PEP 2C BASIC TRAINING FOR THE NRRPT EXAM - PRACTICAL APPLICATIONS Paul Steinmeyer Tom Voss 1
PEP 2CBASIC TRAINING FOR THE
NRRPT EXAM - PRACTICAL APPLICATIONS
Paul SteinmeyerTom Voss
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BASIC TRAINING FOR THE NRRPT EXAM
Basic Training for the NRRPT Exam - Theory February 1, 2015 8:00 AM - 10:00 AM (ET) HPS Mid-Year Meeting
Basic Training for the NRRPT Exam - Practical Applications February 1, 2015 10:30 AM - 12:30 PM (ET) HPS Mid-Year Meeting
Basic Training for the NRRPT Exam - Review of the Applicable CFRs February 1, 2015 2:00 PM - 4:00 PM (ET) HPS Mid-Year Meeting
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A Radiation Protection Technologist is a person engaged in providing radiation protection to the radiation worker, the general public, and the environment from the effects of ionizing radiation.
The Radiation Protection Technologist has a basic understanding of the natural laws of ionizing radiation, the mechanism of radiation damage, methods of detection, and hazards assessment.
The Radiation Protection Technologists' tasks are accomplished by providing supervisory, administrative, and/or physical control, utilizing sound health physics principles in compliance with local and statutory requirements and accepted industry practices.
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The Radiation Protection Technologist mitigates hazards associated with radioactive material and ionizing radiation producing devices, always adhering to the "as low as reasonably achievable" philosophy.
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The 150 question exam covers broad-based radiation protection knowledge of; Accelerators University Health Physics Medical Health Physics Power Reactors Government Radiological Facilities Radioactive Waste Disposal Transportation of Radioactive Material Fundamentals Regulatory Requirements
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The examination consists of 150 multiple choice questions from three general categories identified from a role delineation/task analysis conducted by the NRRPT Board and Panel: Applied Radiation Protection, Detection and Measurements, and Fundamentals. Four hours are allowed for the examination. Contents of past examinations are not released. As the domains may share some common "required knowledge," a general outline has been developed to assist candidates in preparation for the exam.
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Applied Radiation Protection o Survey and Inspectionso Emergency Preparednesso Evaluating Internal and External Exposures and
Controlso Prescribed Dosimetry and Radiation Equipmento Contamination Controlo Radioactive Material Control and Transportationo Guides and Regulationso Procedures and Programs (ALARA)
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Detection and Measurement o Analytical Methodso Instrument Calibration and Maintenanceo Personnel Dosimetryo Equipment Operation
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Fundamentals o Source of Radiationo Biological Effectso Mathematicso Chemistryo Physicso Units and Terminology
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SUGGESTED STUDY MATERIAL Cember, H. 1996. “Introduction to Health Physics,” 3rd edition Gollnick, D. 1994. “Basic Radiation Protection Technology,” 3rd edition Moe, H., et al. 1988. Department of Energy Operational Health Physics Training.” ANL-88-26 US Department of Health, Education and Welfare. 1970. “Radiological Health Handbook.” Turner, J E. 1995. “Atoms, Radiation, and Radiation Protection,” 2nd edition Shleien, B. (revised by). 1992. “The Health Physics and Radiological Health Handbook,” Chart of the Nuclides
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The most recent revisions of: 10 CFR 19, “Notices, Instructions and Reports to Workers; Inspections.” 10 CFR 20, “Standards for Protection Against Radiation.” 10 CFR 30, “Rules of General Applicability to Domestic Licensing of Byproduct Material.” 10 CFR 34, “Licenses for Radiography and Radiation Safety Requirements for Radiographic Operations.” 10 CFR 35, “Medical Use of Byproduct Material.” 10 CFR 835, “Occupational Radiation Protection,” DOE 49 CFR 100-199, “Transportation.”
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PEP 2C - Basic Training for the NRRPT Exam – Practical Applications
Unit Analysis and Conversion 13Counting Statistics 40Sources of Radiation 99Medical Radiation Sources 139Radioactive Source Control 159External Exposure Control 168Contamination Control 211Internal Exposure Control 234Air Monitoring 274Respiratory Protection 314ALARA 346Isotopes Good to Know 378Biological Effects Definitions 379Typical NRRPT Exam Questions 381 12
Unit Analysis & Conversion
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UNITS AND MEASUREMENTS
Units are used in expressing physical quantities or measurements, i.e., length, mass, etc.
All measurements are actually relative in the sense that they are comparisons with some standard unit of measurement.
Two items are necessary to express these physical quantities: a number which expresses the magnitude and a unit which expresses the dimension.
A number and a unit must both be present to define a measurement.
Measurements are algebraic quantities and as such may be mathematically manipulated subject to algebraic rules.
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Fundamental Quantities
All measurements or physical quantities can be
expressed in terms of three fundamental quantities.
They are called fundamental quantities because they
are dimensionally independent.
They are:
• Length (L)
• Mass (M) (not the same as weight)
• Time (T)
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Fundamental Quantities
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Derived Quantities
Other quantities are derived from the fundamental
quantities.
Derived quantities are formed by multiplication and/or
division of fundamental quantities.
For example:
• Area is the product of length times length (width), which
is L × L, or L2.
• Volume is area times length, which is length times
length times length, or L3.
• Velocity is expressed in length per unit time, or L/T.
• Density is expressed in mass per unit volume, or M/L3.
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English System
The system that has historically been used in the United
States is the English System, sometimes called the
English Engineering System (EES).
Many of the units in this system have been used for
centuries and were originally based on common objects
or human body parts, such as the foot or yard.
The base units for length, mass, and time in the English
system are the foot, pound, and second, respectively.
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International System of Units (SI)
International System of Units (abbreviated SI) was adopted by the 11th General Conference of Weights and Measures (CGPM).
The SI, or modernized metric system, is based on the decimal (base 10) numbering system.
First devised in France around the time of the French Revolution, the metric system has since been refined and expanded so as to establish a practical system of units of measurement suitable for adoption by all countries.
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The SI system consists of a set of specifically defined units and prefixes that serve as an internationally accepted system of measurement.
Nearly all countries in the world use metric or SI units for business and commerce as well as for scientific applications.
SI Prefixes
The SI system is completely decimalized and uses
prefixes for the base units of meter (m) and gram (g), as
well as for derived units.
SI prefixes are used with units for various magnitudes
associated with the measurement being made.
Units with a prefix whose value is a positive power of
ten are called multiples.
Units with a prefix whose value is a negative power of
ten are called submultiples.
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SI Prefixes
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SI Units
There are seven fundamental physical quantities in the SI system.
– Length– Mass– Time– Temperature– Electric charge– Luminous intensity– Molecular quantity (or amount of substance)
In the SI system there is one SI unit for each physical quantity.
The SI system base units are those in the metric MKS system.
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Radiological Units
The SI unit of activity is the becquerel, which is the
activity of a radionuclide decaying at the rate of one
spontaneous nuclear transition per second.
The gray is the unit of absorbed dose, which is the
energy per unit mass imparted to matter by ionizing
radiation, with the units of one joule per kilogram.
The unit for dose equivalence is the sievert, which has
the units of joule per kilogram.
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UNIT ANALYSIS AND CONVERSION PROCESS
Units and the Rules of Algebra
A measurement consists of a number and a unit.
Measurement units are subject to the same algebraic
rules as the values.
Measurements can be multiplied, divided, etc., in order
to convert to a different system of units.
In order to do this, the units must be the same.
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Steps for Unit Analysis and Conversion
1) Determine given unit(s) and desired unit(s).
2) Build (or obtain) conversion factor(s)
A conversion factor is a ratio of two equivalent physical
quantities expressed in different units. When expressed
as a fraction, the value of all conversion factors is 1.
Because a conversion factor equals 1, it does not
matter which value is placed in the numerator or
denominator of the fraction.
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Steps for Unit Analysis and Conversion
Also remember that algebraic manipulation can be used
when working with metric prefixes and bases.
3) Set up an equation by multiplying the given units by
the conversion factor(s) to obtain desired unit(s).
When a measurement is multiplied by a conversion
factor, the unit(s) (and probably the magnitude) will
change; however, the actual measurement itself does
not change.
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An air sampler is operated for 1 week at 2 CFM. The sample filter is counted and indicates net 140 CPM. The detector efficiency is 25%.
What is the average airborne concentration for the one week period in uCi/mL and Bq/M3 ?
First Find the total volume of air sampled. Calculate the DPM on the filter. Divide the DPM by the total volume of air sampled.
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Total volume of air sampled
1 week is 7 days at 24 hours per day and there are 60 minutes in one hour. Sample time = 7 days x 24 hours/day x 60 minutes/hour. 7 days x 24 hours/day x 60 minutes/hour days cancel day and hours cancels hour 7 x 24 x 60 minutes = 10,080 minutes
Sampling rate is 2 CFM (cubic feet per minute) Total volume of air is 2 cubic feet/minute x 10,080 minutes 2 cubic feet/minute x 10,080 minutes minutes cancel minutes 2 cubic feet x 10,080 = 20,160 cubic feet 29
Convert cubic feet to mL (remember 28,316 cc/cubic foot) 28,316 mL/cubic foot x 20,160 cubic feet cubic feet cancels cubic feet 28,316 mL x 20,160 = 570,850,560 mL = 5.71E8 mL
Calculate DPM on the filter 140 CPM/0.25 CPM/DPM CPM cancels CPM 140/0.25/DPM = 560 DPM (since the DPM is / / that unit goes on top of the calculation)
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Calculate uCi/mL DPM/mL = 560 DPM/5.71E8 mL Convert DPM to uCi (remember 1 uCi is 2.22 E6 DPM) 560 DPM/2.22 E6 DPM/uCi DPM cancels DPM 560/2.22 E6/uCi = 2.52 E-4 uCi (since the uCi is / / that unit goes on top of the calculation) Divide 2.52 E-4 uCi by 5.71 E8 mL to get Average Airborne Concentration for the week 2.52 E-4 uCi / 5.71 E8 mL = 4.41 E-13 uCi/mL
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Calculate the average airborne concentration in Bq/M3 We could start from the beginning and convert CFM to M3/minute and DPM to Bq. But we can work with our calculated uCi/mL 1 M3 is 1 E6 mL (just memorize this) Total volume is 5.71 E8 mL Divide 5.71 E8 mL by 1 E6 mL/M3 to get total M3
5.71 E8 mL / 1 E6 ml/M3
mL cancels mL and the M3 goes on top Total volume is 5.71 E2 M3
2.52 E-4 uCi x 3.7 E10 Bq/Ci Change 3.7 E10 Bq/Ci to 3.7 E4 Bq/uci 2.52 E-4 uCi x 3.7 E4 Bq/uCi uCi cancels uCi
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2.52 E-4 x 3.7 E4 Bq = 9.32 Bq Now calculate the airborne concentration in Bq/M3 9.32 Bq/5.71 E2 M3 = 1.63 E-2 Bq/M3
Does this make sense ? is 1.63 E-2 Bq/M3 the same concentration as 4.41 E-13 uCi/mL ?
What is the average airborne concentration in the number of DACs if the DAC factor (uCi/mL) is 2 E-12 uCi/mL ?
4.41 E-13 uCi/mL / 2 E-12 uCi/mL / DAC uCi/mL cancels uCi/mL and DAC goes on top 4.41 E-13 / 2 E-12 = 0.221 DAC
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This value is 22 % of a DAC. If the area is to be continuously occupied how should it be posted ?
Unless there are other radiological hazards present this area should only be posted as a Controlled Area. The 10CFR20 definition of an Airborne Area is an area where the airborne concentration could (1) exceed the derived air concentration limits (DACs), or (2) would result in an individual present in the area without respiratory protection exceeding, during those hours, 0.6 percent of the ALI or 12 DAC-hours.
Calculate the DAC-hours if a person were in this area without respiratory protection continuously for 40 hours in that week.
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Calculate the DAC-hours for 40 hours in 0.221 DAC concentration. 0.221 DAC x 40 hours = 8.84 DAC-hours
Calculate an individuals exposure in mRem if they were in that area and inhaled 8.84 DAC-hours. 2.5 mRem/DAC-hour x 8.84 DAC-hours = 22.1 mRem
NOTE: A typical Alpha CAM alarm setpoint is 8 DAC-hours and filters are routinely changed once per week. Since the average concentration is 0.221 DAC and one week is 168 hours the accumulated DAC-hours on the CAM filter would be 37.1 DAC-hours. The alpha CAM would have reached 8 DAC-hours before the end of the second day.
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CAUTION: This individual COULD exceed their periodic limit (weekly, monthly, quarterly, annually) if their exposure continued. Also, any external exposure needs to be considered.
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TEMPERATURE MEASUREMENTS AND
CONVERSIONS
Temperature measurements are made to determine the
amount of heat flow in an environment.
To measure temperature it is necessary to establish
relative scales of comparison.
Three temperature scales are in common use today.
• Fahrenheit
• Celsius
• Kelvin
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Comparison of Kelvin, Celsius and Fahrenheit
Scales
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Equations for Temperature Conversions
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COUNTING ERRORS
AND STATISTICS
Voss Associates
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Assuming the counting system is calibrated correctly,
there are five general sources of error associated with
counting a sample:
1. Self-absorption
2. Backscatter
3. Resolving time
4. Geometry
5. Random disintegration of radioactive atoms (statistical
variations).
GENERAL SOURCES OF ERRORS
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Self-Absorption
When a sample has an abnormally large amount of material
on the sample media, it could introduce a counting error due
to self-absorption, which is the absorption of the emitted
radiation by the sample material itself. Self-absorption
could occur for:
• Liquid samples with a high solid content
• Air samples from a high dust area
• Use of improper filter paper may introduce a type of self-
absorption, especially in alpha counting (i.e., absorption by
the media, or filter).
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Backscatter
Counting errors due to backscatter occur when the
emitted radiation traveling away from the detector is
reflected, or scattered back, to the detector by the
material in back of the sample.
The amount of radiation that is scattered back will
depend upon the type and energy of the radiation and
the type of backing material (reflector).
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The amount of backscattered radiation increases as the
energy of the radiation increases and as the atomic
number of the backing material increases.
Generally, backscatter error is only a consideration for
particulate radiation, such as beta particles.
The ratio of measured activity of a beta source counted
with a reflector compared to counting the same source
without a reflector is called the backscatter factor (BF).
Resolving Time
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Resolving time is the time interval which must elapse
after a detector pulse is counted before another full-size
pulse can be counted.
Any radiation entering the detector during the resolving
time will not be recorded as a full size pulse; therefore,
the information on that radiation interaction is lost.
As the activity, or decay rate, of the sample increases, the
amount of information lost during the resolving time of the
detector is increased.
As the losses from resolving time increase, an additional
error in the measurement is introduced.
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R is the true count rate
R0 is the observed count rate
t is the resolving time of the detector
GEOMETRY
Geometry related counting errors result from the positioning
of the sample in relation to the detector. Normally, only a
fraction of the radiation emitted by a sample is emitted in the
direction of the detector because the detector does not
surround the sample.
If the distance between the sample and the detector is
varied, then the fraction of emitted radiation which hits the
detector will change. This fraction will also change if the
orientation of the sample under the detector (i.e., side-to-
side) is varied.
An error in the measurement can be introduced if the
geometry of the sample and detector is varied from the
geometry used during instrument calibration.47
Random Disintegration
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Gaussian Distribution
Also called the "normal distribution," the Gaussian is
applicable if the average number of successes is
relatively large, but the probability of success is still low.
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Note that the highest number of successes is at the center
of the curve, the curve is a bell shaped curve, and the
relative change in success from one point to the adjacent
is small.
Also note that the mean, or average number of
successes, is at the highest point, or at the center of the
curve.
The Gaussian, or normal, distribution is applied to
counting applications where the mean success is
expected to be greater than 20.
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Poisson Distribution
The Poisson distribution is valid when the probability of
success, P(x), is small.
If we graphed a Poisson distribution function, we would
expect to see the predicted number of successes at the
lower end of the curve, with successes over the entire range
if sufficient trials were attempted.
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DEFINITIONS
Mode -- An individual data point that is repeated the
most in a particular data set.
Median -- The center value in a data set arranged in
ascending order.
Mean -- The average value of all the values in a data
set.
Average --“Mean”
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DETERMINATION OF MODE, MEDIAN, AND MEAN
• Determination of the Mode: In a set of numbers the number
that is repeated more often than any other is the Mode.
• Determination of the Median: In the same set numbers
where one half are below and the other half are above is
the Median.
