:
;

[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 __[ Effective Date: 10/01/2007 | Page 40 of 74

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 6 of 36

A.2.1 STANDARD DEVIATION OF BACKGROUND COUNTS

The data field BKG_COUNTS contained recorded background counts for a sample and was the
preferred source of background count information. When a record’s BKG_COUNTS field was blank,
the background count information was calculated as the product of the T_.BKG_RATE and BKG_MIN
fields. if the BKG_MIN field was blank, but the T_.BKG_RATE field contained a value, the background
count information was calculated as the product of the T_.BKG_RATE and COUNTMIN fields.

The variability in repeat counts, such as background counts, follows the Poisson distribution (Evans
1955, Chapter 26). It is a property of the Poisson distribution that its standard deviation is the square
toot of its mean value. Therefore, in this analysis, the standard deviation of the total background
counts was calculated as the square root of the total background counts.

A.2.2 BACKGROUND COUNT TIME

The data field BKG_MIN contained the recorded background count time for a sample and was the
preferred source of background count time information. If the BKG_MIN field was blank, the time in
the COUNTMIN field was used. This last method of estimating the background count time is likely to
be favorable to claimants because in almost all cases where there is paired data, the COUNTMIN

value is less than the BKG_MIN value. This produces a higher MDA when substituted into Equation
1.

A.2.3 INSTRUMENT EFFICIENCY

Information on instrument efficiencies was contained in two fields. The EFFICIENCY field contained
the percent of disintegrations detected by the instrument. The EFF_FACTOR field contained the
reciprocal of the EFFICIENCY field.

A.2.4 YIELD

The chemical yield for a given radiochemical analysis was contained in the RECOVERY data field.
This value was given in percent recovery.

Many of the records had blank recovery fields. However, due to the repetitive nature of the chemical
extractions used, application of an average, radionuclide-specific, chemical recovery to records with
no recovery value was judged to be acceptable. This approach is supported further by the lack of
variability observed in daily runs on known standards recorded in laboratory logbooks recovered from
the late-1950s through the mid-1970s.

A.2.5 VOLUME OF ALIQUOT ANALYZED
The ALIQUOTVOL field usually contained a value for the amount of material subjected to

tadioanalysis. When this field was blank, it was assumed that the entire sample provided was used in
the analysis. The amount of the sample was recorded in the SAMPLEVOL field.



36 AECU-817

When a disability due to dust actually occurs, there are generally three important factors in-
volved: (1) the concentration of materia! to which the individual has been exposed, {2) the particle
size distribution of the material, and (3) the length of time the individual has been exposed tc the
disabling material.

Medical autopsies have indicated that particles larger than 10 are rarely if ever found in the
lungs. A perusal of the latest literature discloses that the ‘‘critical’’ particle size range is thought
to be from about 0.4 to 2.5 4. By ‘‘critical size” is meant that size range which is most likely to
penetrate deep into the lungs and be retained. .

‘The lymphatic system of drainage in the lungs has been mentioned. Augmenting this are phago-
cytic or scavenger cells that engulf particles which do penetrate deeply into the lungs and remove
them to locations where they may be eliminated by absorption and excretion.

It is due to the overloading of these protective mechanisms by particles in the critical size
range in large concentrations over a long period of time that disabilities from dust occur.

Iam speaking now primarily of fibrotic conditions of the lungs or conditions that affect the lungs
themselves. Iam not referring to toxic materials which enter the body through the lungs and exert
their influence on other organs of the body.

Thave only succeeded in hitting the high spots, and in some cases my statements have been very
broad in discussing aspects of nonactive particulates. I wish to bring them out, however, since it can
be reasonably assumed that in so far as radioactive materials are concerned, their physical proper-
ties will follow some of the physical properties of nonactive materials. These should disperse and
spread in the form of dust, fumes, smokes, and mists in a similar manner to nonactive particulates
and, in general, follow the laws applicable to fine particulates. In other words, conditions would be
expected to be similar, except that we have added radioactivity to our particulates. It is, admittedly,
avery difficult subject to discuss, since there are so many conflicting opinions. The lack of pertinent
biological data has been such that it is impossible to properly evaluate the actual phystologtcal haz-
ard caused by the inhalation of insoluble radioactive particles,

