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FUNDAMENTALS OF RADIATION SAFETY TRAINING MATERIALS Date of Issue: 2002.01.01 Section: Issued By: EH&S - Radiation Safety Revision #: -2- Part: Revision Date: 2006.11.30 Pages: 34 Revised By: -PH- What is Radiation? 2 Atomic Structure and Radioactivity 3 Natural and Man-made Radiation 7 Risk Communication 10 Radiation Biology - Interaction of Radiation and Matter 12 Exposure Reduction and Contamination Control 15 Dosimeters 19 Radiation Detection Instruments 22 References 34 Page 1 of 34
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Page 1: Fundamentals of Radiation Safety Reading Materials

FUNDAMENTALS OF RADIATION SAFETY

TRAINING MATERIALS

Date of Issue: 2002.01.01 Section: Issued By: EH&S - Radiation Safety

Revision #: -2- Part: Revision Date: 2006.11.30

Pages: 34 Revised By: -PH-

What is Radiation? 2

Atomic Structure and Radioactivity 3

Natural and Man-made Radiation 7

Risk Communication 10

Radiation Biology - Interaction of Radiation and Matter 12

Exposure Reduction and Contamination Control 15

Dosimeters 19

Radiation Detection Instruments 22

References 34

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WHAT IS RADIATION? Radiation is energy, in motion, in the form of waves or streams of particles. There are many kinds of radiation. Sound is a familiar form of radiation. Various types of electromagnetic radiation are all around us. Visible light is an example of electromagnetic radiation as are Ultraviolet radiation which produces a suntan, and infrared radiation which is a form of heat energy. High-energy gamma ray and x-rays are one kind of electromagnetic radiation, which can cause changes in atoms creating electrically charged atoms or molecules, which we call ions. Radiation, which has enough energy to produce ions, is called ionizing radiation. All other forms of electromagnetic radiation are called non-ionizing radiation, ultraviolet light, visible, infrared, short-wave, radio, television and higher wavelengths up to the electrical power transmission frequency bands.

Terminology Becquerel - For measuring radioactivity, the unit used is the Becquerel (Bq) to describe the number of atomic disintegrations per second in a radioactive substance.

The term radiation dose is applied to the effects of radiation on any given material. It is necessary to distinguish between exposure, absorbed dose, and dose equivalent.

Exposure - The Roentgen (R) is the old unit of exposure and is defined or measured as the ability of photons to produce ionization in air. The roentgen is no longer used in radiation protection, as it applies only to photons (gamma and x-rays), is related only to their effect in air. The S.I. unit has no name; the units are Coulomb per kilogram. Absorbed Dose - The Gray (Gy) is defined as the amount of energy deposited per unit mass of tissue. The absorbed dose can result from all types of radiation in any material. One gray (Gy) = 1 joule per kg of mass and it is necessary to define the absorbing material when using grays as a unit for absorbed dose. One Gray will produce various levels of tissue damage depending on the type of radiation. Equivalent Dose - The dose equivalent or Sievert (Sv) is the amount of energy absorbed in a given mass multiplied by a quality factor, which is specific for each type of radiation. 1 Sv = 1 Gy x Q Quality factors for various type of ionizing radiation are listed. The higher the quality factor the higher the absorbed energy damage in tissues. Type of radiation Energy range Quality factor Photons all 1 Electrons all 1 Neutrons < 10 keV 5 10 keV to 100 keV 10 >100 keV to 2MeV 20 > 2MeV to 20 MeV 10 Alpha Particles all 20

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ATOMIC STRUCTURE AND RADIOACTIVITY ATOMIC STRUCTURE A simplistic model of an atom will be used to describe the basic properties of radiation and radioactivity. 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 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 (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 much lighter than protons and neutrons. Each electron has a mass of approximately 5.5E-4 AMU. a) Elements and Isotopes Each element has a unique number of protons that determine its chemical properties. The number of protons in an atom is its atomic number and is represented by the symbol “Z”. There is 1 proton in Hydrogen, therefore, Z=1. The atomic number is a subscript, for example 1H, 2He, 3Li, 4Be, 5B, 6C, etc, when the chemical symbol for an element is used with its atomic number. The total number of nuclear particles (protons and neutrons) is the mass number and is represented by the symbol “A”. The mass number is a superscript, e.g. 1H, 2H, 3H, when the chemical symbol for an element is used with its mass number. There are three isotopes of Hydrogen, two isotopes 1H and 2H are stable isotopes, the third 3H is radioactive and is commonly referred to as Tritium since there are three nuclear particles in the nucleus. Each of the elements has various isotopes or nuclides some of which are stable and others which radioactive. There are approximately 1500 known nuclides for the 110 elements. Most are artificially produced and are radioactive, and approximately 270 nuclides are stable. Radioactive nuclides can generally be described as having an excess or deficiency of neutrons in the nucleus. b) Stability and Instability The chart to the right represents all of the known nuclides that exist. Examination of the chart demonstrates that all the known nuclides cluster about a line near the center of the chart. This is list is referred to as the “Line of Stability.” The difference between stability and instability of many atoms depends on the number of protons and neutrons in the nucleus. Unstable atoms attempt to become stable by rearranging the number of protons and neutrons in the nucleus to achieve a more stable ratio. The excess energy from this rearrangement is ejected or emitted fromthe nuclide usually changes atomic numor a proton captures an electron and befor a stable atom ranges from one neutrhigher Z atoms. All nuclides strive for thmaintain a stable state.

the nucleus as kinetic energy. When the rearrangement occurs, ber (for example, a neutron changes into a proton and electron, comes a neutron, etc.) The ratio of neutrons to protons needed on per proton up to one and a half neutrons per proton for the e state at which the least amount of energy is required to

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RADIOACTIVITY AND TYPES OF RADIATION The property of certain nuclides (radionuclides) to spontaneously emit particles, gamma radiation, x-rays following orbital electron capture, or undergo spontaneous fission is known as “radioactive decay.” Each unstable nucleus can disintegrate to form an unstable or stable daughter nucleus. The atom is said to “decay” in this process. Types of Radiation A radioisotope normally emits one or more of the four types of radiation when it decays. a) Alpha Particle Alpha particles are doubly charged helium nuclei (He2+). Sources of alpha radiation are common in nature and include thorium, uranium, and radium. Alpha emitting radioisotopes with one exception (Sm-147) have atomic numbers greater than eighty-two. Alpha emitters such as Pu-239 and Am-241 are mixed with beryllium to create sealed neutron sources. Neutron sources are used to study neutron effects on materials. Alpha particles are highly energetic. Alpha particles have energies of 4 MeV (million electron volts) or greater. A thin absorber such as a sheet of paper or the dead layer of the skin easily stops alpha particles. Such sources do not present a great hazard external to the body. Inside the body, however, alpha emitters are highly significant, because the alpha particle is doubly charged and relatively massive. The alpha particle undergoes many interactions with surrounding atoms, depositing all its energy in a very small volume. Every effort must be taken to prevent alpha emitting sources from entering the body.

b) Beta Particles (negatron or

positrons) There are two decay schemes, which are called beta decay. One scheme involves the emission of a negative beta (negatron) which is an electron, which originates from the nucleus. The second scheme involves the emission of a positron, the "anti-matter" of the negatron. The mass is identical to the beta but unlike the negative beta it has a positive charge. Therefore, a beta particle is an unpaired singly charged electron possessing kinetic energy. The mass of the beta particle is about 1/8000 of the mass of an alpha particle. There are many naturally occurring beta-emitting radioisotopes as well as artificially produced radioisotopes. Beta-emitting radioisotopes are most common in the laboratory and include H-3, C-14, P-32, and S-35. The ability of a beta particle to penetrate matter is a function of its energy. The higher the energy of the particle the greater the ability of the particle to penetrate a material. Low-energy beta particles (less than 0.2 MeV) are easily absorbed in the outer layer of skin; beta-emitting sources external to the body present a somewhat greater threat of penetration when compared to alpha emitters. Sources inside the body within cells or incorporated into biologically active molecules may give significant radiation doses that disable and kill cells.

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c) Gamma and x-ray Gamma and x-ray radiations have no mass, and are uncharged electromagnetic radiation that can pass through matter. Gamma and x-rays create ionization in materials, through which they pass, causing excited electrons to be ejected from the atomic structure. Interactions per unit volume are less than for alpha or beta particles, since gamma and x-rays are without mass or charge, and therefore, are extremely more penetrating than particles. Gamma and x-rays have identical physical characteristics, and differ only in their method of production. Gamma rays result from the emission of energy as an excited nucleus drops from one energy level to a lower energy state; x-rays result from transitions within the electron cloud of an atom or are emitted when a charged particle is quickly decelerated by running into dense material. The penetrating ability of gamma and x-rays makes them biological hazards whether they are external or internal. d) Neutrons Neutron radiation is the least common in the research laboratory. The absorption properties of neutrons are very complex as compared to charged particles and electromagnetic radiation. Neutron absorption is a function of the absorber's atomic weight, neutron to proton ratio, and interaction probabilities with various nuclei. Exposure to neutron radiation is a concern as the biological damage from sources external and internal to the body is greater than equivalent amounts of beta and gamma radiation. DECAY RATE The process of radioactive decay changes a radioisotope into a different isotope. Therefore, over time the amount of a particular radioisotope (parent) in a sample or a stock vial is continually decreasing. Each radioisotope has a unique decay rate. The rate of decay is expressed by the isotopes physical half-life, the time required for an amount of radioisotope to decrease to one-half of its original quantity, see figure to the right. The decrease is exponential, so after each half-life, one-half of the radioactive atoms in the sample will have decayed away. If you start with 1000 atoms at time zero, after one half-life there are 500 atoms left, after two half-lives there are 250 atoms left. After seven half-lives the activity remaining is 1/1024th of the original activity. a) Energy

The energy carried away by the different types of radiation is expressed in units of electron volts (eV). The electron volt is a very small quantity of energy (1.6 x 10-19 Joules). The radiation energy ranges from thousands to millions of electron volts. The units are written as keV, kiloelectron (1000 eV) or MeV, Mega electron volts (1,000,000 eV). The amount of energy involved and the type of radiation emitted determines the penetrability or range of the radiation and consequently the type and the required shield thickness to protect workers form the radiation. All things being equal, the higher the energy, the more penetrating the radiation. Gamma and x-rays are more penetrating than neutrons, alpha, and beta particles.

