Radiation Protection Training Manual (cont)

Chapter 6: RADIATION DETECTION AND MEASUREMENTS

Radiation cannot be perceived by human senses so, we rely upon the use of physical effects of radiation to create an instrument to detect and measuring it. There are two basic types of instruments used for its detection:

  1. Particle counting instruments
  2. Dose measuring instruments

The particle counting instruments measure the number of particles (electrons, alphas, protons, neutrons, etc.) or photons that give a signal in the detector and the result is expressed in counts per minute (cpm) or counts per second (cps). The dose-measuring instruments also measure the number of particles or photons, but the result is given in units of dose (R, rem, Gy, Sv, etc.), or dose rate (R/s, mSv/h, Gy/min, etc.). Sometimes the same instrument is capable of both types of readings.

The internal dose can be estimated from bioassay measurements performed with particle counting instruments.

6.1 Particle counting instruments

Particle counting instruments are used to determine the activity of a sample taken from the environment, to measure the activity of body fluids and can be used as portable survey instruments for contamination monitoring.

The detector in particle counting instruments can either be a gas, a solid or liquid. When ionizing particles pass through the detector, energy dissipation through a burst of ionization occurs. This burst of ionization is converted into an electrical pulse that actuates a readout device, such as a scaler or a ratemeter, to register a count.

6.1.1 Gas-filled detectors

The gas-filled detectors resemble a cylindrical condenser, with a central anode for collecting electrons and an outer cathode for collecting positive ions. The ionizing particle passes through the gas that fills the condenser, creating positive ions and electrons.

Depending on the value of the electrical field applied to the detector, the electrons initially created can acquire enough kinetic energy to produce secondary ionizations.

Fig. 6-1: Schematic design of a gas-filled detector

Depending on its energy, gamma radiation can enter through both the end window and outer wall.

There are various types of particle counting instruments filled with gas:

  1. Ionization chamber counters (no secondary ions are produced)
  2. Proportional counters (secondary ions are produced but the number is proportional to the initial energy of the radiation)
  3. Geiger-Müller counters (secondary ions are produced in large numbers and the number of ions is no longer proportional to the energy of the radiation)

The main difference between these 3 types of counters is the voltage used for charging the condenser.

6.1.2 Solid and liquid scintillations detectors

A scintillation counter is a transducer that changes the kinetic energy of an ionizing particle into a flash of light. The flashes of light are detected by photomultiplier tubes. The output pulses may be amplified, sorted by peak intensity, and counted.

Fig. 6-2: Solid scintillation counter

Different scintillation materials (NaI (Tl), CsI (Tl) crystals, plastics, or liquids) are used to detect different types of radiation.

Scintillation counters are widely used in our radiation protection program for bioassays, swipes, and laboratory experiment samples. Liquid scintillation counters have a very good detection threshold since the scintillation liquid and sample are practically mixed.

6.1.3 The measurement process

Fig. 6-3: Counting statistic

As was discussed in chapter 2.7 dedicated to the definition of radioactivity and the laws of radioactive disintegration, radioactive decay is a statistical process. The probability “p”, of the decay, is proportional to time. The constant of proportionality indicates the dependency on the radionuclide and is called the decay constant, noted as “λ” lambda. 

If the time dt is chosen as 1 s, and if 1 s is much smaller than the half-life of the radionuclide, we can represent on a diagram the frequency of the number of decays per second, as a random event. The histogram represents on the x-axis the number of decays per second and on the y-axis the frequency of each number of decays per second likely to happen. For radionuclides used in our university, which have the half-life of the order of days or more, the number of unstable nuclei is very large, and the number of decays per second larger than 20, the distribution of the number of decays per second can be approximated to a gaussian distribution.

The mean of this distribution is the activity of the sample, and the standard deviation is the square root of the activity. A measurement (counts per a certain time interval called the counting time) is a number proportional to the number of decays per second. The proportionality constant is the detector’s efficiency. If a sample is measured many times with a counting time much smaller than the half-life, a similar histogram to the one presented in figure 6.3 will be obtained, but on the x-axis, the number of counts per counting time will appear.

Fig. 6-4: Sample and background

When measuring a sample we always measure both the activity of the sample and the background. By keeping the detector at a large distance from the sample we can measure the background only. To determine the activity of the sample we need to subtract the background. Both measurements, the sample and the background are statistical events. The two Gaussian distributions will always overlap. The overlap is greater if the measured count rates are close to each other (for example when the sample measurement is for example 60 counts per minute – cpm – and the background is 50 cpm). When interpreting the result of the subtraction of two Gaussian distributions, there are two types of errors: false positive (when we assume that the result belongs to the sample distribution, but is part of the background) or false negative (when we assume that the result belongs to the background distribution but belongs to the sample). In figure 6-4, the superposition takes place at 1.65 standard deviations for both distributions. For this situation, the false-negative and the false-positive are each 5%.

Usually, in the theory of radiation measurements, the 5% errors for both false negative and false positive are considered acceptable. Under this assumption, the lowest number of counts considered statistically significant above the background is called the detection limit (DL) and is given by the formula:

By reducing the background, the detection limit can be decreased. The background can be reduced if the detector is enclosed in a shielding (e.g. a lead castle). 

When the difference between sample measurement and background (net counting rate) is bigger than the detection limit, the measurement is considered statistically significant and the result is expressed as:

The detection efficiency is determined by measuring samples of known activity (calibrated sources) under similar conditions. That is why it’s very important to have the detection system properly calibrated.

If a swipe sample is measured, the detection efficiency includes two terms:

  1. The efficiency of the detector
  2. The efficiency of the swiping process (to be on the conservative side at U of T, we consider a swiping efficiency of 10%)

The uncertainty of measurement is calculated with the formula:

Where CT is the counting time in minutes. When using an instrument set for measuring counting rates in cpm (the majority of the handheld instruments), the sample counts represents the reading of the sample, the BG counts represents the reading of the background, and CT is 1 min.

The final result should be expressed as activity ± uncertainty, in Bq/sample.

When the activity is expressed as Bq per volume of sample or per cm2, as is the case with the swipes, the volume of the sample (or the swiping surface) must also be accounted for in the denominator.

When the difference between sample measurement and background (net counting rate) is smaller than DL, the result of the measurement is NOT considered statistically significant and the result should be expressed as a minimum detectable activity (MDA).