3.6 PERMISSIBLE WASTE-DISPOSAL CONCENTRATIONS
By Forrest Western

Introduction

In Lecture 3,4 a method was given for estimating the maximum permissible quantity of a beta-
gamma~emitting radioisotope which, in the absence of other significant radiation exposure to the
critical tissue, could be taken into the body, Changing the meaning very slightly of two of the sym-
bols used to apply to any case we may wish to consider,

let QV = permissible daily quantity of a radioisotope taken into the body
W = = weight of tissue, measured in grams
F_ = fraction of tsotope which becomes fixed in the critical tissue
T = effective half-life of radioisotope in the critical tissue
E = average energy of radiation per disintegration
an equation derived in Lecture 3.4 becomes

QV = 1.15 x 10°9 We curie/day

In the quantity QV, V will generally represent a volume or mass of material per day, in which case
Q will be the number of curtes of the radioisotope per unit of volume or of mass.

This equation or equivalent equations have been frequently used to compute the maximum per-
missible concentrations of beta-gamma-—emitting radioisotopes in drinking water.

While it appears unlikely that a person might spend enough time in or very near water to be
concerned over external radiation, for the protection of fish, fish eggs, etc., it may be considered
desirable to limit the external radiation dosage rate for bodies submerged in the water to 0.1 rep/24-



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 | Effective Date: 10/01/2007 [ Page 69 of 74 _]

ATTACHMENT A

CALCULATION OF MDAS FROM HISTORICAL DATA

Page 35 of 36

Number of Sample Results Recovered per Year
Between 1945 and 1990 for Pu-239 in Feces

9S pen

Number of Samples (N)

Figure A-17. MDA data for Pu-239 in feces (part 3 of 3).

1E+I aan =
: ‘Sample MDA 7

‘Est. MDA of 0.5 (95%CL on Mean)
= @= 50th Percentile of 0.3

Z1E0

Hy

2

i

=

2

3

2 IE]

1E2

1945 1950 1955 1960 1965 1970 1975 1980
Year

Visual Comparison of Data to Estimated Th-232 MDAs in Feces

1985

1990

Figure A-18. MDA data for Th-232 in feces (part 1 of 3).



00 AECU-817

Otherwise the progress of the job may result in lessening the working time, such as In the case
of an excavation where shielding dirt removal from hot pipe lines produces an increasing radiation
field (as well as introducing a serious contamination problem when such pipe lines have leaked) or
in the case of equipment disassembly where interior parts exposed may be much higher in activity.

Under these conditions it is readily apparent that new working times proper to the stage of work
are needed as the job progresses. In such types of jobs, the health physics representative should
clearly establish the importance of his being currently informed of significant changes produced by
the progress of the work so the job can be resurveyed. All work where such changes can be foreseen
as possibilities should be watched with extra care by the Health Physics Division.

5. Failure to work within designated boundary lines is occasionally encountered, and obviously
if one surveyor surveys location A and the men perform their work at location B or at A and B, then
a survey of location B might very properly be indicative of quite different working times. This
situation results most usually from an imperfect description of the area of action being furmshed
to the Health Physics Division.

If the health physics representative observes that a maintenance job is being carried out in an
area or location different from that initially surveyed, he should certainly point out the need for a
further check to the foreman of the job. It would not be Improper under such circumstances to re-
quest that further work be halted, if possible, until the proper area affected can be surveyed to as-
certain the correct working time.

6. Failure to follow recommendations made by the health physics representative is potentially
serious in every case, and, if this is due to deliberate disregard of necessary precautions, it should
be at once reported to appropriate supervision for whatever corrective or disciplinary action
necessary.

Fortunately such disregard is most often the result of carelessness, forgetfulness, or thought-
lessness. This may show up in numerous ways, from the failure to wear proper protective clothing
or observe given working-time limits to neglect of interposing necessary auxiliary shielding, or
even complete failure on the part of operating supervision to notify maintenance crews of the pro-
visions of the safety work permit.

Cognizance of these six points and the exercise of careful judgment, coupled with maximum
reasonable accuracy in the use of the most appropriate instruments, will assist In the achievement
of a professional standing and record of unexcelled caliber,



UNIT 1. INTRODUCTION*

1,1 A BLUEPRINT FOR THE DEVELOPMENT OF A HEALTH PHYSICIST
(Place of Health Physics in Industry)
By K. Z. Morgant

Dr. Morgan started his lecture by expressing a desire to meet and know the students better. He
assured them that they should feel free to call upon him for any advice or problem they mignt have.

The first point was to give the students a knowledge of what our problem is and how it has been
developed. Dr. Morgan then outlined a brief history of radiation.

