Appendix

Appendix A – Responsibility Chart for the Management of Radiation Safety at the University of Toronto


Appendix B – Responsibility Chart for the Administration of Radiation Safety at the University of Toronto



Appendix C – Responsibilities for Reporting to the CNSC

Appendix C1 – General Reporting Responsibilities to the CNSC

The reporting has two components: an immediate preliminary report upon becoming aware of the reportable incident (by phoning the 24 hr CNSC Duty Officer at 613-995-0479 or the toll-free # 1-844-879-0805), followed by an incident investigation 21-day report.

The immediate report must contain:

  1. the time and location of the incident,
  2. the circumstances of the situation, and
  3. any action that was taken by the University or is proposed to be taken.

The 21-day report must contain the regulatory requirements applicable to the incident.

For:  general reports, safeguards reports, and deficiency in record reports, the requirements from:

http://laws.justice.gc.ca/eng/regulations/SOR-2000-202/page-4.html#docCont must be applied.

For incidents related to radiation devices, the requirements from:

http://laws-lois.justice.gc.ca/eng/regulations/SOR-2000-207/page-5.html#docCont must be applied.

For dangerous occurrences during transportation of nuclear substances, the requirements from:

http://laws-lois.justice.gc.ca/eng/regulations/SOR-2015-145/page-8.html#h-14 must be applied.

When the dose limits are exceeded the requirements from:

http://laws.justice.gc.ca/eng/regulations/SOR-2000-203/page-3.html#docCont must be applied.


Appendix C2 – Notification to the CNSC of Use or More than 10,000 EQ

The Licensee shall obtain written approval from the Commission or a person authorized by the Commission before starting any work requiring the use of more than 10,000 exemption quantities of a nuclear substance at a single time. This is listed as: Condition 2. Project Approval: University of Toronto’s Nuclear Substances and Radiation Devices Licence.


Appendix C3 – Laboratory Classification

The laboratory classification is connected with the ALI. ALI or “annual limit on intake” means the activity, measured in becquerel, of a radionuclide that will deliver an effective dose of 20 mSv during the 50-year period after the radionuclide is taken into the body.

The Licensee shall classify each room, area or enclosure where more than one exemption quantity of an unsealed nuclear substance is used at a single time as:
(a) basic-level if the quantity does not exceed 5 ALI
(b) intermediate-level if the quantity does not exceed 50 ALI
(c) high-level if the quantity does not exceed 500 ALI
(d) containment-level if the quantity exceeds 500 ALI; or
(e) special purpose if approved in writing by the Commission or a person authorized by the Commission.

Except for the basic-level classification, the licensee shall not use unsealed nuclear substances in these rooms, areas or enclosures without the written approval of the Commission or a person authorized by the Commission. This is listed as: Condition 5: Area Classification: University of Toronto’s Nuclear Substances and Radiation Devices Licence.


Appendix C4 – Other CNSC Reporting Requirement

In case of:

  1. a major spill (spill involving more than 100 EQs, or contamination of personnel, or release of volatile material),
  2. any internal contamination of a person,
  3. if more than 1000 Bq of I-125 or 131 is measured in the thyroid,
  4. radioactive material is missing
  5. a radiation device is damaged to an extend that could impair its normal use
  6. the sealed sources is separated from the radiation device when the latter is not being serviced
  7. the sealed source fails to return to the shielded position inside the radiation device
  8. a leakage of a sealed source more than 200 Bq is identified

the SRSO or his delegate will inform the CNSC immediately, and will prepare and send a report to the CNSC within 21 days ( see: NSRD 38 (2)).

In case a package is involved in a dangerous occurrence:

  1. a conveyance caring radioactive material is involved in an accident
  2. a package shows evidence of damage, tampering or leakage
  3. any failure to comply with the CNS Act
  4. radioactive material is lost, stolen or no longer in control of the person that should have the control
  5. radioactive material have escaped from a containment system, a package or a conveyance during transport,

the SRSO or his delegate must inform the CNSC immediately, prepare and send a report to the CNSC within 21 days.

If the dose of radiation received by and committed to a person or an organ or tissue, may have exceeded the dose limits (as identified in chapter 1.1.8.1), the person must stop performing any work that is likely to add to the dose. The person may return to radioactive work only with the CNSC approval.

