will not penetrate the plastic covering on the TLD holder.

The badge is usually worn at the collar, chest or belt level to measure the radiation dose received by the trunk of the body. When not in use, the badge should be left in a safe place on campus away from any radiation sources. (Use the TLD badge rack if one is provided.) Be sure the badge is available for the badge exchange which is done quarterly.

TLD RING

The TLD ring is used to measure dose to the hand. They are issued to individuals who may use millicurie amounts of a gamma or high energy beta emitter. The crystal is mounted in a ring which should be worn on the hand which is expected to receive the larger dose. Wear the ring inside your glove with the label facing towards your palm.

PRECAUTIONS

The radiation doses recorded by your dosimeters become part of your occupational radiation dose record. Make sure that this record is valid and accurate by observing the following precautions:

• Always wear your badge when using radioactive materials or radiation producing machines. Wear your ring when using gamma or high energy beta emitters.
• Keep your dosimeters away from radiation sources when not in use. Do not deliberately expose a dosimeter to radiation or wear your badge when receiving medical or dental x-rays.
• Do not tamper with the TLD packet or remove it from the holder.
• Never wear someone else's dosimeter or let someone else wear yours.
• Avoid subjecting the badge to high temperatures or getting it wet.

Notify the Safety Office if your badge or ring has been damaged or lost, or if you have reason to believe that you or your dosimeter has received an accidental high dose.

FEDERAL NOTIFICATION

The dosimeter vendor and WPI are required by law to report to the United States Nuclear Regulatory Commission (USNRC) any personnel dosimeter which shows a dose higher than the occupational dose limits. It is a violation of the NRC and WPI RHSC regulations and the conditions of our Radioactive Material License to deliberately expose a personnel dosimeter to a radiation source (except when being used as intended). The dose recorded by the dosimeter will become part of the dose record of the individual to whom it was issued unless it can be proven that the individual did not actually receive the dose.

INTERNAL DOSIMETRY

WPI does not currently allow the use of radioactive materials in forms that could lead to biological uptake.

RADIOACTIVE MATERIAL HANDLING AND LABORATORY SAFETY

REDUCTION OF DOSE TO PERSONNEL

The following are ways in which radiation doses can be reduced.

TIME

Carefully plan your activities in order to minimize the time spent handling or in the vicinity of radiation sources.

DISTANCE

Increasing the distance from a radiation source by the use of handling devices will reduce the dose received, since exposure rate decreases as 1/r^2, where r is the distance from a point source. For example:

At 10 cm, a 5 mCi I-125 source has an exposure rate of 75 mR/hr. Moving to 30cm would reduce the exposure rate to

(75 mR/hr)(10/30)^2 = 8.3 mR/hr

Note: The 1/r^2 formula (also known as the inverse square law) does not take into account shielding provided by air. This can be significant for particulate radiation. Even the most energetic alpha particles commonly encountered have a range in air of about 4 inches. A beta from the decay of S-35 has a maximum range in air of about 12 inches.

SHIELDING

As gammas and x-rays pass through an absorber their decrease in number (by the processes discussed in chapter 3) is governed by the energy of the radiation, the density of the absorber medium, and the thickness of the absorber. This can be expressed approximately as

I = Io exp(-ux)

where

Io is the intensity of the initial radiation,

I is the radiation intensity after it has passed through the absorber,

u is a factor called the linear absorption coefficient (The value of u depends on the energy of the incident radiation and the density of the absorbing medium.), and

x is the thickness of the absorber.

TVL & HVL

The thickness of an absorber needed to reduce the radiation intensity by a factor of two and by a factor of ten are called the half-value layer (HVL) and the tenth-value layer (TVL), respectively. Approximate lead TVL's, HVL's, and linear attenuation coefficients for some radionuclides are listed below.

Nuclide	Gamma Energy(MeV)	HVL(mm)	TVL(mm)	u(cm^-1)
I-125	0.035	0.05	0.16	150
Am-241	0.060	0.14	0.45	51
Co-57	0.122	2.0	6.7	3.4
Cs-137	0.662	6.5	21	1.1
Na-22	1.28	9.6	32	0.72
Co-60	1.17 & 1.33	12	40	0.58

Example:

At 30 cm, a 10 mCi Co-60 source produces an exposure rate of about 150 mR/hr. How much lead shielding is needed to reduce the rate to 4 mR/hr?

40 mm (one TVL) will reduce the rate to 15 mR/hr. Auling 12 mm (one HVL) will make it 7.5 mR/hr. One more HVL will put the rate at about 4 mR/hr. So the total lead shielding needed is 40 + 12 + 12 = 64 mm.

SHIELDING CONCERNS

When designing shielding there are several points to be kept in mind.

• Persons outside the shadow cast by the shield are not necessarily protected.
• A wall or partition may not be a safe shield for people on the other side.
• Radiation can be "scattered" around corners.

BREMSSTRAHLUNG

The absorption of high energy beta radiation (e.g. P-32 and Sr-90) in high Z materials such as lead and tungsten may result in the production of electromagnetic radiation (Bremsstrahlung) which is more penetrating than the beta radiation that produced it. Low Z materials such as plastics and glass minimize the production of Bremsstrahlung.

HANDLING PRECAUTIONS

Here are some of the radiological characteristics of and special precautions associated with some radionuclides commonly used on campus. In aulition to the specific precautions for each nuclide, the following general precautions should always be followed when applicable to your work.

• Whenever practical, designate specific areas for radioactive material handling and use. Clearly label the area and all containers. Minimize and confine contamination by using absorbent paper and spill trays. Handle potentially volatile materials in certified fume hoods.
• Do not smoke, eat, or drink in rooms where radioactive materials are used. Do not store food or drink in refrigerators, freezers, or cold rooms used for radioactive material storage.
• Use an appropriate instrument to detect radioactive contamination. Regularly monitor the work area. Always monitor yourself, the work area, and equipment for contamination when your experiment or operation is completed. Decontaminate when necessary.
• Use appropriate shielding when handling millicurie or greater amounts of gamma emitters or high energy beta emitters.
• Wear the dosimeters issued to you while using radioactive materials.
• Wash your hands before leaving the lab, using a telephone, or handling food.