• Determination of the Mean (Average): This is found by
adding all of the numbers in a set together, and dividing by
the number of values in the set.
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VARIANCE AND STANDARD DEVIATION
Using the Gaussian distribution we need to define
the terms "variance" and "standard deviation“.
The amount of scatter of data points around the mean is
defined as the sample variance.
It tells how much the data "varies" from the mean.
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Mathematically the standard deviation is the square root of
the variance.
A term more precise than the variance is standard deviation,
represented by σ (“sigma”).
The standard deviation of a population is defined
mathematically as:
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If most of the data points are located close to the mean
(average), the curve will be tall and steep and have a low
numerical value for a standard deviation. If data points are
scattered, the curve will be lower and not as steep and
have a larger numerical value for a standard deviation.
In a Gaussian distribution, it has been determined
mathematically that 68.2% of the area under the curve falls
within the data point located at the mean ± (plus or minus)
one standard deviation (1σ); 95.4% of the area under the
curve falls between the data point located at ± two
standard deviations (2σ); 99.97% of the area under the
curve falls between the data point located at ± three
standard deviations (3σ).
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CHI-SQUARED TEST
The Chi-squared test (pronounced "ki") is used to determine
the precision of a counting system.
Precision is a measure of exactly how a result is determined
without regard to its accuracy.
It is a measure of the reproducibility of a result, or in other
words, how often that result can be repeated, or how often a
"success" can be obtained.
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This test results in a numerical value, called the Chi-
squared value (Χ2), which is then compared to a range of
values for a specified number of observations or trials.
This range represents the expected (or predicted)
probability for the chosen distribution.
If the Χ2 value is lower than the expected range, this tells
us that there is not a sufficient degree of randomness in
the observed data.
If the value is too high, it tells us that there is too much
randomness in the observed data.
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The Chi-squared test is often referred to as a "goodness-
of-fit" test.
It answers the question: How well does this data fit a
distribution curve?
If it does NOT fit a curve indicating sufficient randomness,
then the counting instrument may be malfunctioning.
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QUALITY CONTROL CHARTS
Quality control charts are prepared using source counting
data obtained during system calibration.
The source used for daily checks should be identical to the
one used during system calibration.
Obviously since this test verifies that the equipment is still
operating within an expected range of response, we cannot
change the conditions of the test in mid-stream.
QC charts, then, enable us to track the performance of the
system while in use.
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Data that can be used for quality control charts include gross
counts, counts per unit time, and efficiency.
Most nuclear laboratories use a set counting time
corresponding to the normal counting time for the sample
geometry being tested.
If samples are counted for one minute, then all statistical
analysis should be based on one-minute counts.
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When the system is calibrated and the initial
calculations performed, the numerical values of the
mean ± 1, 2, and 3 standard deviations are also
determined.
Using standard graph paper, paper designed
specifically for this purpose, or a computer
graphing software, lines are drawn all the way
across the paper at those points corresponding to
the mean, the mean plus 1, 2, and 3 standard
deviations, and the mean minus 1, 2, and 3
standard deviations.
The mean is the center line of the paper.
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SYSTEM OPERATING LIMITS
The values corresponding to ±2 and ±3 standard
deviations are called the upper and lower warning and
control limits, respectively.
The results of the daily source counts are graphed daily
in many countrooms.
Most of the time our results will lie between the lines
corresponding to ±1 standard deviation (68.2%).
We also know that 95.4% of the time our count will be
between ±2 standard deviations and that 99.97% of the
time our count will be between ±3 standard deviations.
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Counts that fall outside the warning limit (±2σ) are not
necessarily incorrect.
Statistical distribution models say that we should get
some counts in that area.
Counts outside the warning limits indicate that
something MAY be wrong.
The same models say that we will also get some
outside the control limits (±3σ).
However, not very many measurements will be outside
those limits.
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We use 3σ as the control – a standard for acceptable
performance.
In doing so we say that values outside of ±3σ indicate
unacceptable performance, even though those values may
be statistically valid.
The typical response is to look for problems if the data is
outside the 2 sigma warning and to repair and/or
recalibrate if the data is outside the 3 sigma control level.
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True randomness also requires that there be no patterns in
the data that are obtained; some will be higher than the
mean, some will be lower, and some will be right on the
mean.
When patterns do show up in quality control charts, they
are usually indicators of systematic error.
For example:
• Multiple points outside two sigma
• Repetitive points (n out of n) outside one sigma
• Multiple points, in a row, on the same side of the mean
• Multiple points, in a row, going up or down.
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COUNTER EFFICIENCY
A detector intercepts and registers only a fraction of the total
number of radiations emitted by a radioactive source.
The major factors determining the fraction of radiations
emitted by a source that are detected include:
• The fraction of radiations emitted by the source which travel
in the direction of the detector window
• The fraction emitted in the direction of the detector window
which actually reach the window.
• The fraction of radiations incident on the window which
actually pass through the window and produce an ionization
• The fraction scattered into the detector window
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The detector efficiency gives us the fraction of counts
detected per disintegration, or c/d.
Since activity is the number of disintegrations per unit
time, and the number of counts are detected in a finite
time, the two rates can be used to determine the
efficiency if both rates are in the same units of time.
Counts per minute (cpm) and disintegrations per
minute (dpm) are the most common with counts per
second (cps) and Bq (disintegrations per second)
becoming more common in the US.
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Thus, the efficiency, E, can be determined as shown
The efficiency obtained in the formula above can be
expressed in decimal or % form.
To calculate the percent efficiency, the value is multiplied
by 100.
For example, an efficiency of 0.25 would mean 0.25 × 100,
or 25%.
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ERROR CALCULATIONS
The error present in a measurement governed by a statistical
model can be calculated using known parameters of that
model.
Nuclear laboratories are expected to operate at a high degree
of precision and accuracy.
However, since we know that there is some error in our
measurements, we are tasked with reporting measurements
to outside agencies in a format that identifies that potential
error.
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ERROR CALCULATIONS
The format that is used should specify the activity units
and a range in which the number must fall.
In other words, the results would be reported as a given
activity plus or minus the error in the measurement.
Since nuclear laboratories prefer to be right more than
they are wrong, counting results are usually reported in a
range that would be correct 95% of the time, or at a 95%
confidence level.
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In order to do this, the reported result should be in this format:
For example, a measurement of 150 ± 34 dpm (2σ) indicates
the activity as 150 dpm; however, it could be as little as 116
dpm or as much as 184 dpm with 95% confidence (at 2σ).
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The calculations of the actual range of error is based on
the standard deviation for the distribution.
In the normal (or Gaussian) distribution, the standard
deviation of a single count is defined as the square root
of the mean, or σ = x.
The error, e, present in a single count is some multiplier,
K, multiplied by the square root of that mean, i.e., some
multiple times the standard deviation, Kσ.
The value of K used is based on the confidence level
that is desired, and is derived from the area of the curve
included at that confidence level.
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Common values for K are:
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BACKGROUND
Determination of Background
Radioactivity measurements cannot be made without
consideration of the background.
Background, or background radiation, is the radiation that
enters the detector concurrently with the radiation emitted
from the sample being analyzed.
This radiation can be from natural sources, either external
to the detector (i.e., cosmic or terrestrial) or radiation
originating inside the detector chamber that is not part of
the sample.
In practice, the total counts are recorded by the counter.
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This total includes the counts contributed by both the sample
and the background.
Therefore, the contribution of the background will produce
an error in radioactivity measurements unless the
background count rate is determined by a separate
operation and subtracted from the total activity, or gross
count rate.
The difference between the gross and the background rates
is called the net count rate.
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This relationship is seen in the following equation:
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The background is determined as part of the system
calibration by counting a background (empty) planchet for
a given time.
The background count rate is determined in the same way
as any count rate, where the gross counts are divided by
the count time.
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Reducing Background
Typically, the lower the system background the more reliable
the analysis of samples will be.
In low-background counting systems the detector housing is
surrounded by lead shielding so as to reduce the background.
On many systems a second detector is incorporated to detect
penetrating background radiation.
When a sample is analyzed the counts detected by this
second detector during the same time period are internally
subtracted from the gross counts for the sample.
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Background originating inside the detector chamber can
be more easily controlled.
The main contributors of this type of background are:
• Radiation emitted from detector materials
• Radioactive material on inside detector surfaces
• Radioactive material on the sample slide assembly
• Contamination in or on the sample planchet or
planchet carrier
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PROPAGATION OF ERROR
The error present in a measurement includes the error
present in the sample count, which contains both sample
and background, and the error present in the
background count.
Rules for propagation of error preclude merely adding
the two errors together.
The total error in the measurement is calculated by
squaring the error in the background and adding that to
the square of the error in the sample count, and taking
the square root of the sum.
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Since we normally use this equation in terms of a count
rate, the formula is slightly modified as follows, and the
error stated as the sample standard deviation (σS):
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If the sample counting time and the background counting
time is the same, the formula can be simplified even
more to:
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IMPROVING STATISTICAL VALIDITY
OF COUNT ROOM MEASUREMENTS
Minimizing the statistical error present in a single sample
count is limited to several options.
If we look at the factors present in the calculation below, we
can see that there are varying degrees of control over these
factors.
The standard deviation is calculated here in terms of count
rate.
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RS+B is the sample count rate. We really have no control
over this. RB is the background count rate. We do have
some control over this.
On any counting equipment the background should be
maintained as low as possible. In most of our counting
applications, however, the relative magnitude of the
background count rate should be extremely small in
comparison to the sample count rate if proper procedures
are followed. This really becomes an issue when counting
samples for free release or environmental samples.
However, some reduction in error can be obtained by
increasing the background counting time.
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TB and TS are the background and sample counting times,
respectively.
These are the factors that we have absolute control over.
In the previous section we talked about the reliability of the
count itself.
We have been able to state that a count under given
circumstances may be reproduced with a certain
confidence level, and that the larger the number of counts
the greater the reliability.
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The total counting time required depends upon both the
sample and background count rates.
For high sample activities the sample count time can be
relatively short compared to the background count time.
For medium count rates we must increase the sample count
time in order to increase precision.
As the sample activity gets even lower, we approach the case
where we must devote equal time to the background and
source counts. By counting low activity samples for the
same amount of time as that of the background
determination, we increase the precision of our sample
result.
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DETECTION LIMITS
The detection limit of a measurement system refers to the
statistically determined quantity of radioactive material (or
radiation) that can be measured (or detected) at a
preselected confidence level. This limit is a factor of both the
instrumentation and technique/procedure being used. The
two parameters of interest for a detector system with a
background response greater than zero are:
LC Critical detection level: the response level at which the
detector output can be considered "above background"
LD Minimum significant activity level, i.e., the activity level
that can be seen with a detector with a fixed level of certainty
90
Two types of statistical counting errors must be considered
quantitatively in order to define acceptable probabilities for
each type of error:
Type I - occurs when a detector response is considered
above background when in fact it is not (Type I errors are
associated with LC)
Type II - occurs when a detector response is considered to
be background when in fact it is greater than background
(Type II errors are associated with LD)
52
91
92
If the two probabilities (areas labeled I and II) are assumed
to be equal, and the background of the counting system is
not well-known, then the critical detection level (LC) and the
minimum significant activity level (LD) can be calculated.
The two values would be derived using the equations LC =
kσB and LD = k2 + 2kσB, respectively.
If 5% false positives (Type I error) and 5% false negatives
(Type II error) are selected to be acceptable levels, i.e.,
95% confidence level, then k = 1.645 and the two equations
can be written as:
93
The minimum significant activity level, LD, is the a priori
(before the fact) activity level that an instrument can be
expected to detect 95% of the time. In other words, it is the
smallest amount of activity that can be detected at a 95%
confidence level. When stating the detection capability of an
instrument, this value should be used.
94
The critical detection level, LC, is the lower bound on the
95% detection interval defined for LD, and is the level at
which there is a 5% chance of calling a background value
"greater than background." This value (LC) should be used
when actually counting samples or making direct radiation
measurements. Any response above this level should be
counted as positive and reported as valid data. This will
ensure 95% detection capability for LD.
If the sample count time (TS) is the same as the
background count time (TB), then equations 16 and 17 can
be simplified as follows:
95
Therefore, the full equations for LC and LD must be used
for samples with count times differing from the
background determination time (95% CL used).
96
The minimum significant activity level, LD, also referred to as
the LLD (Lower Limit of Detection) is calculated prior to
counting samples. This value is used to determine minimum
count times based on release limits and airborne radioactivity
levels.
In using this value we are saying that at a 95% CL, samples
counted for at least the minimum count time calculated using
the LD that are positive will indeed be radioactive (above the
LC).
This also means that 5% of the time samples considered
clean will actually be radioactive.
97
CROSSTALK
Discrimination
Crosstalk is a phenomenon that occurs on proportional
counting systems (such as a Tennelec) that employ electronic,
pulse-height discrimination, thereby allowing the simultaneous
analysis for alpha and beta-gamma activity.
Discrimination is accomplished by establishing two thresholds,
or windows, which can be set in accordance with the radiation
energies of the isotopes of concern.
Recall that the pulses generated by alpha radiation will be
much larger than those generated by beta or gamma.
This makes the discrimination between alpha and beta-
gamma possible.
98
SOURCES OF RADIATION
Voss Associates
99
NATURAL BACKGROUND RADIATION SOURCES
Radioactivity of the Earth
The presence of certain small amounts of radioactivity in the soil adds to the background levels to which man is exposed. The amount of radioactive materials found in soil and rocks varies widely with the locale. The main contribution to the background is the gamma ray dose from radioactive elements chiefly of the uranium and thorium series and lesser amounts from radioactive K-40 and Rb-87.
100
Due to the high concentration of monazite, a thorium mineral, some regions in the world have an extremely high background level.
The majority of the population of the Kerala region in India receive an annual dose greater than 500 mrem.
A small percentage of the inhabitants receive over 2,000 mrem per year and the highest recorded value has been 5,865 mrem in one year.
It is interesting to note that this value is more than what is allowed for a DOE radiation worker.
3
The Minas Garais state in Brazil has an average terrestrial background dose rate of 1,160 mrem per year.
Their maximum recorded dose rate has been 12,000 mrem per year.
In the United States on the average, a square mile of soil, one foot deep, contains one ton of K-40, three tons of U-238 and six tons of Th-232.
102
Radioactivity of Water
Depending upon the type of water supply a number of products may turn up.
Sea water contains a large amount of K-40.
Many natural springs show amounts of uranium, thorium, and radium.
Almost all water should be expected to contain certain amounts of radioactivity.
Since rain water will pick up radioactive substances from air, and ground water will pick up activity present in rocks or soil, radioactivity in water is found throughout the world.
103
The U.S. average of alpha emitters in water is <1 pCi/l.
Some regions contain significantly higher levels of naturally occurring alpha emitters: 40-50 pCi/l may be found in Colorado; and 200 pCi/l may be found in bottled water from Brazil.
The chief source of dose rate from this background factor occurs as the result of uptake of these waters by ingestion.
104
This leads to an internal exposure.
Any estimate of the dose rate from this source is thus included in the estimate of the dose rate from radioactivity in the human body.
The transfer of radioactive substances to the body seems to be mainly by food intake except in cases of very high water concentrations.
105
Cosmic Radiation
Much work has been carried out in the study of cosmic radiation.
This factor in background levels was discovered during attempts to reduce background.
Though detection devices showed a response even in the absence of any known sources, it was assumed this background was due entirely to traces of radioactive substances in the air and ground.
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Thus, if a detector was elevated to a greater height above the earth's surface, the background should be greatly reduced.
The use of balloons carrying ion chambers to great heights yielded data which showed the effect increased, rather than decreased.
This and other data from high altitude aircraft showed that radiation was really coming from outer space.
The name cosmic rays was given to this high energy.
107
Further study has shown that cosmic radiation consists of two parts: primary and secondary. The primary component may be further divided into galactic, geomagnetically trapped radiation, and solar.
PrimaryThe galactic cosmic rays come from outside the solar system and are composed mostly of positively charged particles. Studies have shown that outside the earth's atmosphere, cosmic rays consist of 87% protons, of 11% alpha particles, and about 1% each of other heavier nuclei and electrons at latitudes above 55 degrees. These particles may have energies in the range of about 1 GeV and higher.
108
Secondary
Secondary cosmic rays result from interactions which occur when the primary rays each the earth's atmosphere. When the high energy particles collide with atoms of the atmosphere, many products are emitted: pions, muons, electrons, photons, protons, and neutrons. These, in turn, produce other secondaries as they collide with elements or decay on the way toward the earth's surface. Thus, a multiplication or shower occurs in which as many as 108 secondaries may result from a single primary.