The SRSO or his delegate must notify the CNSC the discovery of

  1. any inaccuracy or incompletes in the Licence Documents
  2. radioactive material not contained in the licence is found
  3. change of the applicant or signing authority for the licence
  4. if a location is no longer used for licenced activities

Appendix D – Designation of Nuclear Energy Workers

The University of Toronto stresses adherence to the ALARA policy of maintaining doses As Low As Reasonably Achievable. All radiation programs are directed towards safety, ensuring that the potential for exposure is minimized. Anyone with a reasonable probability of receiving doses due to radiation greater than the limit for Members of the General Public will be designated a Nuclear Energy Worker (NEW), as defined in the Radiation Protection
Regulations.

The following documents are provided to the worker for information:

  1. Summary of the information regarding dose limitations (Radiation Protection Regulations)
  2. Health Physics Society Position Statement on Radiation Risk in Perspective
  3. Canadian Radiation Protection Association Statement on Radiation Risk

All NEWs must read and understand the information provided, acknowledging their designation by signing the following form. The approved copy of the designation form must be kept by the RPS.

 

University of Toronto                                                                                                                                                                 Radiation Protection Service
NUCLEAR ENERGY WORKER DESIGNATION (Female)                                                                                                             Effective February 1, 2016

As required by the Radiation Protection Regulations of the Canadian Nuclear Safety Commission, this information is being provided to all staff designated as “Nuclear Energy Worker”. The regulation requires the University to designate users of nuclear materials as “Nuclear Energy Workers” if there is a reasonable probability of receiving an effective dose greater than that allowed to members of the general public (1 mSv per annum whole body).

EFFECTIVE DOSE RATES FOR NUCLEAR ENERGY WORKERS GENERALLY:

Effective dose limits for Nuclear Energy Workers, including a pregnant nuclear energy worker, are 50 mSv for any one-year dosimetry period, but must not surpass 100 mSv for any 5 year dosimetry period.

EFFECTIVE DOSE RATES FOR PREGNANT NUCLEAR ENERGY WORKERS:

A pregnant nuclear energy worker must not receive an effective dose of greater than 4 mSv for the balance of the pregnancy. The balance of the pregnancy is defined as “the period from the moment a licensee is informed, in writing, of the pregnancy to the end of the pregnancy”.

A female Nuclear Energy Worker, on becoming aware that she is pregnant, must notify the permit holder and Radiation Protection Service immediately in writing. The licensee shall make any reasonable accommodation to maintain effective doses As Low as Reasonably Achievable (Radiation Protection Regulations, Section 11).

The University of Toronto stresses adherence to the ALARA policy of maintaining doses As Low As Reasonably Achievable. All radiation programs are directed towards your safety, ensuring that the potential for exposure is minimized.

The following documents are provided for your information:

      1. summary of the information regarding dose limitations (Radiation Protection Regulations)
      2. Health Physics Society Position Statement on Radiation Risk in Perspective
      3. Canadian Radiation Protection Association Statement on Radiation Risk

The Radiation Protection Service is available to answer any questions which you may have.

      1. Senior Radiation Protection Officer 416-978-2028
      2. Health and Safety Officers 416-978-6846, 416-946-3265
      3. Director of Environmental Health and Safety 416-978-5944

I have read the information provided regarding my designation as a Nuclear Energy Worker, as defined by the regulations. I understand the risks, my obligations, and the radiation dose limits that are associated with being designated a Nuclear Energy Worker.

I confirm my acceptance of this designation.
Print Name: ___________________________ Department:_____________________
Signature: ____________________________ Date: __________________________

Approved by Radiation Protection Service:
Print Name: __________________________ Title: ___________________________
Signature: ____________________________ Date: ___________________________

Print Female form

 

 

University of Toronto                                                                                                                                                                 Radiation Protection Service
NUCLEAR ENERGY WORKER DESIGNATION (Male)                                                                                                               Effective February 1, 2016

As required by the Radiation Protection Regulations of the Canadian Nuclear Safety Commission, this information is being provided to all staff designated as “Nuclear Energy Worker”. The regulation requires the University to designate users of nuclear materials as “Nuclear Energy Workers” if there is a reasonable probability of receiving an effective dose greater than that allowed to members of the general public (1 mSv per annum whole body).

EFFECTIVE DOSE RATES FOR NUCLEAR ENERGY WORKERS GENERALLY:

Effective dose limits for Nuclear Energy Workers, including a pregnant nuclear energy worker, are 50 mSv for any one-year dosimetry period, but must not surpass 100 mSv for any 5 year dosimetry period.

The University of Toronto stresses adherence to the ALARA policy of maintaining doses As Low As Reasonably Achievable. All radiation programs are directed towards your safety, ensuring that the potential for exposure is minimized.