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b) Activity Activity is the number of radioactive nuclei that change or disintegrate per unit time (for example per

second). The unit of activity used is the Becquerel. One Becquerel represents 1 nuclear disintegration per second. c) Energy Loss Mechanisms

When charged particles (negatrons, positrons, and alpha particles) interact with any absorbing material the energy of the particle is deposited in the absorber (tissue, shielding, walls, etc). There are two mechanisms: which account for all of the energy being deposited in an absorber, collisional energy losses and radiative energy losses call Bremsstrahlung. Collisional Energy Loss The dominant interaction is collisional energy loss. The particle interacts directly with an absorber atom's orbital electrons causing excitation or ionization. Radiative (Bremsstrahlung) A negatron also interacts with the electrical field around the nucleus, which produces a deflection in the path of the charged particle. The particle speed is reduced due to energy loss and there is directional change. The energy loss is emitted as electromagnetic radiation called “bremsstrahlung”. The energy of bremsstrahlung usually falls within the x-ray region of the electromagnetic spectrum hence Brem x-rays. Brem production is proportional to the maximum energy of the charged particle and to the atomic number Z, of the absorber. Brem produced (%) = 3.3 X 10-2 Emax Z Probability of Interaction Photons only interact occasionally and may only release some energy or all of their energy. The chance interaction is often referred to as the probability of an interaction for photons rather than the amount of energy lost over some path length. Photons do not have a range. Gamma and x-rays do not carry an electric charge and therefore can pass through a large number of atoms without any interaction occurring. However, occasionally they will interact with an atomic electron or an atomic nucleus in a number of ways. The following are the most important interactions: Photoelectric Effect: The energy of the x-ray or gamma ray is completely transferred to an atomic electron, which is ejected from its atom. The x-ray or gamma ray no longer exists after the collision.

Compton Effect: The x-ray or gamma ray loses only part of its energy in its interaction with an atomic electron. The electron is ejected from its atom. The x-ray or gamma ray of reduced energy and longer wavelength and the electron fly off in different directions.

Pair Production: Gamma rays with energy greater than about 1.2 MeV may interact with an atomic nucleus to form an electron positron pair. The gamma-ray energy is completely converted into the mass and kinetic energy of the electron and positron. Only a very small amount of energy goes to the nucleus in order to conserve momentum. All of the above reactions result in the production of either energetic electrons, or electrons and positrons. The charged particles produce the ionization, which leads to harmful effects in living tissue.

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Z Dependency An important consideration of photoelectric effect, Compton effect, and pair production is the way in which the probability of the interaction depends on both the photon energy and the atomic number of the absorbing medium. At any given energy a photon has a relative probability of interacting with the shielding materials (lead, Lucite, concrete, etc). The chance of avoiding an interaction decreases exponentially with absorber thickness. In theory only an infinitely thick absorber or shield could stop all incident photons. Sufficient shielding is placed in the radiation path to reduce the dose rate to a reasonable level (ALARA). Two terms commonly used are Half-Value layer (HVL) and Tenth Value Layer (TVL) when discussing the thickness of shielding required to reduce the intensity of the photons. A HVL is a thickness of material which will reduce exposure rate TO one half of the original intensity, and a TVL is a thickness of material which will reduce exposure rate TO one tenth of the original intensity. The Radiation Protection Handbook has individual HVL Values listed for some of the radioisotopes. d) Range of a Charged Particle As a charged particle interacts with an absorber it loses its kinetic energy more or less continuously along its path and then stops. The maximum thickness the beta particles can penetrate in a particular absorber is called the RANGE. For example, the range of P-32 in Lucite is 0.66 cm. The range is very dependent on the maximum energy of the radioisotope. Most of the beta particles emitted from a source are absorbed in distances considerably less than the range. The ranges for a particular radioisotope in selected materials can be found in the Radiation Protection Handbook under individual isotope.

NATURAL AND MAN-MADE BACKGROUND RADIATION NATURAL BACKGROUND RADIATION The discovery of naturally occurring radioactivity and penetrating radiation, and the realization that humans have always lived in an environment with a natural background of radiation, occurred near the turn of the century. The following are three principle sources of natural radiation: 1) Cosmic radiation consists of high-energy particulate radiation produced in the stars and our own sun that bombard the earth and makes the atoms in the upper atmosphere radioactive. To some extent, the lower atmosphere absorbs the radiation produced in the upper atmosphere, so the exposure from cosmic rays depends upon how near one is to outer space and, the outer atmosphere. The cosmic ray intensity doubles for each mile of altitude. Denver, at an altitude of one mile, has a cosmic radiation intensity of about 0.5 mSv/yr compared to an intensity of about 0.25 mSv/yr observed at sea level. Calgary’s cosmic radiation level is approximately 0.40 mSv/yr. Air travel results in an average exposure of approximately 5 uSv per hour of flight. The average Canadian dose equivalent from this “cosmic” radiation is 0.30 mSv per year. 2) Terrestrial (Natural radioactivity) radiation consists of penetrating gamma and x-rays that result from the radioactive decay of naturally occurring, primordial radioactive materials (for example, potassium, uranium, thorium, etc.) found in the earth's crust. Most building materials contain some small amounts of radioactivity. Thus, remaining indoors would only reduce a person's terrestrial dose by about 20%; the remaining 80% of the exposure would emanate from the housing materials. The average Canadian population dose equivalent is 0.35 mSv per year.

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2a) Radon is an inert gaseous element resulting from the decay of uranium. Since radon is a gas, it readily escapes from soil, especially if the soil has been disturbed in any way. Radon-222 has a half-life of 38 days which produces several radioactive elements collectively referred to as “radon daughters” which are not gases and, therefore, precipitate out of the atmosphere by attaching themselves to dust particles in the air. The external radiation hazard due to radon daughters is very small. The radon daughters emit alpha particles which are not very penetrating as the particles only travel a small distance in tissue or air. However, inhaled dust particles bearing the radon daughters may lodge in lungs where sensitive tissue is then directly exposed to the alpha radiation from the radon daughters. Estimates of the average dose equivalent to the lungs vary, but the average effective dose equivalent for Canada is around 1 mSv per year. 3) Internal radiation exposure results from the presence of naturally occurring radionuclides (K-40, C14, H-3, etc.) that were ingested and treated by the body like non-radioactive elements. Potassium is an important part of our body and the food system. Potassium is incorporated into many tissues and contributes more than 95% of the internal dose experienced by the population. Potassium-40 makes up 0.0117% of all potassium. The average dose equivalent due to internal radionuclides is approximately 0.35 mSv per year. MAN-MADE RADIATION SOURCES Man-made exposure sources are primarily due to the application of radiation in the medical and dental treatment area. However, humans have used radioactivity for over one hundred years, and through its use, added to natural occurring radioactive materials. The amounts are small compared to the natural amounts. Since the halting of above ground testing of nuclear weapons the amount of man made radiation source in the environment has been decreasing due to the short half-lives of many of the nuclides. Summaries of the United States and Canadian Dose estimates are shown below

DOSE ESTIMATE AVERAGE DOSE United States population Canadian population Natural Sources Cosmic Radiation 0.27 mSv/yr 0.30 mSv/yr Terrestrial Radiation 0.28 mSv/yr 0.35 mSv/yr Internal Radionuclides 0.39 mSv/yr 0.35 mSv/yr Inhaled (Radon) 2.00 mSv/yr 1.00 mSv/yr Subtotal: 2.94 mSv/yr 2.00 mSv /yr Manmade Sources Medical/Dental 0.53 mSv/yr 0.92 mSv Airline Travel approx. 5 uSv / hour 0.03 mSv- pass. – flightcrew 1.6 mSv/yr Consumer Products 0.10 mSv/yr 0.04 mSv/yr Nuclear Power 0.01 mSv/yr 0.01 mSv/yr Subtotal: 0.63 mSv/yr 1.05 mSv /yr Total 3.54 mSv per year 3.05 mSv per year

The total “background” radiation dose estimate for Canadians is approximately 3.0 mSv/yr. These are averages for the entire population, and there are many instances where the specific local conditions such as elevated levels of uranium and thorium in the soil, or location at a higher altitude may easily increase the contributions from some natural sources by a large factor. The exposure to man-made sources is of course also not uniform, especially for medical diagnoses where there is a very large variation for individual exposures. Considering both the natural and man-made sources, the background radiation can vary by more than a factor of two.

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Naturally occurring background levels of radiation can typically range from 1.0 to 3.5 mSv per year and in some places can be much higher. The highest known level of background radiation is in Kerala and Madras States in India where a population of over 100,000 people receive an annual dose rate which averages 13 mSv, due to the higher levels of terrestrial radiation. Radioactivity of some natural and other materials

Item Activity Isotopes 1 adult human 3,000 Bq K-40, C-14, I-129 1 kg of coffee 1,000 Bq K-40 and various nuclides 1 kg phosphate fertilizer 5,000 Bq Radium and decay products Air in a 100 metre2 home 3,000 Bq Radon 1 household smoke detector 37,000 Bq Am-241 Radioisotope for medical diagnosis 70,000,000 Bq Tc-99m Radioisotope source for medical therapy 100,000,000 million Bq Co-60 1 kg uranium ore 10,000,000 Bq U-235, U-238 1 kg of coal ash 2,000 Bq Thorium and Uranium 1 kg of granite 1,000 Bq Thorium and Uranium Listed below are some uses of radioactive materials historically and presently Americium – 241 Used in smoke detectors, Used to measure levels of toxic lead in dried paint samples,

Uniform thickness gauge in rolling processes like steel and paper production Carbon – 14 Helps in research to ensure that potential new drugs are metabolized without forming

harmful by-products Cesium – 137 Used to treat cancers, Used in flow control in oil pipelines, Used as fill level gauge for

packages of food, drugs and other products. (The products in these packages do not become radioactive.)

Chromium – 51 Used in research in red blood cell survival studies Cobalt – 57 Used in nuclear medicine diagnosis scans of patients' organs, and anemia Cobalt – 60 Used to sterilize surgical instruments, Used to improve the safety and reliability of industrial

fuel oil burners, and to preserve poultry, fruits, and spices Iodine – 123 Used to diagnose thyroid disorders Iodine – 129 Used to check some radioactivity counters, Used in vitro diagnostic testing laboratories Iodine – 131 Used to diagnose and treat thyroid disorders Iridium – 192 Used to test the integrity of pipeline welds, boilers, and aircraft parts Nickel – 63 Used to detect explosives, Used as voltage regulators and current surge protectors Phosphorus – 32 Used in molecular biology and genetics research Technetium – 99m Used for diagnostic studies in nuclear medicine. Different chemical forms used for brain,

bone, liver, spleen and kidney imaging and also for blood flow studies Thoriated tungsten Used in electric arc welding rods in the construction, aircraft, petrochemical, and food

processing equipment industries. Produces easier starting, greater arc stability and less metal contamination

Tritium H – 3 : Used for life science and drug metabolism studies to ensure the safety of potential new drugs, Used for self-luminous aircraft and commercial exit signs, for luminous dials, gauges and wristwatches.

Uranium – 235 Fuel for nuclear power plants and naval nuclear propulsion systems...also used to produce fluorescent glassware, a variety of coloured glazes, and wall tiles

Uranium – 238 Used to make counterbalance weights in some types of civil and military aircraft

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RISK COMMUNICATION HOW MUCH RADIATION DOSE DO WE RECEIVE? The average person in Calgary receives approximately between 2 – 3 mSv every year from naturally occurring background radiation as well as man-made sources. Approximately 80 percent of the dose is from natural sources of radiation. Radon is a major component of that exposure. WHAT IS THE EFFECT OF RADIATION? Radiation causes ionizations in the molecules of living cells. These ionizations result in the removal of electrons from the atoms and form ions. The ions formed go on to react with other atoms and cause damage in the cell. An example of damage would be a gamma ray which passes through the cell. The water molecules near the DNA might be ionized and the ions might react with the DNA and caused it to break. The low doses that we receive every day from background radiation allow the cells to repair the damage rapidly. The cells might not be able to repair the damage at higher doses (up to 1000 mSv), and the cells may either be changed permanently or die. The body can just replace most cells that die. However, cells changed permanently may go on to produce abnormal cells. These abnormal cells may become cancerous under the right circumstances. This is the origin of our increased risk to cancer as a result of radiation exposure. The following table gives a scale of radiation dose levels and the effects Dose Estimate Likely Effects 10,000 mSv (10 Sievert)

A short-term dose would cause immediate illness and subsequent death within a few weeks.