109
Most of the primary rays are absorbed in the upper 1/10 of the atmosphere. At about 20 km and below, cosmic rays are almost wholly secondary in nature. The total intensity of cosmic rays shows an increase from the top of the atmosphere down to a height of 20 km. Although the primary intensity decreases, the total effect increases because of the rapid rise in the number of the secondaries. Below 20 km, the total intensity shows a decrease with height because of attenuation of the secondaries without further increase in their number due to primaries. At less than 6 km of altitude, the highly penetrating muons, and the electrons they produce, are the dominant components.
110
At the earth's surface, the secondary cosmic rays consist mainly of muons (hard component), electrons and photons (soft components), and neutrons and protons (nucleonic component). At sea level about 3/4 of the cosmic ray intensity is due to the hard component.
Because of the earth's magnetic field, cosmic ray intensity also varies with latitude. The energy which is needed for a charged particle to reach the earth's atmosphere at the geomagnetic equator is larger than that needed at other latitudes. The effect is greatest for latitudes between 15 and 50 degrees. Above 50 degrees, the intensity remains almost constant. Thus, the lowest value of the intensity occurs at the geomagnetic equator, and the effect is expressed as the percentage increase at 55 degrees over that at the equator.
111
At sea level, the effect is small for the ionizing component (10%) but is larger for the neutron component. At sea level and high latitudes, the ionization rate, is about 2.1 x 1E6 ion pairs per cubic meter. Using a neutron calculation, the sea level dose would be increased by about 5%.
Taking into account the dose variation with altitude, and the population distribution with altitude, the average yearly dose equivalent rate to the U.S. population from cosmic radiation is estimated to be 27 mrem (270 μSv). This dose equivalent rate would be expected to decrease slightly with latitude and increase with altitude. For example at Denver, the yearly dose would be about 50 mrem (500 μSv).
112
Internal Emitters (Radioactivity of the Human Body)
Since small amounts of radioactive substances are found throughout the world in soil and water, some of this activity is transferred to man by way of the food chain cycle. A number of studies have been made to try to find a correlation between the amounts in soil and that in man. Results have not shown a clear-cut relationship as yet.
113
In the human body, K-40, Rb-87, Ra-226, U-238, Po-210, and C-14 are the main radionuclides of concern. Of these, K-40 is the most abundant substance in man. The amount in food varies greatly, so that intake is quite dependent on diet. However,variations in diet seem to have little effect on the body content. The content of K-40 in body organs of man varies widely. Based on an average content of 0.2% by weight in soft tissue, 0.05% in bone, the yearly dose equivalent rate to the gonads is estimated to be 19 mrem (190 μSv); 15 mrem (150 μSv) to bone surfaces; and 15 mrem (150 μSv) to bone marrow. Rb-87 contributes only a few percent of these values.
114
Most of the Ra-226 which is taken into the body will be found in the skeleton. Much data has been gathered on the concentration in humans, and the present assumed average skeletal concentration is taken as about 0.29 Bq/kg. The skeletal content of Ra-228 is taken as 0.14 Bq/kg. The yearly dose rate produced by these components is estimated to be .5 mrem (5 μSv) to the gonads, 14.6 mrem (146 μSv) to bone surfaces and 2.2 mrem (22 μSv) to bone marrow.
Based upon an average concentration of U-238 of 0.26 Bq/kg in bone, the estimated doses in man are 4.8 mrem (48 μSv) to bone surfaces and 0.9 mrem (9 μSv) to the marrow. From the estimated content in the gonads, the annual dose equivalent is estimated to be about 1 mrem (10 μSv).
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Similarly, the Po-210 decay chain contribution is taken as 2.22 Bq/kg, yielding annualdose equivalents of 24 mrem (240 μSv) to bone surfaces and 4.9 mrem (49 μSv) to bone marrow. The soft tissue concentration is taken as 0.111 Bq/kg, but is about twice that in the gonads. This gives an annual gonad dose equivalent of 6 mrem (60 μSv). The average whole body content of carbon is taken as 23%.
However, C-14 is present in normal carbon only to a very small extent (C-14/C-12 ~10-12), so that only a small amount of C-14 is present. The annual average dose equivalent turns out to be about 1 mrem (10 μSv) total body. In soft tissue, the annual dose is 0.7 mrem (7 μSv). The annual dose to the bone surfaces is 0.8 mrem (8 μSv), and to the bone marrow, 0.7 mrem (7 μSv).
116
The U.S. annual average dose equivalent for all internal emitters (food chain) in the body is 39 mrem (390 μSv) as listed by the NCRP Report No. 93.
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Inhaled Radionuclides (Radioactivity of the Air)
The background which is found in air is due mainly to the presence of radon and thoron gas, formed as daughter products of elements of the uranium and thorium series. The decay of U-238 proceeds to Ra-226. When Ra-226 emits an alpha as it decays, the gas Rn-222 is formed, which is called radon. In the thorium chain, the decay of Ra-224 results in the gaseous product Rn-220, which is called thoron. Since uranium and thorium are present to some extent throughout the crust of the earth, these products are being formed all the time. Since they are gases, they tend to diffuse up through the earth's surface to become airborne. In turn, the decay products of these gases attach themselves to dust in the air.
118
The amount of these gases in the air depends upon the uranium and thorium content of a certain area. In any given area, the weather conditions will greatly affect the concentrations of these gases. It is also common to find that the levels indoors are higher than those outdoors. This is a function of the material of the building and the ventilation rate. In mines and other underground caverns, the concentrations have been found to bequite high.
119
Some homes in Grand Junction and Durango, Colorado, have been found to have high radon levels. This was traced to the use of uranium mill tailings, residues rich in radium, as backfill. This discovery has led to radon measurements in homes in other areas of the country. Some homes in Pennsylvania are situated on land with naturally elevated radium concentrations, giving rise to increased indoor radon levels. Investigations have beenmade of radon levels in homes in the Chicago area. Their results indicated that 6% of the homes studied had radon concentrations comparable to those found at Grand Junction.Because of the potential population dose from this source, much more work on defining this potential problem is being carried out.
120
The major source of exposure from radon in air occurs when the daughter products attach themselves to aerosols and are inhaled. This leads to an internal dose to the lungs. As for external exposure, the external gamma dose rate from Rn-222 and Rn-220 is estimated to be less than 5% of the total external terrestrial dose rate. The contribution of inhaled radon gas to the annual average effective dose equivalent is included as an inhaled radionuclide.
121
Among other radioactive products which are found in air in measurable amounts are C-14, H-3, Na-22, and Be-7. These are called cosmogenic radionuclides, since they are produced in the atmosphere by cosmic rays. None of these products add a significant amount to the background dose rate.
The U.S. annual average dose equivalent for various inhaled radionuclides (primarily radon) is estimated at 200 mrem (2,000 μSv) by the NCRP Report No. 93.
122
Nuclear FalloutThe term fallout has been applied to debris which settles to the earth as the result of a nuclear blast. This debris is radioactive and thus a source of potential radiation exposure to mankind. Radioactive fallout is not considered naturally occurring but is definitely a contributor to background radiation sources.Because of the intense heat produced in a nuclear explosion during a very short time, matter which is in the vicinity of the bomb is quickly vaporized. This includes fission products formed in the fission process, unused bomb fuel, the bomb casing and parts, and, in short, any and all substances which happen to be around. These are caught in the fireball which expands and rises very quickly.
123
As the fireball cools and condensation occurs, a mushroom-shaped cloud is formed, containing small solid particles of debris as well as small drops of water. The cloud continues to rise to a height which is a function of the bomb yield and the meteorological factors of the area. For yields in the megaton range (1 megaton equals an energy release equivalent to one million tons of TNT), the cloud top may reach a height of 25 miles.
124
The fallout which occurs may be described as local or world-wide. The portion of debris which becomes local fallout varies from none (in the case of a high-altitude air burst) to about half (in the case of a contact surface burst). The height at which the bomb goes off is thus quite important in the case of local fallout. If the fireball touches the surface of the earth, it will carry aloft large amounts of surface matter. Also, because of the vacuumeffect created by the rapid rise of the fireball, other matter may be taken up into the rising fireball. This leads to the formation of larger particles in the cloud that tend to settle out quickly. If the width is not too great, the fallout pattern will be roughly a circle around ground zero. Ground zero is the point on the surface directly under, at, or above the burst.
125
If the burst is in the megaton range, the debris is carried into the stratosphere. In this region little mixing will occur, and the absence of rain or snow prevents this matter from being washed down. The time that it takes for this debris to return to the troposphere and be washed down varies. It is a function of both the height in the stratosphere to which the debris is lifted and the locale at which the burst occurs. It may take up to 5 years or more for this debris to return to earth. On the other hand, for bursts in the northern hemisphere in which the debris is confined to only the lower part of the stratosphere, the half-residence time is thought to be less than one year. Half-residence time is the time for one-half of the debris to be removed from the stratosphere.
126
In all, there are more than 200 fission products which result from a nuclear blast. The half-life of each of these products covers the range from a fraction of a second to millions of years. Local fallout will contain most of these products. Because of the time delay in the appearance of world-wide fallout, only a few of these products are important from that standpoint. Since local fallout is confined to a relatively small area, its effect on the human population can be negated by proper choice of test sites, weather conditions, and type of burst. The fallout of interest from the standpoint of possible effects on man due to testing is the world-wide fallout.
127
Because of the associated time delay before world-wide fallout shows up, many fission products and activation products decay out in transit. Others, because they are produced in such small amounts, are diluted so that they do not produce much of an effect. Also, once the fallout does arrive, to be of importance internally, there must be a transfer to the body and absorption into the body organs. All these factors combine to limit the number of fission products which may have an effect on humans. The main contribution comes from Sr-90, Cs-137, I-131, C-14, H-3 with minor contributions from Kr-85, Fe-55, and Pu-239.
NCRP Report No. 93 lists the annual average effective dose equivalent from nuclear fallout exposure at less than 1 mrem (10 μSv). However the total dose commitment, to be delivered over many generations, is 140 mrem (1400 μSv).
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Medical Exposures
The exposure to the U.S. population from X-rays used in medical and dental procedures is the largest source of man-made radiation. It is estimated that more than 300,000 X-ray units are in use in the U.S., and that about 2/3 of the U.S. population is exposed. In 1970, the estimated annual average bone marrow dose equivalent from dental and medical Xrays to the U.S. population was about 78 mrem (780 μSv). In addition to the exposure from X-rays, nuclear medicine programs use radio-pharmaceuticals for diagnostic purposes. Radiologists also use radionuclides for therapy treatment. It has been estimated that more than 10 million doses are administered each year.
129
NCRP Report No. 93 lists the average annual effective dose equivalent in the U.S. for diagnostic Xrays and nuclear medicine as 39 mrem (390 μSv) and 14 mrem (140 μSv), respectively. This gives a combined average annual effective dose equivalent from medical exposures of 53 mrem (530 μSv).
130
Diagnostic X-Rays
There are many different types and styles of X-ray machines used in the medical field. An X-ray machine generally consists of the X-ray tube, an electrical source of high voltage, a type of filament, and radiation shielding to collimate the beam to some limited size and shape. A diagnostic X-ray machine is used to obtain an image of some part of the body on some type of storage material. There are three general types of diagnostic X-ray equipment: radiographic, fluoroscopic, and photofluorographic.
131
Radiography involves the use of an X-ray tube and a photographic plate. The patient is placed between the two and an image is produced on the film of the area exposed. A common "chest X-ray" is an example of a radiographic X-ray.
In a fluoroscopic X-ray machine the film cassette is substituted with an imaging device (image intensifier). This enables the radiologist to observe the part of the body exposed live on a video monitor. A blocking agent, such as barium, is often swallowed by the patient to allow the medical staff to observe internal processes in action.
The photofluorographic process utilizes an X-ray tube, a fluorescent screen and a camera. This practice is similar to radiographic X-rays. Several pictures can be taken on one roll of film of the image on the fluorescent screen.
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Medical Radionuclides
Radionuclides are used in medicine by two general classifications: Nuclear Medicine for diagnostic procedures and Radiation Oncology for radiation therapy.
Since the radioisotope is internally deposited either by mouth or by injection, it should decay by emitting only photons. Isotopes emitting alpha or beta particles would be locally absorbed in the organ and would not contribute to the information signal. Another consideration in radionuclide selection would be the effective half-life. To maintain organ doses ALARA, isotopes with a few hour half-life are optimum. Technetium-99m and Indium-113m are commonly used radiopharmaceuticals.
PET is a relatively new application.
133
Radiation Oncology (study/treatment of tumors) uses radionuclides for tumor treatment. Cobalt-60 has been used for the high activity sealed source. This consists of a mechanical device which moves the source to an opening in a collimator which projects a beam of photons used for treatment. A typical 6,000 curie Cobalt-60 source delivers about 100 rad/minute to a tumor.
134
Consumer Products
In NCRP Report 56, a number of consumer products and miscellaneous sources of radiation exposure to the U.S. population are discussed. In general, two groups of sources have been found:1. Those in which the dose equivalent is relatively large andmany people are exposed.2. Those in which the dose equivalent is small but many peopleare exposed or the dose equivalent is large but only a few people are exposed.
135
Television sets, computer monitors, luminous-dial watches, smoke detectors, static eliminators, tobacco products, airport luggage inspection systems, building materials and many other sources contribute to our exposure.
The estimated annual average whole body dose equivalent to the U.S. population from consumer products is approximately 10 mrem (100 μSv). The major portion of this exposure (approximately 70%) is due to radioactivity in building materials.
136
Nuclear FacilitiesMore than 100 nuclear power plants are operating in the U.S. In addition, over 30 other reactors, classed as non-power reactors, are being operated. In order to provide fuel for these reactors, mining and milling of uranium ore is carried out and fuel fabrication plants are operating. There are many uranium mines, mills and fuel fabrication facilities. Sources of radiation from nuclear reactors consist of prompt neutrons, gamma rays and possible exposures from contamination or environmental releases. The NRC is tasked by the federal government to calculate doses for populations living within 50 miles of a nuclear facility. Three radionuclides released during routine operations, which contribute to the population dose, are H-3, C-14, and Kr-85. Current estimates of the yearly average dose equivalent inthe U.S. from environmental releases is <1 mrem (10 μSv).
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SUMMARY OF RADIATION EXPOSURESAnnual exposure (mrem/yr)
Natural BackgroundTerrestrial 28Cosmic 27Internal Emitters 39Radon 200
Man-Made BackgroundNuclear Fallout <1Medical Exposures 53Consumer Products 10Nuclear Facilities <1
Rounded Total 360
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139
MEDICAL RADIATION
SOURCES
Voss Associates
140
X-ray Tubes
Electronically speaking, an x-ray tube is a vacuum
diode; it consists of two elements inside an evacuated
glass tube.
141
The cathode consists of a tungsten wire which is heated by
passing an electric current through it to cause the release of
electrons through thermionic emission.
An alternating potential difference from a high voltage
source is applied between the cathode and target.
During that part of each cycle when the target is positive
with respect to the cathode, electrons under the influence of
the Coulomb force accelerate across the gap and strike the
target.
142
This leads to the emission of bremsstrahlung radiation
which constitutes most of the energy in the spectrum of x-
rays emitted by the tube.
The intensity of the x-radiation is directly proportional to the
square of the potential difference across the tube.
In almost every modern type of x-ray tube, the applied
voltage is a sine wave.
The x-ray output occurs in “bursts,” one for each cycle of
the input voltage.
143
X-ray spectrum from an x-ray tube
144
X-ray Machine Applications
Medical x-ray machines are divided into two basic groups -
diagnostic and therapeutic.
A diagnostic x-ray procedure is used to obtain an image of
some body part on photographic film or the image is stored
on a computer.
Computerized Axial Tomography (or CAT Scanner) uses a
tiny, highly focused x-ray beam is scanned over a portion of
the patient. The fraction of the beam intensity which is
transmitted through the body part is measured by detectors
placed around the patient. The computer analyzes the
pattern of data points and reconstructs a cross-sectional
view of the body parts to construct a three-dimensional
picture.
145
The other medical use category is therapeutic x-ray
procedures. X-radiation has been found to be useful in
the management of malignancies. Certain forms of skin
cancer respond well to very low energy x-rays, of about
10 to 40 kVp.
Before the common availability of Cobalt-60 gamma ray
sources, higher energy x-rays were used to treat deeper
lying tumors. Machine potential differences of 250 kV
and 400 kV were common.
146
The common photon generating equipment used by
radiation oncology departments for deep tumors today is
the medical linear accelerator.