P-32 INFORMATION
Radioactive half-life	14.3 days
Decay mechanism	Beta emission
Energy	Emax = 1.709 MeV
Contamination monitoring	Thin window Geiger-Mueller detector
Shielding	1 cm Lucite
Dosimetry	TLD badge, TLD ring, urinalysis

• The dose rate on contact on the side of a 1 mCi delivery vial will be on the order of 1000 mrem/hr. If possible, avoid direct hand contact with vials and sources. When working with 100 uCi or more of P-32, work should be done behind a 1 cm Lucite shield.
• One microcurie of P-32 in direct contact with 1 cm^2 of bare skin gives a dose rate to the skin of about 8 rem/hr. Always protect your skin when handling unsealed materials. Wear gloves, lab coats, and shoes.
• An thin window G-M survey meter should always be available. A survey should be made immediately after use and any 'hot spots' should be decontaminated.
• TLD badges must be worn for all P-32 work. TLD rings should be worn for all P-32 work, and are required when handling 1 millicurie or more.
• Handle and store your radioactive waste carefully. The one gallon polyethylene bottles for liquid waste should be placed in a secondary container (e.g. a bucket or tray) to contain spills or leaks. When more than a millicurie is involved, place 1 cm Lucite in front of the container for shielding. The metal barrels for dry waste provide sufficient shielding but be sure to keep the lid on.

S-35 INFORMATION
Radioactive half-life	87.4 days
Decay mechanism	Beta emission
Energy	Emax = 0.167 MeV
Contamination monitoring	Thin window Geiger-Mueller detector liquid scintillation counter for wipe surveys
Dosimetry	Urinalysis

• Radiolysis of S-35 labeled amino acids may lead to the release of S-35 labeled volatile impurities. Delivery vials should therefore be opened in a fume hood.
• The aulition of stabilizers (buffers) will reduce, but not eliminate, the evolution of S-35 volatiles from tissue culture media. Incubators should be checked for contamination after using S-35 methionine or other volatile compounds.
• S-35 may be difficult to distinguish from C-14. If both nuclides are being used in the same laboratory, establish controls to ensure they are kept separate. If 'unknown' contamination is found, treat it as C-14.

I-125 INFORMATION
Radioactive half-life	59.6 days
Decay mechanism	Electron capture (gamma and x-ray emission)
Energy	27-35 keV
Contamination monitoring	Thin crystal NaI detector, liquid scintillation counter for wipe surveys
Shielding	Thin lead
Dosimetry	TLD badge, TLD ring, thyroid scan

• The dose rate at 1 cm from a 1 mCi point source is about 1.5 rem/hr. The dose rate is inversely related to the square of the distance from the source. Thus while a small amount of I-125 held for a short time can result in a significant dose to the hands, a relatively short separation distance reduces the dose rate to an acceptable level.
• The volatility of iodine requires special handling techniques to minimize radiation doses. Solutions containing iodide ions (such as NaI) should not be made acidic or be frozen. Both lead to formation of volatile elemental iodine. Once bound to a protein, the volatility of the radioiodine is tremendously reduced.
• Always work in a fume hood with a minimum face velocity of at least 125 linear feet per minute when working with NaI. The sash should be below the breathing zone.
• Use shoulder length veterinary gloves with short vinyl gloves on top to minimize skin absorption.
• Avoid opening the septum on delivery vials. It is preferable to remove radioiodine using a hypodermic needle and syringe.

  A radiation survey instrument should be available in the immediate area. A low energy scintillation detector is preferable to a G-M detector. You should do a wipe survey in your work areas after each use.

• TLD badges must be worn for all radioiodine work, and finger rings are required when handling 1 mCi or more of I-125.
• Use lead to shield quantities of 1 mCi or more. 1 mm of lead will essentially absorb all of the radiation emitted from I-125.
• Call the Safety Office, x6727, to schedule a thyroid assay after using 1mCi or more of NaI, or in cases of suspected accidental contamination.
• Until waste is picked up by Radiation Safety, it should be kept in the waste containers supplied by Radiation Safety and stored in a fume hood.

H-3 (TRITIUM) INFORMATION
Radioactive half-life	12.4 years
Decay mechanism	Beta emission
Energy	Emax = 18.6 keV
Contamination monitoring	Liquid scintillation counter for wipe surveys
Dosimetry	Urinalysis

• Because the beta emitted has a very low energy, tritium can not be detected with the usual survey meters found in the lab. Therefore, special care is needed to keep the work area from becoming contaminated. Tritium can be detected by doing a wipe survey and counting the wipes in a liquid scintillation counter.
• Many tritiated compounds readily penetrate gloves and skin. Wearing two pairs of gloves and changing the outer pair every fifteen or twenty minutes will reduce the chances of cross contamination and absorption through the skin.

C-14 INFORMATION
Radioactive half-life	5730 years
Decay mechanism	Beta emission
Energy	Emax = 0.156 MeV
Contamination monitoring	Thin window Geiger-Mueller detector, liquid scintillation counter for wipe surveys
Dosimetry	Urinalysis

• Some C-14 labeled compounds can penetrate gloves and skin. Wearing two pairs of gloves and changing the outer pair every fifteen or twenty minutes will reduce the chances of absorption through the skin.
• C-14 may be difficult to distinguish from S-35. If both nuclides are being used in the same laboratory, establish controls to ensure they are kept separate. If 'unknown' Contamination is found, treat it as C-14.

RADIATION SURVEY METERS

INTRODUCTION

There are several types of portable radiation survey instruments in use on campus. Various types have different qualities and can therefore have very different detection capabilities.

As a user of radioactive materials or radiation producing machines, you are expected to be able to use the survey meters in your laboratory. During your initial training, you will learn how to operate the instruments in your lab. You should know their capabilities and limitations and be able to interpret the meter readings.

GEIGER-MUELLER DETECTOR

The Geiger-Mueller (G-M) counter is the most common radiation detection instrument on campus. In this type of meter, an ionization in the detector results in a large output pulse that causes meter and audio responses. Because of the inherent characteristics of the detector, all initial ionizing events produce the same size output pulse. Therefore, the meter does not differentiate among types or energies of radiation.

Most G-M detectors have a thin mica film 'window' at one end. This window is very fragile. Always use the thin end window for detecting pure beta emitters and low energy photons (e.g. P-32, S-35, C-14, Fe-55, I-125, and x-rays less than 40 keV). The aluminum side wall should be used only for the detection of penetrating x-rays and gamma radiation.

Very low energy beta emitters such as H-3 and Ni-63 are not detectable since their betas do not have enough energy to penetrate the window. They are best detected by using liquid scintillation counting techniques. C-14 and S-35 emit betas energetic enough to pass through the thin window. However, covering the window with plastic wrap or paraffin film will stop most or all of their betas from entering the detector.