1,000 mSv (1 Sievert)

A short-term dose would probably cause (temporary) illness such as nausea, but not death, and in 5 of every 100 persons exposed to this dose would probably cause cancer many years later.

50 mSv/yr. and 100mSv in a 5 year period

Dose limit for nuclear energy workers. The limit includes all individuals who work with radioactive materials in a University setting.

1 mSv/yr. The dose limit for members of the general public. 2 mSv/yr. Normal background radiation in Calgary from natural sources. 0.23 mSv/yr. The average dose estimate for radioactive material users at the University of Calgary 0.05 mSv/hr. The design target for maximum annual radiation at the perimeter fence of a nuclear

electricity generating station. The actual dose is less. The biological effects are so small that the effects for low levels of radiation exposure cannot be detected with certainty. Radiation protection standards assume that the effect is directly proportional to the dose. According to this “linear” theory of radiation effects, the effect is halved if the dose is halved. HOW IS RADIATION RISK DETERMINED? Scientific committees started in the 1950s first evaluated risk estimates for radiation. The most recent of these committees were the Biological Effects of Ionizing Radiation Committee Five (BEIR V). Like previous committees, this one was charged with estimating the risk associated with radiation exposure. The BEIR IV Committee established risks exclusively for radon and other internal alpha-emitting radiation, while BEIR V Committee concentrated primarily on external radiation exposure data. Risks from radiation are difficult to estimate. Most of the radiation exposures that humans receive are very close to background levels. In most cases, the effects of radiation are not distinguishable from normal levels. Unfortunately the researchers and users of radiation in the early part of the century were not as careful as researchers are today. Most of what we know about radiation and its effects on humans

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comes from medical applications and from the survivors of the atomic bomb survivors in Japan. The following are some limitations or radiation risk estimates of BEIR V: 1. The dose rates were much higher than normally received today. 2. The actual doses received by the atomic bomb survivors group and some of the medical treatment

cases have had to be estimated and are not known precisely. 3. Other factors like ethnic origin, natural levels of cancers, diet, smoking, stress and researcher biases

affect the estimates. WHAT ARE THE RADIATION EXPOSURE RISKS? The linear, no-threshold dose-response model (line 1 in the Figure) implies that radiation exposure carries some long-term risk. One estimate is, a single whole body radiation exposure of 10 mSv would represent a risk of about 4 in 10,000 of developing a fatal cancer. A linear model would predict that a worker who receives a 50-mSv exposure would have a risk of 20 in 10,000 of developing a fatal cancer. A worker who receives a 1-mSv exposure would have a risk of 0.4 in 10,000 of developing a fatal cancer. The calculated theoretical radiation risk should not be taken as the only risk facing workers. Some potential cancers are easily cured (i.e. thyroid) while other are incurable at the present time. The natural incidence of cancer is approximately 30 percent; the incidence of fatal cancers is approximately 20 percent. The United States cancer fatality rate is approximately, one in five persons (2,000 in 10,000) will die of cancer induced from one of many possible environmental causes (smoking, food, alcohol, drugs, pollutants, natural background radiation, inherited traits, etc.). Extrapolating the risk estimate with the natural incidence for a group of 10,000 radiation workers, who each receive 10-mSv radiation exposure, 2,004 fatal cancers would be expected. The problem facing regulators is that the occupational radiation cancers, if produced, are of such a low frequency that they are indistinguishable among the high background rate of natural cancers.

To complicate matters even more, if the same total exposure is received over a longer period of time, there is lower risk by one-half to one-quarter, than from a single exposure event.

The Biological Effects of Ionizing Radiation committee V (BEIR V) suggested that the risk of cancer death is 0.08 percent per 10 mSv for acute doses and might be 2 - 4 times less than that for chronic, low-level, low Linear Energy Transfer doses. There is significant uncertainty associated with the estimate, because these estimates are an average for all ages, sexes, and all forms of cancers. Other agencies have suggested estimates that differ primarily because of the different assumptions and risk models used in the calculation. Although the radiation effects from high radiation exposures are well known and documented. There is no measurable increase found in the number of cancers or genetic effects in persons occupationally exposed to the allowable limits for a life-time of radiation work. Another way to look at radiation risk is to compare the projected average number of days of life expectancy lost per 10-mSv (1-rem) exposure to other health risks. The average United States radiation worker exposure in 1992 was 3 mSv (0.3 rem) and the University of Calgary worker average annual exposure (of those who received an actual reportable dose) approximately 0.16 mSv in 2005. A 3 mSv radiation exposure per year from age 18 to 65 would estimate a life expectancy loss of 15 days. Lost of life expectancy and health risks are compared with other risks in Tables A and B below.

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HEALTH RISKS vs. LIFE EXPECTANCY INDUSTRIAL ACCIDENT vs. LIFE EXPECTANCY TABLE A TABLE B

Smoking 20 cigarettes a day 6 years All industries 60 days Overweight (by 15%) 2 years Agriculture 320 days Alcohol consumption (US) 1 year Construction 227 days Motor vehicle accidents 207 days Mining / Quarrying 167 days Home accidents 74 days Transportation and Utilities 160 days Natural disaster (earthquake, flood) 7 days Government 60 days Medical diagnostic radiation 6 days Manufacturing 40 days 3 mSv per year from 18 to 65 15 days Trades 27 days 10 mSv per year from 18 to 65 51 days Services 27 days

These are estimates taken from the NRC Draft Guide DG-8012, and were adapted from B.L. Cohen and I.S. Lee, "Catalogue of Risks Extended and Updates", Health Physics, Vol. 61, September 1991.

Each individual must balance the risks with the benefits. The benefit is that countries can have a source of power, or the ability to do scientific research, or receive medical treatments. The risks are a small increase in cancer risk. Risk comparisons show that radiation is a small risk factor when compared to everyday risks taken by individuals. Radiation has been studied for over 100 years. Radiation is not a mysterious source of disease, but a well-understood phenomenon. Radiation is better understood than almost any other cancer-causing agent to which humans are or can be exposed.

RADIATION BIOLOGY INTERACTION OF RADIATION AND MATTER

BIOLOGICAL EFFECTS OF RADIATION Radiation biology is the study of interactions between any type of radiation and living organisms or organic materials. The energy absorption initiates the effects that are observed in living irradiated systems. No effects are produced, if the radiation is transmitted, scattered or reflected, or if the energy is otherwise not absorbed. Bioeffects are roughly proportional to the amount of absorbed energy. When any agent (chemicals, radiation, excessive heat, etc.) damages a cellular component, a multitude of measurable effects can result. The type of cell damage will depend upon what the specific agent is that the cell is exposed to. Also, the amount of damage will be related to how much of the agent reaches that particular kind of cell. The changes maybe initially restricted to either a single cell or a few types of cells. Whole organs or organ systems may be affected in time due to the absence of a required function that upsets the equilibrium or control of the whole-interrelated system. Any alteration of the genetic information carried by the genes, or of the processes associated with mitosis can result in either a permanent change in the nature of the cell (mutation), or in the cell's death. Biological effects from radiation are produced as a result of the transfer of energy from the radiation to the cells through ionization and excitation as described in the next section. Initial Chemical Reaction Radiation passing through living cells causes ionization or excitation of atoms and molecules contained in the cell. Ionization and excitation occurs in any material. The human body is comprised of large amounts of water. Water molecules are the most likely targets for being hit by photons or charged particles. The reaction, which occurs when this happens, is the formation of an ion pair. The reaction is shown below.

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A positive ion (H2O+) and an electron (e-); The H2O+ is rapidly hydrated to form: The free electron will also react with a water molecule.

H2O+ + H2O --> H30+ + OH* e- + H2O --> OH- + H* Here the OH* is a "free radical", a species After the electron slows down from bumping into other that contains an unpaired orbital electron, water molecules to yield another free radical, this time and is a highly reactive chemically. hydrogen:

The overall reaction is:

H2O + e- --> OH* + H* The products are separated by a considerable distance so that immediate back reactions to form water are not favoured. Such radicals can combine with each other and with dissolved oxygen to give a variety of potent oxidizing agents such as hydrogen peroxide, superoxide, molecular oxygen, and the perhydroxy radical. Both the initial radicals and these products can migrate to biologically important molecules (like DNA - the structural material of genes) and cause bond breakage and/or oxidation of attached groups. The energy of the radiation is transferred to biologically significant molecules, changing their structure. The energy-transfer is known as the INDIRECT EFFECT and can account for an appreciable fraction of damage. Additionally, to the indirect effect, radiation may itself cause ionization in DNA or other biological molecules. The energy of ionization is far greater than the bond energy in organic molecules, thus causing bond breakage. The amount of this DIRECT EFFECT occurring depends on the number of a particular type of molecule in the cells, and its size. The larger a molecule is, the better target it makes. Since DNA is the largest molecule in the cell as well as the site of all DNA information, its response has a central role in the mediation of radiation effects. Depending on how it is damaged, different results may occur. Cells have developed a sophisticated approach to the integration, of the initial sensing and subsequent repair of DNA as a means of ensuring genomic stability. A double strand break may fail resulting in cellular death if the inherent repair mechanisms can not repair the damage. A break in one of the basepairs will transmit different genetic data during subsequent division resulting in some kind of a mutation. Both direct and indirect effects contribute to the overall number of such damaging events to the DNA and will vary for individual cell types. Law of Bergonie and Tribondeau Early observations show that changes in different types of cells to radiation or the radiosensitivity of a particular cell depends on a number of factors. The "Law of Bergonie and Tribondeau" states "the radiosensitivity of a tissue is directly proportional to the reproductive activity and inversely proportional to the degree of differentiation". Tissues consisting of rapidly dividing stem cells (like blood or sperm cell precursors) are quite sensitive to radiation whereas cells that do not divide or only rarely divide (like nerve or muscle cells) are considerably more resistant. Microscopic examination shows that cells appear to be stuck in the division process and never successfully complete division after radiation exposure, which is consistent with the "Law" above. Other factors involved include metabolic rate, state of nourishment, oxygen level, and presence of particular enzymes within the cell. The latter are most likely involved with repair of radiation damage. The most radiation-sensitive state of any individual is during embryonic development. Humans are most radiosensitive during the first sixteen weeks of gestation. Radiation sensitivity is due to the presence of only a few cells at this stage of gestation, which will ultimately develop into a particular tissue or organ. Cells damaged at this stage of development cannot be replaced or regenerated.