The machines produce high energy electron beams in a
microwave waveguide.
The electrons are then directed onto a tungsten target and
the resulting bremsstrahlung radiation used for treatment.
A typical dose rate at 100 cm treatment distance is 300 or
400 rad/min.
147
Average US Doses from Medical X-rays
148
Typical Patient Doses from Medical Procedures
149
The other common use of radionuclides in a radiation
oncology setting is for implant therapy or brachytherapy.
A variety of radioisotopes, including Ra-226, Rn-222,
Cs-137, Ir-192, I-125, Pd-103 and Au-198, are used to
treat malignancies by placing the source near or inside the
affected tissues.
More than 100,000 brachytherapy treatments are
performed each year.
150
Positron emission tomography (PET) is a test that uses a
gamma camera and a radioactive tracer to look at organs
in the body.
During the test, the tracer liquid is put into a vein. The
tracer moves through your body, where much of it collects
in the target organ or tissue.
The tracer gives off tiny positively charged particles
(positrons).
When the positrons collide with an electron they emit two
511 keV gamma rays.
The gamma camera turns the image into pictures on a
computer.
151
PET scan pictures do not show as much detail as CAT scans
or MRI imaging because the pictures show only the location
of the tracer.
The PET picture may be matched with those from a CAT
scan to get more detailed information about where the tracer
is located.
A PET scan is often used to find cancer, to check blood flow,
or to see how organs are working.
152
Several iodine isotopes are frequently used in diagnostic
(and occasionally therapeutic) medical procedures and
in many biological research applications.
These include 131-I, 123-I and 125-I.
The metabolic behavior of these isotopes are is
identical, and handling procedures are similar.
153
I-125 Emissions
Rem/hr per Ci Ingestion Inhalation
at 30 cm ALI DAC Factor
3.1 0.04 mCi 3E-8 uCi/mL
154
Iodine compounds are rapidly absorbed into the blood
stream following surface contact on human skin.
It is common practice to require double gloves when
handling single quantities of 125-I in excess of 1 mCi
and to change the outer pair of gloves every 10 minutes.
155
Rem/hr per Ci Ingestion Inhalation
at 30 cm ALI DAC Factor
I-125 3.1 0.04 mCi 3E-8 uCi/mL
I-123 0.7 3 mCi 3E-6 uCi/mL
I-131 3.1 0.03 mCi 3E-8 uCi/mL
Tl-201 0.1 20 mCi 2E-6 uCi/mL
Tc-99m 1.4 80 mCi 6E-5 uCi/mL
Xe-133 0.5 Immersion 1E-4 uCi/mL
ANZAI EZ SCOPE GAMMA CAMERA
The EZ Scope is a handheld, light-weight, CZT-based
gamma camera that produces both planar and
tomographic images.
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157
158
RADIOACTIVE SOURCE
CONTROL
Voss Associates
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160
A radioactive source is material used for its emitted
radiation. Sources are constructed as sealed or unsealed
and are classified as accountable or non-accountable .
Radioactive sources are used for response checks in the
field, functional checks, calibration of instruments and
monitors to traceable standards. To ensure the safety and
welfare of all personnel it is important to maintain control of
radioactive sources.
Radioactive sources are controlled to minimize the
potential for:
Spread of contamination
Unnecessary exposure to personnel
Loss or theft
Improper disposal
161
In accordance with 10 CFR 835, Subpart M, the following
provisions apply to sealed sources:
A. Sealed radioactive sources shall be used, handled, and
stored in a manner commensurate with the hazards
associated with operations involving the sources.
B. Each accountable sealed radioactive source shall be
inventoried at intervals not to exceed six months. This
inventory shall:
1. Establish the physical location of each
accountable sealed radioactive source;
2. Verify the presence and adequacy of associated
postings and labels; and
3. Establish the adequacy of storage locations,
containers, and devices.
162
C. Except for sealed sources consisting solely of
gaseous radioactive material or tritium, each accountable
sealed radioactive source shall be subject to a source
leak test upon receipt, when damage is suspected, and
at intervals not to exceed six months. Source leak tests
shall be capable of detecting radioactive material
leakage equal to or exceeding 0.005 μCi.
D. An accountable sealed radioactive source is not
subject to periodic source leak testing if that source has
been removed from service. Such sources shall be
stored in a controlled location, subject to periodic
inventory, and subject to source leak testing prior to
being returned to service.
163
E. An accountable sealed radioactive source is not subject
to periodic inventory and source leak testing if that source
is located in an area that is unsafe for human entry or
otherwise inaccessible.
F. An accountable sealed radioactive source found to be
leaking radioactive material shall be controlled in a manner
that minimizes the spread of radioactive contamination.
164
Sources are controlled using the following precautions:
1. Each source is to be inspected before each use.
2. Remove damaged sources from service.
3. Fingers, whether gloved or not, or other objects should
never be allowed to touch the active surface of unsealed
sources.
4. Protect the source from being contaminated when used
in a surface contamination area.
165
Sealed radioactive sources not in storage containers or
devices and not labeled by the manufacturer must be
clearly marked with a radiation symbol and have a durable
label/ tag containing the following information:
a. Radionuclide
b. Amount of activity
c. Name of manufacturer
d. Date of assay
e. Model and serial numbers (where available)
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1. Accountable Sealed Radioactive Source means a
sealed radioactive source having a half-life equal to or
greater than 30 days and an isotopic activity equal to or
greater than the corresponding value provided in Appendix
E of 10CFR 835.
2. Radioactive Material Area means any area within a
controlled area, accessible to individuals, in which items or
containers of radioactive material exist and the total activity
of radioactive material exceeds the applicable values
provided in Appendix E to 10 CFR 835.
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3. Sealed Radioactive Source means a radioactive
source manufactured, obtained, or retained for the
purpose of utilizing the emitted radiation. The sealed
radioactive source consists of a known or estimated quantity
of radioactive material contained within a sealed capsule,
sealed between layer(s) of nonradioactive material, or firmly
fixed to a non-radioactive surface by electroplating or other
means intended to prevent leakage or escape of the
radioactive material.
4. Source Leak Test means a test to determine if a
sealed radioactive source is leaking radioactive material.
168
External Exposure Control
Voss Associates
169
Basic Methods for Exposure Reduction
The radiological control organization shall make whatever
reasonable efforts it can to reduce exposure to the lowest
levels.
There are four basic methods available to reduce external
exposure to personnel:
– Reduce the amount of source material (or reduce emission
rate for electronically-generated radiation).
– Reduce the amount of time of exposure to the source of
radiation.
– Increase the distance from the source of radiation.
– Reduce the radiation intensity by using shielding between
the source and personnel.
170
Basic Methods for Exposure Reduction
In order to use the basic methods for controlling
exposure, the worker must be able to determine the
intensity of the radiation fields. The following equations
are used to make this determination.
A "rule-of-thumb" method to determine the radiation
field intensity for simple sources of radioactive material
is the "curie/meter/rem" rule. (Co-60)
1 Ci @ 1 meter = 1 R/hr
171
Basic Methods for Exposure Reduction
To determine the gamma radiation field intensity for a
radioactive point source
I1ft = 6CEN
where:
I1ft = Exposure rate in R/hr at 1 ft.
C = Activity of the source in Ci
E = The gamma energy in MeV
N = The number of gammas per disintegration
172
Basic Methods for Exposure Reduction
– This equation is accurate to within +20% for gamma
energies between 0.05 MeV and 3 MeV.
– If N is not given, assume 100% photon yield (1.00
photons/disintegration).
– If more than one photon energy is given, take the
sum of each photon multiplied by its percentage, i.e.:
[(γ1)(%1) + (γ2)(%2) + ··· + (γn)(%n)]
173
Basic Methods for Exposure Reduction
For distances in meters:
I1m = 0.5CEN
For short distances greater than 1 foot from the source, the
inverse square law can be applied with reference to the dose
rate at 1 foot, resulting in the following equation:
I = 6CENx12
d2
where: d = distance in feet
174
Example
Determine the exposure rate at 10 ft for a 8 Ci point
source of Co-60 that emits a 1.173 and 1.332 MeV
gamma, both at 100% of the disintegrations.
175
Source Reduction
The first method that should be employed to reduce personnel
external exposure is source reduction. If a source can be
eliminated or if its hazard potential can be significantly
reduced, then other engineering means may not be necessary.
Various techniques are employed to accomplish external
exposure reduction using source reduction.
Allow natural decay to reduce source strength
– If the radioisotopes involved are short-lived, then waiting to
perform the task may significantly reduce the hazard.
– By waiting for natural decay to reduce the source strength, a
considerable savings in external exposure can be achieved.
176
Source Reduction
Move the source material to another location
– Decontaminate the equipment or material throughmechanical or chemical means to remove the sourcematerial prior to working in the area or on the equipment.
– Reduce the source material in the system by flushingequipment with hot water or chemical solutions and collect itin a less frequently occupied area.
– Discharge or remove the resin or filtering media prior toworking in the area or on the system.
– Move the radioactive source (e.g., a drum, barrel orcalibration source) to another location prior to starting work.
177
Time Savings
Personnel working in radiation fields must limit their exposure
time so that they do not exceed their established permissible
dose limits and are able to keep exposures ALARA.
The longer the time spent in the radiation field, the greater
the exposure to the individual; therefore, the amount of time
spent in radiation fields should be reduced.
The Radiological Control Technician needs to be aware that
radiation exposures are directly proportional to the time spent
in the field. If the amount of time is doubled, then the amount
of exposure received is doubled.
178
Time Savings
Analyze and train using mock-ups of the work site
– A particular task can be analyzed on a mock-up of the
system to determine the quickest and most efficient
method to perform the task.
– The team of workers assigned to the task can rehearse,
without radioactive materials, so that problems can be
worked out and the efficiency of the team increased prior
to any exposure.
– By determining the most efficient method and rehearsing
the task, the amount of time, and therefore the exposure,
can be reduced.
179
Time Savings
Pre-job briefings are an important part of any good ALARA
program
– Discussions at the pre-job briefing with the individuals
assigned to the task can identify any potential problems not
previously identified.
– Identifying personnel responsibilities and the points at
which various individuals are required to be present can
reduce the overall time required to perform the job.
Review job history files -- Review the files from previously
completed tasks of the same nature to identify previous
problems and spots where time could be saved.
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Time Savings
Pre-stage all tools and equipment -- All tools should be
staged prior to entry to prevent the worker from waiting in a
radiation field for a tool to be brought.
Pre-assemble equipment and tools outside the area
– Equipment that can be preassembled should be
preassembled prior to any entry into the radiation field.
– Tools that require assembly, pre-testing, and/or calibration
should be performed outside the radiation field.
181
Time Savings
Use time limiting devices -- Time limitations for workers
can be monitored and limited using various devices
such as stopwatches, alarming dosimeters, or radio-
transmitting dosimeters.
Use communication devices such as walkie-talkies
– Poor communication can lead to incorrect or poor
quality work and prolonged waiting in the radiation
field while supervisors or experts are contacted.
– Communication devices such as walkie-talkies or
radio headsets can alleviate these problems and
reduce the amount of time that is spent in the
radiation field.
182
Time Savings
Use a team of workers instead of allowing one individual to
receive all of the exposure
– Even if the task requires a minimum amount of time, if it
causes one individual to receive an exposure greater than
allowable, a team of workers should be used to reduce the
individual exposures.
– If a team of workers is used, good communications are
necessary to ensure the total exposure for the job does
not increase significantly.
183
Use experienced personnel
- The total time required to perform a job is reduced if experts
are used instead of inexperienced personnel.
- Inexperienced personnel should not be trained in significant
radiation fields
Time Savings
184
Exposure Calculation
The exposure received by personnel will increase as
the time spent in the radiation field increases.
185
Stay Time Calculation
When the time allowed in a radiation field is calculated
to prevent a worker from exceeding an allowable dose
equivalent, it is called "stay time."
Stay time is calculated as follows:
186
Distance
The intensity of the radiation field decreases as the distance
from the source increases.
Therefore, increasing the distance will reduce the amount of
exposure received.
In many cases, increasing the distance from the source is
more effective than decreasing the time spent in the radiation
field.
187
Distance
The radiation intensity for a point source decreases
according to the Inverse Square Law which states that
as the distance from a point source changes the dose
rate decreases or increases by the square of the ratio of
the distances from the source.
The inverse square law becomes inaccurate close to
the source (i.e., about 10 times the diameter of the
source).
For a point source, if the distance is doubled, the
radiation intensity will be reduced by a factor of (2)2 or 4.
188
Distance
Remote handling tools/remote control devices – Tools, such as tongs or long-handled tools, are an
effective means of increasing the distance from a pointsource to a worker.
– For very high radiation fields, remote control devices maybe appropriate, especially if the task is performedfrequently.
Remote observation by cameras or indicators– Gauges or meters can be moved to a location remote from
the source of radiation.– Closed-circuit television and video cameras can be used
to allow observation of work activities or systemoperations from a location remote to the source ofradiation.
189
Distance
Move work to another location
– If the source of radiation can not be reduced, then
possibly the work can be moved to a low exposure
area.
– For example, if a pump or valve needs reworking,
then an exposure savings could be achieved by
removing the component from the system and
performing all repair work in a lower exposure area.
190
Distance
Posting of areas -- Posting of radiological areas based
on radiation level is a method for increasing the
distance between the workers and the radiation source.
Extendable Instruments -- Extendable radiation survey
instruments, such as the Eberline Teletector or RO-7,
can reduce the exposure to the surveyor by increasing
the distance.
191
Distance
– For workers or inspectors not actively engaged in the workactivity in the radiation field, moving to a lower exposurerate "waiting" area can be effective.
– Identifying "low dose rate waiting areas" can notify workersof the location of the lowest exposure rate in an area orroom.
– Be aware of the location of radiation sources at theworksite and locate the worker at a point farthest from thesource.
– Work at arm's length and do not lie on or hug radioactivecomponents.
192
Inverse Square Law
The exposure rate is inversely proportional to the
square of the distance from the source. The
mathematical equation is:
(I2)(d2)2 = (I1)(d1)
2
where:
I1 = Exposure rate at distance (d1)
I2 = Exposure rate at distance (d2)
d1 = First distance from the source
d2 = Second distance source
193
Inverse Square Law
A somewhat useful variation on the inverse square law
is the use of a quadratic equation when the actual
distance to the source is not known.
How does the inverse square law apply to alpha, beta,
and neutron radiation ?
194
Inverse Square Law
The inverse square law holds true only for point
sources; however, it gives a good approximation when
the source dimensions are smaller than the distance
from the source to the exposure point.
Some sources, such as a pipe or tank, cannot be
treated as a point source unless the distance to the
source is greater than the length of the source.
195
Line Source Calculations
The actual calculations for a line source involve
calculus; however, the mathematics can be simplified if
the line source is treated as a series of point sources
placed side by side along the length of the source.
If the line source is treated in this manner, the
relationship between distance and exposure rate can be
written mathematically as:
I1d1 = I2d2
196
Line Source Calculations
The exposure rate is inversely proportional to the distance from the source
Assuming the source material is distributed evenly along the line
Assuming the point at which the exposure rate is calculated is on a line perpendicular to the center of the line source
Assuming the width or diameter of the line is small compared to the length
197
Planar Or Surface Sources
Planar or surface sources of radiation can be any type of
geometry where the width or diameter is not small compared
to the length.
When the distance to the plane source is small compared to
the longest dimension, then the exposure rate falls off a little
slower than 1/d (i.e. not as quickly as a line source).
As the distance from the plane source increases, then the
exposure rate drops off at a rate approaching 1/d2
198
Mass Attenuation Coefficient
The linear attenuation coefficient ( µ ) is dependent on:
–photon energy
–chemical composition of the absorber (Z)
–the physical density of the absorber
The linear attenuation is the probability of a photon
interaction per path length and has units of (length)-1
(typically cm-1).
The linear attenuation coefficient will change depending
on the physical density of the absorber.
199
Mass Attenuation Coefficient
Mathematically:
µm = µl/ρ
where:
µm = mass attenuation coefficient
µl = linear attenuation coefficient
ρ = physical density
When the units of the linear attenuation coefficient are
cm-1 and the units of physical density are mg/cm3 then
the units of mass attenuation coefficient become
cm2/mg.
200
Density-Thickness
The mass attenuation coefficient µm is used in the
attenuation equation
I = I0e-µmx
where:
I = shielded (attenuated) radiation intensity
I0 = unshielded radiation intensity
µm = mass attenuation coefficient
x = density-thickness value
201
Density-Thickness
When µm is used in the attenuation equation, x must
have units of mg/cm2 to cancel out the units of µm. With
these units, x is called density-thickness.