The efficiency of a meter for a specific source of radiation is given by the ratio of the meter count rate to the actual disintegration rate of the source (cpm/dpm). Some examples of approximate G-M efficiencies through the end window at 1 inch from a point source are given below:

H-3	not detectable
C-14, S-35	0.2% - 0.8%*
P-32	3% - 8%
I-125	0.01% - 0.03%

* Not detectable if the detector window is covered with paraffin film, plastic wrap, or other material.

Example:

Your G-M counter reads 5000 cpm at one inch from a small spot of P-32 contamination on the bench. What is the total activity of the contamination?

actual disintegration rate = (5000 cpm)/(0.05 cpm/dpm) = 100,000 dpm = 1700 dps = 1700 Bq = 45 nCi

Because of the randomness of radioactive decay, the meter reading at low count rates often fluctuates widely. For this reason, the audio speaker is sometimes a better indicator of small amounts of radioactivity than the meter reading. At higher count rates, the speaker response is often faster than the meter reading. It is better, therefore, to have the speaker on when using a G-M counter.

Very high radiation fields may temporarily overload the detector circuit resulting in a partial or complete loss of meter or audio response. If this happens, remove the meter and yourself from the area and push the reset button or turn the meter off then back on. The meter should resume normal operation. Always turn on a survey meter before entering an area that might have high radiation fields.

SCINTILLATION DETECTOR

Scintillation detectors which incorporate a sodium iodide crystal are used in some laboratories for the detection of low energy gamma emitters such as I-125. Some survey meters allow the use of either a G-M detector or a scintillation detector. The efficiency of a low energy scintillation probe for the detection of I-125 is about 5% at one inch -- over a hundred times better than a G-M probe.

ION CHAMBER

Ionization chambers are suitable for measuring radiation exposure rate or cumulative radiation exposure at high radiation intensities. They are not especially useful at low radiation intensities or for detecting small quantities of radioactive material.

CALIBRATION

Most survey meters have scales that read in milliRoentgen per hour (mR/hr) and/or counts per minute (cpm) or counts per second (cps). After detector efficiency is taken into consideration, the cpm or cps scales give an indication of the quantity of radioactivity. The mR/hr scales give an indication of the radiation exposure rate. There is an important difference in these measurements. Exposure rate measurements are only valid for electromagnetic radiation.

Radiation Safety calibrates all of the portable radiation survey instruments on campus. We use two general types of calibration procedures -- one for meters that are used for detection and measurement of particulate radiation, and another for meters used for detection and measurement of penetrating electromagnetic radiation. The two procedures are explained briefly below so that you will know what to expect.

Survey meters used in biology and chemistry research labs are calibrated for the detection and measurement of particulate radiation. These meters are calibrated using a pulse generator so that the cpm or cps scales read correctly (i.e. one pulse in = one meter count). If the meter reads only in cpm or cps, we may place an aulitional calibration tag on the instrument giving the mR/hr equivalent of the count rate reading for penetrating electromagnetic radiation. If the meter also reads in mR/hr, those readings may not be accurate for the measurement of electromagnetic radiation. We will indicate a correction factor.

Survey meters that are used for radiation exposure measurements are calibrated with a comparable radiation source. The mR/hr scale will read correctly when the detector is exposed to electromagnetic radiation greater than 100 keV.

RADIOACTIVE WASTE DISPOSAL

INTRODUCTION

Due to Federal legislation, out-of-state low-level radioactive waste (LLRW) disposal facilities are no longer accepting Massachusetts generated LLRW. Unfortunately, litigation and other delays have kept Massachusetts from developing its own LLRW disposal facility. In response to these events, the Radiation Health and Safeguards Committee has approved storage of radioactive waste under the supervision of the Radiation Safety Office. The purpose of this storage is to allow time for the decay of short-lived radionuclides and to facilitate the proper disposal of all radioactive waste.

WASTE MINIMIZATION

Since all radioactive waste must be stored on campus until it decays or until it can be shipped to an authorized LLRW disposal facility, it is important that the amount of waste generated be kept to a minimum. Radiation Safety has a limited area to store radioactive waste. Some ways to minimize waste are listed below.

• Design experiments to use as little radioactive material as possible.
• Use proper handling techniques. This will reduce the chance of contamination.
• When practical, use techniques which do not involve radioactive materials. There are many new techniques and products available which can be used in place of radioactive materials.
• Monitor for contamination and dispose of as little as possible. If there is a spot of contamination on a piece of absorbent paper, cut out that spot and dispose of it rather than the whole piece. Don't automatically place your gloves in the radioactive waste. Monitor them. If there is no detectable contamination, throw them in the regular trash.
• Liquid radioactive waste includes the radioactive material and the first rinse of its experimental container. After the first rinse, the container can be washed in the sink.

SEGREGATION BY HALF-LIFE

All radioactive waste must be segregated according to radionuclide half-life. The three categories for segregation are:

• Half-life less than 15 days (P-32)
• Half-life between 15 and 90 days (S-35, Cr-51, I-125)
• Half-life greater than 90 days (H-3, C-14, Ca-45)

Waste containers are marked with the category of waste they are intended for. It is very important that waste is placed in the proper container.

If waste contains two different radionuclides, place it in the container appropriate for the longer half-life.

LABELING OF WASTE

Be sure to include the following items on the outer container:

• Primary laboratory user
• Radionuclide disposed
• Date and estimated activity for each significant disposal.

PROHIBITED ITEMS

All radioactive labels, markings, and tape must be defaced or removed before being put in a regular waste container.

Solid waste can not be picked up by the RSO if it contains any of the following:

• Hazardous material (e.g. lead, toxins)
• Biohazard bags or other hazardous material markings
• Radioactive markings
• Sharps (e.g. needles, razor blades)

Liquid radioactive waste must be readily soluble or dispersible in water. It must not contain any hazardous materials such as solvents or scintillation fluid.

LEAD PIGS/SHIELDING

Lead shipping containers and other lead shielding should not be disposed of as ordinary trash or placed in solid radioactive waste containers. Lead which is boxed and identified will be picked up by the RSO when requested.

GELS

If a gel is very solid at room temperature, it may be disposed of as solid waste. If it is soft or semi-solid at room temperature, use a solubilizer to liquefy it and dispose of it as liquid waste.

