Practical Whole Body Counting and Internal Dosimetry Tim

Practical Whole Body Counting and Internal Dosimetry Tim Kirkham – Chesapeake Nuclear Services

Reasons for Study α Increase knowledge of applied fundamentals. α Enhance professionalism of radiological protection technicians.

Objectives 1. 0 2. 0 3. 0 4. 0 Describe the basic design/theory of operation of a standard standup whole body counter Identify artifacts present in: γ Background Count γ QCC count γ Personnel count Explain why Transuranic’s and Hard-To-Detects are included in internal dose calculations Know why internal dosimetry is required and the appropriate regulations and standards

Objectives 5. 0 Describe basic in-vitro sampling 6. 0 Explain basic Internal dose calculations 7. 0 Be able to draw conclusions concerning Human Relations as it pertains to the Dosimetry Program

Applicable Regulations and Guidance documents α 10 CFR 20. 1204 α Regulatory Guide 8. 9 – Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program

10 CFR 20. 1204 α a) For purposes of assessing dose used to determine compliance with occupational dose equivalent limits, the licensee shall, when required under § 20. 1502, take suitable and timely measurements of-α (1) Concentrations of radioactive materials in air in work areas; OR α (2) Quantities of radionuclides in the body; OR α (3) Quantities of radionuclides excreted from the body; OR α (4) Combinations of these measurements.

Part 20 α. 1204 (b) Unless respiratory protective equipment is used, as provided in § 20. 1703, or the assessment of intake is based on bioassays, the licensee shall assume that an individual inhales radioactive material at the airborne concentration in which the individual is present. α Allowed to: β Adjust DAC or ALI to actual characteristics (requires prior NRC approval) β Delay reporting if Class “Y” material β When a mixture of radionuclides exist in the air, may disregard certain nuclides in the mixture

Regulatory Guide 8. 9 α Describes practical and consistent methods acceptable to the NRC for estimating intake α References ICRP 30, ICRP 54 and NUREG-4884 α Frequency of routine measurements β Baseline β Periodic β Termination

Regulatory Guide 8. 9 α Special monitoring β High levels of facial β Entry without controls or into an area with unknown quantities of airborne activity β Suspected ingestion/inhalation/wound β Failure of respiratory device α Estimating Intakes β Evaluation levels β Investigation levels α Type of Measurement β In-vivo β In-vitro

Other “help” α ICRP-66 replaces ICRP-30 α ICRP-68 α ICRP-78 replaces ICRP-54 because of ICRP publication 68 (1994) based on ICRP 60 β New dose coefficients for intakes α NUREG-4884 is very useful

Theory of Operation (for most standup counters) α Two 4” x 16” detectors α Sodium Iodide scintillators α 4” low cobalt steel α Designed to count accurately 90% of all heights/weights

More operations stuff α α α Count is performed by each detector Signals are amplified (separately) Signals are converted to digital (separately) Signals are processed into a graphic spectrum (separately) Signals are summed

More operations stuff α Software searches according to the library, α Software searches lastly on the spectrum, α Software cannot separate all peaks so sometimes calls one peak - two radioisotopes, or call it one isotope with extra activity, α Assigns activity based upon counts, efficiency at that energy, and the gamma per second of that radionuclide

Dose calculations by the software α Calculates a dose based upon the “date of intake” input at the time of the count. α Accurate assuming no other radionuclides α Does not include HTD’s β Dose calculation could easily double due to these radionuclides β Include tritium as well as alpha emitters

Administration α α Background counts QCC counts Blind counts Calibration

Background Counts α Helps to ensure the counter is not contaminated (or source is present) α Sources of Background include β radon daughters β personnel in room at time of count β noble gases β electronic noise β Induced background α Background subtracted from personnel counts

Spectrum Artifacts

Quality Control Counts α Performed normally twice per day α Ensures the spectrum has not drifted α Compares itself to anticipated centroid locations and activities

Spectrum Artifacts

Blind Counts α Administered by a third party α Unknown radioisotopes and activities α Must meet a predefined limit to pass

Calibration α Normally annually α Energy/FWHM calibration α Efficiency calibration

Phantom α Used to simulate a body (for calibration, blind counts, etc. ) β Determines that software is calculating correctly α Many different types β Japan Atomic Energy Research Institute (JAERI) α Tissue equivalent α Some are only anthropomorphic from a given direction

Spectrum Artifacts

Accuracy α Very geometry dependent α Can increase activity by 50% α Studies/QA program indicates a ± 20% accuracy: allowed -25% to +50%. α Obviously influenced by statistics

In-Vitro bioassay α Secondary form of bioassay for most radionuclides, α Verifies HTD’s - TRU’s, electron conversions, and beta (tritium mainly) α Used mainly in DOE for TRU’s and H-3 α Can be more sensitive depending upon the radionuclide

In-Vitro Protocol α Urine samples work best if over a 24 -hour period α Fecal samples - same restriction α Gamma emitters - normally use fecal α TRU’s - normally use urine

Data Gathering α Personnel Contamination Report β Work Location β SSN β Start time for exposure α Dosimetric Assessment of Personnel Contamination β Type of analysis required (In-vivo, In-vitro)

Data Gathering α Decontamination β Must ensure the counter is detecting only internal contamination β Obtain documentation indicating decontamination (starting and ending levels) α Whole Body Counter β Intake Date & Time β Nuclides Identified β Activity for each nuclide

Data Gathering α Air samples α Swipes α Lapel

Scaling for HTD’s α Based upon “easy-to-detects” α Normally Co-60 or Cs-137 α Used for: β alpha emitters β pure beta emitters β low activity gamma emitters

Internal Dose Calculation α Example Dose Calculation β Gather Applicable Data β Identify isotopes and activity β Scale in Hard-to-Detect radionuclides β Compute intake β Compute WB and Organ dose (several different models)