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Dose-response Relationships The biological effects are due to the formation of ions in the cellular materials, which make up the body's tissues. Energy is transferred from the radiation to the material, and the absorbed dose is the amount of energy absorbed per unit mass of the material. The basic assumption in radiation protection is that the harmful effects of ionizing radiation are related to the absorbed dose. One of the questions that is often asked is whether the effect of low-level ionizing radiation is cumulative or whether it is harmful only if it exceeds some threshold value. Most of the evidence now appears to indicate that there are effects of both kinds. Some effects, such as cataracts, do not occur below a certain threshold level of radiation. Others, effects, the induction of cancer, do not appear to have any threshold and the risk of inducing such an effect increases with increasing dose. See figures a) dose response and stochastic effects b) dose response and deterministic effects. Stochastic Effects Stochastic effects are health effects that occur randomly. The probability of the effect occurring, rather than the severity, is assumed to be a linear function of the radiation dose received. There is no evidence of a threshold dose, such as the induction of cancer, it is known that if a group of people are exposed to radiation then a certain fraction will show the effect, for example develop cancer. The severity of the cancer does not depend on the radiation dose. There is no way at present of predicting which individuals will be affected. Any individual, who is exposed to radiation, can be thought of as an increased probability that cancer will occur. For example, the likelihood of developing cancer increases with increased exposure to ionizing radiation. These effects are called stochastic and it is assumed for protection purposes that the risk of stochastic effects from exposure to radiation depends directly on the absorbed dose. Deterministic Effects The term “deterministic” is used to describe effects in an individual; the severity of the effects depends on the ionizing radiation dose. For deterministic effects to be seen there is a threshold level radiation dose. Examples of deterministic effects are cataracts, skin damage, depression of red blood cell formation, and decreased fertility. Skin damage, similar to a burn, for example does not occur in any individuals as a consequence of radiation exposure except for rather large doses and the larger the dose above the threshold the greater the damage to the individual. Radiation exposures sufficiently high to produce most deterministic effects are not very likely at the University of Calgary except possibly in case of an accident. Genetic Effects Stochastic effects are genetic effects which show up in the children of a parent who has been exposed to radiation before the children were conceived. These effects should not be confused with the possible harmful effects due to exposure of the mother during pregnancy. The estimates of genetic risk to humans is based entirely on animal studies since no genetic effects due to radiation exposure have ever been demonstrated in human populations. It does not mean that genetic effects do not exist, but rather genetic effects are small and difficult to find. The difficulty with identifying radiation induced inherited abnormalities in the human population is that the normal incidence of such abnormalities is quite high. Most of the information on the normal incidence of inherited disorders comes

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from epidemiological studies done in British Columbia and Northern Ireland. The population survey in British Columbia reported that at least 9% of all live born humans would be seriously handicapped at some time during their lifetimes due to genetic disorders.

EXPOSURE REDUCTION AND CONTAMINATION CONTROL

Two radiation exposure categories that are of concern are external exposures and internal exposures. External exposures result from radiation sources, which are located outside of the body. Internal exposures result from radiation sources, which have been internalized or are located inside the body. Internal exposures have the potential of being much more significant from a biological point of view than external exposure. CONTROL OF EXTERNAL EXPOSURE Time, distance, shielding, and identification are four basic radiation protection principles that MUST be used to minimize external radiation exposures. Time - Radiation doses are directly proportional to the time spent in the radiation field. • The workers dose is halved, when the time spent in a given radiation field is halved. • Preplan the procedures to reduce the time in an exposure field. • Develop a detailed plan for carrying out the steps before beginning any new procedures. • Complete as much preparation work as possible before incorporating any radioactive materials are

into the experimental procedures • Practice the techniques required for the procedure to reduce the time of exposure. • Prior to beginning the procedure organize and have readily available all the necessary materials,

tools, and equipment

Distance – Radiation doses are inversely proportional to the distance squared • The amount of exposure received will be

reduced by, increasing the distance from a radiation source.

• Remote handling devices (manipulators) typically the use of forceps, tongs, or tweezers can greatly reduce the exposure to hands or extremities

• Do not work directly over open containers of radioactive materials

• Practical methods for maximizing distance from radiation sources include 1. Positioning the major portion of your

body, including the head, as far as possible from radiation sources while working.

2. Work at arms length from radiation sources. 3. Use tools (forceps, tongs) to handle primary vials and other sources, whenever possible

Shielding – Radiation doses can be greatly reduced by the use of the appropriate type of shields • Store and work with high energy beta emitters like 32-P behind Lucite shielding. • Keep gamma emitters behind lead shielding. Ensure that all surfaces of the shielding containers (top,

bottom, front, back, & sides) are appropriate. • Do not assume that the floor or a wall is adequate shielding. Personnel working in adjacent spaces

may receive unnecessary and/or unmonitored exposures due to inadequate shielding. • Radioactive material and samples when not in use shall be returned to appropriate shielded storage

areas. • Use beta shields on pipettes for hand protection. • Wear plastic safety eyewear to protect the cornea of your eye from beta radiation.

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• When loading or removing samples from “96 Well Plates”, place a piece of Lucite on the wellplate to shield the beta radiation

Identification – Knowing the type and energy of the emitter will determine the appropriate type of shielding, the assay method, and technological advances for the procedure being completed. • Isotopic Substitution – Can a less energetic nuclide be substituted? For example P-33 can be

substituted for P-32 with a significant reduction in the energy of the beta emission, from 1710 keV to 265 keV.

• Non-radioactive techniques – Can a non-radioactive assay or labelling technique get the desired result with the same sensitivity?

• Technology advances are making certain procedures obsolete, from a radiation safety perspective. (Hybridization ovens have almost replaced hybridization bags and bag sealers, which were a major source of personal contaminations in the past.)

CONTAMINATION CONTROL – (As Low As Reasonably Achievable) ALARA The ALARA principle is a concept to reduce exposures to radioactive materials and exposure fields. The state of technology, the economics of equipment improvements in relations to benefits of the health and safety of the workers and the public are taken into account. It is a prudent practice to make every reasonable effort to reduce radiation exposures, remove all contamination if possible, and to prevent releases of radioactive materials to unrestricted areas. CONTROL OF RADIOACTIVE CONTAMINATION AND INTERNAL RADIATION EXPOSURE Radioactive material located inside the body may result in an internal radiation dose. Internal doses are hazardous and more difficult to detect and quantify than external doses. Therefore, significant emphasis is placed upon preventing the internalization of contamination. What is radioactive contamination? “Radioactive contamination” refers to the presence of radioactive material in areas or on surfaces where it is not desired. Radioactive contamination is actual physical material that typically cannot be seen, but can be transferred from one surface to another through contact. Radioactive contamination can be airborne, or deposited directly on surfaces. Contamination on surfaces can be either “loose” (removable) or “fixed” (non-removable). “Loose surface contamination" can be described as contamination that is easily removed from a surface. Loose contamination can be easily spread to other surfaces through contact. “Fixed surface contamination” remains adhered or stuck to surfaces and are not further reduced using normal decontamination techniques such as wiping or scrubbing. Areas with non-removable radioactive contamination will generally be controlled because of the radiation levels produced by the contamination. Potential Routes of Intake The following are four potential routes of intake: absorption, ingestion, inhalation, and wounds • Absorption - Absorption takes place when direct skin contact is made with a contaminated surface or

when skin is exposed to airborne contamination. Absorption is a common route for H-3 intake, since the molecule is so small. Absorption through the skin is the most common route of intake at the University of Calgary or any University research facility.

• Inhalation - The leading means of inhalation is through the use of volatile substances outside of the fumehood. Tritiated, some Sulfur and Iodine compounds can have significant volatile byproducts of decomposition present in the vial, always break the seal of the vial in the fumehood. However, re-suspension of loose surface contamination can also result in inhalation; however, this is rare and requires vigorous scrubbing of a dry surface with contamination. The use and replacement of benchkote will greatly reduce the potential of re-suspension of loose contamination.

• Ingestion – Ingestion is the accidental consumption of contaminated food or liquids. A potential intake can occur because of nervous habits like nail biting, chewing on pens, or rubbing your face.

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• Wounds – The accidental injection of radioactive materials with needles or sharps is remotely possible. When you are working with animals ensure that you have appropriate protective equipment to prevent scratches and bites which will possibly inject radioactive materials under the skin.

Steps to Control Contamination 1. Define and label the work area.

a) All radioactive work areas must be identified with radioactive warning tape on all four sides, on the front of the bench, both sides and across the back.

b) The radioactive work area should be a bench area that is away from high traffic areas and away from desk areas.

c) Bench areas next to the fumehood should be used if part of a procedure incorporates the use of the fumehood, or has the potential to generate volatile gases or compounds.

d) Radioactive work areas must be kept as small as possible, while still allowing sufficient room to carry out procedures.

e) The warning tape must be visible and shall fully enclose the radioactive work area. f) Radioactive work areas and waste trays are required to be covered with appropriate

ABSORBENT bench top covering. The absorbent side shall be facing up. g) With daily use, the covering should be changed, as a minimum, weekly. h) When working with radioactive liquids, if a spill was to occur would the bench cover absorb the

spill, if not work in a spill tray. i) Sinks used for radioactive work or the cleaning of contaminated glassware should be located in or

near radioactive work areas. One sink should be labelled for radioactive work. j) Warning tape shall completely enclose the designated sink.

2. Label equipment.

a) Large equipment (waterbaths, centrifuges, incubators, etc.) used in conjunction with radioisotopes must be labelled with warning tape.

b) Small equipment (micropipettes, glassware, pencils etc.) within radioactive work areas must be labelled with warning tape.

c) Equipment and glassware are required to be dedicated to radioactive work. d) Decontaminate and monitor for contamination before removing any items from a radioactive work

area. 3. Use personal protective equipment.

a) Gloves – Protect your hands by double gloving, if you contaminate a glove you can remove the glove and continue with the experiment.

b) Laboratory coats – Protect your clothes from contamination. c) Appropriate clothing – Wear appropriate clothing. Pants, which cover your legs, will provide

additional protection if a spill occurs. No shorts! d) Shoes – Shoes must cover your feet completely. Sandals, thongs, or clogs do not offer

appropriate protection in the event of a spill. e) Fumehood – A fumehood prevents airborne contaminates from being inhaled by redirecting or

drawing the contaminates away from your breathing zone. Fumehoods are to be used with all volatile radioisotopes like iodine, some sulfur compounds and tritium labelled organic solutions.

4. Perform a dry run.

a) Perform a dry run before beginning any new procedures. b) A dry run will identify any missing equipment, improper layout, and potential spill problems. c) Areas to minimize exposure will be identified by performing a dry run. d) Have a copy of the protocol available as a reference, but ensure that it cannot become

contaminated. 5. Perform procedure using good work practices.

a) Complete the protocol as intended, DO NOT USE shortcuts or changes the order of the procedures.

b) NEVER mouth pipette radioactive solutions.