The following documents are provided for your information:

      1. summary of the information regarding dose limitations (Radiation Protection Regulations)
      2. Health Physics Society Position Statement on Radiation Risk in Perspective
      3. Canadian Radiation Protection Association Statement on Radiation Risk

The Radiation Protection Service is available to answer any questions which you may have.

      1. Senior Radiation Protection Officer 416-978-2028
      2. Health and Safety Officers 416-978-6846, 416-946-3265
      3. Director of Environmental Health and Safety 416-978-5944

I have read the information provided regarding my designation as a Nuclear Energy Worker, as defined by the regulations. I understand the risks, my obligations, and the radiation dose limits that are associated with being designated a Nuclear Energy Worker.

I confirm my acceptance of this designation.
Print Name: ___________________________ Department:_____________________
Signature: ____________________________ Date: __________________________

Approved by Radiation Protection Service:
Print Name: __________________________ Title: ___________________________
Signature: ____________________________ Date: ___________________________

Print Male form

 


Appendix E – Table of Unit Conversions

THE RAD (rad) IS REPLACED BY THE GRAY (Gy)
1 kilorad (krad) = 10 gray (Gy)
1 rad (rad) = 10 milligray (mGy)
1 millirad (mrad) = 10 microgray (μGy)
1 microrad (μrad) = 10 nanogray (nGy)

 

THE GRAY (Gy) REPLACES THE RAD (rad)
1 gray (Gy) = 100 rad (rad)
1 milligray (mGy) = 100 millirad (mrad)
1 microgray (μGy) = 100 mikcrorad (μrad)
1 nanogray (nGy) = 100 nanorad (nrad)

 

THE REM (rem) IS REPLACED BY THE SIEVERT (Sv)
1 kilorem (krem) = 10 sievert (Sv)
1 rem (rem) = 10 millisievert (mSv)
1 millirem (mrem) = 10 microsievert (μSv)
1 microrem (mrem) = 10 nanosievert (nSv)

 

THE SIEVERT (Sv) REPLACES THE REM (rem)
1 sievert (Sv) = 100 rem (rem)
1 millisievert (mSv) = 100 millirem (mrem)
1 microsievert (μSv) = 100 microrem (μrem)
1 nanosievert (nSv) = 100 nanorem (nrem)

 

THE CURIE (Ci) IS REPLACED BY THE BECQUEREL (Bq)
1 kilocurie (kCi) = 37 terabecquerel (TBq)
1 curie (Ci) = 37 gigabecquerel (GBq)
1 millicurie (mCi) = 37 megabecquerel (MBq)
1 microcurie (μCi) = 37 kilobecquerel (kBq)
1 nanocurie (nCi) =  37 becquerel (Bq)

 

THE BECQUEREL (Bq) REPLACES THE CURIE (Ci)
1 terabecquerel (TBq) = 27 curie (Ci)
1 gigabecquerel (GBq) = 27 millicurie (mCi)
1 megabecquerel (MBq) = 27 microcurie (μCi)
1 kilobecquerel (kBq) = 27 nanocurie (nCi)
1 becquerel (Bq) = 27 picocurie (pCi)

 


Appendix F – Common Radionuclides Used in U of T

Below you can find radiation safety data sheets for the following radionuclides: H-3, C-14, P-32, P-33, S-35, Ca-45, Cr-51, Fe-59, Ni-63 and I-125. For more radiation safety data sheets please visit:

http://www.nuclearsafety.gc.ca/pubs_catalogue/uploads/Radionuclide-Information-Booklet-2018-eng.pdf

TRITIUM

3H

Radioactive half-life (T1/2) 12.4 years
Principal emission 18,6 keV beta (maximum)
Monitoring for contamination Swipes counted by liquid scintillation
Biological monitoring Urine samples
Annual Limit on Intake by ingestion or inhalation 1 x 109 Bq (17 mCi) (tritiated water)
Maximum range in air 6 mm
Shielding required None

Special Considerations for Open Sources
Tritium, because of its low beta-energy, cannot be monitored directly and therefore special care is needed to keep the working environment clean and tidy. Regular monitoring by counting swipes is advisable in areas where this nuclide is used.

Tritium can be absorbed through the skin. Volatile compounds containing tritium, tritiated water and tritium gas should be handled in a fume hood.

External contamination, although not causing a radiation dose itself, should be kept as low as possible as it can lead to internal and hence hazardous contamination; it can also interfere in experimental results.

DNA precursors (e.g. tritiated thymidine) are regarded as more toxic than tritiated water partly because activity is concentrated into cell nuclei. This is reflected by lower ALI’s for the material in this form.

Bioassays may be required for handling high amounts. Consult permit.