DISPOSAL PROCEDURES

Disposal procedures are described in the Radiation Safety Manual.

RADIATION SAFETY FOR X-RAY UNITS

NATURE OF ANALYTICAL X-RAYS

Analytical x-ray machines produce intense beams of ionizing radiation that are used for diffraction and fluorescence studies. The most intense part of a beam is that corresponding to the K emission of the target material and is called characteristic radiation. In aulition to the characteristic radiation, a continuous radiation spectrum of low intensity is produced ranging from a very low energy to the maximum kV-peak setting. This is referred to as 'Bremsstrahlung' or white radiation. Undesirable wavelengths may be filtered out using a monochromator.

X-ray diffraction wavelengths (w) are selected so as to roughly correspond to the inter-atomic distances within the sample, and to minimize fluorescence. Wavelengths commonly used are 1.54 Å (Cu targets), 0.71 Å (Mo targets), 0.56 Å (Ag targets), and 2.3 Å (Cr targets). The relationship between wavelength and x-ray photon energy is determined by the equation

E = hc/w

where

E = energy in ergs (1eV = 1.6E-12 erg)

h = Planck's constant = 6.614E-27 erg/sec

c = velocity of light = 3E10 cm/sec

w = wavelength in cm (1Å = 1E-8 cm)

X-rays emitted from an open, uncollimated port form a cone of about 30 degrees. The x-ray flux can produce a radiation field at one meter on the order of 10,000 R/hr. A collimator reduces the beam size to about 1 millimeter diameter.

X-RAY HAZARDS AND BIOLOGICAL EFFECTS

X-rays produced by diffraction machines are readily absorbed in the first few millimeters of tissue, and therefore do not contribute any dose to the internal organs of the body. However, the lens of the eye can receive a dose from x-rays of this energy. Overexposure of lens tissue can lead to the development of lensopacities and cataracts.

Absorbed doses of a few hundred Rad may produce a reulening of the skin (erythema) which is transitory in nature. Higher doses -- 10,000 Rad and greater -- may produce significant cellular damage resulting in pigment changes and chronic radiation dermatitis. Exposure to erythema doses may not result in immediate skin reulening. The latent period may be from several hours to several days.

(Note: X-rays used for medical diagnosis are about one order of magnitude shorter in wavelength. Diagnostic rays are designed for tissue penetration and are carefully filtered to avoid x-ray damage to the skin caused by the longer, more readily absorbed wavelengths).

SOURCES OF IONIZING RADIATION

The primary beam is not the only source of ionizing radiation. Any high voltage discharge is a potential source of x-rays. Faulty high-voltage vacuum-tube rectifiers may emit x-rays of twice the voltage applied to the x-ray tube. Other sources of ionizing radiation are:

• Secondary emissions and scattering from the sample, shielding material, and fluorescent screens.
• Leakage of primary or scattered x-rays through gaps and cracks in shielding.
• Penetration of the primary beam through or scattering from faulty shutters, beam traps, or collimator couplings.

SAFETY PRECAUTIONS AND NOTES

The shielding, safety equipment and safety procedures prescribed for x-ray diffraction equipment are applicable only for up to 75 kV-peak x-rays. Aulitional or greater precautions are necessary for machines operating at higher voltages.

The Principle Investigator has the basic responsibility for providing a safe working environment by ensuring that equipment is operationally safe and that users understand safety and operating procedures.

The equipment operator is responsible for his own safety and the safety of others when using an analytical x-ray machine.

Prior to removing shielding or working in the sample area, the operator must check both the warning lights and the current (mA) meter on the console. Never trust a warning light unless it is on! Always use a survey meter to check that the shutters are actually closed if current is still being supplied to the tube. It is possible for a shutter to be stuck partially open even when the indicator shows that it is shut. The best way to avoid an accidental exposure is to turn the machine off before working in the sample area.

Never put any part of the body in the primary beam. Exposure of any part of the body to the collimated beam for even a fraction of a second may result in damage to the exposed tissue.

A person not knowledgeable about x-ray equipment should not attempt to make repairs or remedy malfunctions. If you suspect a machine is malfunctioning, turn it off or unplug it. Place a note on the control panel and inform the PI or his designated representative.

Repairs to the high voltage section must not be made unless the primary leads are disconnected from the high voltage transformer and a signed and dated notice is posted near the x-ray ON switch. Turning off a circuit breaker is not sufficient.

Bare feet are not permitted in the laboratory or around electrical equipment. Even slightly moist skin is an excellent electrical conductor and contact with faulty, ungrounded equipment may result in severe injury or death.

Do not attempt to align x-ray cameras without first consulting an experienced person. Alignment procedures require special training and knowledge.

Special care is required when one power supply is connected to more than one x-ray tube.

EYE PROTECTION

The use of safety glasses or prescription lenses is encouraged when working with analytical x-rays. While glasses cannot be depended upon to provide complete protection to the eyes, they can reduce x-ray exposure. Glass provides about 10 times the protection of plastic. Neither, however, will adequately protect the eye from direct exposure to the primary beam.

FLUORESCENT SCREENS

It is unsafe to inspect an x-ray beam with a fluorescent screen without special precautionary measures. Notify the Safety Office before performing a procedure using a fluorescent screen.

TUBE STATUS INDICATORS

There must be a visual indication located on or near the tube head to indicate when x-rays are being produced This is usually an assembly consisting of two red bulbs, wired in parallel and labeled X-RAYS ON. If one of the lights is burned out, the operator should either replace it before leaving the room, or leave a note on the light assembly indicating that the bulb is burned out. An unlit warning bulb does not necessarily mean that x-rays are not being produced. Always check the control panel.

SAFETY DEVICES

Interlock switches are used to prevent inadvertent access to the beam. They should not be bypassed. Interlocks should be checked periodically to insure that they are functioning properly.

Interlocks and other safety devices and warning systems are not foolproof or fail-safe. A safety device should be used as a back-up to minimize the risk of radiation exposure -- never as a substitute for proper procedures and good judgment.

ACCELERATOR SAFETY

Accelerator facility designs, operating procedures, and safety protocols (including requirements for installed safety devices) are reviewed and approved by the Radiation Health and Safeguards Committee. They must also satisfy the requirements of the Commonwealth of Massachusetts.

In research environments such as the Olin accelerator facilities, a major responsibility for safety is placed on the users. Whenever you are in one of the accelerator facilities, you must be aware of and follow the safety protocols of that facility.