MDA is not a characteristic of the sample measured but it is a characteristic of the instrument’s limit in detecting radioactivity. That is why results should be expressed as “less than MDA (Bq)”. Such a result indicates that the radioactivity of the sample is less than the capability of the instrument for detecting radioactivity.

Nevertheless, this kind of result can be quite informative. It will indicate whether an instrument is appropriate for specific applications. For example, an instrument’s MDA might be equal to or smaller than the release criteria for contaminated surfaces. Thus, any contaminated surface measured using this instrument cannot be considered “clean”.

Example:

Tritium contamination occurred in a laboratory. A swipe of 100 cm2 was taken and measured with a liquid scintillation counter for 1 minute.

  1. Considering a swipe efficiency of 10% and a detection efficiency of 55%. Determine the result of a measurement when the number of counts for the sample is 750 with a background of 41.
  2. Decide to clean the area and swipe again. In the same conditions, determine the result of the measurement if the new reading is now 60 counts and the new background is 39 counts in one minute.

Answer:

  1. The detection limit DL is:

The difference between the sample (750 cpm) and the background (41 cpm) is 709 cpm, which is more than 32.8 cpm. Therefore, the result is statistically significant. The activity and the uncertainty can now be calculated as:

Therefore, the result is (2.15 ± 0.09) Bq/cm2

    2. The new detection limit is:

The difference between the new sample (60 cpm) and the background (39 cpm) is now 21 cpm. Because 21 is smaller than 32, the measurement is not statistically significant. In this case, the MDA is:

Therefore, the result is expressed as “less than 0.10 Bq/cm2” (< 0.10 Bq/cm2) and indicates that the surface is not contaminated with an activity of 0.10 or more Bq/cm2.

6.2 Dose measuring instruments

To measure the radiation dose, the instrument’s response must be proportional to the energy absorbed from radiation. A “radiation flux” measuring instrument does not necessarily measure the dose. To effectively measure dose, efficiency for different types of radiation and energy must be taken into account. Pocket dosimeters, film badges, and personal thermoluminescent dosimeters are all used to measure a personal dose. They are based on the effects of the accumulated energy of radiation in the detector. Changes in the detector’s material structure can be stored until the reading process takes place. With appropriate calibration, the personal dose can be determined.

Thermoluminescence is a property of some crystals (i.e. LiF, CaF2:Mn, etc) that when high energy radiation creates electronically excited states, these states are trapped by localized defects or imperfections in the crystal’s lattice.

According to the quantum mechanic, these states are stationary and they do not have time dependence, however, they are not energetically stable. Heating the material enables the trapped states to rapidly decay into lower-energy states, causing the emission of photons in the process.

Fig. 6-5: Thermoluminiscent principle

Radiation badges are used at U of T for measuring doses received by people working with radiation and for area dosimetry. A certified (by the federal regulatory agency) company provides this service to the university. There are two kinds of radiation badges used in U of T:

  1. Whole-body dosimeter. This dosimeter should be worn somewhere on the person’s trunk.
  2. Extremity dosimeter (a ring dosimeter). This special dosimeter should be worn on the hand that is used the most during radioactive work.

Area dosimeters are identical to whole-body dosimeters and are used to measure the exposure in the area around the workplace.

All radiation badges (whole body, ring or area dosimeters) are limited to measuring high energy beta radiation and gamma/x-ray exposure. These dosimeters cannot measure exposure received from low-level energy beta emitters like H-3, C-14, S-35 or Ni-63.

Fig. 6-6: Radiation badges

Electronic dosimeters are also used for direct reading of dose. They are very useful for work in high radiation fields because of the alarm system they are equipped with. Alarms can be set for total dose, dose rate, and superficial or deep dose.

Survey meters are particle counting instruments that have been calibrated to measure the dose. They are highly specialized and can only be used for the type of radiation (X-ray, gamma-ray or neutrons) for which they have been calibrated. These instruments should never be used to measure doses outside the energy range or type of radiation for which they were calibrated. The calibration of these instruments must be done according to a procedure approved by the federal regulator.

6.3 Internal dosimetry

As was mentioned before, inhalation and ingestion are the main paths for internal irradiation of those working with radioactive materials. Therefore, special methods are required for measuring the internal irradiation of personnel. At U of T, there are two internal dosimetry programs: thyroid bioassays and urinalysis.

6.3.1 Iodine measurement

The iodine bioassay program is based on the well-known fact that iodine radionuclides used in our laboratories (I-125 and I-131) tend to accumulate inside the thyroid. Both radionuclides are gamma emitters. Therefore, a gamma detector can be used to measure the iodine content of the person’s thyroid. Proper calibration of the instrument is done using ‘phantoms’ that mimic human body composition. After gathering information about thyroid activity (in Bq) and the moment of iodine use, it is possible to estimate iodine intake and uptake. Since the level of radioiodine in the thyroid decreases after 5 days, the measurement must be done between 1 and 4 days after usage. The amount of radioiodine in the thyroid is compared with the annual limit on intake (ALI) and the dose received by the contaminated person can be estimated.

6.3.2 Urinalysis

The scientific basis for this type of analysis is the fact that most radionuclides tend to be eliminated in body fluids. By measuring activity content in urine, it is possible to estimate the intake and uptake of a specific radionuclide. The dose is estimated by comparing the intake with the ALI for that particular radionuclide.

Example:

  1. During an iodine bioassay of a person using I-125, the result of a 2-minute measurement is 1200 counts (or 600 cpm). The background for a 2-minute measurement performed immediately before was 950 counts (or 475 cpm). Knowing the detection efficiency of the instrument for I-125 to be 1.42 % for a person without fat tissue, determine the content of I-125 in the person’s thyroid.
  2. What is the content of I-125 for a different person measured immediately after the first one, if the measured value is now 980 counts (or 490 cpm)?

Answer:

  1. The detection limit for a 475 (cpm) background measurement is:

The difference between the person’s measurement (600 cpm) and the background (475 cpm) is 125 cpm. This is more than 75 cpm. Therefore, measurement is statistically significant. The activity of I-125 in the thyroid is:

The uncertainty is:

The result is: (147 ± 27) Bq

    2. The second measurement is taken immediately after the first one. The assumption is made that the background of the instrument remains the same. Therefore, the detection limit is the same DL = 150 counts for a 2-minute measurement.