 

CARBON-14

14C

Radioactive half-life (T1/2) 5730 years
Principal emission 0.156 MeV beta (maximum)
Monitoring for contamination Swipes counted by liquid scintillation
Thin end-window Geiger-Müller detector
Biological monitoring Urine samples; breathe measurements (CO2)
Annual Limit on Intake by ingestion or inhalation 4 x 107 Bq (1.1 mCi)
Maximum range in air 24 cm
Shielding required  1 cm Perspex/Plexiglas. Thinner Perspex/Plexiglas down to 3 mm, although adequate to reduce doses, does not have good mechanical properties. Glass containers, although not generally recommended for shielding of beta radiation, are effective for small quantities of 14C.

Special Considerations for Open Sources
There is a possibility that some organic compounds can be absorbed through gloves.

Care needs to be taken not to generate carbon dioxide which could be inhaled.

Work with volatile compounds or those likely to generate carbon monoxide or carbon dioxide in fume hood.

 

PHOSPHORUS-32

32P

Radioactive half-life (T1/2) 14.3 days
Principal emission 1.709 Mev beta (maximum)
Monitoring for contamination Swipes counted by liquid scintillation
Geiger-Müller detector
Biological monitoring Urine samples
Annual Limit on Intake by ingestion or inhalation 1 x 107 Bq (0.27 mCi)
Maximum range in air 790 cm
Dose rate from 1 MBq (27 μCi) in 1 ml 210 mSv/h (21 rem/h) at surface
2.5 μSv/h (0.25 mrem/h) at 1 m
Shielding required Plexiglas or similar plastic (at least one cm)

Special Considerations for Open Sources
Phosphorus-32 is the highest energy beta emitting radionuclide commonly encountered in research laboratories and as such requires special care. Avoid exposure as much as possible (e.g. do not hold tubes containing even small quantities of 32P any longer than necessary – use a stand or holder).

If quantities greater than a 50 MBq (1.35 mCi) are used, whole body and ring dosimeters must be worn. The use of lead-impregnated rubber gloves is also recommended.

Even with low-density materials (for example, Perspex/Plexiglas) the absorption of the beta-particles gives rise to relatively high energy Bremsstrahlung which may require some lead shielding when quantities greater than a few hundred MBq (or tens of millicuries) are being handled.

Specific Precautions for the Handling of Phosphorus-32

Solutions containing more than 1 mCi (37 MBq) of 32P or carrier-free solutions of 32P require specific handling precautions. Carrier-free material is readily absorbed by the skin and will contribute significant doses to the bone where it is preferentially deposited. Careful handling can avoid high radiation doses to the hands while working with this material.

      • follow all general radioisotope safety precautions(Sect. 1.1)
      • double glove (disposable), changing the outer pair frequently during the procedure
      • plexiglass shielding should be used as shielding for all 32P handling and must be used with quantities in excess of 1 mCi (37 MBq). The half-value layer (HVL) thickness for 32P is 1 cm of plexiglass. Lead or other high density material may be used as secondary shielding
      • safety glasses or goggles should be used when handling 32P. This will reduce the external irradiation of the eye and skin as well as prevent the high radiation doses which accompany accidental contamination by splashing
      • ring radiation dosimeters as well as whole body dosimeters must be worn if handling quantities of 1.35 mCi (50 MBq) or larger
      • more than one person should be present during handling involving more than 1 mCi (37 MBq)
      • due to the high dose rates encountered, work should never be carried out above an open container of 32P or other high energy beta emitter

A solution of phosphate buffer is most effective in removing 32P contamination from surfaces.

 

PHOSPHORUS-33

33P

Radioactive half-life (T1/2) 25.4 days
Principal emission 0.249 Mev beta (maximum)
Monitoring for contamination Swipes counted by liquid scintillation
Geiger-Müller survey meter with pancake detector
Biological monitoring Urine samples
Annual Limit on Intake by ingestion or inhalation 80 x 106 Bq (2 mCi)
Maximum range in air 89 cm
Dose rate from 1 MBq (27 μCi) in 1 ml 30 mSv/h (3 rem/h) at surface
3.6 μSv/h (0.36 mrem/h) at 1 m
Shielding required Plexiglas or similar plastic (at least one cm)

 

Special Considerations for Open Sources
Phosphorus-33 is moderate energy beta emitting radionuclide, commonly encountered in research laboratories. Laboratory coats and gloves are the principle protection since skin dose and contamination are the primary concerns – approximately 14% of P-33 beta particles can be transmitted through the skin.

Drying can form airborne P-33 contamination.