X rays were discovered in 1895 by Roentgen, and in 1896 Becquerel discovered the activity of
uranium, That very year Grubble, a manufacturer of X-ray tubes in Chicago, reported an erythema
on his hands. This was the first recorded case of radiation damage. Following that recognition of
radiation injury of the man’s hands, there have been many cases of radiation damage recorded. Vari-
ous types of radiation damage were discovered later, but most of the early damage was observed in
medical doctors, radiologists, dentists, and radium technicians, and there was damage in varying de-
grees. The damage that was first observed was the erythema, but it was soon recognized that the
blood frequently sustained a more serious damage. Unfortunately radiation damage occurs in a rather
insidious manner. If you walk up too close to a hot plate, you recoil with pain. On the other hand,
there is no sixth sense to indicate that we are receiving a radiation exposure which will account for
damage or even death, and such an exposure may be received without our knowing it. The first symp-
tom of a serious overexposure is drowsiness the next day, perhaps followed by illness 3 or 4 days
later. Doses of as much as 100,000 r (roentgens) would kill a person in a matter of minutes, The last
stages of a case following a fatal radiation exposure (of, for example, 600 r) would result in much suf-
fering. The medical profession ts only able to ease the pain of such an unfortunate individual,

The object of Health Physics is not treatment after exposure, but to make certain that no one re-
ceives exposures of consequence.

Other effects of radiation include the following: The body may be made more susceptible to some
diseases, It has been found to be rather definite that if a person is exposed to radiation, it reduces
his expected life. For example, if a person receives our currently proposed tolerance level (which is
0.3 r/week) for 35 years, it has been calculated that he would lose about 1.5 years of his expected life.
We are not sure of this exact figure, but we are sure that radiation does reduce the expected life span
on the average. Radiation causes cancer, and ordinarily cancer does not show up until 10 or 15 years
after the exposure, It would be necessary to see these persons 10 or 15 years after certain radiation
exposures in order to see that any damage had been done. Genetic effects may not show up for several
generations. Thus we see that in most cases it is not possible to observe radiation effects immediately.

‘This was the sort of picture of the past that we had when the Plutonium Projects began in 1942.
The medical men, in looking over the records, found that man, in working with 2 Ib of radium, had
produced a great deal of damage; some died from it, and others suffered because of the radiation, A.
Compton, Fermi, Stone, and others actually considered whether it was advisable to go ahead with this

*This series of lectures was sponsored by the Radiation Survey -Monitoring Section of the Health
Physics Division, Oak Ridge National Laboratory and conducted under the direction of T. H. J. Burnett,
Technical Adviser,

tNotes taken by C. Scott; edited by K. Z. Morgan.

2 AECU-817



[Document No. ORAUT-OTIB-0034 | Revision No. 00 __| Effective Date: 12/13/2005 | Page 20 of29 |

* annual bioassay
vatue (50th percentile)
* — not used in fit
fit

Figure A-25. Assumed “‘Am intake, 1985 to 1988, 50th-percentile results,
Type M.

* — annual bioassay

value (50th percentile)
© not used in fit
fit

Figure A-26. Predicted “*'Am intake, 1968 to 1988, 50th-percentile composite
results, Type M.

Table A-12. Type M“*Am intake periods and rates.

“Am intake rate (dpmiday)

Start date | Stop date | 50th percentile | 84th percentile | GSD
1968 1984 6.673 14.61 2.19
1985 1988 2.207 6.659 2.99



[Document No. ORAUT-OTIB-0034 | Revision No. 00 _| Effective Date: 12/13/2005 | Page 24 of 29 |

* annual bioassay
value (50th percentile)
© not used in fit
fit

Figure A-16. Assumed “"U intake, 1983 to 1987, 50th-percentile results, Type S.

* annual bioassay
value (50th percentile)
*  notused in fit
fit

Figure A-17. Predicted uranium excretion rate from independently fit intakes,
1951 to 1987, 50th-percentile, Type S.



‘ Special Exposure Cohort Petition U.S. Department of Health and Human Services

under the Energy Employees Occupational Centers for Disease Control and Prevention
lilness Compensation Act National Institute for Occupational Safety and Health
OMB Number: 0920-0639 Expires: 09/20/2013

Special Exposure Cohort Petition — Form B Page 1 of 7

General Instructions on Completing this Form (complete instructions are available in a separate packet):

Except for signatures, please PRINT all information clearly and neatly on the form.