Before you start an operation or enter the accelerator or target rooms, be sure to have planned what you are going to do, and understand the safety precautions you will need to take. In particular:

• Know the location of the installed x-ray/gamma and neutron detectors and how to interpret their associated radiation meters adjacent to the control console.
• Portable Geiger-Mueller, ion chamber, and fast neutron detectors are available. Know their capabilities and limitations and how to interpret their readings. If a detector is not working or if you have doubts about its accuracy, report it to a responsible person so that it can be repaired or calibrated as soon as possible.
• Before you enter the accelerator or target room, always check the console to see if terminal voltage is being generated and if the meters from the installed detectors indicate the presence of radiation.
• If you need to enter the accelerator or target room when there is voltage on the accelerator high voltage terminal:
  • notify the operator.
  • wear your TLD badge.
  • take an operating survey meter with you appropriate for the type and level of radiation you could encounter.
  • During accelerator conditioning, hazardous radiation fields may exist near the high voltage terminal. You should avoid being in the vicinity of the tank at any time during the conditioning period.
  • In aulition to radiation hazards, accelerators generally involve high voltage power supplies and various kinds of moving machinery that may constitute serious hazards for electrical shock or injury.
  • Safety devices and installed radiation detectors must be maintained in an operating condition. Inoperative equipment affecting safety must be reported to the Principal Investigator immediately.

The National Council on Radiation Protection and Measurements (NCRP) publishes a series of reports dealing with various aspects of the use of radioactive materials and radiation producing machines, and the design of facilities in which they are used. Their recommendations, while not law, are used by federal and state agencies as the criteria for regulatory licensing and inspections.

APPENDICES

DOSE CONCEPTS

INTRODUCTION

This discussion is provided as an aulitional source of information to those who desire a more in-depth understanding of radiation dose concepts.

Changes to the federal radiation protection regulations took effect in January, 1994. These changes were based on reports and recommendations by the International Commission on Radiological Protection (ICRP), the National Council on Radiation Protection and Measurements (NCRP), and other organizations involved with radiation protection.

TOTAL DOSE CONCEPT

Previously, the radiation doses received from external radiation sources and internally deposited radioactive materials were treated separately. Limits on internal uptake of radioactive materials were based on the dose to a "critical organ" and could not be compared to the 'whole body' dose received from an external source.

The external dose number was and still is related to the risk of stochastic effects (primarily cancer). For a stochastic effect, the higher the dose received, the greater the chance of developing the effect. The new regulations have a mechanism for determining the increased risk of stochastic effects from an intake of radioactive material. The dose calculated is based on a variety of factors such as the biological half-life of the material, the distribution of the material in the body, and the type and energy of the radiation. The result is that both the external dose and the internal dose are related to the risk of stochastic effects and thus can be auled to obtain a total dose.

ORGAN DOSE

For a few radionuclides, the limits on intake are based on nonstochastic effects rather than stochastic effects. For a nonstochastic effect, the higher the dose received, the more severe the effect. However, unlike stochastic effects, there is a threshold dose, i.e. a certain dose, below which the effect will not occur. Limits on the internal intake of radioactive materials are set to keep organ doses well below the thresholds. Even in these cases, however, the aulitional risk of stochastic effects must also be determined.

The dose limit for external exposure of the lens of the eye is also based on prevention of a nonstochastic effect (lens opacities).

DEFINITIONS

Absorbed Dose

The energy imparted by ionizing radiation per unit mass of irradiated material.

Dose Equivalent

The product of the absorbed dose in tissue, quality factor, and all other necessary modifying factors at the location of interest.

Deep-dose Equivalent (ulE)

The dose equivalent at a tissue depth of 1 cm. (Applies to external whole-body exposure.)

Shallow-dose Equivalent

The dose equivalent at a tissue depth of 0.007 cm. (Applies to external exposure of the skin or an extremity.)

Eye Dose Equivalent

The dose equivalent at a tissue depth of 0.3 cm. (Applies to the external exposure of the lens of the eye.)

Committed Dose Equivalent (CDE)

The dose equivalent to organs or tissues of reference that will be received from an intake of radioactive material by an individual during the fifty-year period following the intake.

Weighting Factor

For an organ or tissue, the proportion of the risk of stochastic effects when the whole body is irradiated uniformly.

Committed Effective Dose Equivalent (CEDE)

The sum of the products of the weighting factors applicable to each of the body organs or tissues that are irradiated and the CDE to these organs or tissues.

Total Effective Dose Equivalent (TEDE)

The sum of the deep-dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures). TEDE = ulE + CEDE

Total Organ Dose Equivalent (TODE)

The sum of the ulE and the CDE to an organ or tissue.

Annual Limit on Intake (ALI)

The derived limit for the amount of radioactive material taken into the body of an adult worker by inhalation or ingestion in a year. ALI is the smaller value of intake of a given radionuclide in a year by the reference man that would result in a CEDE of 5 rem or a CDE of 50 rem to any individual organ or tissue.

EXAMPLE DOSE CALCULATIONS

Fortunately, the NRC has already determined the ALIs for all of the radionuclides and listed them in a table. This makes calculating CEDEs and CDEs fairly simple. Table 2 shows the ALIs for several of the radionuclides used at WPI.

EXAMPLE 1

P-32 in most chemical forms has an ALI for ingestion of 600 uCi. This is listed as a stochastic ALI, which means that ingesting 600 uCi of P-32 would result in a CEDE of 5 rem.

If a worker accidentally ingests 10 uCi of P-32, the CEDE would be (10 uCi) (5 rem/600 uCi) = 0.083 rem = 83 mRem.

EXAMPLE 2

I-125 has a nonstochastic ALI for inhalation of 60 uCi. This means that inhaling 60 uCi of I-125 would result in a CDE to the thyroid of 50 rem. The stochastic ALI for inhalation of I-125 is 200 uCi.

If a worker accidentally inhales 3 uCi of I-125, the CDE to the thyroid would be (3 uCi) (50 rem/60 uCi) = 2.5 rem. The CEDE would be (3 uCi) (5 rem/200 uCi) = 0.075 rem.

Suppose this worker also received an external dose from working with a high energy gamma emitter. Evaluation of his TLD badge showed a ulE of 50 mRem. The TEDE would then be 50 mRem + 75 mRem = 125 mRem.

EMBRYO/FETUS DOSE

The dose limit to the embryo/fetus of a declared pregnant woman is 0.5 rem. Efforts must also be made to avoid a dose substantially higher than 0.05 rem in one month. A declared pregnant woman means a woman who has voluntarily informed the Safety Office, in writing, of her pregnancy and the estimated date of conception.