The difference between the person’s measurement (490 cpm) and background (475 cpm) is now 15 cpm. This number is less than 75 cpm. Therefore, the result is NOT statistically significant. The MDA of the instrument is:

The result is expressed as “less than 88 Bq” (< 88 Bq), and the person’s thyroid is ‘clean’.

Chapter 7: RADIATION SAFETY

In recent years, extensive efforts have been made to reduce the risk of irradiation from radioactive sources at work by controlling the permitted levels of radioactivity. This was achieved by applying the three principles of radiation protection:

 JustificationAny decision that alters the radiation exposure should do more good than harm.

Optimization of ProtectionThe number of people exposed, and the magnitude of their doses should be kept as low as reasonably achievable (ALARA), taking into account economic and social factors.

Application of Dose limitsThe total dose to any individual from regulated sources in planned exposure situations, other than the medical exposure of patients, should not exceed the appropriate limits.

Note that the optimization principle requires actual operational dose limits for any work with radiation sources to be more restrictive than the maximum regulated dose limits.

7.1 Reduction of dose to personnel

Once radioactive materials are being considered as part of one’s research, an application for a radioisotope permit must be completed. During the process of obtaining the permit, the radioisotope work procedures will be examined. Other considerations include the applicant’s training, previous work experience with radioactive materials, adequacy of workplace preparation and equipment, types of dosimeters used, protective equipment, etc.

No radioactive material can be ordered until the radioactive permit has been approved. Subsequently, only those isotopes listed on the permit can be ordered within the prescribed limits. However, new isotopes can be added and/or their limits changed, provided a written request by the permit holder is sent to the Radiation Protection Service (RPS) for approval.

It is better to order radioactive materials as close as possible to the date of the experiment, from both experimental and ALARA perspectives. This will reduce the risks associated with the long-term storage of labelled materials, as well as reducing the possibility of source leakage, external irradiation, etc.

7.2 Protection against external exposure

Protection against external exposure can be done by following three fundamental methods:

  1. Time
  2. Distance
  3. Shielding

7.2.1 Time

Because radiation is roughly emitted at a constant rate from its source, the radiation dose will be proportional to the amount of time spent in proximity to the source. Therefore, the worker should try to reduce the time spent working with radioactive materials as much as possible.

Good work practice is to run a mock experiment first, without any radioactive material, to get used to the procedures. Then, perform the first experiment with the smallest amount of radioactive material that will provide a readable result. After becoming familiar with the procedures and safe handling of these materials, the quantities can be increased if necessary.

Store the bulk of the radioactive material away from the work area or keep behind shielding. Only take the necessary amount of radioactive material for the experiment on the bench, returning the stock solution to the storage area. All radioactive waste should also be kept away from the work area or stored behind shielding.

Example:

  1. Radiation is used in ferrokinetic studies of blood plasma. An injection of ferric chloride containing 18.5 MBq (0.5 mCi) of Fe-59 is to be administrated to a calf. The dose rate at the surface of the syringe is 0.40 mSv/minute. The syringe is handled for 1.5 minutes. Estimate the dose to the fingers.
  2. If the procedure can be performed in half the time noted above, then the radiation dose received will be reduced. Estimate the new dose to the fingers.

Answers:

  1. Dose = 0.40 mSv/min * 1.5 min = 0.60 mSv
  2. Dose = 0.40 mSv/min * 0.75 min = 0.30 mSv

7.2.2 Distance

As described in Chapter 4, gamma and X-ray exposure decreases with distance from the source, according to the inverse square law (for point-source). As the potential for exposure decreases, so does the potential dose.

Dose ≈ 1/r

An effective method of protection from gamma radiation is to maintain a distance as great as possible from the source. It is good practice to work with radioactive materials at arm’s length, minimizing the radiation field to the trunk of the body. When handling sources of high activity, the use of long-handled tools is required. This will reduce exposure to the hands and fingers. For high activity sources such as greater than 50 MBq (1.35 mCi) of P-32 in the research laboratory, whole-body and ring radiation badges are required.

Example:

The gamma field from a particular Co-60 source is 0.10 mSv/hr at 0.5 meters.

What is the dose at 3 meters?

Answer:

According to the inverse square law:

Dose (3 m) = Dose (0.5 m) * 0.52 / 32 = 0.10 * 0.25 / 9 = 0.0028 mSv/hr

Distance is also very effective in reducing the dose received from alpha and beta radioisotopes. The resulting decrease in dose is even greater than in the case of gamma and X-ray sources due to the absorption of alpha and beta radiation in the air.

When maintaining a sufficient distance from beta and gamma radiation is not feasible, shielding becomes necessary.

7.2.3 Shielding

As explained previously, no shielding is required for alpha radiation unless beta and/or gamma emission is also associated with the alpha emission.

When working with beta radiation sources, X-rays also appear in the shielding material due to bremsstrahlung. It is recommended to use low atomic mass number (Z) materials such as plastics (e.g. plexiglass) or Aluminium, for shielding against beta radiation. If the radionuclide has gamma emissions associated with beta radiation (e.g., I-131), protection against gamma and X-rays is also required.

Due to their weak interaction with matter, shielding against gamma and X-rays is the most difficult. High atomic mass number (Z) materials (lead, steel, or even high-density concrete) are used. The thickness of the material depends on the intensity and energy of the X- and gamma radiation (for high energies, a thicker layer is necessary).

Interaction of gamma and X-rays with the shielding material usually produces secondary radiation. A build-up factor, specific to each type of material and radiation, estimates the resulting increases in the radiation field.

A simple approach to selecting shielding against gamma and X-ray is by using the HVT (Half Value Thickness) of a specific material. This is the thickness of the material required to reduce X- and gamma radiation to half intensity. Another important value is the TVT (Tenth Value Thickness). It’s the thickness of the material that reduces the intensity of the gamma field to one-tenth of its initial value. The most commonly used material for gamma and X-ray shielding is lead. The following table lists lead HVT and TVT values for different radioisotopes used at U of T.

Lead HVT and TVT values (in cm) for different radionuclides
  Tc-99m I-131 Cs-137 Ir-192 Co-60
HVT (cm) 0.03 0.72 0.65 0.55 1.10
TVT (cm) 0.09 2.40 2.20 1.90 4.00

 

 

 

 

 

Example:

A Co-60 radiation source creates a gamma field of 20 Sv/hr. Find the lead thickness that will reduce the gamma field to 2 mSv/hr?

Answer:

TVT value for Co-60 lead is 4.0 cm.