Please read each of Parts A — G in this form and complete the parts appropriate to you. If there is more
than one petitioner, then each petitioner should complete those sections of parts A — C of the form that apply
to them. Additional copies of the first two pages of this form are provided at the end of the form for this pur-
pose. A maximum of three petitioners is allowed.

If you need more space to provide additional information, use the continuation page provided at the end of
the form and attach the completed continuation page(s) to Form B.

If you have questions about the use of this form, please call the following NIOSH toll-free phone number and
request to speak to someone in the Division of Compensation Analysis and Support about an SEC petition:
1-877-222-8570.

| | A Labor Organization, Start atD on Page 3

Ifyou |_an Energy Employee (current or former), StartatC on Page2
are:

BS Survivor (of a former Energy Employee), StartatB on Page 2

iA Representative (of a current or former Energy Employee), StartatA on Page1

Representative Information — Complete Section A if you are authorized by an Employee or
Survivor(s) to petition on behalf of a class

Are you a contact person for an organization?

[Jno (Go to A.3)

A.2 Organization Information:

Name of Organization

Position of Contact Person
A3 Name of Petition Representative:

Mr/Mrs./Ms. First Name Middle Initial Last Name

A4 Address:
Street Ant # P.O. Box
City State Zip Code

A5 Telephone Number: ( ” 2

A6 Email Address:

AT Check the box at left to indicate you have attached to the back of this form written authorization to
petition by the survivor(s) or employee(s) indicated in Parts B or C of this form. An authorization

Name or Social Security Number of First Petitioner:



96 AECU-817

The wearing of protective clothing will not control the spread of contamination. However, the
removal of protective clothing is an appropriate method for personnel to decontaminate where con-
tamination control is to be effected.

Contamination zones of three types may be established to control the wearing of discardable
clothing (protective clothing).

1. Cold’? zone in which contamination above background (natural) is impossible.

2, “Low-risk’’ zone in which immaculate conditions will normally be attained, but within which
exists a possibility of accidental contamination, Such contamination is to be cleaned up immediately
upon its occurrence. :

3. “‘High-risk’’ zone into which entry without acquiring contamination is unlikely.

High-risk zones are exclusion areas to which casual access of personnel is prohibited. Pro-
tective covering shall be worn in high-risk zones which will prevent contact of the body with the
contaminating materials to be encountered. All clothing worn in a high-risk zone shall be removed
or passed by appropriate frisking with radiation-detection instruments before the wearer may leave
the boundary of that zone.

The wearing of protective clothing in a low-risk zone is recommended. Contaminated garments
and other contaminated items are subject to confiscation, Supervision is to maintain contamination
levels below that which could be a health hazard for a continuous exposure of 8 hr per day throughout
all low-risk zones.

Protective clothing worn in either high- or low-risk zones shall be excluded from any cold zones.
Measures should be instituted to ensure that no contamination which might become a legal liability
will be permitted to pass into a cold zone.

Cold zones may include counting rooms, instrument shops, urinalysis laboratories, off-plant
areas, cafeteria, etc.

Details of zone boundaries are to be established for local areas by the supervision responsible
for directing work in these areas. This supervision wil! also be responsible for adequately labeling
the boundaries. A high-risk zone may suddenly develop within a low-risk area. A small high-risk
zone may be established about the immediate vicinity of certain operations, Although too small to
admit a body, any portion of a person entering a high-risk zone shall be governed by high-risk zone
procedures,

Supervision of areas in which high-risk zones occur should provide protective clothing, change
room, and contamination checking facilities to accommodate all workers on jobs for which they are
responsible.

It is evident that a decontamination-type change room should be located at a zone boundary.
Whether compulsory showers or merely slip-over garment procedures are required depends upon
the nature of work to which personnel will be assigned. Arrangement should be such that correct
procedure in the change facilities is easy to follow and difficult to avoid §Crosstracking over areas
where work shoes are worn by bare feet and shoes for the clean side should be avoided. If low-risk
clothing will be worn more than once, lockers separate from those for street clothing should be
provided for each person.

Contamination control is needed to preclude genera! air-borne radioactive contamination, as
well as to prevent the accumulation of hazardous sources of penetrating radiation. The limitation
based upon possible intake of radioactive material into the body is not always the most stringent,
The need to prevent nuisance contamination which may foul instruments and thwart experiments
provides an additional motive that favors the radiation safety program.

