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.

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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).

115

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.

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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.

180

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

PEP 2C QUESTIONS ?

Paul Steinmeyerprstein@radpro.com

Tom Voss

jtvoss@newmexico.com

407