Density-thickness is a value equal to the product of the
density of the absorbing material and its thickness.
This value is given in units of mg/cm2.
The density of any material is a measure of its mass per
unit volume, as compared to the density of water. Water
has a density of 1 g/cm3, or 1000 mg/cm3.
202
Density-Thickness
According to ICRP 15, the density of soft human tissue is
equal to 1000 mg/cm3. Using this value we can calculate
density-thickness values for various depths that radiation
may penetrate into the human body and cause damage.
For purposes of reporting radiation dose the tissue depths of
concern are the skin (shallow dose), the lens of the eye, and
the whole body (deep dose).
The concept of "density-thickness" is important to
discussions of radiation attenuation by human tissue, as well
as detector shielding and windows, and dosimetry filters.
203
Density-Thickness
Although materials may have different densities and
thicknesses, if their density-thickness values are the same,
they will attenuate radiation in a similar manner.
For example, a piece of mylar used as a detector window
with a density of 7 mg/cm2 will attenuate radiation similar to
the skin of the human body.
These values can be used to design radiation detection
instrumentation such that detector windows and shields have
the same or similar density-thickness values.
204
Density-Thickness
Any radiation passing the detector window would also
penetrate to the basal layer of the skin on the human
body and deposit energy in living tissue.
External dosimetry can be designed around these
values such that dose equivalent is determined for the
skin of the whole body, lens of the eye, and whole body.
For example, a dosimeter filter may be designed as
1000 mg/cm2. Any radiation passing this filter would
also pass through the skin of the whole body and
deposit energy in vital human organs.
205
Density-thickness Values for Human Body
206
Shielding Calculations
The simplest method for determining the effectiveness of the shielding material is using the concepts of half-value layers (HVL) and tenth-value layers (TVL).
One half-value layer is defined as the amount of shielding material required to reduce the radiation intensity to one-half of the unshielded value.
The symbol µ is known as the linear attenuation coefficient and is obtained from standard tables for various shielding materials.
207
Shielding Calculations
One tenth-value layer is defined as the amount of
shielding material required to reduce the radiation
intensity to one-tenth of the unshielded value.
Both of these concepts are dependent on the energy of
the photon radiation and a chart can be constructed to
show the HVL and TVL values for photon energies.
208
Half-Value Layers
209
HVL Equation
The basic equation for using the HVL concept is:
Where:
210
TVL Equation
The basic equation for using the TVL concept is:
Where
CONTAMINATION CONTROL
Voss Associates
211
212
Contamination is defined as radioactive material in an
unwanted location, e.g., personnel, work areas, etc.
Two types of contamination are possible, fixed and
removable (loose).
Fixed contamination is radioactive surface
contamination that is not easily transferred to either personnel or equipment through normal contact.
Removable contamination is radioactive surface
contamination that is easily transferred to either personnel or equipment through normal contact.
213
Fixed contamination is measured by use of a direct survey
technique using a portable radiation survey instrument.
This technique, commonly referred to as "frisking" or
“scanning”, indicates the total contamination on a surface
apparent to the detector from both fixed and removable
contamination.
When non-removable levels are to be recorded, the
removable level must be subtracted from the total.
214
Removable contamination is measured by a transfer test
using a suitable sampling material. Common materials used
for the monitoring are the standard paper disk smear or cloth
smear. The standard technique involves wiping approximately
100 cm2 of the surface of interest using moderate pressure. A
common sampling practice used to ensure a 100 cm2 sample
is to wipe a 16 square inch "S" shape on the surface (i.e., 4
inches by 4 inches).
Qualitative, large area wipe surveys may be taken using other
materials, such as Masslinn cloth or Kimwipe, to indicate the
presence of removable contamination. Large area swipes
are commonly used when exact levels of contamination are
not required.
215
Constant (Continuous) Monitoring
There are various types of constant monitoring
instruments throughout the facilities to warn personnel of
radiation and contamination hazards.
Some instruments are permanently installed, and some
instruments are portable to allow movement from place to
place as deemed appropriate by the radiological control
staff.
216
Continuous Air Monitor (CAM)
These instruments continuously sample the air for
radioactive contamination in specific locations.
The air being sampled is typically drawn through a moving
particulate filter which is then monitored by a detector
system or through an internal detector to directly identify
radioactive materials present.
A CAM can give both a visual and audible alarm to warn
personnel of the presence of airborne contamination.
217
Process Monitoring Systems
Process monitoring systems monitor certain operations in
various facilities to alert operators of abnormal conditions
which might lead to the release of excessive amounts of
radioactivity to the facility or environment.
Process monitors include temperature, pressure, and flow
sensors.
218
Area and Equipment Surveys
Area and equipment surveys are conducted routinely
throughout the facilities to locate sources of radiation and
contamination and to detect potential changes in
radiological conditions.
Pre-job surveys are performed prior to work in radiological
areas in order to evaluate the hazards and determine work
limitations and physical safeguards.
Direct surveys with portable radiation detection
instruments and removable contamination surveys with
smears or swipes are used.
219
External Personnel Surveys
Personnel surveys are either performed by the individual
(self-monitoring) using handheld or automated
instruments or by a radiological control technician.
Self-monitoring is typically performed upon exiting a
contaminated area at established boundary points.
Personnel monitoring by a RCT is usually conducted
whenever contamination of the body or clothing is
suspected, or as required by exit monitoring when self-
monitoring is not feasible (remote location) or not
allowed.
220
Personnel Internal Monitoring
A routine program of internal contamination monitoring is
conducted as a final check on contamination control
procedures.
The program consists of external whole/partial body
counting and/or urinalysis.
221
Once the presence of radioactive material has been
located, the basic goal underlying any effective
contamination control program is to minimize
contaminated areas and maintain contamination levels as
low as reasonably achievable.
In some situations, this is not always possible due to:
• Economical conditions: Cost of time and labor to
decontaminate a location(s) outweighs the hazards of
the contamination present.
• Radiological conditions: Radiation dose rates or other
radiological conditions present hazards which far
exceed the benefits of decontamination.
• Operating conditions: Some areas, e.g., hot cells, will
be contaminated due to normal operations.
222
Other means of control must be initiated when
decontamination is not possible.
Engineering control (ventilation and containment),
administrative procedures (RWPs), and personnel protective
equipment are alternatives for the control of contamination.
In Fixed Contamination Areas the contamination may be
covered by paint, floor tiles, etc. when decontamination is not
possible.
223
CONTAMINATION CONTROL MEASURES
• Access/Administrative Controls
• Engineering Controls
• Personnel Protective Measures
• Decontamination
• Preventive Methods
224
Access/Administrative Controls
Once contamination has been located and quantified
and radiological areas have been determined, access
control to these areas must be adequately established.
Work Authorization, Radiological Posting, and
Radiological Work Permits are the primary administrative
controls.
225
ENGINEERING CONTROLS
Ventilation and Containment (Confinement) are the main
engineering controls.
Ventilation is a process of providing adequate air flow to
keep the radiological area well ventilated and to keep
the potential airborne radioactivity from migrating to
other non-radiological areas. This is done thru negative
pressure, much like some businesses use to keep their
shops temperature and humidity controlled without trying
to air condition the outside environment.
Containment (or confinement) is simply keeping
radioactive materials in enclosures adequate to prevent
the materials from getting outside the container.
226
In vitro involves counting an excreted sample, such as urine.
The amount of material present in the body is estimated
using the amount of materials present in excretions or
secretions from the body.
Samples could include urine, feces, blood, sputum, saliva,
hair, and nasal discharges.
Calculation of dose requires knowledge and use of
metabolic models which allow sample activity to be related
to activity present in the body.
PERSONNEL PROTECTIVE MEASURES
227
Dosimetry Terms
Absorbed Dose (D): Energy absorbed by matter from
ionizing radiation per unit mass of irradiated material at
the place of interest in that material. The absorbed dose is
expressed in units of rad (or gray) (1 rad = 0.01 gray).
Dose Equivalent (H): The product of the absorbed dose
(D)(in rad or gray) in tissue, a quality factor (Q), and all
other modifying factors (N). Dose equivalent is expressed
in units of rem (or sievert) (1 rem = 0.01 sievert).
Deep Dose Equivalent (DDE): The dose equivalent
derived from external radiation at a tissue depth of 1 cm in
tissue (1000 mg/cm2).
228
Shallow Dose Equivalent (SDE): The dose equivalent
derived from external radiation at a depth of 0.007 cm in
tissue (7 mg/cm2).
Whole Body: For the purposes of external exposure,
head, trunk (including male gonads), arms above and
including the elbow, or legs above and including the
knee.
Extremity: Hands and arms below the elbow or feet and
legs below the knee.
229
Committed Dose Equivalent (CDE): The dose equivalent
calculated to be received by a tissue or organ over a 50-
year period after the intake of a radionuclide into the body.
It does not include contributions from radiation sources
external to the body. Committed Dose Equivalent is
expressed in units of rem (or sievert).
Committed effective dose equivalent (H E,50)— The sum
of the committed dose equivalents to various tissues in the
body (HT,50), each multiplied by the appropriate weighting
factor (WT) - that is HE,50=ΣWTHT,50.
Committed effective dose equivalent is expressed in units
of rem (or sievert).
230
Weighting factor (WT)—The fraction of the overall health
risk, resulting from uniform, whole-body irradiation,
attributable to specific tissue (T).
The dose equivalent to tissue (HT) is multiplied by the
appropriate weighting factor to obtain the effective dose
equivalent contribution from that tissue.
231
232
Total Effective Dose Equivalent (TEDE) -- The sum of the
effective dose equivalent (for external exposures) and the
Committed Effective Dose Equivalent (for internal
exposures).
Annual Limit on Intake (ALI) -- The limit for the amount of
radioactive material taken into the body of an adult worker
by inhalation or ingestion in a year. ALI is the smaller value
of intake of a given radionuclide in a year by the reference
man (ICRP Publication 23) that would result in a
Committed Effective Dose Equivalent of 5 rems (0.05
sievert) or a Committed Dose Equivalent of 50 rems (0.5
sieverts) to any individual organ or tissue.
233
Derived Air Concentration (DAC) -- The airborne
concentration that equals the ALI divided by the volume
breathed by an average worker for a working year of 2000
hours (assuming a breathing volume of 2400m3).
Bioassay -- The determination of kinds, quantities, or
concentrations, and, in some cases, locations of radioactive
material in the human body, whether by direct measurement
or by analysis, and evaluation of radioactive materials
excreted or removed from the human body.
Declared pregnant worker -- A woman who has voluntarily
declared to her employer, in writing, her pregnancy for the
purpose of being subject to the occupational dose limits to
the embryo/fetus in accordance with 10CFR835.
234
Internal Exposure Control
Voss Associates
235
Entry Of Radioactive Materials Into The Body
Modes of Entry
Inhalation: Materials enter the body in the air that is
breathed.
Ingestion: Materials enter the body through the mouth.
Absorption: Material enters the body through intact skin.
Entry through wounds:
– Penetration: Materials enter (passively) through
existing wounds which were not adequately covered.
– Injection: Materials enter (forcefully) through wounds
incurred on the job.
236
Preventive Measures
Inhalation -- assessment of conditions, use of
engineering controls, respiratory protection equipment
Ingestion -- proper radiological controls and work
practices
Absorption -- assessment of conditions and protective
clothing
Entry through wounds -- not allowing contamination
near a wound by work restriction or proper radiological
controls if an injury occurs in a contaminated area.
237
Preventive Measures
Note that the preventive measures are designed to do
one of two things:
– Minimize the amount of radioactive materials present
which are available to enter the body, or
– Block the pathway from the source of radioactive
materials into the body.
238
Annual Limit On Intake
Derived Air Concentration
Assimilation of radioactive materials in the workplace occurs
most often as a result of inhalation of airborne radioactive
contaminants. With some nuclides, specifically tritium,
absorption through the skin is also a major concern.
Two limiting values have been calculated and are available
for use in limiting the inhalation of radioactive materials.
These limiting values are:
–Annual Limit on Intake (ALI)
–Derived Air Concentration (DAC)
239
Annual Limit On Intake
Derived Air Concentration
Annual Limit on Intake is the quantity of a single radionuclide
which, if inhaled or ingested in one year, would irradiate a
person, represented by reference man (ICRP Publication
23), to the limiting value for control of the workplace.
ICRP 68 methodology is being applied to 10CFR835 DAC
factors but have not been adopted by 10CFR20.
Derived Air Concentration is the quantity obtained by dividing
the ALI for any given radionuclide by the volume of air
breathed by an average worker during a working year
(2.4E3m3).
240
Annual Limit On Intake
Derived Air Concentration
According to ICRP 23, reference man breathes at an
average rate of 20 liters per minute, or 0.02 m3/min. In
the course of one working year, the total volume
breathed would be:
241
Annual Limit On Intake
Derived Air Concentration
The DAC is equal to the ALI divided by the volume of air
breathed by the average worker during a working year:
242
Annual Limit On Intake
Derived Air Concentration
During routine operations, the combination of physical
design features and administrative controls shall provide
that:
a) The anticipated occupational dose to general employees
shall not exceed the limits.
b) The ALARA process is utilized for personnel exposures
to ionizing radiation.
243
Annual Limit On Intake
Derived Air Concentration
Monitoring of airborne radioactivity shall be performed:
1) Where an individual is likely to receive an exposure of 40
or more DAC-hours in a year; or
2) As necessary to characterize the airborne radioactivity
hazard where respiratory protective devices for protection
against airborne radionuclides have been prescribed.
Real-time air monitoring shall be performed as necessary to
detect and provide warning of airborne radioactivity
concentrations that warrant immediate action to terminate
inhalation of airborne radioactive material.
244
Annual Limit On Intake
Derived Air Concentration
Measures used to minimize the concentration of airborne contaminants remain the primary means of minimizing potential exposure.
Minimizing the concentrations to below DAC values helps insure that workers could not exceed the ALI even if they were in the area continuously for long durations and breathing air at those concentrations.
An Airborne Radioactivity Area is any area where the concentration of airborne radioactivity exceeds or is likely to exceed the DAC value or where an individual present in the area without respiratory protection could receive an intake exceeding 12 DAC-hours in a week.
245
Annual Limit On Intake
Derived Air Concentration
Posting of airborne radioactivity areas controls access to
minimize exposure.
Minimize the stay time of workers in airborne areas to short
periods of time.
Augment installed engineering controls with respiratory
protection equipment to further reduce the concentration of
contaminants in the air the workers are actually breathing.
246
Annual Limit On Intake
Derived Air Concentration
The limitations imposed in terms of dosage to exposed workers are expressed as an annual limit.
Concentrations of contaminants in the air are monitored by continuous monitoring equipment and are supplemented by grab sampling as required.
Engineering controls are augmented with respiratory protection equipment when airborne contaminants exceed or potentially exceed DAC values.
247
Movement of Radioactive Materials
Through the Body
There is no simple device that can be placed on or in the body to determine the quantities of radioactive materials in the body or the dose received by the individual as a result of irradiation of body tissues by these materials.
When radioactive material enters the body, the assessment methods must be based on what happens to the materials, or what the body does with them.
The body does not possess the ability to differentiate between a nonradioactive atom and a radioactive atom of the same element. In terms of metabolic processes, the material is handled the same way.
248
Movement of Radioactive Materials
Through the Body
Once the material is in the body, then its behavior is
governed by the chemical form, its location in the body,
and the body's need for that material.
–Chemical form - solubility
–Location - pathways
–Body's need - intake and incorporation vs. elimination
249
Intake and Uptake
Intake: the amount of radionuclide taken into the body
Uptake: the amount of radionuclide deposited in the body
which makes its way into the body fluids or systemic
system (i.e., blood)
250
Inhaled Radioactive Materials
General Pathways
– Deposition in lungs with eventual transfer to GI tract
or retention
– Transport to body fluids
– Transfer to lymph nodes with eventual movement to
body fluids
– Retention in lymph nodes
251
Inhaled Radioactive Materials
Once in the bodily fluids, possibilities include:
– Transfer to specific organ
– Filtration and elimination by kidneys
– Transport and removal from body fluids through
circulatory systems (perspiration)
252
Inhaled Radioactive Materials
Insoluble particulates
– Lung retention time based on particle size and density
– Removal in mucous to digestive tract
– Elimination in fecal waste
253
Inhaled Radioactive Materials
Soluble particulate materials
– Retention in lungs based on size and density – someexhalation
– Some removed to GI tract for elimination or to body fluids
– Transfer to body fluids via lymph nodes or directly fromlungs
– Some retention in lymph nodes
– Body fluids to tissue or organ of interest
– Excretion
254
Metabolic Pathways
255
Ingested Radioactive Materials
For elements not used by the body, absorption by
ingestion is poor, and most materials will pass straight
through the body.