To reduce the gamma field from 20 Sv/hr to 0.002 Sv/hr (four orders of magnitude), a layer of lead with thickness 4*TVT= 4*4.0 cm = 16.0 cm is necessary. 

The best value of the thickness of the shielding material needed can be calculated using computer codes. The Radiation Protection Services offers a web application to make these calculations that can be accessed at:

https://ehs.utoronto.ca/our-services/radiation-safety/useful-calculations/gamma-radiation-shielding-calculations/

The best shielding against neutron radiation is a material with abundant Hydrogen content such as water, paraffin, wax, or concrete.

A large enough quantity of any material can shield all types of radiation.

Always remember to check the effectiveness of your shielding first. The shielding used for one experiment may not be appropriate for the next experiment, especially when larger quantities of radioactive material or different radioisotopes are used.

7.3 Protection against internal radiation

As discussed previously, all types of radiation become more hazardous once inside the human body. The greatest internal hazard is alpha radiation, which is the least dangerous when outside the body.

Protection against internal radiation is the same as with other biological or chemical hazardous materials. Suitable handling precautions, good working procedures and protective equipment are necessary.

Eating or drinking, and even storage of food or drinks, are not permitted inside any designated radioactive laboratory. Lab coats and the use of disposable gloves are required when handling radioactive materials. Eyes or face protection (goggles or plastic face shield) is also recommended when using open sources of radioactive materials.

7.4 Radionuclides used at U of T

The Periodic Table links to some of the most commonly used radionuclides at U of T along with their half-life, disintegration mode, type and the energy of emitted radiation, necessary protective equipment, etc.

Chapter 8: REGULATORY REQUIREMENTS

8.1 Introduction

The possession, storage, use and disposal of radioactive materials is highly regulated at the international, federal, provincial, municipal and local (U of T) levels.

The United Nations established the International Atomic Energy Agency (IAEA) in 1956, with headquarters in Vienna (Austria). An important IAEA activity is to establish rules for international transport, use, and disposal of radioactive materials. However, their publications serve mostly as guidelines for each country. The agency has the power to control the enforcement of international treaties banning the use of nuclear technology for military purposes. Their safeguard inspectors also have the right to visit U of T and ask questions about the use, storage, and disposal of strategic radioactive materials (like compounds containing Pu, U and Th).

In 1928, an international committee was formed to recommend radiation protection measures and to propose limits on radiation exposure. In 1950, the committee was reorganized and named the International Commission on Radiological Protection (ICRP). Recommendations from the ICRP are used as the basis for radiation protection programs in countries all over the world.

In Canada, the federal body with regulatory powers on all aspects of radiation safety is the Canadian Nuclear Safety Commission (CNSC). It exercises control over radioactive materials through the tenets of the Nuclear Safety and Control Act and enforces its mandate through license report submissions and regular inspections and audits. The CNSC also establishes dose exposure standards for the public and nuclear energy workers all across Canada.

Other federal and provincial agencies are involved in the control of radioactive materials. These include Transport Canada, Health Canada, Environment Canada, Ontario Ministry of Health, and Ontario Ministry of Labour.

8.2 Licensing and permits

The main method for controlling the use of radioactive materials employed by the CNSC is the licencing system. The University of Toronto possesses a Nuclear Substances and Radiation Devices Licence. This type of licence allows considerable flexibility, which is required in a dynamic research centre such as U of T. It requires the University to maintain a well-managed and documented radiation protection program to ensure that radioactive materials are used safely.

The licence contains information about the federal and provincial acts and regulations for the use of radioactive materials. The Radiation Protection Service maintains copies of the licence. Any company that delivers radioisotopes to the U of T must have a copy of this licence.

8.3 University of Toronto Radiation Protection Authority – UTRPA

The University of Toronto Radiation Protection Authority (UTRPA) is a committee composed of academics and administrators who are appointed by the U of T Governing Council. The UTRPA oversees the radiation protection program at U of T. It is the responsibility of the UTRPA to establish policies and procedures to ensure the safe use of radioactive materials following federal and provincial legislation.

Under the terms of the Nuclear Substances and Radiation Devices Licence, the UTRPA is responsible for authorizing qualified individuals to use radioactive materials and has the power to issue internal Radioisotope Permits.

8.3.1 UTRPA responsibilities and duties

The UTRPA oversees all aspects of the radiation protection program at the U of T. Their control of the program is complete and all-embracing. The UTRPA has all necessary executive power delegated to it by the Governing Council to enforce and maintain the required standards of radiation protection necessary for a complex teaching and research institution.

The responsibility of the UTRPA includes all sources of ionizing and non-ionizing radiation (from both materials and machines) on all properties owned or controlled by the U of T.

The UTRPA has the power to enforce and set standards through policies, inspections and disciplinary action, if necessary. It issues, administers, reviews, and amends Radioisotope Permits.

Membership consists of academic personnel (including active permit holders) and administrative staff (see current membership). It meets regularly to discuss and implement new radiation protection initiatives in the interest of improving the overall radiation safety program.

The UTRPA maintains the radiation safety-training program to ensure that all users of radioactive materials receive appropriate training.

The UTRPA considers and advises on, the establishment of radiation emergency measures within the university, as well as co-operation and integration with other authorities (CNSC, Health Canada, etc.). 

8.3.2 RPS responsibilities and duties

The UTRPA executes its mandate through the University of Toronto Radiation Protection Service (RPS). The RPS resides administratively within the Office of Environmental Health and Safety. The function of the RPS is to carry out the daily operation of the radiation safety program, as directed by the UTRPA.

