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)

Comparisons between models Intake Estimate (Bq) Committed Lung Dose (m. Sv) Whole Body Committed Dose (m. Sv) ICRP 30 ICRP 66 810 1373 200 104 51 19

Human Interactions α α 10/90 rule Who is asking? “Oh my gosh” Empathy