Materials pass through stomach to small intestine
where transport of soluble materials to body fluids will
occur.
From body fluids, materials go to the organs and/or are
removed through normal biological elimination
processes.
256
Ingested Radioactive Materials
Soluble materials
– Transfer to body fluids in intestines
– Circulation, absorption, incorporation in tissues and organs
– Elimination in urine
Insoluble materials
– Passes straight through
– Elimination in feces
257
Absorbed Radioactive Materials
Many radioactive nuclides are absorbable through the skin.
These nuclides include tritium, iodine, and some of the
transuranics in an acidic form.
Most of these do not pose any considerable concern
because of the relative percentages absorbed as opposed to
entry through inhalation.
The most important of these is tritium as water vapor.
Once absorbed into the body, tritium exchanges freely with
hydrogen, disperses throughout the body almost
immediately, and irradiates tissues throughout the body.
258
Target Organs
Some elements are collected in target organs. As an example, iodine is collected by the thyroid gland.
Major dose to the thyroid could be expected as a result of gamma and beta interactions emitted by iodine collected in the thyroid gland.
The radiation emitted from iodine in the thyroid also can irradiate other nearby parts of the body. Gamma radiation can penetrate tissue very easily and cause interactions in parts of the body in which no iodine is located.
259
Target Organs
Other elements are processed differently.
– Some are distributed freely throughout body fluids.
– Some are collected in specific organs such as the kidneys,
spleen or bone.
– Sr, U, Pu are concentrated in the bone
Some materials which enter as particulate materials may
spend the majority of their stay in the body in the lungs.
260
Elimination Processes
Normal Biological Elimination
Radioactive materials incorporated into body tissues and organs are eliminated from the body as are their non-radioactive counterparts.
Eliminated through exhalation, perspiration, urination, and defecation.
Each element has a measurable biological half-life - the time required to reduce the amount of material in the body to one-half of its original value.
The biological half-life is independent of the physical or radiological half-life.
261
Elimination Processes
Radioactive Decay
Each radioactive nuclide has a distinctive decay rate that is not influenced by any physical process, including biological functions.
The amount of time required for one half of the material in the body to decay is called the radiological half-life.
Radioactive decay will result in reduction of the quantity of the original nuclides deposited in the body.
It is important to remember that the progeny of these nuclides may also be radioactive.
It is possible that the progeny could introduce completely different concerns for internal dose assessments.
262
Effective Half-life
The combined processes of biological elimination and
physical decay result in the removal of radioactive
materials at a faster rate than the individual reduction
rate produced by either method.
This means that:
Te < Tb, Tp
263
Effective Half-life
The removal rate as a result of the combined processes
is measured as an effective half-life and is calculated
using the following formula:
264
Effective Removal Constant
Another way that this is expressed is the effective
removal constant, λe, which is the composite of the
physical decay constant λp and the biological
elimination constant λb.
λe = λb + λp
265
Example Calculations
Determine the effective half-life of tritium if the biological
half-life is 10 days and the physical half-life is 12.3 years.
Determine the effective half-life of 59-Fe if the biological
half-life is 2000 days and the physical half-life is 44.56 days.
266
Blocking Agents
A blocking agent saturates the metabolic processes in a
specific tissue with the stable element and reduces uptake
of the radioactive forms of the element.
As a rule, these must be administered prior to or almost
immediately after the intake for maximum effectiveness and
must be in a form that is readily absorbed.
The most well known example of this is stable iodine, as
potassium iodide, which is used to saturate the thyroid
gland, thus preventing uptake of radioactive iodine in the
thyroid.
267
Diluting Agents
A diluting agent is a compound which includes a stable
form of the nuclide of concern. By introducing a large
number of stable atoms, the statistical probability of the
body incorporating radioactive atoms is reduced.
Diluting agents can also involve the use of different
elements which the body processes in the same way. This
type of treatment is called displacement therapy.
The compound used must be as readily absorbed and
metabolized as the compound that contains the
radioisotope.
268
Mobilizing Agents
A mobilizing agent is a compound that increases the
natural turnover process, thus releasing some forms of
radioisotopes from body tissues.
Usually most effective within 2 weeks after exposure;
however, use for extended periods may produce less
dramatic reductions.
269
Chelating Agents
A chelating agent is a compound which acts on
insoluble compounds to form a soluble complex ion
which can then be removed through the kidneys.
Commonly used to enhance elimination of transuranics
and other metals.
Therapy is most effective when begun immediately after
exposure if metallic ions are still in circulation and is
less effective once metallic ions are incorporated into
cells or deposited in tissue such as bone.
270
Chelating Agents
Common chelating agents include EDTA and DTPA
–CaNa2 EDTA - commonly used in cases of lead
poisoning. It is also effective against zinc, copper,
cadmium, chromium, manganese, and nickel.
–CaNa3 DTPA - transuranics such as plutonium and
americium. If the chelating agent is administered within
a few hours of the intake the residual transuranics in the
body can be reduced by a factor of 10.
271
Diuretics
Diuretics increase urinary excretion of sodium and water.
Used to reduce internal exposure, however its use has
been limited. Applications include 3-H, 42-K, 38-Cl and
others.
Can lead to dehydration and other complications if not
performed properly.
272
Expectorants and Inhalants
Used to increase flow of respiratory tract excretions.
Inhalants have been used in removing radioactive
particles from all areas of lungs.
273
Lung Lavage
Involves multiple flushing of lungs with appropriate fluid to
remove radioactive materials in the lungs.
Usually limited to applications where resulting exposures
would result in appearance of acute or subacute radiation
effects.
The procedure has been proven effective in cases of “black
lung” disease exhibited by coal miners where the procedure
is used to remove some of the small coal particles that were
deposited at the alveolar region of the lungs.
AIR MONITORING
Voss Associates
274
275
Inhalation of radioactive particles is the largest cause of
internal dose.
Airborne radioactivity measurements are necessary to
ensure that the control measures are effective and continue
to be effective.
Regulations govern the allowable effective dose equivalent
to an individual.
276
The effective dose equivalent is determined by combining
the external and internal dose equivalent values.
Typically, airborne radioactivity levels are maintained well
below allowable levels to keep the total effective dose
equivalent small.
277
Annual Limit on Intake (ALI) - The quantity of a single
radionuclide which, if inhaled or ingested in one year,
would irradiate a person, represented by reference man
(ICRP Publication 23), to the limiting value for control of
occupational exposure.
Derived Air Concentration (DAC) - The concentration of
a radionuclide in air that, if breathed over a period of a
work year, would result in the ALI for that radionuclide
being reached. The DAC is obtained by dividing the ALI by
the volume of air breathed by an average worker during a
working year.
278
The primary objectives of an air monitoring program are:
• To measure the concentration of the radioactive
contaminant(s) in the air by collection and analysis
• To identify the type and physical characteristics of the
radioactive contaminant
• To help evaluate the hazard potential to the worker
• To evaluate the performance of airborne radioactivity
control measures
• To assess air concentration data in order to determine if
bioassay sampling should be initiated to verify whether an
exposure has occurred, and if so, to determine the
magnitude of the exposure
279
Air sampling is performed;
• To establish the need for posting of airborne radioactivity
areas and to determine the need for respiratory protection
of workers
• To assess unknown hazards during maintenance on
systems contaminated with radioactive material or when
there is a loss of process controls
• To assist in determining the type and frequency of
bioassay measurements needed for a worker
• To provide an estimate of worker exposures for
situations where bioassay measurements may not be
available or their validity is questionable
280
• To develop baseline airborne radioactivity levels and verify
containment integrity as necessary during startup of a new
facility or new operation within an existing facility
• Where respiratory protection devices for protection against
airborne radionuclides have been required
• Real-time air monitoring shall be performed as necessary
to detect and provide warning of airborne radioactivity
concentrations that warrant immediate action to
terminate inhalation of airborne radioactive material
281
THE NATURE OF AIRBORNE RADIOACTIVITY
Airborne radioactive contaminants are generally divided
into three categories, based on the physical state of the
contaminant.
• Particulates
• Gases
• Vapors
The physical properties of airborne radioactive particles
can affect inhalation deposition, their dynamical properties
in air, and particle solubility in the lung.
282
Particulates
Particulate contaminants are solid and liquid particles,
ranging upward from molecular sizes (approximately 10-3
μm), suspended in the air.
Solids may be subdivided into fumes, dusts, and smokes.
Liquids are subdivided into mists and fogs.
The term "aerosols" is used to collectively refer to relatively
stable suspensions of either solid or liquid particles in a
gaseous medium.
Generally, particulates are more readily retained in the lungs
than are gases, but retention of particulates is highly
dependent on particle size and solubility in the lung.
283
Particulate airborne contaminant sampling should measure
particle size, however this is not usually done on a routine
basis.
The size range of particles retained in the respiratory tract is
generally 1-10 μm.
The retention of inhaled radioactive particles after deposition
in the pulmonary region of the lung is strongly influenced by
the dissolution characteristics of the particles.
Dissolution in the lungs allows clearance into the blood and
the rest of the systemic circulation.
284
For this reason the various chemical forms of radioactive
particles are classified with respect to their potential
solubility in the lungs.
Specifically, these are:
• Class Y for the very insoluble particle that takes years to
clear from the lungs
• Class W for the somewhat more soluble particles that
take weeks to dissolve and clear into the systemic
circulation
• Class D for the relatively soluble particles that dissolve
in a matter of days in the lung
285
Gases
Gases are substances that, under normal conditions of
temperature and pressure, exist in the gaseous phase.
The retention of the gases in the body from inhalation is
poor so radioactive gases are usually treated as an external
source of exposure.
Radioactive gases typically found are the fission product
gases, such as xenon and krypton, and naturally occurring
radon.
While the gases contribute primarily to external exposure,
the particulate progeny to which they decay can contribute
to internal exposure.
286
Vapors
Vapors are considered the gaseous phase of a substance
that is a solid or liquid under normal conditions of
temperature and pressure.
Airborne vapor sampling is most commonly done for
radioiodine and tritium.
The contaminant may be dispersed in vapor form at
abnormal conditions of temperature and pressure.
However, as the temperature and pressure conditions return
to "normal," the contaminant will return to its normal solid or
liquid form, or become an aerosol.
287
REPRESENTATIVE AIR SAMPLES
To ensure that the sample is representative of the actual
conditions:
• The airborne radioactivity concentration entering the
sample line must be representative of the airborne
radioactivity concentration in the air near the sampling
device.
• The airborne radioactivity concentration entering the
sampling inlet must be representative of the airborne
radioactivity concentration at the point of concern, or
the air that is breathed, i.e., breathing zone.
288
When obtaining an air sample, care must be taken to
ensure that the sample obtained is representative of the
air around the sampling device.
This is particularly important for sample lines that directly
sample an air flow, such as a stack or duct monitor.
Air flow into sampling lines needs to be balanced with
respect to the flow of air around the probe or sample inlet.
To ensure the sample is representative, the flow velocity
in the sample line or inlet must be the same as the flow
velocity in the system, such as the duct or stack.
289
BASIC SAMPLING METHODS
Basically, three types of samples are collected:
1. A volumetric sample in which part of the atmosphere is
isolated in a suitable container, providing the original
concentration of the contaminant at a particular place and
time.
2. An integrated sample which concentrates the
contaminant on some collecting medium, providing an
average concentration over the collection time.
3. A continuous sample where the sample air flow is
directed past or through a detection device providing a
measurement of the activity per unit volume of air.
290
There are six general methods for obtaining samples or
measurements of airborne radioactivity concentrations.
• Filtration
• Volumetric
• Impaction/impingement
• Adsorption
• Condensation/ dehumidification
• In-line/flow-through detection
291
Filtration
Filter samplers draw air through a removable filter medium
at a known flow rate for a known length of time as the
method of concentrating the airborne radioactive particulate
(aerosol) contaminants.
Filtration is the most common sampling method employed
for particulates because it is relatively simple and efficient,
but sampling for gases and vapors requires other methods.
292
The filtration medium selected for a sample depends on
several factors:
the collection efficiency required,
the flow resistance of the medium,
the mechanical strength of the filter,
pore size,
the area of the filter,
the background radioactive material of the filter,
cost,
self-absorption within the filter, and
chemical solubility.
293
The most common types of filters are:
• Cellulose-asbestos filters
• Glass fiber filters
• Membrane filters are manufactured with various pore sizes
and many can be dissolved in organic solvents or ashed and
then analyzed in a counter.
Filter samples may also be evaluated by direct radiation
counting.
Filters may be mounted into different types of holders such
as those with open faces for direct sampling and those with
in-line enclosure for sampling through a sampling line.
294
Volumetric
Volumetric samplers employ a sample container into which
the sample is drawn. Several methods are employed to
draw the sample into the container.
• The container may be evacuated by a vacuum pump and
isolated away from the sample location. The container is
opened at the sample location to draw the air into the
container.
• A sample pump may be used at the sample location.
• The container could be filled with water, then when the
water is poured out of the container the air sample is drawn
into the container.
295
Impaction/Impingement
Impingers or impactors concentrate particulate contaminants
on a prepared surface by abruptly changing the direction of
the sample air flow at some point in the sampler.
Particles are collected on a selected surface as the
airstream is sharply deflected. Due to their inertia, the
particles are unable to follow abrupt changes in airstream
direction.
The surface on which the particles are collected must be
able to trap the particles and retain them after impaction.
Several methods are commonly used to trap the particles,
such as:
• Coating the collection surface with a thin layer of grease or
adhesive.
• Immersing the collection surface in a fluid, such as water or
alcohol. The liquid is then analyzed after the sample is
collected.
296
Adsorption
Adsorber sampling devices concentrate the contaminants by
causing them to adhere to the surface of the adsorption
medium.
Adsorption is the adhesion of a substance to the surface of
another substance through chemical bonding.
The adsorption medium is granulated or porous to increase
the surface area available for trapping of the contaminant.
Adsorbers, such as activated charcoal, silica gel, and silver
zeolite, are commonly used to collect organic vapors and
non-reactive gases and vapors.
297
Some uses of each type of absorption methods are:
• Activated charcoal is used primarily for radioiodine and
radon sampling, but does trap noble gases, such as xenon,
krypton and argon.
• Silica gel is primarily used for tritium oxide vapor sampling.
• Silver zeolite is used for radioiodine sampling when
trapped noble gases would interfere with the radioiodine
analysis.
A paper filter upstream of the absorption cartridge may be
used to filter out particles to prevent interference during the
analysis of the media.
298
Condensation/Dehumidification
Condensation or dehumidifier sampling devices employ a
"cold trap" such as liquid nitrogen or a refrigeration unit to
condense water vapors in the sampled atmosphere and
provide a liquid sample for further analysis.
The collected water is frequently analyzed using a liquid
scintillation counter.
Calculations must include the relative humidity and
temperature of the air at the time the sample is taken to
determine the concentration of water vapor per unit volume
of air.
This technique is normally only applied for sampling tritium
oxide vapor (HTO or T2O).
299
In-Line/Flow Through Detection
In-line or flow-through samplers direct the sample air
flow through or past the detection device.
This method is employed for radionuclides which are
difficult to collect or detect by other means.
Because the air flow passes directly outside the detector
or actually through the inside of the detector, the air
should not contain other particulates or vapors that could
accumulate on or in the detector.
Flow-through detectors are employed for radionuclides
such as tritium which emit low energy radiation, which
could not otherwise pass through the detector window.
300
Multi-Purpose Samplers and Monitors
The various sampling methods may be combined into
one sampler.
Some samplers employ the filtration method for
particulates, the adsorption method for vapors and the
volumetric grab-sample method for gases (in that
order).
301
PRIMARY TYPES OF AIR SAMPLERS
The five primary types of airborne radioactivity
samplers/monitors are:
• Personal air samplers (breathing zone)
• High volume/flow rate air samplers
• Low volume/flow rate air samplers
• Portable Continuous Air Monitors (CAMs)
• Installed Continuous Air Monitoring systems
302
Personal Air Samplers
Personal air samplers (PASs) provide an estimate of the
airborne radioactivity concentration in the air the worker is
breathing during the sampling period.
PASs may also be used to determine if the protection factor
for respiratory equipment is exceeded, to compare with
other workplace air samples, and to verify the effectiveness
of engineered and administrative controls.