 

SULPHUR-35

35S

Radioactive half-life (T1/2) 87.4 days
Principal emission 0.167 MeV beta (maximum)
Monitoring for contamination Swipes counted by liquid scintillation
Thin end-window Geiger-Müller detector
Biological monitoring Urine samples
Annual Limit on Intake by ingestion or inhalation 2 x 108 Bq (5 mCi)
Maximum range in air 26 cm
Shielding required  1 cm Perspex/Plexiglas. Thinner Perspex/Plexiglas down to 3 mm, although adequate to reduce doses, does not have good mechanical properties. Glass containers, although not generally recommended for shielding of beta radiation, are effective for small quantities of 35S.

Special Considerations for Open Sources
Note that organic compounds are often strongly retained and no limits of exposure have been set for them.

Be careful not to generate sulphur dioxide or hydrogen sulphide which could be inhaled.

Radiolysis of 35S-amino acids during storage and use may lead to the release of 35S-labelled volatile impurities. Handle such material in fume hood. Although the level of these impurities is small (typically less than 0.05%), contamination of the internal surfaces of storage and reaction vessels may occur. Vials should be opened and used in fume hoods.

 

CALCIUM-45

45Ca

Radioactive half-life (T1/2) 163 days
Principal emission 0.257 MeV beta (maximum)
Monitoring for contamination Swipes counted by liquid scintillation
End-window Geiger-Müller detector
Biological monitoring Urine
Annual Limit on Intake by ingestion or inhalation 1 x 107 Bq (0.27 mCi)
Maximum range in air 52 cm
Shielding required  1 cm Perspex/Plexiglas cuts out all betas. Glass containers, although not generally recommended for shielding of beta radiation, are effective for small quantities of 45Ca.

Special Considerations for Open Sources
In general Calcium-45 does not require any special precautions over and above those necessary for any beta-emitting radionuclide of this energy of emissions.

The majority of Calcium-45 is deposited in the bone: retained with a long biological half-life.

 

CHROMIUM-51

51Cr

Radioactive half-life (T1/2) 27.7 days
Principal emission 0.32 MeV gamma (9.8%); 5 keV X-ray (22% V-51 K X-ray)
Monitoring for contamination Swipes counted by liquid scintillation
End-window Geiger-Müller detector
Biological monitoring Whole body
Annual Limit on Intake by ingestion or inhalation 7 x 108 Bq (20 mCi)
Dose rate from 1 GBq (27 mCi) 4.7 μSv/h (0.47 mrem/h) point source at 1 m
First half value layer  3 mm lead

Special Considerations for Open Sources
In general Chromium-51 does not require any special precautions over and above those necessary for any radionuclide of this energy of emissions.

Chromium-51 in the form of chromate is not selectively absorbed by any organ in the body.

 

IRON-59

59Fe

Radioactive half-life (T1/2) 44.6 days
Principal emission 1.292 Mev gamma; 1.099 Mev gamma; 0.466 Mev beta (maximum); 0.273 beta (maximum)
Monitoring for contamination Swipes counted by liquid scintillation
Thin end-window Geiger-Müller detector
Biological monitoring Urine samples
Annual Limit on Intake by ingestion or inhalation 5 x 106 Bq (0.14 mCi)
Dose rate from 37 MBq (1 mCi) 6.1 μSv/h (0.61 mR/h) at 1 meter
Shielding Lead
First half value layer 9.7 mm lead

Special Considerations for Open Sources
Near an unshielded 59Fe source, dose rates from beta radiation can be much higher than dose rates due to gamma radiation.

Store 59Fe behind lead shields.

Avoid direct eye exposure by interposing transparent shields or indirect viewing.

Urinalysis to determine uptake is only effective from 4 to 24 hours after handling 59Fe.

Wear extremity and whole body dosimeters while handling more than 1.35 mCi (50 MBq) quantities.

Handle potentially volatile compounds and powder in fume hoods.

 

NICKEL-63

63Ni

Radioactive half-life (T1/2) 100 years
Principal emission 0.066 MeV beta (maximum)
Monitoring for contamination Swipes counted by liquid scintillation
Biological monitoring Urine samples
Annual Limit on Intake by ingestion or inhalation 2 x 107 Bq (0.5 mCi)
Maximum range in air 5 cm
Shielding required Perspex/Plexiglas if necessary. Glass containers, although not generally recommended for shielding of beta radiation, are effective for small quantities of 63Ni.

Special Considerations for Open Sources
Millicurie quantities of 63Ni do not represent a significant external exposure hazard since the low energy betas emitted cannot penetrate the outer skin layer.