Duties of the Radiation Protection Service include:

  1. Functioning as a link between the UTRPA and radioisotope users at the University of Toronto;
  2. Reviewing the radiation safety manual, at least every two years;
  3. Having major input in matters of:
    1. facility and equipment design
    2. work practices and procedures
    3. waste storage and disposal management
    4. evaluation, issuance and enforcement of internal Permits
    5. disciplinary actions;
  4. Organizing radiation safety training;
  5. Preparing an Annual Report to the CNSC, as required by Regulatory Guide R-80 (or update);
  6. Acting as the contact for the institution concerning licencing matters with the CNSC;
  7. Being available for advice to radioisotope users on a full-time basis;
  8. Establishing, implementing and maintaining a radiation safety control and assessment program in conjunction with the UTRPA;
  9. Systematically and periodically reviewing survey programs for radiation and contamination levels in all areas where radioactive materials are found;
  10. Ensuring the proper operation of the personnel monitoring program, including the bioassay program;
  11. Ensuring that radiation safety instruments are available to the RPS in sufficient number and are calibrated and serviced as required;
  12. Conducting a quarterly review of occupational radiation exposures and recommend ways of reducing exposures in the interest of the ALARA principle;
  13. Supervising decontamination procedures, as necessary;
  14. Ensuring that waste disposal procedures are under the conditions of the radioisotope license;
  15. Ensuring that the necessary leak testing of sealed sources is performed;
  16. Controlling the purchasing, use and disposal of radioactive materials through the issuance of internal Permits and the enforcement of Permit conditions;
  17. Ensuring that appropriate radiation protection training is provided regularly for all users and for those who regularly come into contact with radioactive materials;
  18. Maintaining required records;
  19. Ensuring that each internal Permit is amended when needed by changes to facilities, equipment, policies, isotopes, conditions of use or procedures;
  20. Co-ordinating the development of plans to be used in the case of an emergency involving radioactive materials;
  21. Investigating all overexposures, accidents and losses of radioactive materials and reporting them to the CNSC, when necessary;
  22. Liaising with radioisotope users to ensure that doses of radiation are consistent with the ALARA principle.

8.4 Responsibilities and duties of the Permit Holder

The Internal Permit Holders are responsible at all times for all aspects of radiation safety in areas under their supervision. They must provide adequate training of persons under their supervision in the proper use, handling, and storage of radioactive materials.

An applicant for a radioisotope permit must be a person experienced in working with radioactive materials. He/she must be a U of T faculty member, or otherwise acceptable to the UTRPA. Work with radioactive materials must be performed on the University properties or other premises under the control and authority of the U of T. A chart listing the criteria necessary to become a permit holder, along with an application form and other necessary information for obtaining a radioisotope permit can be found at https://ehs.utoronto.ca/our-services/radiation-safety/guide-completion-radioisotope-permit-application/

After completing the form, supporting documents establishing previous work experience with radioisotopes should be attached to the form. Both the applicant and Chair of the department under which the work will be performed must sign the form, and the signed form forwarded to the RPS. The Radiation Protection Officer in charge of new radioisotope permit approval and commissioning will visit the applicant and discuss the application. After all additional information (if any) is clarified, the application is sent to the Chair of the UTRPA. After approval of the permit, purchasing of radioactive materials under that permit will be approved and work with radioactive materials can begin. At the next UTRPA meeting, all applications for new radioisotope permits will be analyzed. If more information is necessary, it will be obtained from the applicant and the permit will be amended, if necessary.

Permit Holders must:

  1. Be responsible for the safe use of radioactive materials being stored and handled in radioisotope laboratories under their supervision;
  2. Comply with all CNSC and UTRPA policies, procedures and permit conditions;
  3. Ensure that staff and students are familiar with all relevant aspects of radiation safety, procedure and policy;
  4. Provide specific training as required;
  5. Be available to supervise students and oversee the daily operation of their radioisotope laboratories;
  6. Establish specific work and storage areas for radioactive materials;
  7. Provide personal protective equipment and dose surveillance, including Thermoluminescent Dosimeters where required, functional survey meter(s), appropriate shielding, laboratory coats, gloves and other personal safety supplies;
  8. Keep an inventory of all radioactive materials and maintain a contamination monitoring program;
  9. Report incidents.

8.5 Responsibilities and duties of the radioisotopes users

Students, research associates, research assistants, technicians, etc. must:

  1. Work in compliance with all CNSC and UTRPA policies and procedures;
  2. Understand and implement all relevant aspects of radiation safety, procedure and policy;
  3. Ensure that work with radioactive materials does not create a hazard to themselves or others;
  4. Use personal protective equipment such as laboratory coats and gloves;
  5. Wear thermoluminescent dosimeters, when required;
  6. Participate in the bioassay program, where required;
  7. Perform contamination monitoring: Survey work areas and equipment for contamination after work, or at maximum within 7 days after use.
  8. Record contamination survey results;
  9. Record daily use of radioactive materials;
  10. Follow all waste disposal procedures, ensuring that there are no releases of radioactive materials to the environment;
  11. Record date of disposal;
  12. Report any defective equipment or situations that may endanger themselves or others.

Chapter 9: U OF T POLICIES, STANDARDS, AND PROCEDURES FOR RADIATION SAFETY

9.1 U of T Radiation safety policies

The ALARA concept has been adopted by the UTRPA as the basic philosophy governing the use of radioactive materials at the U of T.

The UTRPA radiation safety policies are presented in the “University of Toronto Ionizing Radiation Safety Procedures and Policies Manual”. It is the responsibility of all persons who supervise work with radioactive materials to become familiar with the information presented in this manual.

In addition to the information and requirements set out in the “Manual”, the UTRPA may require additional compliance as necessary. Each policy will be approved by the UTRPA and notification sent to each permit holder. The policies are effective upon approval by the UTRPA.

The main policies are:

  1. Disciplinary action (4-step policy)
  2. Security of radioisotope laboratories
  3. Decommissioning
  4. Laboratory decontamination
  5. Foodstuff in radioactive laboratories
  6. Counting facilities
  7. Interrupted laboratories

9.2 U of T Radiation internal standards

According to ICRP recommendations, the CNSC has established the following limits for effective dose levels in Canada:

  1. Nuclear energy worker (NEW)
    1. for a one-year dosimetry period: 50 mSv
    2. for a five-year dosimetry period: 100 mSv
  2. Pregnant NEW (for the balance of the pregnancy): 4 mSv
  3. A person who is not a nuclear energy worker (one calendar year): 1 mSv

Equivalent dose levels have also been established for various organs:

  1. The lens of the eye, NEW, one-year dosimetry period: 150 mSv
  2. The lens of an eye, any other person, one calendar year: 15 mSv
  3. Skin, NEW, one-year dosimetry period: 500 mSv
  4. Skin, any other person, one calendar year: 50 mSv
  5. Hands and feet, NEW, one-year dosimetry period: 500 mSv
  6. Hands and feet, any other person, one calendar year: 50 mSv

The UTRPA established administrative investigation levels for the dose received by a person working with radioactive materials at U of T. The role of the investigation level is to allow for intervention in preventing further exposure. Each time the radiation badge reading results are received by the RPS, they are checked against the investigation levels. Action is taken if these investigation levels are exceeded.

