Practical Whole Body Counting and Internal DosimetryTim Kirkham – Chesapeake Nuclear Services

Reasons for Study

α Increase knowledge of applied fundamentals.

α Enhance professionalism of radiological protection technicians.

Objectives

1.0 Describe the basic design/theory of operation of a standard stand-up whole body counter

2.0 Identify artifacts present in: γ Background Countγ QCC countγ Personnel count

3.0 Explain why Transuranic’s and Hard-To-Detects are included in internal dose calculations

4.0 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 10CFR20.1204

α Regulatory Guide 8.9 – Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program

10CFR20.1204

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; ORα (2) Quantities of radionuclides in the body; ORα (3) Quantities of radionuclides excreted from the body; ORα (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” materialβ 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 measurementsβ Baselineβ Periodicβ Termination

Regulatory Guide 8.9α Special monitoring

β High levels of facialβ Entry without controls or into an area with unknown quantities of

airborne activityβ Suspected ingestion/inhalation/woundβ Failure of respiratory device

α Estimating Intakesβ Evaluation levelsβ Investigation levels

α Type of Measurementβ In-vivoβ 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 60β New dose coefficients for intakes

α NUREG-4884 is very useful

Theory of Operation (for most standup counters)

α Two 4” x 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 Count is performed by each detectorα Signals are amplified (separately)α Signals are converted to digital (separately)α Signals are processed into a graphic spectrum (separately)α 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 softwareα 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’sβ Dose calculation could easily double due to these radionuclidesβ Include tritium as well as alpha emitters

Administration

α Background countsα QCC countsα Blind countsα Calibration

Background Counts

α Helps to ensure the counter is not contaminated (or source is present)α Sources of Background include -

β radon daughtersβ personnel in room at time of countβ noble gasesβ electronic noiseβ Induced background

α Background subtracted from personnel counts

Spectrum Artifacts

Background Spectrum

Channel

Co

un

ts

A

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

QCC Count

Channel

cou

nts

A

B

C

C

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 typesβ Japan Atomic Energy Research Institute (JAERI)

α Tissue equivalentα Some are only anthropomorphic from a given direction

Spectrum Artifacts

Worker Spectrum

Channel

Co

un

ts

B

A

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 Reportβ Work Locationβ SSNβ Start time for exposure

α Dosimetric Assessment of Personnel Contamination β Type of analysis required (In-vivo, In-vitro)

Data Gathering

α Decontaminationβ Must ensure the counter is detecting only internal contaminationβ Obtain documentation indicating decontamination (starting and

ending levels)

α Whole Body Counterβ Intake Date & Timeβ Nuclides Identifiedβ 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 emittersβ pure beta emittersβ low activity gamma emitters

Internal Dose Calculation

α Example Dose Calculationβ Gather Applicable Dataβ Identify isotopes and activityβ Scale in Hard-to-Detect radionuclidesβ Compute intake β Compute WB and Organ dose (several different models)

Comparisons between models

Intake Estimate (Bq)

Committed Lung Dose (mSv)

Whole Body Committed Dose

(mSv)

ICRP 30 ICRP 66 ICRP 30 ICRP 66 ICRP 30 ICRP 66

810 1373 200 104 51 19

Human Interactions

α 10/90 ruleα Who is asking?α “Oh my gosh”α Empathy