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6. Use appropriate housekeeping practices. a) Ensure that all radioactive materials are stored properly, upon completion of all radioactive work. b) Ensure that containers are labelled with the following: radioactive warning tape, isotope, activity,

date, name of person whose is responsible for the material, and record the storage location on the inventory sheet.

c) Dispose all wastes into the appropriate waste containers. d) Record the amount of radioactive materials used on the inventory sheet. e) Record the amount and location of radioactive waste on the inventory sheet. f) Swipe check all work surfaces and equipment used during the radioactive procedure, if

contamination is found decontaminate as required and re-swipe check. g) Monitor your hands, arms, feet, and labcoat when your procedure is completed or if

contamination is suspected. h) Remove labcoat and wash hands when your work is complete to remove any possible

contaminates, prior to leaving the laboratory. Internal Dose Reduction The equivalent dose received, as a result of an internal uptake of radioactive material is proportional to the effective half-life. Although the physical half-life component is unalterable, in some cases the biological half-life can be reduced. Isotopic dilution, metabolism stimulation, and chelation therapy are techniques, which have proven effective in reducing the biological impacts of internalized radioisotopes. These treatments must be supervised by a physican specializing in Health Physics. Isotopic dilution floods the area of concern with the stable form of that same material. Metabolic stimulation refers to the speeding up of normal body process with the intent of causing the internally deposited compounds which contain radionuclides to eliminated from the body more rapidly. Chelation therapy involves the administering of a chelating agent, which is a chemical compound that will bind metal ions into a soluble complex, which can be excreted through the kidneys.

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DOSIMETERS Radiation Safety monitors a radiation worker's external radiation exposure with a personal dosimeter badge. These devices store the radiation energy over a specified period of time. There are several types of dosimeters (wrist, ring, and whole body) used at the University of Calgary.

PERSONAL DOSIMETRY PROGRAM The dosimeter badge (at right) used at the University of Calgary has Aluminum oxide chips to record the exposure to ionizing radiation. Unlike active detectors, this dosimeter has no display or readout. Radiation workers wear the dosimeter for a given period (monthly or quarterly) and return the dosimeter to EH&S for processing by Landauer. Therefore, the worker learns of their radiation exposure only several weeks after it has occurred. Dose estimate reports are sent to workers who actually received a dose on the Dosimeter badge. The CNSC requires personnel monitoring for all workers likely receive in excess of 10 percent of the one year dose limit. Individuals are required to wear Dosimeter badges when handling or using any radioactive material which decays by beta emission with a maximum energy of emission above 200 keV. The Dosimeter badge will not register exposure to beta radiation with energy less than 150 keV. Extremity dosimeter badges are used to monitor exposure to a worker's hands and fingers. The extremity dosimeter badges are ring badges with a single TLD chip. Radiation workers who have been issued a dosimeter badge to monitor radiation exposure should follow a few simple rules to insure that the dosimeter accurately records radiation exposure. • Wear only your assigned dosimeter badge. • Never wear another workers dosimeter badge. • Wear your whole body dosimeter badge between your collar and waist. • Wear your ring badge underneath your gloves with the label of the Ring on

the palm side of the hand, the surfaces that handles the radiation sources. • Place your dosimeter badge inside your apron, when it is necessary to wear a

lead apron. • Do not store your dosimeter badge near radiation sources or heat sources. • Store your dosimeter badge on the badge board when the dosimeter badge is

not in use. • Do not leave your badge attached to your lab coat (when not wearing your lab coat). • If contamination on your dosimeter badge is suspected, telephone EH&S at 220-5333 immediately. • Never intentionally expose your badge to any radiation. • Do not wear dosimeter badge when receiving medical radiation exposure (e.g., x-rays, nuclear

medicine treatments, etc.). • Return the dosimeter badge to the dosimeter badge coordinator for exchange at the end of the

wearing period. The laboratory is responsible for any lost or missing dosimeter badges. TYPES OF DOSIMETRY USED ON CAMPUS Whole Body Dosimeter The whole body dosimeter is worn on the upper torso between the neck and the waist; usually people clip the dosimeter to the labcoat breast pocket. The whole body dosimeters are exchanged every three months. Extremity Dosimeters - Ring Dosimeters: Thermoluminescent dosimeters in the form of finger rings are worn on the NON-dominant hand index or middle finger inside gloves to monitor hand exposure to radioactive materials. It is important to ensure

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that the chip is facing the inside of the hand prior to handling radioactive materials. The dosimeters are exchanged every month due to the higher potential for exposures experienced by the hands. Extremity Dosimeters - Wrist Dosimeters: Thermoluminescent dosimeters in the form of dosimeter with a wrist strap are worn on the dominant hand to monitor extremity exposure to radioactive materials. The dosimeter badge holder must face the source of radiation. It important to ensure the holder is in place prior to using radioactive materials. The dosimeters are exchanged every three months. Distribution of dosimeter badges Dosimeters are issued by EH&S, based on procedures used and the type of equipment used. Badges are exchanged monthly, or quarterly, depending upon the type of dosimeter. Dosimeter (TLD) – Exposure Determination The dosimeters containing Aluminum oxide (Al2O3) are used as a personnel monitor. Exposure of these materials to ionizing radiation results in the storage of the radiation energy. The energy is transferred to the electron structure of the Aluminum oxide which traps the electrons in higher energy state. The amount of radiation exposure is measured by stimulating the Al2O3 material with green light from either a laser or light emitting diode source. The resulting blue light emitted from the Al2O3 is proportional to the amount of radiation exposure. Both high and low-energy photons and beta particles can be measured with this technique. The amount of light released is measured and is proportional to the exposure of the dosimeter to radiation. These materials are x-ray, beta, and gamma sensitive and exposure is reported as being either shallow energy penetration or whole body organ dose. Dosimetry Records Dosimetry records are maintained for active workers by EH&S. Dose estimate reports are sent to individuals who receive an exposure on their dosimeter badge. The Radiation Safety Officer will notify individuals of high exposure readings and will investigate the cause, if at any time your exposure exceeds the CNSC guidelines or is unusually high.

Putting Your Dose Estimates in Perspective

Radiation Protection Branch of Health Canada Occupational Doses To help put the numerical dose estimates listed in your exposure reports in perspective, the table below contains some representative dose values to which can be used as a comparison. Background Doses The average Canadian receives approximately 2.0 mSv per year from natural background radiation. This exposure is made up of 0.3 mSv from cosmic radiation from space, 0.35 mSv from terrestrial gamma radiation from the soil beneath our feet, 0.35 mSv from natural radionuclides that are incorporated into our bodies, and 1.0 mSv from radon gases in the air we breathe. Cosmic ray doses received by airline passengers vary with flight duration, altitude and latitude, as well as solar activity. A typical radiation dose received from a trip from Toronto to Vancouver is 0.04 mSv. Annual doses to flight crewmembers on this route have been estimated to be in the 5.0 mSv per year range.

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Average Diagnostic Radiation Doses Doses to patients receiving diagnostic x-rays can range from about 0.10 mSv for a chest x-ray to 5.4 mSv for a barium enema and 7.0 mSv for a CT bodyscan. In Canada, the annual average effective dose per person (total effective dose from medical procedures divided by total population) has been found to be 0.94 mSv for x-ray procedures and 0.16 mSv from diagnostic nuclear medicine procedures. Summary The various radiation exposures discussed differ in terms of who benefits and how much choice we have about the exposures. While, in principle, we could all choose to live on coastal plains, where natural background radiation levels are at a minimum, in practice we tend to consider natural background as not being under our control. On the other hand, we view artificial radiation sources as entirely under our control and we have a right to expect that there is a benefit from such exposures that will not cause any detrimental effects. In the case of medical exposures, the benefit should accrue directly to the exposed individual; e.g., a patient should expect improved diagnosis and medical care to result from an x-ray, or that a cancer will be cured and the patient's life extended via radiation therapy. In the case of diagnostic procedures, there is an ongoing effort to minimize patient dose while maintaining the quality of the diagnostic information, in other words, to maximize the ratio of benefit to risk. With occupational exposures, the associated benefits don't go to the exposed individuals. Instead, there is either a societal benefit, e.g., clean electrical power from nuclear generating stations or improved safety due to radiographic inspections of welds, or there is a benefit to specific groups of individuals, e.g., doses received by nuclear medicine technicians benefit the patients. There are ongoing efforts to improve the ways we use radiation so that occupational doses are minimized and the benefit-to-risk ratio is maximized for individuals and society. Selected occupational groups in Canada received radiation doses during 1995 as shown in the following table:

Job Category below Report Threshold (Note 1)

Average Reported Doses (Note 2)

Range of Maximum Doses(Note 3)

Office Staff 79.7% 0.49 mSv 2 - 5 mSv Industrial Radiographer 47.4% 5.23 mSv 20 - 50 mSv Lab Technician (Industrial) 78.3% 0.78 mSv 5 - 20 mSv Scientist/Engineer (Lab) 74.3% 0.46 mSv 5 - 20 mSv Dental Assistant 98.9% 0.56 mSv 5 - 20 mSv Dentist 98.3% 0.48 mSv 5 - 20 mSv Lab Technician (Medical) 90.5% 0.43 mSv 5 - 20 mSv Medical Radiation Technologist 88.6% 0.49 mSv 20 - 50 mSv Nuclear Medicine Technologist 33.8% 1.64 mSv 5 - 20 mSv Nurse 89.8% 0.42 mSv 2 - 5 mSv Diagnostic Radiologist 85.8% 0.61 mSv 5 - 20 mSv Veterinarian 89.4% 0.40 mSv 5 - 20 mSv Note 1: For example, 98.3% of dentists had exposures below the report threshold of 0.2 mSv. Note 2: For example, the remaining 1.7% of dentists had an average of 0.48 mSv reported. Note 3: For details, please obtain a copy of the source document identified below.

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RADIATION DETECTION AND MEASUREMENT RADIATION DETECTORS Radiation is detected using instrumentation, which measure the amount or number of ionization’s or excitation events that occur within the detector's sensitive volume. A radiation detection system can be either passive or active, depending upon the device and the mechanism. Passive devices are usually processed at a special processing facility before the amount of radiation exposure can be reported. Examples of passive devices are Thermoluminescent Dosimeters (TLD’s) used in determining individual radiation exposure. Active devices provide an immediate indication of the amount of radiation or radioactivity present and consist primarily of portable radiation survey meters and laboratory sample counting devices. Gas-Filled Radiation Detectors Schematic at right illustrates the basic principle used by portable radiation survey instruments for the detection and measurement of ionizing radiation. The Geiger-Müeller detector is a gas-filled, cylindrical tube with a long central wire that has a 900-volt positive charge applied to it and is then connected, through a meter, to the walls of the tube. Radiation enters the sensitive volume of the detector and produces ions in the gas. The electron part of the ion pair is attracted to the positively charged central wire where it enters the electric circuit. The meter then shows this flow of electrons (i.e., the number of ionizing events) in counts per minute (cpm). The only requirement for radiation detection by this type of detector is that the radiation must have enough energy to penetrate the walls of the detector tube and create ion pairs in the gas. Particulate (alpha and beta) radiations have limited ranges in solid materials. Radiation detectors designed for these radiations must be constructed with thin walls that allow the radiation to penetrate. The most common types of gas-filled radiation survey meters are ion chamber, (gas-flow) proportional and Geiger-Müeller (GM) detectors. a) Ion Chamber Survey Meters Ion chamber survey meters are radiation detection devices designed to collect all of the ion pairs produced in the detector tube and then measure the current flow. These meters are primarily used to measure gamma or x-ray exposure in air and the readings are usually expressed as milliRoentgen per hour (mR/hr) or Roentgen per hour (R/hr). Because research labs use only small quantities of predominantly beta emitters, they do not use ion chamber survey meters. Ion chamber are the most often used for measuring gamma and x-ray exposure, because the system is stable to within plus-or-minus 0.1% over several years, it is used to reliably measure calibration sources. b) Geiger- Müeller (GM) Contamination Monitor A Geiger-Mueller detector is characterized by the fact that almost all radiation incidents on the sensitive volume are detected. Any incident particle that ionizes at least one molecule of the gas will institute a succession of ionization’s and discharges in the counter that causes the central wire to collect a multitude