PASs are small, portable battery-powered devices which
sample the air in the breathing zone of the worker's
environment.
The sample cartridge is removed and counted after the
worker finishes the task.
303
High Volume/Flow Rate Samplers
High volume/flow rate samplers provide an estimate of the
airborne radioactivity concentration at a particular location
in a short period of time.
Portable high flow rate samplers are used to collect
aerosols on a filter paper (filtration) or on by impaction.
Portable high flow rate samplers can also be used to collect
radioiodine samples using activated charcoal cartridges.
These samplers do not have installed detectors and the
sample must be removed from the sampler to be analyzed.
304
The high volume/flow rate samplers may be used to:
• Provide a routine "slice of time" estimate of the general
area airborne radioactivity
• Verify boundaries of areas posted for airborne
radioactivity
• Monitor the airborne radioactivity related to a specific
work activity
High volume samplers typically use flow rates of 30 cubic
feet per minute (cfm).
Although these samplers are noisy and not intended for
continuous duty, the shorter sample times allow for
greater sensitivity by collecting a large volume of air.
305
Low Volume/Flow Rate Samplers
Low volume/flow rate samplers provide an estimate of
airborne radioactivity concentrations averaged over a
longer period of time at a particular location.
Portable low volume/flow rate samplers are used much like
High Volume Samplers. The Low Volume Sampler needs
to operate longer than the High Volume Sampler to collect
a similar volume of air.
306
Portable Continuous Air Monitors
Portable Continuous Air Monitors (CAMs) provide an
estimate of airborne radioactivity concentrations, and
provide immediate readout and alarm capabilities for preset
concentrations.
These air monitors are typically semi-portable monitoring
systems, containing the necessary sampling devices and
built-in detection systems to monitor the activity on the
filters, cartridges, planchettes and/or chambers in the
system.
Typical CAMs provide information on alpha and/or
beta/gamma particulates (filtration), radionuclide activity on
absorption cartridges, noble gas activity (volumetric
chamber or in-line detector), and tritium activity.
307
Installed Continuous Air Monitors
Installed continuous air monitoring systems (CAMs) provide
an estimate of airborne radioactivity concentrations at a
fixed location and provide immediate local and remote
readout and alarm capabilities for preset concentrations.
These air monitors are fixed flow rate sampling systems,
and contain the necessary sampling devices and built-in
detection systems.
The system may provide a local and remote recording
system for data, and computer functions such as data
trending, preset audible and visual alarms/warning levels
and alerts for system malfunctions.
308
Installed CAM applications include:
• Fixed installations capable of sampling several locations
through valved sample lines
• Work area monitors
• Stack monitors
• Duct monitors
309
Factors affecting the accuracy of airborne radioactivity
measurements include:
• Sample is not representative of the atmosphere being
sampled
• Sample is not representative of the air being breathed
by the worker
• Incorrect or improperly installed sampling media for the
selected sampler, causing leaks or improper flow rates
• Malfunctioning, miss-operated, or improperly calibrated
sampling device, causing errors in flow rate
measurements
310
• Accuracy and operation of the timing device, causing
errors in the time value
• Accuracy and operation of the flow rate measuring
device, causing errors in the flow rate value
• Mishandling of the sample media causing cross-
contamination or loss of sample material
• Changes in the collection efficiency of the medium due
to sample loading, humidity or other factors
311
• Improper use or selection of analysis equipment
• Inherent errors in the counting process due to sample
geometry, self-absorption, resolving time, backscatter and
statistical variations
• Mathematical errors during calculations due to rounding of
numbers and simple mistakes
• Incorrect marking of samples and inaccurate recording of
data
312
BASIC AIR SAMPLE CALCULATIONS
Once the air sample is collected and analyzed,
calculations must be performed to determine the amount
of activity per unit volume.
This value may be corrected for decay for the time period
between when the sample was taken to when it was
analyzed. This is especially true for short-lived
radionuclides.
313
Exercise:
Calculate the dose a worker would receive if they
were exposed to 300 mCi/m3 of HTO for 30 minutes.
Solution:
1 DAC is 20 μCi/m3,
so 300 mCi/m3 is 15,000 DAC for 0.5 hours (30 min)
which is 7,5000 DAC-hours.
One DAC-hr is 2.5 mrem, so 7,500 DAC-hrs is 7,500
x 2.5, or 18,750 mrem, or 18.75 rem.
RESPIRATORY
PROTECTION
Voss Associates
314
315
The Occupational Safety and Health Standard, 29 CFR, Part
1910.134, specifies the minimal acceptable respiratory
protection program must contain or address the following:
• Written standard operating procedures governing the
selection and use of respirators shall be established.
• Respirators shall be selected on the basis of hazards to
which the worker is exposed.
316
• The user shall be instructed and trained in the proper
use of respirators and their limitations.
• Respirators shall be regularly cleaned and
disinfected. Those issued for the exclusive use of one
worker should be cleaned after each day's use, or
more often if necessary.
317
• Respirators shall be stored in a convenient, clean, and
sanitary location.
• Respirators used routinely shall be inspected during
cleaning. Worn or deteriorated parts shall be replaced.
• Respirators for emergency use such as self-contained
devices shall be thoroughly inspected at least once a
month and after each use.
• Appropriate surveillance of worker area conditions and
degree of employee exposure or stress shall be maintained.
• There shall be regular inspection and evaluation to
determine the continued effectiveness of the program.
• Persons should not be assigned to tasks requiring use of
respirators unless it has been determined that they are
physically able to perform the work and use the equipment.
The local physician shall determine what health and
physical conditions are pertinent. The respirator user's
medical status should be reviewed periodically (for
instance, annually).
318
• Approved or accepted respirators shall be used when
they are available. The respirator furnished shall
provide adequate respiratory protection against the
particular hazard for which it is designed in accordance
with standards established by competent authorities.
ANSI Z88.2-1992 further specifies the minimal
acceptable program for industries involved in the use
of radioactive material.
319
320
TABLE 1—ASSIGNED PROTECTION FACTORS 5
Type of Quarter Half Full Helmet/ Loose
respirator 1,2 mask mask face- Hood fitting
piece face-
piece
1. Air-Purifying Respirator …………. 5 …….310 ….. 50 .............................
2. Powered Air-Purifying …………………………………….….
Respirator (PAPR) ........................................ 50 … 1,000 … 425 …... 25
3. Supplied-Air Respirator (SAR)
or Airline Respirator.
• Demand mode ........................................... 10 …….50 ............................
• Continuous flow mode .................................50 …1,000 …. 425 …… 25
• Pressure-demand or other
positive-pressure mode ................................ 50 … 1,000 ..........................
321
TABLE 1—ASSIGNED PROTECTION FACTORS 5
Type of Quarter Half Full Helmet/ Loose
respirator 1,2 mask mask face- Hood fitting
piece face-
piece
4. Self-Contained Breathing
Apparatus (SCBA).
• Demand mode ........................................... 10 …….50 ..... 50 .................
• Pressure-demand or other
positive-pressure mode
(e.g., open/closed circuit)...................................... 10,000 . 10,000 ……….
322
1 Employers may select respirators assigned for use in
higher workplace concentrations of a hazardous substance
for use at lower concentrations of that substance, or when
required respirator use is independent of concentration.
2 The assigned protection factors in Table 1 are only
effective when the employer implements a continuing,
effective respirator program as required by this section,
including training, fit testing, maintenance, and use
requirements.
3 This APF category includes filtering facepieces, and half
masks with elastomeric facepieces.
323
4 The employer must have evidence provided by the
respirator manufacturer that testing of these respirators
demonstrates performance at a level of protection of 1,000
or greater to receive an APF of 1,000. This level of
performance can best be demonstrated by performing a
WPF or SWPF study or equivalent testing. Absent such
testing, all other PAPRs and SARs with helmets/hoods are
to be treated as loose-fitting facepiece respirators, and
receive an APF of 25.
5 These APFs do not apply to respirators used solely for
escape. For escape respirators used in association with
specific substances covered by 29 CFR 1910 subpart Z,
employers must refer to the appropriate substance-specific
standards in that subpart. Escape respirators for other IDLH
atmospheres are specified by 29 CFR 1910.134 (d)(2)(ii).
324
APPENDIX A TO 10CFR20—ASSIGNED PROTECTION
FACTORS FOR RESPIRATORSA Assigned
Operating Protection
mode Factors
I.Air Purifying Respirators [Particulate b only] c:
Filtering facepiece
disposable d. Negative Pressure .... (d)
Facepiece, half e Negative Pressure .... 10
Facepiece, full Negative Pressure .... 100
Facepiece, half Powered air-purifying
respirators. ………… 50
Facepiece, full Powered air-purifying
respirators. ………… 1000
Helmet/hood Powered air-purifying
respirators. ………… 1000
Facepiece, loose-fitting. Powered air-purifying
respirators. ………… 25
325
II. Atmosphere supplying respirators particulate, gases,
1. Air-line respirator: and vaporsf
Facepiece, half Demand ............................ 10
Facepiece, half Continuous Flow ............... 50
Facepiece, half Pressure Demand ……..... 50
Facepiece, full Demand ............................ 100
Facepiece, full Continuous Flow ............... 1000
Facepiece, full Pressure Demand ……...... 1000
Helmet/hood Continuous Flow ................ 1000
Facepiece, loose-fitting. Continuous Flow ................ 25
Suit Continuous Flow ………..... (g)
2. Self-contained breathing Apparatus (SCBA):
Facepiece, full Demand ............................ h100
Facepiece, full Pressure Demand ……….. i10,000
Facepiece, full Demand, Recirculating.…. h100
Facepiece, full Positive Pressure Recirculating.. i10,000
326
III. Combination Respirators: Assigned protection
Any combination of air-purifying and factor for type and
atmosphere-supplying respirators. mode of operation
as listed above.
327
a These assigned protection factors apply only in a
respiratory protection program that meets the requirements
of this Part. They are applicable only to airborne radiological
hazards and may not be appropriate to circumstances when
chemical or other respiratory hazards exist instead of, or in
addition to, radioactive hazards. Selection and use of
respirators for such circumstances must also comply with
Department of Labor regulations.
Radioactive contaminants for which the concentration values
in Table 1, Column 3 of Appendix B to Part 20 are based on
internal dose due to inhalation may, in addition, present
external exposure hazards at higher concentrations. Under
these circumstances, limitations on occupancy may have to
be governed by external dose limits.
328
d Licensees may permit individuals to use this type of
respirator who have not been medically screened or fit
tested on the device provided that no credit be taken for
their use in estimating intake or dose. It is also recognized
that it is difficult to perform an effective positive or negative
pressure pre-use user seal check on this type of device. All
other respiratory protection program requirements listed in
§ 20.1703 apply. An assigned protection factor has not
been assigned for these devices. However, an APF equal
to 10 may be used if the licensee can demonstrate a fit
factor of at least 100 by use of a validated or evaluated,
qualitative or quantitative fit test.
329
e Under-chin type only. No distinction is made in this
Appendix between elastomeric half-masks with
replaceable cartridges and those designed with the filter
medium as an integral part of the facepiece (e.g.,
disposable or reusable disposable). Both types are
acceptable so long as the seal area of the latter contains
some substantial type of seal-enhancing material such as
rubber or plastic, the two or more suspension straps are
adjustable, the filter medium is at least 95 percent
efficient and all other requirements of this Part are met.
330
f The assigned protection factors for gases and vapors are
not applicable to radioactive contaminants that present an
absorption or submersion hazard. For tritium oxide vapor,
approximately one-third of the intake occurs by absorption
through the skin so that an overall protection factor of 3 is
appropriate when atmosphere-supplying respirators are
used to protect against tritium oxide. Exposure to
radioactive noble gases is not considered a significant
respiratory hazard, and protective actions for these
contaminants should be based on external (submersion)
dose considerations.
331
g No NIOSH approval schedule is currently available for
atmosphere supplying suits. This equipment may be used
in an acceptable respiratory protection program as long as
all the other minimum program requirements, with the
exception of fit testing, are met (i.e., § 20.1703).
h The licensee should implement institutional controls to
assure that these devices are not used in areas
immediately dangerous to life or health (IDLH).
332
i This type of respirator may be used as an emergency
device in unknown concentrations for protection against
inhalation hazards. External radiation hazards and other
limitations to permitted exposure such as skin absorption
shall be taken into account in these circumstances. This
device may not be used by any individual who
experiences perceptible outwardleakage of breathing
gas while wearing the device.
333
RESPIRATORY PROTECTION EQUIPMENT
Air Purifying, Particulate-Removing Filter Respirators
These are often called "dust," "mist," or "fume" respirators
and by a filtering action remove particulates before they
can be inhaled. Single use, quarter mask, half mask, full
facepiece, and air powered hood/mask are the five types
of respirators that work by the particulate removal method.
Air purifying respirators generally operate in the negative
pressure (NP) mode; that is, a negative pressure is
created in the facepiece during inhalation. An exception is
a special type of powered air purifying respirator (PAPR)
that operates by using a blower to drive the contaminated
air through an air purifying filter or sorbent canister.
334
Air Purifying, Chemical Cartridge and Canister
Respirators for Gases and Vapors
Vapor and gas-removing respirators use cartridges or
canisters containing chemicals (i.e., sorbents) to trap or
react with specific vapors and gases and remove them
from the air breathed. The basic difference between a
cartridge and a canister is the volume of the sorbent. A
particulate filter may be combined with a chemical
cartridge to form a combination cartridge.
335
Atmosphere Supplying Respirators - Supplied Air
Supplied air respirators use a central source of breathing
air that is delivered to the wearer through an air supply
line or hose. The respirator type is either a tight-fitting
facepiece (half face or full) or loose-fitting hood/suit.
There are essentially two major groups of supplied air
respirators - the air-line device and the hose mask with
or without a blower. The operating modes are demand,
pressure demand, and continuous flow.
336
In a demand device, the air enters the facepiece only on
"demand" of the wearer, i.e., when the person inhales.
During inhalation, there is a negative pressure in the mask,
so if there is leakage, contaminated air may enter the mask
and be inhaled by the wearer.
The pressure demand device has a regulator and valve
design such that there is a flow (until a fixed static pressure is
attained) of air into the facepiece at all times, regardless of
the "demand" of the user. The airflow into the mask creates a
positive pressure.
The continuous-flow air line respirator maintains a constant
airflow at all times using an airflow control valve or orifice
which regulates the flow of air. The continuous flow device
does not guarantee a positive pressure in the facepiece.
337
To utilize the Protection Factor (PF) assigned to air
supplied hoods, a delivery flow rate of at least 6 CFM but
not greater than 15 CFM must be obtained. The individual
user's air flow valves should not be altered to maintain a
minimum delivery flow rate of 6 CFM as this violates the
NIOSH/MSHA approval. Taping or otherwise securing the
airflow valves in the fully open position does not void the
NIOSH/MSHA approval provided the valve is not
permanently altered or made so that it would be
impossible to increase or decrease the air flow by the
user.
338
Atmosphere Supplying Respirators - Self-Contained
Breathing Apparatus (SCBA)
The self-contained breathing apparatus (SCBA) allows the
user to carry a breathing air supply and does not need a
stationary air source such as a compressor to provide
breathable air. The air supply may last from 3 minutes to 4
hours depending on the nature of the device.
There are two groups of SCBAs - the closed circuit and the
open circuit. Another name for closed circuit SCBAs is
"rebreathing" device. The air is rebreathed after the
exhaled carbon dioxide has been removed and the oxygen
content restored by a compressed oxygen source or an
oxygen-generating solid. These devices are designed
primarily for 1-4 hours use in toxic atmospheres.
339
An open circuit SCBA exhausts the exhaled air to the
atmosphere instead of recirculating it. A tank of
compressed air carried on the back, supplies air via a
regulator to the facepiece. Because there is no recirculation
of air, the service life of the open circuit SCBA is shorter
than the closed circuit system. The pressure demand open
circuit SCBA has a regulator and a valve design which
maintains a positive pressure in the facepiece at all times
regardless of the "demand" of the user. Because of the high
degree of protection provided by the pressure-demand
SCBA, this type of unit is recommended for emergency use,
escape and rescue.
340
AIR QUALITY TESTING
An air quality testing program for all sources of respirable
air is required. Compressed breathing air shall meet at
least the quality specification for Grade D breathing air as
described in Compressed Gas Association Commodity
Specification G-7.1-1989.
341
342
343
344
345
346
ALARA
Voss Associates
347
ALARA Philosophy
The assumption is that a proportional relationship exists
between dose and effect for all doses; this is the basis for
ALARA.