The critical organ for 63Ni is the bone. The elimination rate of 63Ni depends on the chemical form. A few percent of most compounds taken into the body are eliminated via the urine.

Handle 63Ni compounds which are potentially volatile or in powder form in fume hoods.

Many 63Ni compounds cannot be detected with sufficient sensitivity by liquid scintillation counting (LSC) of small volume urine samples. If insoluble compounds are handled, 24-hour urine samples should be periodically collected and analyzed (LSC) to ensure that controls are adequate.

 

IODINE-125

125I

Radioactive half-life (T1/2) 59.6 days
Principal emission 35 kev gamma (7% emitted, 93% internally converted); 27.32 keV X-rays (140 % Te K X-ray)
Monitoring for contamination Swipes counted by liquid scintillation
End-window Geiger-Müller detector
Biological monitoring Thyroid scans (scintillation detector NaI)
Annual Limit on Intake by ingestion or inhalation 2 x 10Bq (54 μCi)
Dose rate from 1 GBq (27 mCi) 41 μSv/h (4.1 mrem/h) point source at 1 meter
First half value layer 0.02 mm lead

Special Considerations for Open Sources
Volatilization of iodine is the most significant problem with this isotope. Simply opening a vial of sodium [125I] iodide at high radioactive concentration can cause minute droplets of up to 100 Bq to become airborne. Solutions containing iodide ions should not be made acidic nor stored frozen: both lead to formation of volatile elemental iodine.

As some iodo-compounds can penetrate surgical rubber gloves, it is advisable to wear two pairs, or polythene (polyethylene) gloves over rubber.

In the event of suspected or actual significant contamination of personnel the thyroid should be blocked by administration of stable iodine as tablets of potassium iodate (170 mg) or potassium iodide (130 mg) which are available at hospitals.

To render any spilled Iodine-125 chemically stable the area of the spill should be treated with alkaline sodium thiosulphate solution prior to commencing decontamination. Note, however, that the quantity of radioiodine in normal RIA kits (usually <2 MBq or 54 microCi) is such that these can be handled safely with reasonable care on the open bench.

Specific Precautions for the Handling of Radioiodine

  1. follow all general radioisotope safety practices (Sect. 1.1)
  2. users of radioiodine must participate in the thyroid bioassay program (Sect. 1.1.8)
  3. background bioassay must be conducted prior to beginning use of radioiodine
  4. bioassays of the thyroid must be performed within four days after radioiodine use
  5. contact the RPS for information on this service
  6. double glove (disposable), changing the outer pair frequently during the radioiodine procedure
  7. ensure that the radioiodine container has been properly checked for leakage upon receipt
  8. vials containing radioiodine should be opened only in a fume hood, and containers of radioiodine should be kept closed when not required
  9. carry out all work involving volatile forms of radioiodine in a fume hood
    1. a properly functioning VentAlert alarm system will warn users if the fume hood does not have a proper air exhaust in the range of 100-200 linear feet per minute. Contact the RPS if there is any doubt as to the proper operation of the fume hood
    2. charcoal filtration of the exhaust may be required for large quantities of radioiodine
  10. direct contact with unshielded containers of radioiodine should be avoided
  11. shielding material of sheet lead will reduce doses received from external gamma radiation
  12. minimizing the time near radioiodine sources will reduce doses from external radiation
  13. radioactive waste contaminated with volatile radioiodine should be kept in the fume hood
  14. shielding may be necessary to reduce radiation fields near the waste
  15. radioiodine solutions with a pH of 8 or more are less likely to produce vapours
  16. during the experiment and afterwards, monitor the area with appropriate detection equipment.

A solution consisting of 0.1 M sodium iodide, 0.1 M sodium hydroxide and 0.1 M sodium thiosulphate is effective in cleaning radioiodine spills.

Wash hands immediately following a radioiodine procedure.

Contact the RPS immediately in case of spill of free radioiodine.


Appendix G – Sealed Sources Leak Test Procedure

Introduction
At the University of Toronto, a wide variety of sealed sources are used in various different applications.  These range in size from large sources containing hundreds of TBq (kCi) to very small sources of less than a few tens of kBq (µCi).  The Canadian Nuclear Safety Commission’s regulations require that leak tests are performed on most sealed sources.  The exceptions are gaseous sources, sources of tritium or any other source containing a radioactive prescribed substance which is less than 50 MBq.