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of additional electrons. This tremendous avalanche of charge (about 109 electrons) produces a signal of about 1-volt. Ionizing radiation enters the chamber and strikes a filling gas molecule or gamma and x-ray photons interact with the wall material, kicking electrons into the gas to cause secondary ionization. Ion pairs produced accelerate toward their respective electrodes due to high voltage potential. Secondary ions are produced due to the rapid movement of the initial ion pair towards the electrodes so the entire sensitive volume of the tube is ionized (Townsend avalanche). The ions reaching electrodes are neutralized and produce the voltage pulse, which will be measured by the electronic processing unit. GM systems are pulse counters but, are independent of the energy of incident radiation and are relatively independent of applied voltage. However, because of this characteristic, a GM tube tells nothing about the energy of the radiation. GM systems are primarily used as contamination meters because of their high sensitivity. Although practically every Beta particle that reaches the counter gas will cause a discharge and produce a count, because gamma photons are less densely ionizing only a small fraction of the gamma will interact with the walls and a much smaller fraction interacts with the gas. The production, collection and neutralization of the ion pairs require time. This time period is called the instrument's resolving time and is the time required to attain the ion field and collect it. During this period, the GM is incapable of responding fully to a second radiation interaction. This resolving time is the sum of two other time elements. The dead Time is the time required for the positive ions created by the Townsend avalanche to move to the anode to be neutralized. The recovery time is the time interval between dead time and full recovery. During the recovery phase, an output pulse from ion avalanche is not large enough to pass the meter's discriminator and be counted. Thus, the resolving time is the minimum time that must elapse after detection before a second event can be detected. For GM tubes it is generally about 10 - 1000 microseconds. A potential problem of older GM systems related to resolving time is saturation. This may occur to a tube exposed to a high radiation field. In such fields, the ionizing events are interacting with the gas in the GM tube with an average separation in time much shorter than the meter's dead time. If a new ionizing event occurs in the sensitive volume when the tube still has not fully recovered, a pulse much smaller than normal (or none at all) is produced. In saturation, the instrument will show a momentary upswing of the meter needle followed by a return of the needle to a point near zero. Thus, the meter can be indicating background when the operator is in an extremely hazardous radiation field. This is why it's important to turn your meter on before entering a potentially high radiation field. Most new GM systems, when saturated, will fail upscale so the operator will know there is a high field. The gas in a GM tube is usually argon or helium and is kept at less than atmospheric pressure. Decreased pressure is used so excessively high voltages (and risk of saturation) are not required. A quenching gas is used to stop the Townsend avalanche effect by absorbing the characteristic x-rays produced by the filling gas when the positive ions reach the cathode. Thin-window detectors are either pancake or end-window variety. The window is usually a very thin (e.g., 1.5 - 4 mg/cm2) sheet of mica or Mylar. This allows for the penetration of both alpha and Beta particulate radiation. Normally, the pancake probe is a little more sensitive than an end-window probe, especially for low energy radiation. The reason is that in end-window detectors, there is usually a small dead space just behind the window where the local electric field of the central wire is too small to attract the electrons so ion pairs recombine. The pancake probe has several circular electrodes so the electric field is relatively uniform. Beta particles that enter the sensitive volume will be counted. The wide range of beta energies results in a wide range of efficiencies for the same sample geometry. Obviously, higher energy beta particles will have greater range so source absorption and absorption by the GM window will be less, and the efficiency higher.

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In summary, GM survey meters are radiation detectors used to detect radiation or to monitor for radioactive contamination. These detectors may have a variety of window thicknesses, however, if the radiation cannot penetrate the window it will not be detected. Radioactive materials that emit these types of radiation (e.g., C-14, Na-22, P-32, S-35, and Cs-137) can usually be detected using GM survey meters. GM meters at the University are usually read in units of counts per minute (cpm) for particle radiation.

c) Scintillation Detectors Scintillation is a process by which energy deposited by ionizing radiation is absorbed and converted to light photons. There are many types of scintillators including organic crystal scintillators (e.g., anthracene), organic liquid (LSC) or solid scintillators, inorganic crystal scintillators (e.g., Nal (Tl), GeLi), and noble gas scintillators (e.g., xenon, helium). Scintillation detectors are superior to gas filled detectors because the number of light photons produced is proportional to the energy absorbed in the scintillator. The purpose of the scintillation crystal is to stop the incident photon and convert the radiation energy into visible light. While incident photons can interact with the crystal by photoelectric, Compton, or by pair production, photoelectric interactions are preferred because the photon loses all of its energy in one interaction, hence the light produced in the scintillator will be proportional to the energy of the photon. Other interactions may result in only partial energy absorption (e.g., several Compton interactions, followed by a photoelectric interaction). The energized electrons are ejected from the regions in the crystal, which they had occupied and travel a short distance transferring energy to other electrons along the way. Sodium iodide (NaI) by itself does not produce much light in a pure NaI crystal the electrons move around and the energy transferred appears in the form of heat. Therefore, NaI crystals are purposely flawed with thallium (Tl) ions, which initially trap the energized electrons and subsequently increase light output, by a factor of ten at room temperature. The NaI (TI) crystal is hydroscopic, so it is placed in a hermetically sealed container. If the crystal were left unsealed, it would dissolve in about one week. The excess energy of the thallium-trapped electrons is released as light photons in the 3 eV energy range. Approximately thirty light photons, each of 3 eV is produced per keV of energy transferred to the crystal. The crystal is transparent to light photons with energies around 3 eV so these light photons pass freely through the crystal. The photocathode is selected to be maximally sensitive to light at frequencies of 410 nm. Just as with air filled detectors, it takes time for the photon to be absorbed, the electrons to de-excite and give off the light photons. For NaI (TI) this dead time is approximately 0.25 microseconds. For best results, do not count samples with activities greater than 2 MBq (~54uCi). The photomultiplier tube then takes a pulse of light and converts it into a pulse of electrons and amplifies the pulse of electrons into a measurable electric current. The photocathode is a thin, semi- transparent layer on the inside of the tube that is facing the crystal and is a substance which will emit electrons when exposed to visible light. Cesium- antimony (CsSb) is the most common material used for NaI (Tl) crystals. Light photons from the crystal interact with electrons in the photocathode causing the electrons to be ejected from their orbits as photoelectrons. The number of electrons removed from the photocathode is proportional to the energy deposited in the crystal by a gamma photon. These photoelectrons, with the aid of a focusing grid, will go to the first dynode. A dynode is an electronic device, which acts alternately as an anode and as a cathode. Seven to thirteen metal electrodes coated with a material similar to the photocathode are arranged in a special geometric pattern such that each succeeding dynode will have more positive voltage applied to it than the one before. Each photoelectron strikes the dynode and dislodges several

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secondary electrons from its surface. These secondary electrons are in turn accelerated toward the second dynode, the process continues through the dynode chain where at each stage, a variable number of secondary electrons, averaging between 2 and 4 are released for each incident electron. The overall "gain" ranging from 100,000 to 100,000,000 electrons produced per photoelectron. The anode is the last dynode of the tube. The photoelectrons are collected here and then flow through a load resistor to form a voltage pulse, which is the output signal from the PM tube. This output signal will be in the millivolt range. When the signal leaves the preamp, the signal is proportional to the energy deposited in the crystal. Finally, the signal is sent through a pulse height analyzer which, quantitatively measures the maximum amplitude reached during an electrical impulse and records the accepted pulses as counts. On campus several types of scintillation counters are normally used, low energy gamma survey meters, liquid scintillation counters (LSC), and gamma counters. Survey meters are radiation detection systems used to monitor radionuclides that emit low energy gamma radiation (e.g., Cr-51, I-125). They can not detect alpha particles nor low energy beta particles. They can detect radionuclides that emit high-energy gamma (Na-22) and/or high-energy beta (e.g., P-32 however pancake type GM detectors are normally more efficient) radiation. The meter is usually read in counts per minute. Liquid scintillation counting is a method of assaying a radioactive sample by dissolving it in a chemical solution called scintillation fluid or cocktail. When alpha or beta radiation energy is absorbed in the cocktail, it emits light. The light flashes are converted to electrical signals in the photomultiplier tube (PMT) and the electrical signals are related to the absorbed energy allowing the sample to be quantified. Liquid scintillation counters (LSC) are usually used to quantify radioactivity and to measure removable radioactive contamination. They are ideal for counting radionuclides that decay by alpha and beta particle emission (e.g., H-3, C-14, P-32, etc.) and may also be used to measure some gamma emitters (e.g., I-125, Cr-51) which emit auger electrons as part of their decay. Radiation Detection and Measurement Techniques Radioactive materials pose a potential long term risk to users. All personnel who work with radioactive materials must understand how to use the various types of radiation detectors. This is to verify that their work areas are free of contamination. To detect and measure radiation, a worker must understand how a detector works and then how to use the detector in the work place. a) Portable Survey Meter External Controls All portable survey meters have certain components in common. • The detector (or probe) produces electrical signals when exposed to

radiation. It usually has a window through which the radiation can penetrate its cavity.

• The dial (or readout) is a meter, which indicates the amount of detected radiation present. It may have two scales, mR/hr and/or cpm, at the University, only the cpm scale is used.

• The selector switch is used to turn the meter On-Off, check batteries, or select a scale (range) multiplier. The scale multiplier is the number (e.g., 0.1, 1.0, 10, etc.) by which the dial readings must be multiplied to calculate the number of counts per minute.

• The reset button allows the meter reading to be zeroed. When the radiation level causes the number of counts per minute (cpm) to exceed the highest reading on a particular scale, switch the scale multiplier to a higher range, and push the reset button. This causes the readout needle to reset to zero so the user can accurately determine count rate.

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• The response button adjusts the response time of the meter. When this switch is pushed towards the meter (fast), the meter will have a fast response but the dial needle will be less stable (i.e., rapid and great fluctuations). For slow response times, the readout's needle is slow to move, but changes are not as erratic.

• The speaker is an audible device connected to the radiation monitor. The speaker is in-line with the detector so each count produces an audible click on the speaker.

b) Radiation Survey Meter Operating Procedures Read the following instructions to become familiar with the controls and operating characteristics. • Check the meter for physical damage. • Meters are required to be calibrated at least once a year. • Check the batteries. • Turn selector switch to the BAT position. • The readout's needle must move into the BATT OK range. If not, the batteries are weak and must be

replaced. • Turn off the meter and speaker when not in use. When storing the survey meter for extended periods,

remove the batteries. • Check the detector response. • Determine the background count-rate. With the meter turned on and the selector switch on its lowest

scale, point the detector away from any radiation fields and measure the background count-rate. Note that the meter reading must be multiplied by the selector switch scale (e.g., x 0.1, x 1, x 10, etc.). This result is the background reading. Normal background for thin-window GM meters is between 20 - 40 cpm and is about 200 - 350 cpm for Scintillation meters.