The investigation levels at U of T are:

  1. 4 mSv for effective whole-body dose
  2. 10 mSv for extremity dose

The bioassay program enables the RPS to immediately detect the possible intake of radioactive materials. For radio-iodine, the reporting limit set by the CNSC is 10,000 Bq from a person’s thyroid. However, at U of T, the internal administrative investigation level is set at 1,000 Bq.

Loose contamination limits, determined by swipes, are established by the CNSC for each radionuclide in controlled and public areas. These limits vary from 3 Bq/cm2 to 30 Bq/cm2 for the radioisotopes used as open sources in U of T labs. Under the ALARA principle, in U of T, any detectable loose contamination must be removed whenever possible. The minimum detectable activity (MDA) for the method used to determine the contamination must be 0.05 Bq/cm2 for alpha-emitting radionuclides, and 0.5 Bq/cm2 for all other radionuclides. If the removal is not possible, the surface should be covered to prevent the spread of radioactive contamination.

CNSC regulations require that signs indicating the presence of radioactive materials must be posted when there is a reasonable probability that a person in the area, room or vehicle, will be exposed to an effective dose rate greater than 25 mSv/h. According to ALARA, U of T standard requires that the signs should be posted at a level 10 times lower: 2.5 mSv/h (0.25 mrem/h).

9.3 U of T Radiation procedures

9.3.1 Procedures for ordering, receiving and transferring radioactive materials

9.3.1.1 Procedure for obtaining radioactive materials

The Radiation Protection Service must be notified of all radioisotope orders, transfers, and gifts before receipt. Permit holders, authorized staff and students can obtain radioactive materials for storage and use in their designated radioisotope laboratories only.

Currently, all radioisotope orders must be processed using one of the following options:

  1. the computerized Financial Information System/Administrative Management System (AMS), which is accessed by either the departmental business officer or the permit holder;
  2. the U-source, which is accessed by the permit holder/lab member.

All orders for radioactive materials submitted via the AMS and U-source systems are automatically routed to Radiation Protection Service for approval. The purchaser must provide ALL of the following information for approval of the order:

  1. permit number;
  2. permit holder’s name;
  3. radioisotope (e.g., P-32);
  4. chemical form (e.g., dCTP);
  5. total activity per vial ordered (e.g., 250 microCi);
  6. number of vials ordered (e.g., 2 vials);
  7. supplier;
  8. date of request;
  9. delivery location;
  10. name of lab member/authorized user requesting;
  11. laboratory telephone number.

The link to the purchase request form is: Purchase request

N.B.: Since the AMS and U-source systems are designed specifically for the processing of new purchases, the Radiation Protection Service must be notified of any gifts, donations, exchanges and transfers of radioactive materials by email to sandu.sonoc@utoronto.ca before receipt or transfer. All radioactive materials must be received, used, and disposed of in designated radioisotope laboratories under the same permit.

The Radiation Protection Service must be contacted for assistance whenever radioactive materials need to be transported between buildings or to external institutions. The transport of radioactive materials between buildings is strictly prohibited without prior permission from the RPS.

9.3.1.2 Receiving radioactive materials

Caution must be exercised when receiving and opening radioisotope shipments. The packing slip and label information should be compared with the original order to ensure that the correct compound has been delivered. A dose rate meter should be used to check the dose rate being emitted by the package and compare the reading with the value identified on the radiation warning labels (if applicable). The following table describes the dose rate limits for each type of radiation warning label:

Label Maximum radiation field
  In contact At 1 m
White I < 0.005 mSv/hr  
Yellow II > 0.005 mSv/hr but < 0.05 mSv/hr < 0.01 mSv/hr
Yellow III > 0.05 mSv/hr but < 2 mSv/hr < 0.1 mSv/hr

9.3.1.3 Radiation warning labels: (what they mean)

  1. White I: maximum radiation level < 0.005 mSv/hr at any location on the external surface of the package;
  2. Yellow II: maximum radiation level > 0.005 mSv/hr but < 0.5 mSv/hr on surface of the package and maximum radiation level < 0.01 mSv/hr at 1 m away from the package;
  3. Yellow III: maximum radiation level > 0.50 mSv/hr but < 2 mSv/hr on the surface of the package and maximum radiation level < 0.1 mSv/hr at 1 m away from the package.

9.3.1.4 Procedure for receiving radioisotopes packages

  1. Receive radioisotope packages in a designated radioisotope laboratory;
  2. Wear laboratory coat and gloves;
  3. If the package shows evidence of leaking, (decolouration), tampering, or if it is damaged:
    1. Inform RPS and your PH immediately;
    2. Store package in a fume-hood in a secure place;
    3. Control the spread of contamination;
    4. Identify any contaminated areas;
    5. Mark the contaminated area;
    6. Inform all lab personal about the possible area being contaminated;
    7. Clean the contaminated area;
    8. Check the effectiveness;
    9. Record the results.
  4. Check radiation dose rates (if applicable) and compare the result with the type of warning label, as well as the written value. Dose rates exceeding the described limits may suggest an incorrect shipment or leakage from the internal container;
  5. Open the package in a fume hood, if contents are volatile;
  6. Wipe test the package and radioisotope container – contact RPS if contamination is found;
  7. Avoid direct contact with the radioactive material; shield it if necessary;
  8. Compare the information on the packing slip with the container label – contact RPS if there are any inconsistencies;
  9. Confirm the receipt of the package on the web application of the EHS database;
  10. File a copy of the packing slip in the lab (inventory binder);
  11. Deface or remove radiation labels from packaging and check for contamination before disposing of. Non-contaminated packaging should be disposed of as regular waste, while contaminated packaging must be disposed of in the solid radioactive waste container. Do not put non-radioactive materials in radioactive waste containers.

9.3.2 Procedures for working with radioactive materials

General work procedures with radioactive materials are presented during the regular training delivered by the RPS to all workers before they start working with radioactive materials in U of T.

9.3.2.1 Specific work procedures

Each Permit Holder develops a specific set of working procedures. This is a condition of obtaining and keeping a radioisotope permit. The Permit Holder is responsible for training students and staff under his/her supervision in these working procedures. Any change in the procedures and/or radioisotopes (or any increase in the amount used) should be reported in writing to the RPS, with a formal request to amend the permit.