This document outlines the procedures used for sampling of removable contamination and the analysis of the samples that are taken.  The procedures and equipment described in this document are intended to meet the expectations of the CNSC as outlined in the Appendix of the CNSC/NSRD Licence Application Guide (http://www.nuclearsafety.gc.ca/eng/nuclear-substances/licensing-nuclear-substances-and-radiation-devices/index.cfm).  The University of Toronto conducts leak test sampling and measurement of sealed sources under this procedure.

In this document, the sampling methods are provided followed by the measurement methods available.  A table of sources and their locations is maintained by the Radiation Protection Service. Finally, a sample copy of a completed leak test certificate is provided.

1 Sampling Procedure
1.1 General Description of the Method of Wipe Sampling
1.1.1 All samplings are to be conducted by staff of the Radiation Protection Service, unless otherwise authorized in writing. Staff of the Radiation Protection Service is familiar with the sources, their use and the hazards associated with the radiation field that may be encountered near the source. The person taken the samples must be familiar with this procedure.
1.1.2 Survey the surrounding area of the source with a suitable radiation monitor and ensure that there are no excessive radiation field readings.
1.1.3 During the sampling, the radiation monitor should be on and checked regularly to ensure that the radiation does not increase to unacceptable levels.
1.1.4 Follow the procedure of the method for wipes sampling determined by the source type. Samples are collected dry unless otherwise noted.  This conforms to the procedure used throughout the University of Toronto for contamination monitoring.
1.1.5 In order for each sample not to be contaminated by any other samples, each envelope should be clearly marked and contain only one sample.
1.1.6 After the sample is taken, the source should be returned to its proper storage position.
2 Specific Sampling Procedures
2.1 Gammacell AECL 20, 220, Nordion 1000 and Nordion 40 Exactor
2.1.1 Procedure of the Method for Wipe Sampling.
2.1.1.1 Move the source into the irradiation position using the control panel
2.1.1.2 After one minute return the source to the storage position using the control panel.
2.1.1.3 Turn off the control panel.
2.1.1.4 Open the shielding to access the sample chamber.
2.1.1.5 For swiping, use one quarter of a circular piece of filter paper 9 cm in diameter.  The filter paper is used because it removes any free particles form the surface of the source.
2.1.1.6 Hold the filter paper with a pair of tweezers.
2.1.1.7 Swipe the surface of the irradiator opening with the filter paper.
2.1.1.8 Place the filter paper into a marked envelope which will then be taken to the measuring facility.
2.1.1.9 Make the measurement looking for Co-60 (Gammacell 220) or Cs-137 (Gammacell 20, Nordion 1000 or Nordion 40 Exactor).
2.2 Gamma and Neutron Sources (Applies to sources that can be removed from container).
2.2.1 Procedure of the Method for Wipe Sampling.
2.2.1.1 Place absorbent surface liner near the source container.
2.2.1.2 Remove the source from the storage position using a pair of tweezers or a rod and place it on the absorbent surface liner.
2.2.1.3 For swiping, use one quarter of a circular piece of filter paper 9 cm in diameter.  The filter paper is used because it removes any free particles from the surface of the source.
2.2.1.4 Hold the filter paper with a pair of tweezers.
2.2.1.5 Swipe the surface of the source with the filter paper.
2.2.1.6 Place the filter paper into a marked envelope which will then be taken to the measuring facility.
2.3 Gamma Calibration Source (Applies to sources that cannot be removed from the container).
2.3.1 Procedure of the Method for Wipe Sampling.
2.3.1.1 Remove the screws which hold the shielding of the source in place and then remove the shielding.
2.3.1.2 For swiping, use one quarter of a circular piece of filter paper 9 cm in diameter.  The filter paper is used because it removes any free particles from the surface of the source.
2.3.1.3 Hold the filter paper with a pair of tweezers.
2.3.1.4 Swipe the surface of the source with the filter paper.
2.3.1.5 Place the filter paper into a marked envelope which will then be taken to the measuring facility.
2.4 Beta Calibration Source
2.4.1 Procedure of the Method for Wipe Sampling.
2.4.1.1 Place an absorbent surface liner near the source container.
2.4.1.2 Remove the source form the storage position using a pair of tweezers and place it on the absorbent surface liner.