The dose to an embryo/fetus is the sum of the deep-dose equivalent to the declared pregnant woman and the dose from internally deposited radionuclides in the embryo/fetus and in the woman.

DOSE REPORTING

Each worker who is monitored must be advised annually of his or her dose.

DOSE LIMITS

A summary of dose limits set by the revised regulations is shown in Table 1. The WPI Radiation Health and Safeguards Committee has established the general policy that planned radiation doses shall not exceed NRC limits for adult radiation workers.

The dose limit for an individual member of the public is 0.1 rem/year TEDE.

Table 1

Revised Occupational Dose Limits

Dose Category	Adult Occupational Dose Limit
Total Effective Dose	5 rem/year* Equivalent (TEDE)
Total Organ Dose	50 rem/year to any individual
Equivalent (TODE) organ or tissue except the lens of the eye*
Eye Dose Equivalent	15 rem/year*
Shallow Dose Equivalent	50 rem/year*
Embryo/Fetus Dose	0.5 rem for the entire gestation period

*Occupational dose limit for minors is 10% of the adult limit

Table 2

Annual Limit on Intake (ALI) for Radionuclides Commonly Used at WPI

Radionuclide	Form	ALI for ingestion	ALI for inhalation
(uCi)	(uCi)
H-3	gas	8E8
H-3	other	8E4	8E4
C-14	most compounds	2E3	2E3
P-32	most compounds	6E2	9E2
P-33	most compounds	6E3	8E3
S-35	most compounds	8E3 stochastic, 1E4 nonstochastic	2E4 stochastic
Ca-45	all compounds	2E3	8E2
Cr-51	most compounds	4E4	5E4
I-125	all compounds	4E1 nonstochastic, 1E2 stochastic	6E1 nonstochastic, 2E2 stochastic

RADIATION RULES OF THUMB

ALPHA PARTICLES

An alpha energy of at least 7.5 MeV is required to penetrate the protective layer of the skin (0.07mm).

BETA PARTICLES

A beta energy of at least 70 keV is required to penetrate the protective layer of the skin (0.07mm).

The average energy of a beta-spectrum is approximately one-third the maximum energy.

The range of beta particles in air is about 12 ft per MeV. (e.g. The maximum range of P-32 betas is 1.71 MeV x 12 ft/MeV = 20 ft).

The skin dose rate from a uniform thin deposition of 1 uCi/cm2 is about 9 rem/hr for energies above 0.6 MeV.

For a beta emitter point source, the dose rate in rem/hr at one foot is approximately 300 x Ci where Ci is the source strength in curies. This calculation neglects any shielding provided by the air, which can be significant. For example, the maximum range in air for a beta from S-35 is less than one foot, so the dose rate at one foot is zero for any size S-35 source.

GAMMAS AND X-RAYS

For a point source gamma emitter with energies between 0.07 and 2 MeV, the exposure rate in R/hr at 1 foot is approximately 6 x C x E x n, where C is the activity in curies; E is the energy in MeV; and n is the number of gammas per disintegration.

Gammas and x-rays up to 2 MeV will be attenuated by at least a factor of 10 by 2 inches of lead.

SI UNITS AND CONVERSION FACTORS

SI Units for Radioactive Materials

Prepared by:

U.S. Council for Energy Awareness

Committee on Radionuclides and Radiopharmaceuticals

Suite 400

1776 I Street, N.W.

Washington, D.C. 20006-3708

202/293-0770

SI (Systeme International) units are now being used in many countries as the primary measurement system, including measurement of radioactivity, and the system is coming into use in the United States. Many journals (including those published by the American Medical Association) now require the use of SI units, and U.S. regulatory agencies are beginning to use SI units as well as conventional units in regulations. It is the policy of the United States Government that regulations should not impede the transition to SI units.

The U.S. Trade Act of 1988 includes a provision establishing federal policy to designate the metric system as the preferred measurement system for U.S. trade and commerce. It also requires all federal agencies to adopt the metric system for business-related activities by 1992, except where it proves impractical.

USCEA's Committee on Radionuclides and Radiopharmaceuticals is seeking to familiarize users of radioactive materials with SI units and to facilitate their use in the United States. The SI unit for radioactivity is the Becquerel (Bq), and is defined as one nuclear transformation per second. It is a small unit when compared to the curie (Ci), and it is convenient to use multiples of the unit (see listing later in this brochure). It does have the convenience however of relating directly to count rate once corrections have been made for counting efficiency.

Most suppliers of radioactive materials including the National Institute of Standards Technology (NIST-formerly NBS) have been using dual units (curies and Becquerels) in catalogs, product literature and labeling for some time and plan to do so for the foreseeable future. The European Economic Community (EEC) has stated that it will accept only SI units for radioactivity after 1999, and it is anticipated that all suppliers of radioactive products will be using only SI units at that time. In Canada, Atomic Energy Control Board documents produced since 1985 have been in SI units only, and conversion of regulations is in progress.

Other SI radiation measurement units are as follows:

Exposure and Exposure Rate

The roentgen (R) is the traditional unit of measurement for exposure, the charge produced in air by gamma or x-rays. The SI unit of exposure is coulombs per kilogram (C/kg) of air.

1 C/kg = 3876 R

1 R = 2.58E-4 C/kg

No special name has been given to this SI unit (C/kg) and since there is no convenient conversion to other SI units, it is seldom used. Instead, the observed dose rate in air, that is the air kerma rate, is typically being used as the SI measurement to replace exposure rate. An example of the use of air kerma rate is to define the radiation output from a sealed radioactive source in SI units. The SI units usually used to express air kerma rate are grays/second. In traditional units, exposure rate from a sealed source has typically been expressed in roentgens/hour at a distance of 1 meter from the source.

Charge as defined in exposure (charge produced in air by gamma and X-radiation) does not include ionization produced by bremsstrahlung arising from absorption of electrons (beta particles). Apart from this difference, which is significant only with high energy beta particles, exposure is the ionization equivalent of air kerma. For a further discussion of air kerma see ICRU (International Commission on Radiation Units and Measurements) Report 33, 1980.

Absorbed Dose

This is the amount of energy imparted to matter, and the Rad has been the unit of measurement. The SI unit for absorbed dose is the gray (Gy).