9.3.2.2 Wearing radiation badges

Persons handling radioisotopes other than H-3, C-14 and S-35, of activity of more than 1.35 mCi per vial/container, must wear whole body and ring radiation badges. The radiation badges are the primary source of information for personal exposure, as it measures the accumulated personal dose. The radiation badges are replaced and analyzed quarterly for the open-source workers and Nuclear Energy Workers (NEWs). All personal dose reports are sent to the NEWs.  

A whole-body radiation badge records the dose to the skin and body, while an extremity radiation badge records the superficial dose to the hands and extremities. The radiation badges should be stored away from sources of ionizing and UV radiation when not in use.

9.3.2.3 Bioassay requirements

Bioassays are performed on individuals to determine whether there has been an uptake of radioisotopes in the body. Users of radio-iodine (I-125 or I-131) working with activities greater than 2 MBq without containment, or greater than 200 MBq in a fume hood during 24 hours, must register for the thyroid bioassay. The bioassay measurement determines the amount of radio-iodine in the person’s thyroid. Other radionuclides are monitored for uptake by urinalysis. This type of bioassay is necessary for all users of unbound, volatile radionuclides (e.g. tritiated water).

The radioisotope and it’s biological half-life determine the frequency of bioassay monitoring:

  1. Within four days of usage of I-131 and I-125
  2. Within four days of usage for tritium users

Since the bioassay requirements for tritium depends on the chemical form of the labelled material, users of large quantities of H-3 (0.96 GBq or 26 mCi at a time) must consult with Radiation Protection Service to determine whether registration in the tritium bioassay program is necessary. All recorded uptakes are investigated to verify if safe work procedures are being followed and that the fume hood and experimental apparatus are working properly.

9.3.2.4 Actions taken to protect a pregnant worker

To protect the fetus, pregnant women working with radioactive materials should inform their supervisors in writing, indicating the expected date of birth. The supervisor will contact the university RPS and the following actions will be taken:

  1. An RSO will contact the pregnant worker and analyze the working procedures;
  2. An estimate of the dose for the remaining period of the pregnancy will be performed, with special attention to the possible internal and external irradiation of the abdomen;
  3. If a dose above 0.4 mSv is expected, a change in procedures will be recommended;
  4. If a change in procedures is not possible, it will be suggested that non-radioactive work be assigned to the worker until completion of the pregnancy;
  5. An electronic dosimeter may be issued to the worker who decides to continue to work with radioactive materials. The electronic dosimeter allows for direct reading (at any moment) of external dose;
  6. An action level is established for each specific case and communicated to the worker. In the event of reading above this action level, the worker must notify the RPS immediately.

9.3.3 Procedure for disposal of radioactive materials

Radioactive waste can be characterized into different categories (described in further detail below). All radioactive waste must be segregated and disposed of into the proper containers. All material that is determined to be contaminated should be treated and disposed of as radioactive waste.

All efforts must be made to prevent the disposal of non-radioactive waste, such as paper and packaging, into the radioactive waste stream. The Environmental Protection Services technicians provide labs with radioactive waste supplies including jars, bags, and waste tags. Any request for waste supplies should be directed to the technicians by e-mail or by phone:

eps.hazdisposal@utoronto.ca       Phone: 416 946-3473

9.3.3.1 Dry / solid radioactive waste

  1. Dispose of materials such as contaminated gloves, filter and auto-radiography paper, gels, and lead-free radioisotope source containers into the yellow solid radioactive waste container;
  2. Do not dispose of counting vials or free liquids in the solid waste container;
  3. Place broken glass and sharps into a durable box or securely wrap before disposing into the solid waste container to prevent puncture and potential injury to the Environmental Protection Services technicians;
  4. Complete the required information on the waste tag when disposing of solid waste.

9.3.3.2 Liquid waste

  1. Dispose of all liquids and buffers into the appropriate liquid waste containers;
  2. Complete the required information on the waste tag when disposing of liquid waste;
  3. Liquid radioactive waste is segregated into three categories (see below) to facilitate the “decay-in-storage” of shorter-lived radioisotope species;
  4. Dispose of mixed radioisotope liquid waste into the container designated for the longer-lived species.

9.3.3.3 Three liquid waste categories / containers

  1. for isotopes with half-life < 30 days (i.e., P-32, P-33, I-131, Cr-51)
  2. for isotopes with half-lives > 30 days but < 90 days (i.e., S-35, Fe-59, I-125)
  3. for isotopes with half-lives > 90 days (i.e., H-3, C-14, Ca-45)

9.3.3.4 Liquid scintillation vials

  1. Collect used counting vials into a durable, leak-proof cardboard box, durable bags or lined buckets;
  2. Separate plastic and glass vials;
  3. Label with radiation tape, mark as “waste vials”, and place beside the solid waste container;
  4. Vial waste does not need to be characterized for documentation purposes.

9.3.3.5 Animal carcasses and bedding

  1. Identify a freezer or storage bin in a designated location that can be used for the storage of animal carcasses and bedding;
  2. Place waste materials into durable, leak-proof plastic bags;
  3.   Complete the required information on the waste tags and notify the radiation service technicians of the storage location.

9.3.3.6 Lead radioisotope shipping pots

  1. Check lead shipping pots for contamination and deface radiation labels;
  2. Collect lead shipping pots into a durable box and label box with radiation tape – place beside solid waste container for pickup by the waste technicians;
  3. Clean the contaminated lead shipping pots and proceed like above;
  4. Lead shipping pots should be disposed of as radioactive waste only if they are contaminated and could not be cleaned.

9.3.3.7 Shipping boxes

  1. Check shipping boxes and packaging for contamination and deface radiation labels;
  2. Dispose of non-contaminated packaging as regular waste.

9.4 Laboratory compliance checklist

9.4.1 Signs, labels and housekeeping

  1. Make certain that the current radioisotope permit is posted in all designated radiation laboratories;
  2. All benches, equipment, containers and storage areas used for radioactive materials must be labelled with radiation tape or stickers;
  3. CNSC rule card must also be posted, along with the radioisotope permit, in all designated labs;
  4. The laboratory must be kept neat and tidy. Active areas for the use of radioactive materials must be free of extraneous equipment and supplies.

9.4.2 Lab classification and supervision

  1. All locations being used for handling or storing radioactive materials must be indicated on the permit;
  2. All radioisotopes in storage, and use, must be within delivery rate limits as indicated on the permit;
  3. The activity of isotopes handled on the bench and/or fume hood must be within laboratory designation limits;
  4. The permit holder or designate must be available to supervise. For any absence of more than a month, the permit holder must notify the RPS before leaving and inform the name of another permit holder who will be responsible during his/her absence.