2.4.1.3 For swiping, use one quarter of circular piece of filter paper 9 cm in diameter.  The filter paper is used because it removes any free particles from the surface of the source.
2.4.1.4 Hold the filter paper with a pair of tweezers.
2.4.1.5 Swipe the source environment with the filter paper.
2.4.1.6 Place the filter paper into a 5 ml liquid scintillation sample vial.
2.4.1.7 Place the vial into a marked envelope which will then be taken to the measuring facility.
2.5 Sources within an instrument.
2.5.1 Procedure of the Method for Wipe Sampling.
2.5.1.1  Find the orifice closest to the source.
2.5.1.2 For swiping, use one quarter of a circular piece of filter paper 9 cm in diameter.  The filter paper is used because it removes any free particles from the surface of the source.
2.5.1.3 Hold the filter paper with a pair of tweezers.
2.5.1.4 Swipe the source environment with the filter paper.
2.5.1.5 Place the filter paper into a 5 ml liquid scintillation sample vial.
2.5.1.6 Place the vial into a marked envelope which will then be taken to the measuring facility.
3 Measuring Procedures.
3.1 Gamma Spectrometry.
3.1.1 Analyzer Unit – Multichannel Analyzer
Detector Na(Tl) 3″ x 3″
High voltage: Tennelec TC 948; Position 2.49
ADC gain: 1024
Efficiency: 11.2% for Co-60 (area under 1.173 MeV photopeak only; channels: 14.5% for Cs-137 (area under 0.661 MeV photopeak only; channels: 385-485)
3.1.2 Instructions to set-up, operate and measure samples.
3.1.2.1 Turn on the system and wait 2 hours for stabilization.
3.1.2.2 Cover the detector with a thin plastic foil.
3.1.2.3 Set the timer for 600 seconds.
3.1.2.4 Verify the channels under the peak of interest using an appropriate calibration source.
3.1.2.5 Highlight the channels band corresponding to the photon’s energy.
3.1.2.6 Place the sample directly over the detector’s center.
3.1.2.7 Close the top of the shielding.
3.1.2.8 At the end of the measurement register the “gross” counts of the selected peak.
3.1.2.9 Remove the sample and close the shielding back.
3.1.2.10 Measure the background counts under the same conditions.
3.1.2.11 Subtract the background counts from the peak counts.
3.1.2.12 Estimate the activity using the efficiency of the detector.
3.1.3 Description of test using check sources
3.1.3.1 A Co-60 or Cs-137 source are used for calibrating the instrument1
3.1.3.2 The calibration source is used to check the instrument on an annual basis and after instrument repair or adjustment.
3.1.3.3 The instrument is capable of analyzing to a minimum of 12 Bq.
3.2 Liquid Scintillation Counting
3.2.1 Analyzer Unit: Liquid Scintillation Counter
Model: Perkin-Elmer TriCarb 2800 TR
Serial No. DG05072579
Efficiency: 45% for Tritium
3.2.2 Instructions to set up, operate and measure samples.
3.2.2.1 At the measuring facility, add 5 ml of the liquid scintillation cocktail2 into the vial.
3.2.2.2 Place the filled vial into the sample tray for measuring.
3.2.2.3 Set up channel A from 0 – 18.6, channel B from 0 – 156 and channel C from 0 – 2000
3.2.2.4 Set the counting time to ten minutes.
3.2.2.5 Place the sample tray containing the sample vial and control vial in the counting position and start to count.
3.2.2.6 Subtract the background counts from the sample counts.
3.2.2.7 Estimate the activity using the efficiency of the counter.
3.2.3 Description of the tests using check sources.
3.2.3.1 Use a Tritium and a Carbon – 14 calibration source to calibrate the instrument. Unquenched H-3 and C-14 samples are available for this assessment.
3.2.3.2 Calibration sources will be measured annually and a record kept for the results of the measurement
3.2.3.3 The instrument is capable of analyzing to a minimum of 10 Bq.

1   Amersham C0-60 gamma reference source IU690, 427 kBq (1 December 1981);  Amersham Cs-137 gamma reference source IS739, 391 kBq  (1 December 1981)

2   “Ultima Gold” of Perkin-Elmer or equivalent


Appendix H – Summary or Changes from the January 2017 Edition of the Manual

  • Chapter 1.1.1 paragraph 1 was changed
  • Chapter 1.1.2 – Radiation Labeling and Signs, was changed
  • Chapter 1.1.6 – Radiation Signage/Posting/Labeling, was changed
  • Chapter 1.1.9.1 – Thyroid Bioassay, was changed
  • Chapter 1.3.2.3 – Measurement of the External Radiation Field, was changed
  • Chapter 1.4 – Sealed Sources Leak Testing, was changed
  • Chapter 2 – Radioactive Waste Handling Procedures, was changed
  • Chapter 3 – Emergency Procedures, was changed
  • Chapter 4.8 – Records Management, was changed
  • Appendix C1 – General Reporting Responsibilities to the CNSC, was changed
  • Appendix G – Sealed Sources Leak Test Procedure, was added.