1 Gray (Gy) = 100 Rad

1 Rad = 0.01 Gy

One roentgen of X-radiation in the energy range of 0.1-3 MeV produces 0.96 Rad in tissue.

Dose Equivalent

The dose equivalent is the absorbed dose multiplied by modifying factors such as a quality factor (accounts for the biological effect of different types of radiation) and the dose distribution factor. The rem is the unit of measurement that has been used, and the SI unit is the Sievert (Sv).

1 Sv = 100 rem

1 rem = 0.01 Sv

We are giving advance notice of the change to SI Units to allow users time to become familiar with the new units. Do not hesitate to contact your supplier of radioactive materials or USCEA should you have any questions concerning SI units or the implementation of the change.

CONVERSION TABLE FOR RADIOACTIVITY

Curie Units to Becquerel Units
uCi	kBq	mCi	MBq	Ci	GBq
0.1	3.7	0.25	9.25	0.5	18.5
0.75	27.75	1	37	2	74
3	111	5	185	7	259
10	370	20	740	25	925

Curie Units to Becquerel Units
uCi	MBq	mCi	GBq	Ci	TBq
50	1.85	60	2.22	100	3.7
200	7.4	250	9.25	500	18.5
800	29.6	1000	37	2500	92.5

To convert from one unit to another, read across from one column to the other ensuring the units are in the same line of the column headings. For example:

From the first table:

0.1 mCi = 3.7 MBq

0.1 Ci = 3.7 GBq

From the second table:

50 mCi = 1.85 GBq

3.7 MBq = 100 uCi

SI Units

1 Becquerel (Bq) = 1 disintigration/second

1 Becquerel = 2.7027E-11 curie or about 27 picocuries (pCi)

To convert becquerels to curies, divide the Becquerel figure by 37E9 (alternatively multiply the Becquerel figure by 2.7027E-11)

1 curie (Ci) = 3.7E10 disintigrations/second or 37 gigabecquerels (GBq)

To convert curies to becquerels, multiply the curie figure by 37E9

Curie units that are frequently used:

1 Curie (Ci) = 1000 mCi

1 millicurie (mCi) = 1000 uCi

1 microcurie (uCi) = 1000 nCi

1 nanocurie (nCi) = 1000 pCi (picocuries)

Becquerel units that are frequently used:

1 kilobecquerel (kBq) = 1000 Becquerels (Bq)

1 megabecquerel (MBq) = 1000 kBq

1 gigabecquerel (GBq) = 1000 MBq

1 terabecquerel (TBq) = 1000 GBq

1 Ci = 37 GBq

1 mCi = 37 MBq

1 uCi = 37 kBq

1 nCi = 37 Bq

11/14/91

GLOSSARY OF TERMS

ABSORBED DOSE:

The energy imparted by ionizing radiation per unit mass of irradiated material.

ABSORPTION:

The process by which radiation imparts some or all of its energy to any material through which it passes.

ACTIVITY:

The rate of decay (disintigrations/time) of a given amount of radioactive material.

ALARA:

An acronym for As Low As Reasonably Achievable. The principal that radiation doses should be kept as low as reasonably achievable taking into account economic and social factors.

ALPHA PARTICLE (a):

A strongly ionizing particle emitted from the nucleus during radioactive decay which is equivalent to a helium nucleus (2 protons and 2 neutrons).

ANNIHILATION RADIATION:

The two 511 keV photons produced when a positron combines with an electron resulting in the annihilation of the two particles.

ANNUAL LIMIT ON INTAKE (ALI):

The derived limit for the amount of radioactive material taken into the body of an adult worker by inhalation or ingestion in a year. ALI is the smaller value of intake of a given radionuclide in a year by the reference man that would result in a CEDE of 5 rem or a CDE of 50 rem to any individual organ or tissue.

ATOMIC MASS UNIT (AMU):

One-twelfth the mass of a neutral atom of C-12. (1 AMU = 1.66E-24 g)

ATOMIC NUMBER (Z):

The number of protons in the nucleus of an atom.

ATTENUATION:

Process by which a beam of radiation is reduced in intensity when passing through material -- a combination of absorption and scattering processes.

AUTORADIOGRAPH:

Record of radiation from radioactive material in an object, made by placing the object in close proximity to a photographic emulsion.

BACKGROUND RADIATION:

Ionizing radiation arising from sources other than the one directly under consideration. Background radiation due to cosmic rays and the natural radioactivity of materials in the earth and building materials is always present.

BECQUEREL (Bq):

The SI unit of activity equal to one disintegration per second. (1 Bq = 2.7E-11 Ci).

BETA PARTICLE (B):

A charged particle emitted from the nucleus of an atom, having a mass equal to that of the electron, and a single positive or negative charge.

BIOLOGICAL HALF-LIFE:

The time required for the body to eliminate by biological processes one-half of the amount of a substance which has entered it.

BREMSSTRAHLUNG:

X-rays produced by the deceleration of charged particles passing through matter.

CARRIER FREE:

An adjective applied to one or more radionulcides of an element in minute quantity, essentially undiluted with stable isotope carrier.

COMMITTED DOSE EQUIVALENT (CDE):

The dose equivalent to organs or tissues of reference that will be received from an intake of radioactive material by an individual during the fifty-year period following the intake.

COMMITTED EFFECTIVE DOSE EQUIVALENT (CEDE):

The sum of the products of the weighting factors applicable to each of the body organs or tissues that are irradiated and the CDE to these organs or tissues.

COMPTON SCATTERING:

The elastic scattering of a photon by an essentially free electron.

CONTAMINATION:

The deposition of radioactive material in any place where it is not desired, particularly in any place where its presence may be harmful.

COUNT:

The external indication of a device designed to enumerate ionizing events.

CURIE (Ci):

The unit of activity equal to 3.7 x 1010 disintigrations per second.

DEEP-DOSE EQUIVALENT (ulE):

The dose equivalent at a tissue depth of 1 cm from external radiation.

DOSE:

A general term denoting the quantity of radiation or energy absorbed in a specified mass.

DOSE EQUIVALENT:

The product of the absorbed dose in tissue, quality factor, and all other necessary modifying factors at the location of interest.

EFFECTIVE HALF-LIFE:

Time required for a radioactive nuclide in the body to be diminished fifty percent as a result of the combined action of radioactive decay and biological elimination.

EFFICIENCY:

The ratio of the count rate given by a radiation detection instrument and the actual disintegration rate of the material being counted.