Devise Protection from Penetrating Radiation

When radioactive materials other than alpha emitters are to be cared for, protection from the
radiation must be arranged in addition to the control of contamination.

Alpha radiation is excepted because its low penetrability makes any practical contamination-
control barrier enough to shield live tissues completely. Energetic alpha rays will only penetrate a
few centimeters of air and are likewise stopped completely by one or two sheets of bond paper.



AECU-817

Auswer: Response time {s not a consideration when using an alpha proportional survey counter
a8 it is when using an ionization meter, but if one is looking for small amounts of activity, it is
necessary to scan the surface slowly.

45



(Document No, ORAUT-TKBS-0012-5 | Revision No. 01__| Effective Date: 10/01/2007 | Page 45of74_]

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 11 of 36

Number of Sample Results Recovered per Year
Between 1945 and 1990 for Am-241 in Urine

700

600 +

500 +

‘Number of Samples (N)

200

1963

Figure A-1. MDA data for americium-241 in urine (part 3 of 3).

TEH
> Sampte MDA
TET Est MDA 0f02 (95%CL on Met)
= 50th Percentile of 0.)
VE {4 . _
LE +

Activity (dpm/24 h sample)

1E-3
1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Visual Comparison of Data to Estimated Cm-244 MDAs in Urine

Figure A-2. MDA data for curium-244 in urine (part 1 of 3) (some later tables
are “in feces”).



[“Docoment No. ORAUT-OTIB-0034 | Revision No. 00 _| Effective Date: 12/1 3/2005 | Page 27 of29 |

Figure A-22. Predicted TM*Pu intake, 1951 to 1988, 50th-percentile composite
results, Type M.

Table A-10. Type M Pu intake periods and rates.

TM"Bu intake rate (dpmiday) "Bu intake rate (dpmiday)

Year | 50th percentile | 84th percentile GSD Year | 50th percentile | 84th percentile GSD
1951 40.75 109.8 2.69 1971 7.35 22.45 3.05
1952 40.75 409.8 2.69 1972 7.35 22.45 3.05
1953 10.98 86.42 787 1973 7.35 18.4 2.50
1954 10.98 86.42 7.87 1974 7.35 18.4 2.50
1955 10.98 86.42 7.87 1975 7.35 18.4 2.50
1956 10.98 86.42 7.87 1976 7.35 18.4 2.50
1957 10.98 86.42 787 1977 7.35 18.4 2.50
1958 10.98 86.42 7.87 1978 7.35, 18.4 2.50
1959 10.98 86.42 7.87 1979 7.35 18.4 2.50
1960 10.98 22.45 2.04 1980. 7.35 18.4 2.50.
1961 10.98 22.45 2.04 1981 7.35 18.4 2.50
1962 10.98 22.45 2.04 1982 7.35 18.4 2.50
1963 10.98 22.45 2.04 1983 7.35 18.4 2.50
1964 10.98 22.45 2.04 1984 7.35 18.4 2.50
1965 10.98 22.45 2.04 1985. 4.614 6.811 4.22
1966 10.98 22.45 2.04 1986 1.614 6.811 4.22
1967, 10.98 22.45 2.04 4987 1.614 6.811 4.22
1968 10.98 22.45 2.04 1988 1.614 6.811 4.22
1969 7.35, 22.45 3.05

1970 7.35 22.45 3.05



56 AECU-817

instrument Hmitations, but be diplomatic, Do a selling job of your services by the correct usage and
interpretation of your instrument,

$.5 SURVEY INSTRUMENT USAGE AND SELECTION (i)
By G. R, Patterson

1am supposed to have given you a list of any references 1 use in this lecture, along with the
abstract which you got belatedly. I shall give you those references now in case you might want to
check my mistakes and really learn something about the instruments.

The most important general reference is the Instrument Manual compiled by D. M. Davis. I
was lucky in that I received a copy of it, by coincidence, the same day I was notified that I was to
speak to you.

‘The other references are: CP~3316, a paper on the Poppy, a pulse-type detector for alpha and
other radiations; CP-3317, a paper on the Walkie Poppy, which is a portable model of the same type
of detector; and MonH-448, a paper dealing with the conversion of the Poppy for detecting the soft
beta radiation such as from C34,

‘The instruments we are considering today are the Lauritsen electroscope, the Poppy Probe as
used for alpha and C4 beta detection, and the Fish pole probe.