The effects of low-level doses over extended periods of
time are not definitively characterized and the risk is difficult
to quantify.
Studies of atomic bomb survivors and individuals involved
in nuclear incidents show the relationship between dose
and effects is well known only at high doses.
The benefit of completing a task must be compared to the
risk of the exposure received.
348
Objectives of ALARA Programs
There should not be any occupational exposure of
workers to ionizing radiation without the expectation of
an overall benefit from the activity causing the exposure.
Personal radiation exposure shall be maintained ALARA.
Radiation exposure of the work force and public shall be
controlled such that radiation exposures are well below
regulatory limits and that there is no radiation exposure
without commensurate benefit.
349
ALARA Concerns
Program
Engineering features
Discharge of radioactive liquid to the environment
Control of contamination
Efficiency of maintenance, decontamination and
operations should be maximized
Components should be selected to minimize the buildup
of radioactivity
350
ALARA Concerns
Program
Support facilities should be provided for donning and
removal of protective clothing and for personnel monitoring
Shielding requirements
Ergonomics consideration
Access control designed for hazard level
Surfaces that can be decontaminated or removed
Equipment that can be decontaminated
351
ALARA Concerns
Area arrangement
Traffic patterns to allow access yet prevent unnecessary
exposure
Equipment separation
Valve locations
Component laydown/storage areas
352
ALARA Concerns
Operations
Inspection tour - access, mirrors, visibility
Inservice Inspections - use of remote control equipment,
TV, Snap on insulation, platforms, etc.
Remote readout instrumentation
Remote valve/equipment operators
Sampling stations, piping, valving, hoods, sinks
353
ALARA Concerns
Maintenance needs
Adequate lighting, electric outlets, other utilities
Removal and storage areas for insulation/shrouding
Relocation of components to low dose areas
Workspace for maintenance personnel
Lifting equipment
Conditions that could cause or promote the spread of
contamination, such as a leaking roof or piping need to be
identified and corrected on a priority basis.
354
ALARA Concerns
Radiological Control Needs
Access control
Shielding adequacy and access plugs
Temporary shielding and support structures
Adequate ventilation
Breathing air
355
ALARA Concerns
Radiological Control Needs
Contamination control - drip pans, curbs, drains, and routing
Decontamination facilities
Radiation monitoring equipment
Communications
356
Collective Dose Philosophy
Control of the collective dose to the work force.
Collective dose is defined as the total individual doses in a
group or a population.
Spreading dose among more workers versus higher
individual exposures for fewer workers is an ALARA issue.
357
Spreading dose
The linear model states that the less exposure a worker
receive the less chance they will receive harmful
biological effects.
Lower collective dose is a good indicator of an effective
ALARA Program
358
Higher Individual Exposure
Exposure to fewer individuals means that the risk to the
rest of the work force has been minimized.
Merely controlling maximum dose to individuals is not
sufficient, collective dose must be controlled as well.
Reducing radiological risks should not result in higher
risks for other hazards.
Reduction in radiological risk should be reasonably
achievable based on the current state of technology,
economic factors, and social conditions
359
Scope of ALARA Program
Establish a program to maintain exposures ALARA.
Design and modify facilities and select equipment with
ALARA concepts integrated into the processes.
Establish radiological control programs, plans and
procedures.
Make available equipment, instrumentation and facilities
necessary for ALARA program implementation.
360
Scope of ALARA Program
Train facility workers, management, and radiological control
personnel in ALARA programs and reduction techniques.
Applies equally to the reduction of external and internal
exposure.
The ALARA program must be incorporated in everyday,
routine functions as well as non-routine, higher risk tasks.
The involvement and commitment of all facility personnel, not
just radiological control personnel, is necessary to achieve
reduction of external and internal exposures.
361
Scope of ALARA Program
To justify activities that could result in exposure to ionizing
radiation, the following conditions should be satisfied:
–The risks associated with projected radiation exposures
should be small when compared to the benefit derived.
–Further reduction in projected exposure is evaluated
against the effort required to accomplish such reduction.
–The risks from occupational exposure or to the public
should not exceed everyday or accepted risks.
362
Ownership
Each individual involved in radiological work must
demonstrate responsibility and accountability through
an informed, disciplined and cautious attitude toward
radiation and radioactivity.
366
Management Responsibilities
Design and implement ALARA program
Provide resources such as tools, equipment, and
adequate personnel
Create and support ALARA Review Committee
Approve ALARA goals
Design and implement worker training
367
Radiological Control Technician Responsibilities
Perform the functions of assisting and guiding workers in
the radiological aspects of the job
Knowledge of conditions at the work site
Knowledge of work activities to be performed
Identification of protective clothing and equipment
requirements
365
Radiological Control Technician Responsibilities
Identification of dose reduction techniques
During work conduct, maintaining awareness of conditions
Correction of worker mistakes
Response to abnormal events
366
ALARA "Group“
(Including Facility/RC Supervision/Management)
Evaluate worker suggestions and provide feedback in a
timely manner
Participate in pre-and post-work meetings
Keep abreast of ALARA techniques pertinent to
operations on site
Track facility performance in comparison to stated goals
367
Pre-job ALARA Reviews
For every task involving radiological work, sufficient
radiation protection controls should be specified in
procedures and work plans to define and meet
requirements.
Applicable ALARA practices shall be factored into the
plans and procedures for each task or type of task. The
practices shall be communicated to the workers in ways
that ensure that the employee is able to maintain their
exposure ALARA.
Proposed ALARA protective measures shall be
evaluated to ensure the costs are justified.
368
Pre-Job Briefing
Pre-job briefings are held with employees who will be
involved in work activities involving unusual radiological
conditions.
Identify effective dose reduction measures.
RC needs are communicated to workers. Worker needs
are communicated to RC.
Procedures are verified.
369
Pre-Job Briefing
Worker qualifications are verified.
Emergency procedures are discussed.
At the end of the meeting, everyone should know what
is expected of them, how to do it, and the conditions
under which it is to be done.
370
Pre-Job Briefing
ALARA pre-job briefing checklists
–Scope of work to be performed
–Radiological conditions of the workplace
–Procedural and RWP requirements
–Special radiological control requirements
–Radiologically limiting conditions, such as contamination
or radiation levels that may void the RWP
–Radiological Control Hold Points
–Communications and coordination with other groups
–Provisions for housekeeping and final cleanup
–Emergency response provisions
371
Post-Job ALARA Reviews
Jobs determined to require a ALARA review shall
undergo a post-job review to ensure the overall
effectiveness of job planning and implementation.
Unusual exposure events are investigated to determine
the root cause.
Recommendations are made and corrective actions are
then taken to prevent future reoccurrences of these
events.
372
Post-Job Debriefing
Although post-job debriefings will not affect the dose
already received for a particular job, they can be
effective in reducing the doses received the next time
that job is performed.
Information discussed at post-job meetings include
discussions of what went wrong and what could have
been done differently to reduce the exposures received.
Post-job meetings rely heavily on the input of each
radiation worker for information on how best to reduce
exposure the next time that job is performed.
373
Post-Job Debriefing
Typical questions asked could include:
– Were there any problems performing the job in
accordance with the procedure?
– Did you have the tools and equipment needed to perform
the work? Could special tools ease the job?
– Were there any unexpected conditions noted during the
work? Could these conditions have been anticipated?
– Were there any unexpected delays in the performance of
the job? What was the cause of the delay?
– Was temporary shielding used? Could the use of
temporary shielding reduce exposures received for this
job?
374
Responsibilities of the RCT
Pre-job ALARA reviews
Pre-job briefings
Radiation hold points identified
Tool and equipment requirements/need for special tools
–Pre-fabrication of temporary shielding
–Removal of component to low dose areas
–Previous job evolutions, previous survey conditions
375
Responsibilities of the RCT
Area Set-up
–Access to and from work area
–Service lines available - air, electric, ventilation, lighting
–Staging areas - low radiation areas, tool preparation andpersonnel waiting areas
–Communications - equipment, lines, TV monitoring
–Radiological controls - anticipation of conditions during jobwith identification of controls required, surveys completed, high and low dose areas identified, contamination control requirements, airborne
376
Responsibilities of the RCT
Worker preparation
– Experienced workers
– Specialized training - mock-ups, photographs,
rehearsals, etc.
– Briefings - conditions, needs of RC personnel, what
to expect, abnormal conditions
– Pre-work check off packages
377
Responsibilities of the RCT
Conduct of the job
– The technician is tasked with assisting other workers inmaintaining their exposures ALARA.
– The technician can not lose sight of their own exposurereduction needs.
– The RCT is expected to observe the workers to ensurethat the radiological control requirements pertinent to thehazards present are taken and followed properly.
– If the RCT observes the workers not following goodradiological work practices, on the spot correctionsshould be made.
Isotopes Good to Know
Mo-99 Tc-99m Tc-99 Co-57 Co-59 Co-60 Ni-58 Fe-54 Fe-58 Fe-59 Mn-54 Mn-55 Mn-56 H-3 C-14 N-16 Sr-90 Y-90 P-32 I-125 I-131 Xe-131m Xe-130 Cs-137 Ba-137m U-238 decay chain U-235 decay chain Th-232 decay chain
378
BIOLOGICAL EFFECTS
Teratogenic effects - are malformations and other growth and structural changes that result from irradiation of the embryo and fetus.
Genetic effects - are hereditary effects observed in the progeny of persons whose germ cells were irradiated and affected.
Stochastic effects – health effects that occur randomly and for which the probability of the effect occurring, rather than its severity, is assumed to be a function of dose without threshold. Examples: hereditary effects and cancer incidence.
379
Non-stochastic effects – health effects, the severity of which varies with the dose and for which a threshold is believed to exist. Example: radiation induced cataracts.
Somatic – effects in the exposed individual.
380
Typical NRRPT Examination Questions
1. In general, the body cells most susceptible to damageby radiation are those found in:
A. rigid or semi rigid tissues B. muscle tissues C. rapidly dividing tissues D. highly specialized tissues E. nerve tissues
381
2. In a picocurie of any radioactive substance, thedisintegration rate is:
A. 2.22 dpm B. 2.22 x 106 dpm C. 37,000,000 dpm D. 3.7 x 104 dps E. 3.7 x 1010 dps
382
3. Which of the following radionuclides cannot bedetected by gamma spectrometry pulse heightanalysis?
A. Hydrogen-3 B. Iodine-131 C. Cerium-144 D. Ruthenium-106 E. Cesium-137
383
4. The elemental symbols for Boron, Beryllium,Cadmium, and Calcium are;
A. Bo, B, Ca, C B. B, By, Cd, Ca C. Bo, Be, Cd, Ca D. B, Be, Cd, Ca E. B, Br, Ca, Cl
384
5. Which of the following radionuclides is most suited toin-vivo measurements?
A. Hydrogen-3 B. Carbon-14 C. Strontium-90 D. Iodine-131 E. Plutonium-239
385
6. How long must a sample with a count rate of 300 cpmbe counted to give a total count rate standarddeviation of 1%?
A. 3.5 min B. 17 min C. 30 min D. 33 min E. 65 min
386
7. At what radius would you post a radiation area aroundan 8 curie Cesium 137 (662 Kev photon energy and aphoton yield of 0.85 photons/disintegration) pointsource?
A. 10 feet B. 74 feet C. 145 feet D. 53 feet E. 101 feet
387
8. An air filter with a collection efficiency of 99.97% isbeing used in a decontamination effort. Calculate thedecontamination factor for this filter.
A. 9997 B. 0.9997 C. 3000 D. 10,000 E. 3333
388
9. During an emergency in a DOE regulated facility, withknown or potential high radiation fields, exposure topersonnel must be voluntary if it is anticipated thatsuch exposure may exceed a whole body exposureof:
A. 5 rem B. 10 rem C. 25 rem D. 75 rem E. 100 rem
389
10. A worker is to perform maintenance on a ReactorCoolant pump under the following radiological conditions; Dose rate on contact with the pump - 350 mrem/hr, Dose rate at 30 cm from the pump (working area dose rate) is 85 mrem/hr, and an airborne concentration of 0.45 DAC. She will spend a maximum of 14 hours in this area during the week. According to 10CFR20, how is this area to be posted?
A. Danger High Radiation Area, Airborne Radioactivity Area B. Caution Radiation Area, Airborne Radioactivity Area C. Caution High Radiation Area, Airborne Radioactivity Area D. Caution Airborne Radioactivity Area E. Caution Radiation Area
390
11. For an exclusive use vehicle that is transportingradioactive materials, radiation levels on contact with
any external surface of the vehicle must not exceed: A. 0.01 mSv/hour B. 0.02 mSv/hour C. 0.1 mSv/hour D. 2.0 mSv/hour E. 10.00 mSv/hour
391
12. Two categories of ionization are:
A. alpha and beta B. direct and indirect C. microwave and infrared D. charged and uncharged E. molecular and atomic
392
13. Intrinsic efficiency of a detector expresses the:
A. probability that a count will be recorded if radiation enters the sensitive volume. B. ability of an instrument to count different energies. C. percent of gamma energy producing ion pairs. D. total detector counts minus the background. E. total beta/gamma counts by a tissue equivalent
detector
393
14. The antiparticle of a positron is a:
A. proton B. neutrino C. electron D. meson E. neutron
394
15. Forms of the same chemical element that contain different numbers of neutrons are called: A. isobars B. isomers C. radionuclides D. isotones E. isotopes
395
16. An atom of a radionuclide that has a low neutron toproton ratio, and an atomic rest mass energy that is 1.02 Mev greater than the product atom's rest mass energy may decay by which of the following?
A. Either positron emission or electron capture B. Annihilation C. Beta minus emission D. Isomeric transition E. Internal conversion
396
17. Which radioactive decay series includes Ra-226 as one of its decay products? A. Thorium B. Uranium C. Actinium D. Neptunium E. Polonium
397
18. An individual who receives an acute, whole body(DDE) radiation exposure of approximately 8 Gy will likely suffer symptoms of up to which level of the Acute Radiation Syndrome?
A. Subclinical B. Hemopoietic C. Gastrointestinal D. Central Nervous System E. Not enough exposure to classify
398
19. The term "isokinetic sampling" refers to the procedure of using sampling velocity that is exactly equal to the: A. velocity of the gas stream at the point of sampling B. velocity at the center of the main gas stream corrected for temperature and pressure C. velocity at the center of the main gas stream D. velocity of the gas stream adjacent to the duct wall E. average velocity of the main gas stream
399
20. In which of the following radioactive decays will the daughter product be an isobar of the parent? A. alpha decay B. gamma decay C. neutron decay (elastic scatter) D. positron decay E. neutron decay (inelastic scatter)
400
21. The respiratory protection device of choice for entryinto an atmosphere immediately dangerous to life and
health is a (an): A. supplied air hood B. air-purifying respirator equipped with a high efficiency
filter C. air-purifying respirator, full face piece, equipped with organic vapor canister D. self-contained breathing apparatus equipped with a pressure demand regulator E. self-contained breathing apparatus equipped with a
demand type regulator
401
22. The average distance of travel in a medium between interactions, describes a photon's: A. mass energy absorption coefficient B. mean free path C. linear attenuation coefficient D. Compton cross section E. linear energy transfer
402
23. The Bragg-Gray principle is based upon the relationship of: A. secondary charged particle equilibrium requirements and the thickness of the wall material of the chamber. B. ionization in an air-filled ionization chamber to the dose in air C. ionization of the gas in an ionization chamber to the dose in the wall material D. ionization in a gas-filled ionization chamber to the dose in the gas E. scatter of low energy photons to the probability of
ionization in the chamber
403
24. Given a gamma-energy value of 0.662 Mev, and aphoton yield of 0.85 per decay, the exposure rate at 2 yards from an unshielded 10 mCi Cs-137 point source is:
A. 1.10 R/hour B. 0.55 R/hour C. 5.50 R/hour D. 0.55 mR/hour E. 0.94 mR/hour
404
25. A radionuclide has a decay constant of 0.1314 years,a gamma energy (per disintegration) of 2.50 Mev, and will produce a dose rate of approximately 30 R/hr at one foot from a 2 Curie source. Calculate the radiological half life of this nuclide:
A. 5.27 years B. 229 years C. 3.93 years D. 30.1 years E. 0.0231 years
405
ANSWERS TO EXAMPLE QUESTIONS
1 – C 11 – D 21 – D 2 – A 12 – B 22 – B 3 – A 13 – A 23 – C 4 – D 14 – C 24 – E 5 – D 15 – E 25 – A 6 – D 16 – A 7 – B 17 – B 8 – E 18 – C 9 – A 19 – A 10 – E 20 – D
406