9.4.3 Training and knowledge

  1. All staff and students must have completed the U of T Radiation Safety Course before handling radioactive materials;
  2. Radioisotope users must demonstrate adequate knowledge of safe work practices and have a clear understanding of all regulatory requirements.

9.4.4 Security

  1. Laboratories must be locked when unattended;
  2. Storage areas must be secured or locked when unattended.

9.4.5 Food prohibition

  1. Do not eat, drink, store food, smoke, or apply make-up in radioactive laboratories;
  2. There must be no evidence of food consumption or storage of food utensils or containers in designated radioisotope laboratories;
  3. There must be no disposal of food or food containers in laboratory waste receptacles.

9.4.6 Inventory

  1. All open sources of radioactive materials in use and storage must have corresponding inventory records. All new orders have an inventory record automatic generated by the EHS database;
  2. A separate inventory form must be prepared and maintained whenever an open-source is diluted, processed or separated into different products that are subsequently utilized;
  3. The radioisotope storage location and a unique identification number must be recorded on all inventory records;
  4. Daily usage, remaining quantities and final disposal dates must be recorded on the inventory forms;
  5. The Radiation Protection Service must be notified of any relocation of sealed sources or planned disposal;
  6. Inventory records must be kept for a minimum of three years.

9.4.7 Contamination control and detection criteria

  1. Documented contamination surveys must be done within seven days of work with radioactive materials;
  2. Survey locations must be identified in contamination records and include all active benches, equipment and floors;
  3. A contamination results binder must be maintained in all permitted rooms;
  4. A copy of the contamination survey results must also be kept in shared radioisotope laboratories whenever open sources of radioisotope are used in these locations;
  5. Contaminated areas must be cleaned and re-monitored. Results from contamination clean-up must be recorded both before and after decontamination;
  6. The radiation survey technique must be appropriate and adequate for the type of isotopes used (meets the threshold criteria of 0.5 Bq/cm² for beta and gamma contamination and 0.05 Bq/cm² for alpha contamination)
  7. Survey meters must be calibrated annually and must be in good working condition. Instruments should be given pre-operational checks before each use (e.g., checking the battery)(link to operational checks)
  8. Liquid scintillation counters and well-crystal gamma counters should be routinely serviced and calibrated according to the manufacturer’s specifications;
  9. Count and record a blank and standard (e.g. H-3, C-14) with each set of wipes;
  10. Monitoring records from LSC must be kept for a minimum of three years.

Dose rates due to fixed contamination that exceeds 2.5 mSv/h (0.25 mrem/h) must be posted (post the reading, the unit and the date and time of the reading)

9.4.8 Personnel dosimetry

  1. Persons handling radioisotopes other than H-3, C-14 and S-35 of activity more than 50 MBq (1.35 mCi) must wear whole-body and ring dosimeters;
  2. The radiation badges must be stored away from any source of radiation;
  3. Radiation badges must be returned to the supplier for analysis on time.

9.4.9 Lab and personnel safety

  1. Areas used for work with radioactive materials must be properly identified, contained, prepared, and sequestered whenever possible;
  2. Appropriate shielding must be available and used properly when needed;
  3. Laboratory coats, gloves and other appropriate protective equipment must be worn by radioisotope users;
  4. Dose rates from any source exceeding 2.5 mSv/h (0.25 mrem/h) must be posted;
  5. The fume hood must be functioning properly;
  6. Laboratory necessities must be readily available (absorbent pads, wipe test paper, decontamination solution, etc.)

9.4.9.1 Bioassays

  1. Persons working with radio-iodine in quantities greater than 50 MBq (1.35 mCi) must participate in the thyroid bioassay program;
  2. Persons working with more than twice ALI quantities of radionuclides at a time without containment must participate in the urinalysis bioassay program (criteria is case specific)

9.4.9.2 Radioactive waste disposal

  1. All radioactive materials must be deposited into appropriate waste containers and the required information must be recorded on the waste tags;
  2. Radioactive waste containers must be adequately shielded or stored in a location that minimizes potential exposures to all personnel;
  3. Proper procedures for waste disposal must be followed at all times (i.e., sharps boxed or wrapped before being disposed of into solid waste container, liquid waste disposed of into appropriate liquid waste containers);
  4. Radiation symbols on lead/plastic pots or radioisotope containers must be defaced when re-used for non-radioactive work;
  5. Containers re-used to store radioisotope must be re-labelled with a description of the current contents.

9.4.10 Room commissioning and decommissioning

  1. A formal written request should be sent to the RPS for the addition of a new room to the radioisotope permit;
  2. An RSO will visit the room, fill in the commissioning form and submit it for review by the building manager or equivalent, to ultimately be approved by the Senior Radiation Safety Officer;
  3. After approval, the Permit Holder will be notified and a change to the permit will be implemented before work with radioactive materials commences in the new room;
  4. A formal written request must also be submitted to the RPS for room decommissioning;
  5. An RSO will visit the room to make certain that all radioactive materials have been disposed of, confirm contamination control by swipes and direct radiation monitoring, and remove all radioactive signs;
  6. A room decommissioning form will then be filled out by the RSO and approved by the Senior Radiation Safety Officer;
  7. After approval, the Permit Holder will be notified and a change to the permit will be implemented.

9.4.11 Inventory and leak testing of sealed sources

  1. A detailed inventory of all sealed sources is kept for each permit by the Permit Holder as well as by the RSO in charge of sealed sources;
  2. Any sealed source over 50 MBq (1.35 mCi) must be tested for leakage every year. It must be tested every two years when in storage and immediately before used again.

9.4.12 Decommissioning of devices with radioactive sources

  1. A formal written request must be sent to the RPS;
  2. The RPS will arrange to remove and dispose of any radioactive source from the device;
  3. An RSO will conduct a contamination check of the device;
  4. A formal report will be sent to the Permit Holder and records of the device decommissioning will be kept in the RPS files;
  5. The permit will then be amended.

9.4.13 Decommissioning of instruments and furniture used for radioactive work

  1. A formal written request must be sent to the RPS;
  2. An RSO will conduct a contamination check of the object;
  3. A formal report will be sent to the Permit Holder and records of the decommissioning will be kept in the RPS files.

CONTINUE