• With speaker on, point the probe window at the area or equipment you wish to monitor for radiation or radioactive contamination. Unless contamination is expected, place the selector switch on the lowest scale.

• When surveying or entering contaminated areas with unknown radiation levels, turn the meter on outside the area, place the selector switch on the highest range setting and adjust the switch downward to the appropriate scale. Multiply the meter reading by the switch setting.

c) Procedures for performing these surveys • Identify the areas where radioactive material is used and/or stored. • Document survey locations in a contamination logbook. • Contamination logbooks should include a floor plan of the room and a listing of the areas. • Follow the Operating Procedures for Radiation contamination Meters. • Hold the window of the probe within 1 cm of the area or equipment you wish to monitor. Pay special

attention to telephones, logbooks, instrument handle(s) and computer keyboards (all of which should remain contamination free).

Record all information on the survey form, including: • Date survey is performed • Background radiation count rate • Initials of the person conducting the survey • Meter information (make, model, type, and serial number) Slowly move the detector over each the designated area. • With speaker on, move detector about 5 centimeters per second, listening to the speaker s clicking. • Make sure you do not contaminate the probe, use care to prevent contamination. • Turn the meter off when completed or when the meter is not in use. • Areas or locations with meter count rates exceeding 200 cpm above background must be further

investigated (e.g., decontaminated, shielded, etc.). If the exposure is due to radioactive contamination, the contamination must be cleaned and the successful decontamination must be documented and verified using swipe checks.

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Liquid Scintillation Counting (LSC) Liquid scintillation counting is an analytical technique, which is defined by the incorporation of the radioactive (radiolabelled) sample into uniform distribution within a liquid medium capable of converting the kinetic energy of the nuclear emissions into light photons. Although the liquid scintillation counter is a sophisticated laboratory counting system used to quantify the activity of radioisotopes emitting alpha, beta and gamma radiation. The steps can be broken down into relatively simple processes. a) Liquid Scintillation Principles The Figure provides a graphic illustration of the way the emitted radiation (for example a β-particle) interacts with the liquid scintillation cocktail (a solvent and a scintillator) leading to a flash of light (photon) being recorded by the Photomultiplier tube (PMT) and the system converts that information into numerical value (cpm) on a display or printout. • Beta particle emitted as part of the radioactive decay process. To assure efficient transfer of energy

between the beta particle and the cocktail, the cocktail must be a solvent for the sample material. • In the liquid, the beta particle only travels a short distance before all of its kinetic energy is dissipated

or transferred to the solvent. Typically a beta particle will take a few nanoseconds to dissipate all its energy.

• Energy of the excited solvent is emitted as UV light and the solvent molecule returns to ground state. The excited solvent molecules can transfer energy to each other and to the solute (Figure below). The solute is a Fluor. An excited solvent molecule, which passes its energy to a solute molecule, disturbs the orbital electron cloud of the solute raising it to a state of excitation. As the excited orbital electrons of the solute molecule return to the ground state, a radiation results, in this case a photon of UV light. The UV light is absorbed by Fluor molecules which emit blue light flashes upon return to ground state. Nuclear decay events produce approximately 10 photons per keV of energy. The energy is dissipated in approximately 5 –10 nanoseconds. The total number of photons from the excited Fluor molecules constitutes the scintillation. The intensity of light in the scintillation is proportional to the beta particle's initial energy.

• Blue light flashes hit the photocathode of the PMT. Electrons (proportional in number to the blue light pulses) are ejected producing an electrical pulse that is proportional to the number of blue light photons. A LSC normally has two PMT's. The signal from each PMT is fed into a circuit which measures the amplitude of the pulses from each PMT; an output is obtained which is proportional to the total intensity of the scintillation process.

• The amplitude of the electrical pulse is converted into a digital value and the digital value, which represents the beta particle energy, passes into the analyzer where it is compared to digital values for each of the LSC's channels. Each channel is a specific address for a memory slot in the multi-channel analyzer, many storage slots or channels covering the energy range from 0 - 2000 keV.

• Number of pulses in each channel is printed out or displayed on a monitor. The sample is analyzed and the spectrum can be plotted to provide information about the energy of the radiation or the amount of radioactive material dissolved in the cocktail.

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b) LSC Terminology • Chemiluminescence - Random single photon events, which are generated as a result of the

chemical interaction of the sample components. • Chemical Quenching - A reduction in the scintillation (light) intensity as seen by the photomultiplier

tubes due to light absorbing chemicals present in the scintillation solution. The result is fewer photons per keV of beta particle energy and usually a reduction in counting efficiency.

• Cocktail or counting fluid - The scintillation counting fluid, a mixture of 3 chemicals (solvent, emulsifier, and Fluor) which produces light flashes when it absorbs the energy of particulate radioactive decay.

• CPM - Counts per minute. The number of counts the LSC registered per minute. • DPM - The sample's activity in units of nuclear decays per minute. • Efficiency - The ratio of measured counts per unit time to the actual number of decays which

occurred during the measurement time. • Emulsifier - A chemical component of the liquid scintillation fluid that works to keep the radioactive

sample suspended in the cocktail. • Fluor - A chemical component of the liquid scintillation cocktail that absorbs the UV light emitted by

the solvent and re-emits a light at a frequency that the PMT can detect. • Fluorescence - The emission of light resulting from the absorption of incident radiation and persisting

only as long as the stimulating radiation is continued. • Luminescence - A term for the emission of light by a material. • Optical Quenching - A reduction in the scintillation intensity seen by the photomultiplier tubes due to

absorption of the scintillation light either by materials present in scintillation solution or deposited on the walls of the sample container or optic (e.g., dirt). The result is fewer photons per keV of beta particle energy and usually a reduction in counting efficiency.

• PMT - The Photo-Multiplier Tube is an electron tube that detects the blue light flashes from the Fluor and converts them into an electrical pulse.

• Phosphor- A luminescent substance or material capable of emitting light when stimulated by radiation.

• Photoluminescence - Delayed and persistent emission of single photons of light following activation by photons of light such as ultraviolet.

• Quench - Anything which interferes with the conversion of the sample's radioactive decay energy into light photons that the PMT can detect. Quenching results in a reduction in counting efficiency.

• Secondary Scintillator - Material in the scintillation fluuid, which absorbs the emitted light of the primary scintillator and remits it at a longer wavelength, nearer the maximum spectral sensitivity of the photomultiplier tubes. It is added to improve the counting efficiency of the sample.

• Solvent - A chemical component of the liquid scintillation cocktail that dissolves the sample material and absorbs the emitted radiation excitation energy and emits UV light, which is absorbed by the Fluor.

c) Considerations in Isotopic Analysis The beta particle must have sufficient energy to produce at least 2 photons in the cocktail and one must interact with each PMT. The photocathode of a PMT is not 100% efficient. The conversion efficiency from a photon to a photoelectron is only about 30%. The use of LSC for beta-emitters is attractive because it offers counting efficiencies of up to 100% and simplicity of sample preparation. Almost all the kinetic energy associated with a beta emission is given up to the media in a relatively short distance. The relative scintillation yield from this

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depends upon specific ionization. The higher the specific ionization, the lower the relative number of photons produced. The organic scintillators used in LSC’s have a lower gamma ray absorption coefficient than inorganic (NaI) scintillators. The photoelectric effect is small when E gamma > 300 keV and Compton scattering becomes the main absorption process. The pulse thus depends upon gamma energy. For E gamma < 20 keV, the photoelectric effect, in which all gamma energy transferred to a single electron, predominates. For 20 keV < E gamma < 100 keV, both the photoelectric and Compton effects contribute. And, for 100 keV < E gamma < 3000 keV the Compton effects predominates. But for 125-I, counting efficiency can be as high as 76% in a typical LSC. d) Quench Quench is a reduction in system efficiency as a result of energy loss in the liquid scintillation solution. Because of quench, the energy spectrum detected from the radionuclide appears to shift toward a lower energy (see figure stripped area). There are three major types of quench encountered photon, chemical, and colour quenching. Photon quenching is the incomplete transfer of beta particle energy to solvent molecule. Chemical (or impurity) quenching causes energy losses in the transfer from solvent to solute. Optical or colour quenching causes the attenuation of photons produced in solute. Chemical quenching absorbs beta energy before it is converted to photons while colour quenching results from the passage of the photons through the medium. Colour quenching depends on the colour of interfering chemical and the path length that the photon must travel. In a chemically quenched sample, all energy radiations appear to be equally affected whereas, for a coloured sample, events that take place close to one PMT will give rise to a large pulse and a smaller pulse in the other PMT. By summation, the pulses are added so the resultant pulse height may be as large as from unquenched, only the # of events will be significantly reduced. Thus, at equal quench levels, the pulse height of coloured samples are spread over a wider range than for chemical quench samples. Quenching affects the efficiency of sample detectors quench could have a significant impact on your LSC results. Quench is important. You must understand the impacts of quenching and how the system you are using indicates the problem if you want to obtain accurate results. e) Chemiluminescence / Photoluminescence Luminescence is a single photon event and is registered as a count due to the probability of having coincidence events at high luminescence activity. Although LSC’s employ a coincidence circuit, luminescence becomes a problem when the production of single photons occurs at a rate sufficiently great that separate luminescence events stimulate each PMT within the resolving time of the coincidence circuits.

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Chemiluminescence is the production of light as a result of a chemical reaction between components of the scintillation sample in the absence of radioactive material. This most typically occurs in samples of alkaline pH and/or samples containing peroxides when mixed with emulsifier-type scintillation cocktails, when alkaline tissue solubilizers are added to emulsifier-type scintillation cocktails, or in the presence of oxidizing agents in the sample. Reactions are usually exothermic and result in the production of a large number of single photons. Photoluminescence results in the excitation of the counting fluid or the counting vial by UV light (e.g., exposure to sunlight or UV light sources). Chemiluminescence has a relatively slow decay time (from 20 minutes to one-day depending on the temperature) while photoluminescence decays more rapidly. The luminescence spectrum has a pulse height distribution, which overlaps the H-3 spectrum. The maximum pulse height corresponds to approximately 6 keV and the spectrum is (chemical) quench independent. Cooling the luminescent scintillation samples reduces the photon intensity to low levels, but the interference is still present and provides a false indication of luminescence control. f) Static Electricity Static electricity on liquid scintillation vials is a single photon event with pulse height limited to about 10 keV in energy. Many of the items (plastic racks, plastic vials) used in the liquid scintillation counter process can contribute to development of static electricity. As a general rule, glass vials generate less static than plastic vials; small vials in adapters or carrier vials can contribute to a static charge build-up. Most of the newer counters have a static discharge device or an electrostatic eliminator built in to reduce this problem. Alternatively, to raise the relative humidity within the LS counter fill a number vials with water and leave these vials in a rack inside the counter. g) Sample Volume / Dual Phase Samples As the sample volume decreases, light output falls on less efficient areas of the PMT, so energy detection becomes less efficient with low volumes. When 2 phases are present, each phase will have a specific counting efficiency. h) Sample Activity “DPM” Determination Count your samples, the counts per minute and the quench level are printed out for each sample. Look up efficiency for each sample at its quench level from the calibration curve plot. To determine the activity (dpm) from the reported counts per minute (cpm), divide the number of counts by the efficiency (i.e., dpm = cpm / eff). i) Operating Procedures for LSC Counters Read the instruments operating manual to gain familiarity with the controls and operating characteristics.