ELECTRON CAPTURE:

A mode of radioactive decay involving the capture of an orbital electron by its nucleus resulting in conversion of a proton to a neutron.

ELECTRON VOLT (eV):

A unit of energy equal to the amount of energy gained by an electron passing through a potential difference of 1 volt.

ERYTHEMA:

An abnormal reulening of the skin due to distention of the capillaries with blood.

EXPOSURE:

A measure of the ionizations produced in air by x-ray orgamma radiation. Sometimes used to mean dose.

EYE DOSE EQUIVALENT:

The dose equivalent at a tissue depth of 0.3 cm from external radiation at the eye.

TLD badge:

A packet of photographic film in a holder used for the approximate measurement of radiation dose.

GAMMA:

Electromagnetic radiation (photon) of nuclear origin.

GEIGER-MUELLER (G-M) COUNTER:

A radiation detection and measurement instrument.

GRAY (Gy):

The SI unit of absorbed dose equal to 1 Joule/kilogram.

HALF VALUE LAYER:

The thickness of any specified material necessary to reduce the intensity of an x-ray or gamma ray beam to one-half its original value.

HEALTH PHYSICS:

The science concerned with the recognition, evaluation, and control of health hazards from ionizing radiation.

ION:

Atomic particle, atom, or chemical radical bearing an electrical charge, either negative or positive.

IONIZATION:

The process by which a neutral atom or molecule acquires either a positive or a negative charge.

IONIZATION CHAMBER:

A radiation detection and measurement instrument.

IONIZING RADIATION:

Any electromagnetic or particulate radiation capable of producing ions, directly or indirectly, by interaction with matter.

ISOTOPES:

Nuclides having the same number of protons in the nuclei, and hence having the same atomic number, but differing in the number of neutrons, and therefore in mass number. Almost identical chemical properties exist among isotopes of a particular element.

LABELED COMPOUND:

A compound consisting, in part, of radioactive nuclides for the purpose of following the compound or its fragments through physical, chemical, or biological processes.

LINEAR ENERGY TRANSFER (LET):

Average amount of energy lost per unit track length by the individual particles or photons in radiation passing through an absorbing medium.

MASS NUMBER (A):

The number of protons and neutrons in the nucleus of an atom.

NUCLIDE:

An of atom characterized by its mass number, atomic number, and energy state of its nucleus.

POSITRON:

A particle having a mass equal to that of an electron and a charge equal to that of an electron, but positive.

QUALITY FACTOR (Q):

The LET-dependant modifying factor that is used to derive dose equivalent from absorbed dose.

Rad:

The unit of absorbed dose equal to 100 erg/gram (or 0.01 Joule/kilogram).

RADIATION:

Energy propagated through space or a material medium.

RADIOACTIVE DECAY:

Disintegration of the nucleus of an unstable nuclide by the spontaneous emission of charged particles, neutrons, and/or photons.

RADIOACTIVE HALF-LIFE:

The time required for a radioactive substance to lose fifty percent of its activity by decay.

RADIOACTIVITY:

The property of certain nuclides of spontaneously disintegrating and emitting radiation.

RADIONUCLIDE:

An unstable (radioactive) nuclide.

RADIOTOXICITY:

The potential of a radioactive material to cause damage to living tissue by radiation after introduction into the body.

Rem:

The unit of dose equivalent equal to the absorbed dose in Rad multiplied by any necessary modifying factors.

ROENTGEN (R):

The unit of radiation exposure in air equal to 2.58E-4 coulombs/kilogram.

SCINTILLATION COUNTER:

A radiation detection and measurement instrument in which light flashes produced in a scintillator by ionizing radiation are converted into electrical pulses by a photomultiplier tube.

SHALLOW-DOSE EQUIVALENT:

The dose equivalent at a tissue depth of 0.007 cm from external exposure of the skin or an extremity.

SIEVERT (Sv):

The SI unit of dose equivalent equal to 1 Joule/kilogram.

SPECIFIC ACTIVITY:

Total activity of a given radionuclide per unit mass or volume.

SYSTEME INTERNATIONAL (SI):

A system of units adopted by the 11th General Conference on Weights and Measurements in 1960 and used in most countries of the world.

THERMOLUMINESCENT DOSIMETER (TLD):

A dosimeter made of a crystalline material which is capable of both storing energy from absorption of ionizing radiation and releasing this energy in the form of visible light when heated. The amount of light released can be used as a measure of absorbed dose.

TOTAL EFFECTIVE DOSE EQUIVALENT (TEDE):

The sum of the deep-dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures). TEDE = ulE + CEDE

TOTAL ORGAN DOSE EQUIVALENT (TODE):

The sum of the ulE and the CDE to an organ or tissue.

WEIGHTING FACTOR:

The proportion of the risk of stochastic effects for an organ or tissue when the whole body is irradiated uniformly.

X-RAY:

Electromagnetic radiation (photon) of non-nuclear origin having a wavelength shorter than that of visible light.

REFERENCES

Basic Radiation Biology. D. J. Pizzarello and R. L. Witcofski; Lea & Febiger, 1967.

Basic Radiation Protection Criteria. National Council on Radiation Protection and Measurements (NCRP) Report No. 39; NCRP, 1971.

The Dictionary of Health Physics & Nuclear Sciences Terms. R. J. Borders; RSA Publications, 1991.

The Effects on Populations of Exposure to Low Levels of Ionizing Radiation: 1980. Committee on the Biological Effects of Ionizing Radiations (BEIR); National Academy Press, 1980.

Health Effects of Exposure to Low Levels of Ionizing Radiation. Committee on the Biological Effects of Ionizing Radiations (BEIR); National Academy Press, 1990.

The Health Physics and Radiological Health Handbook, Revised Edition. B. Shleien; Scinta, 1992.

The Health Physics and Radiological Health Handbook. B. Shleien; Nucleon Lectern Associates, 1984.

Influence of Dose and its Distribution in Time on Dose-Response Relationships for Low-LET Radiations. NCRP Report No. 64; NCRP, 1980.

Introduction to Health Physics. H. Cember; Pergamon Press, 1969.

The Physics of Radiology. H. E. Johns, and J. R. Cunningham; Thomas, 1978.

Radiation Biophysics. H. Andrews; Prentice-Hall, 1961.

Radiation Protection Design Guidelines for 0.1 - 100 MeV Particle Accelerator Facilities. NCRP Report No. 51; NCRP, 1977.

Radiation Protection Training Manual. A. Zea; University of Southern California.