• Place samples into LSC vials and add the correct amount (the smallest volume possible) of liquid scintillation cocktail (e.g., 3-4 ml for mini vials and 10 ml, for maxi vials). Include a background vial, which contains scintillation cocktail and the entire non-radioactive sample similar in make-up (i.e., geometry) to your radioactive samples.

• Place your sample vials with the background vial into the LSC tray (or belt) and place into the LSC.

• Set count time to at least 2 minutes, shorter times give poor counting statistics. • Select the isotopes for your counting needs and begin counting. • Calculate the true radioactivity of the sample in units of dpm by dividing the sample cpm by the

counter efficiency for that energy of sample (i.e., dpm = cpm/eff). As discussed above, the counter efficiency may be different for different vials depending on the amount of quenching present.

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j) Beckman LSC Considerations All LSC’s operate in the same manner, but different manufacturers use different terminology or offer more options than others. Some systems allow the user to select the regions of interest by selecting a keV range of interest. Others offer several options (channel or keV). Most of the comments made so far apply to systems, where the user selects energy regions. If you are using a Beckman, usually the channel option is the default option for the window setting. Beckman LSC’s have 1000 channels and the energy is related to the channel by the equation: Channel # = 72 + 280 log 10X(Emax): where Emax is in keV. k) Cerenkov Counting Some beta emitting isotopes can be analyzed on an LSC without using any cocktail. The literature of several manufacturers discusses counting high energy (Emax > 800 keV) beta emitters without cocktail or with only a little water, using a technique called Cerenkov counting. When high-energy beta particles travel faster than the speed of light in the medium they are traversing (e.g., water, plastic, glass, etc.) Cerenkov radiation (light) is produced. Cerenkov radiation is the blue light you see when you look into a reactor pool. Cerenkov radiation allows some beta emitting radionuclides to be analyzed with a liquid scintillation counter without using any cocktail. For Cerenkov radiation to be produced, the beta particle must exceed a minimum threshold energy (Eth) which is calculated by the formula Eth =511*n / (n2-1)1/2 - 511 In this equation, 511 is the rest mass of an electron in keV and n is the refractive index of the medium (n glass = 1.5, n water = 1.33). Consider, for example, using water instead of cocktail. Then, for water, Eth = 263 keV. If you were counting filter papers in glass vials, then Eth = 175 keV. Given these energy constraints, P-32, Cl-36l and Sr-90 / Y-90 have sufficient energy to be analyzed using Cerenkov counting. From a practical point of view, the only beta emitting radionuclide likely to be analyzed by Cerenkov counting is P-32 which emits a beta particle with Emax = 1,710 keV. Due to the spectrum of energies emitted from beta particles, approximately 86% of the 32-P beta particles have energies exceeding the Eth = 263 keV for counting in water. With proper LSC adapters, researchers can directly analyze their samples in 0.5 and 1.5 ml microfuge tubes. Consider the following example of Cerenkov counting of a P-32 labeled compound. An aliquot was placed in a 20 ml glass vial and counted with various quantities of water added (see below). The activity used was estimated by counting an identical sample in LSC cocktail and assuming 100% efficiency. As shown, counting P-32 in a 20-ml glass vial, with 4 - 12 ml of added water gives optimum efficiency. However, note that relatively good efficiencies were obtained for all samples. Typically the counting efficiency of P-32 in 4 - 12 ml of water is expected to be approximately 40 - 50% compared to the efficiency obtained by using LSC cocktail for the same P-32 sample of nearly 100%. Cerenkov Counting Efficiencies ml of water 0 1 2 4 8 12 16% efficiency 30.8 42.2 44.1 48 46.8 46.9 46.3 As with any counting method, Cerenkov counting has advantages and disadvantages. Advantages

• simple sample preparation (i.e., only add water, the volume is not too critical) • less expensive, (no LSC cocktail used) • sample can usually be recovered • no chemical quench (light is given up directly to the medium, no cocktail is employed) • Waste can be treated as solid if no water was used.

Disadvantages • lower efficiency, • higher colour quench • Volume dependence (particularly if using less than 2 ml of water) • Medium dependence (glass, plastic vials, water, etc.) • Need approximately 1000 Becquerels of activity as a minimum • CAN NOT be used for counting swipe checks.

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The biggest factor preventing universal use of Cerenkov counting is beta energy. In order to achieve adequate efficiency, the average beta energy (E avg. to the Emax) must be greater than the required threshold energy, Eth. Thus, from a practical point of view, this criterion limits Cerenkov counting to beta emitters with maximum energies greater than I MeV. The only commonly used radionuclide fitting these criteria is P-32. Removable Contamination Swipe Survey Techniques Loose surface contamination is radioactive material in a form that is easily spread and is in a place where we don't want it to be or don't know it is there. If persons walk through a contaminated area, some of the radioactive material will be picked up by their shoes and spread around as they go about their business. Loose surface contamination can be cleaned up using conventional janitorial methods although the rags and other cleaning materials will then have to be controlled as contaminated materials. Removable contamination poses potential problems, it may be inadvertently ingested if not quickly discovered and decontaminated. It can be spread beyond the laboratory and cause undue stress to families and friends of the workers involved. It may become airborne and become a potential inhalation hazard. If loose contamination is absorbed into or worked into surfaces, it becomes more difficult to remove. Although it would now appear to be fixed contamination, it may once again become loose contamination due to abrasive actions or may simply leach from the surface. Labs in which radionuclides are used and/or stored must be surveyed for removable radioactive contamination by a swipe or smear survey. At a minimum, surveys must be done weekly when a lab has radioactive materials in use. The swipe survey is performed by rubbing areas with a small piece of filter paper or cotton swab. The survey is performed over an area of approximately 100 cm2 because that is the approximate surface area that would be brushed by a person walked through the lab. Even though this area is equivalent to a square approximately 10-centimeters on a side, the preferred method of performing this survey is to swipe an area in an S-shaped pattern over a distance of about 100 centimeters (see Figure). If the item to be surveyed is small and does not have 100 cm2 of surface to swipe, attempt to swipe the entire surface and divide the results by the approximate total surface area. Swipe surveys are usually analyzed on low background, high efficiency laboratory equipment such as liquid scintillation counters, or gamma counters, as appropriate. LSC’s are routinely used to analyze for alpha, beta, and auger electron emitting radionuclides (H-3, C-14, P-32, P-33, S-35, Cl-36, Cr-51, I-125, etc.). Gamma counters are used to measure gamma-emitting radionuclides (Cr-51, Co-57, Tc-99m, I-125). a) Swipe Survey Terminology Gamma Counter A laboratory radiation detection instrument specially adapted to detect the presence of radio nuclides, which emit gamma or X-rays (e.g., Cr-51, Co-57, Tc-99m, I-125, Ce-141, etc.). Contamination The presence of radioactive material where it is not supposed to be, certain areas (desks, floors, telephones, doorknobs, etc.) are expected to be contamination free. Non-Removable Contamination Radioactive contamination present on a surface that cannot be removed or reduced using routine cleaning methods. Removable Radioactive Contamination which can be removed using routine cleaning methods.

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Survey A deliberate evaluation of the presence or radiation / contamination related to the production, use, release, disposal, or presence of sources of radiation under a specific set of condition Action Levels When the following levels of Removable Surface Contamination for beta and gamma radiation emitters are encountered, the surfaces are required to be decontaminated and resurveyed to ensure the surface has been decontaminated.

For non-radioactive work areas 0.5 Bq / cm2

For radioactive work areas 5.0 Bq / cm2

b) Swipe Survey Procedures A weekly swipe survey must be performed as noted above. A meter survey is usually performed first to identify radiation levels in the laboratory and potentially contaminated areas. After performing the meter survey, a swipe survey at the same points must be performed. Wear lab coat, safety glasses, and disposable gloves. • Protective gloves should be worn while taking swipes. • Identify locations where radioactive material is used/stored and the equipment used for radioactive

work. • Key these locations to the room's floor diagram with letters or numbers (Figure ). • Moisten pieces of filter paper, cotton-tipped swabs, or Kimwipe. • Key each swab to the identified locations on the floor diagram (e.g., label the lid of each vial into

which they will be placed). • Swipe an area of at least 100 cm2 (e.g., 10 x 10, 100 x 1, etc.) at each identified location (Figure) or

piece of equipment. • It is preferable to swipe a larger area, but use only one swab per area. Once taken, the swipe is

considered radioactive. • Handle swipes so that you avoid cross-contaminating the swipe samples. • Do not place them in your pocket as they may contaminate your clothing. • Place each swipe into its appropriate vial, tube, or planchet. • For LSC counting, add the appropriate volume of liquid scintillation cocktail into the vial. • Place and secure caps on vials. • The first sample must be a background sample, that is a sample with the swipe material, but it has

not swiped any laboratory surfaces. • Places the vials in racks and load the racks into the liquid scintillation counter. • Ensure that the counting program is set to count the isotope you are using and a wide open counting

window. • To ensure good counting statistics, set the count time for at least two minutes and the count the

samples. • Review the results for any indication of contamination. • Areas with removable in excess of the maximum permissible levels must be decontaminated and

then re-swiped. • The print out from the counter must be placed in the contamination logbook for the laboratory. The

weekly survey must include the date of the survey; initials of the person who performed the survey; the areas which were surveyed.

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References Atomic structure pictures, Page 4,5 - Triumf http://www.triumf.ca/EHS/RadiationSafetyLessons/Atomic%20Structure/AtomicStructure5.htm Figure Page 3 - Line of Stability, Pearson Education, Inc. publishing as Addison Wesley Figure page 5 - exponential decay http://www.astronomynotes.com/solarsys/decay.gif Decay rate graph Page5 - Decay rate Graph – HyperPhysics. ©C.R. Nave, 2006 Figure 3 a b http://documents.triumf.ca/docushare/dsweb/Get/Document-645/TSN_6.2.1.PDF http://www.cc.umanitoba.ca/faculties/medicine/radiology/presentations/hjohnson/K2RadiobiologyText.doc http://www.ryerson.ca/cehsm/forms/rad_safety/Radiation Safety Fundamentals Workbook.doc Radiation Protection Bureau, Publication 97-EHD-210, Occupational Radiation Exposures in Canada - 1996, available on the Internet in PDF format at: http://www.hc-sc.gc.ca/datahpb/dataehd/English/catalog/rpb_pubs/ Canada: Living with Radiation, Atomic Energy Control Board, 1995. Canada Communication Group Publishing, Ottawa, Canada K1A 0S9.

B.J. Lewis et al, Measurement of neutron radiation exposure of commercial airline pilots using bubble detectors. Nuclear Technology Volume 106, June 1994, pages 373-383.

Radiation Doses from Medical Diagnostic Procedures in Canada, Advisory Committee on Radiological Protection, INFO-0670, 1997.

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