In the beginning let me say that, though I was exposed to physics, including some radio theory,
in college, I know practically nothing about electronics and instruments. I cannot tell you why they
work or how they work. Through my experience here at the laboratory Ican tell you what they are
intended to do, whether they work, and how they are used.

‘The Lauritsen electroscope is our most accurate portable instrument for measuring hard gamma
radiation. In a purely gamma field it is as high as 99 per cent accurate, It is not calibrated to inter-
pret beta ionization in rep or mrep, but it does respond to this ionization. Therefore, in field usage
where radiation is usually a composite of gamma and beta radiation with a variety of energies, the
electroscope is not quite 50 accurate. It is more accurate, however, than our other portable instru-
ments, and it is much less subject to fluctuations in accuracy.

Its range of calibration is from 1 to 1000 mr/hr hard gamma. It will indicate intensities of less
than 1 mr/hr, but it will not tell how much less it is.

It ts a simple instrument, and it is easy to use. One simply charges the chamber with a push-
button switch and times the fiber drift with a stop watch, The instrument requires no warm-up
period.

Its main survey use is checking isotope shipments for conformance with maximum permissible
radiation levels for shipping. A model of the Lauritsen electroscope using a boron-coated chamber
is used in conjunction with a standard Lauritsen for computing thermal neutron flux.

Tt does have limitations. Since the operator must look directly into the microscope eyepiece at
very close range, it cannot be used in small places where, for instance, the chamber of a Cutie Pie
or a Fish pole probe could be thrust at arm’s length.

The time-lag feature when making a reading makes it unsatisfactory for routine building or lab
surveys, Where perhaps many readings may be taken in a short time, or in scanning an area for hot
spots or general background, This same time-lag feature, together with the fact that the surveyor
is of necessity in the same radiation as the instrument, makes it undesirable for use in high radia-
tion fields. Even though the fiber drift time is very short at the high limit of calibration and the
maximum limit of the instrument is only 1000 mr/hr, it is much more satisfactory and desirable
to use an instrument which can be read directly or instantaneously from a distance and which can
be extended into the radiation.

The fact that the fiber drift time as the upper limit of calibration is approached is very short
makes accurate timing with the stop watch very difficult at these higher levels,

‘The main cause of inoperation with this instrument is burned-out or very weak batteries. Other
technical faults show up, also, such as dirty insulators or damaged insulators or fibers or flaking
off of the conductive coating of the fiber, etc., but these are far less frequent.



Special Exposure Cohort Petition U.S. Department of Health and Human Services
under the Energy Employees Occupational Centers for Disease Control and Prevention

Illness Compensation Act National Institute for Occupational Safety and Health
OMB Number: 0920-0639 Expires: 09/20/2013
Petitioner Authorization Form Page 1 of 2

Use of this form and disclosure of Social Security Number are voluntary. Failure to use this form or disclose
this number will not result in the denial of any right, benefit, or privilege to which you may be entitled.

Instructions:

If you wish to petition HHS to consider adding a class of employees to the Special Exposure Cohort and you
are NOT either a member of that class, a survivor of a member of that class, or a labor organization
representing or having represented members of that class, then 42 CFR Part 83, Section 83.7(c) requires
that you obtain written authorization. You can obtain such authorization from either an employee who is a
member of the class or a survivor of such an employee. You may use this form to obtain such authorization
and submit the completed form to NIOSH with the related petition. Please print legibly.

For Further Information: If you have questions about these instructions, please call the following NIOSH

toll-free phone number and request to speak to someone in the Division of Compensation Analysis and
Support about an SEC petition: 1-877-222-8570.

Authorization for Individual or Entity to Petition HHS on Behalf of a Class of Employees for

Addition to the Special Exposure Cohort

Name of Class Member or Survivor

Street Address of Class Member or Survivor Apt. # P.O. Box

City, State, Zip Code of Class Member or Survivor

do harahy anthariza:

Name of Petitioner
Address of Petitioner Apt. # P.O. Box

City, State and Zip Code of Petitioner

to petition the Department of Health and Human Services on behalf of a class of employees
that includes:

Name of Class Member (employee, not the employee's survivor)
for the addition of the class to the Special Exposure Cohort, under the Energy Employee’s
Occupational Illness Compensation Program Act (42 U.S.C. §§ 7384-7385).

In providing this authorization, | recognize that the petitioner named above will have all the rights
of a petitioner as provided for under 42 CFR Part 83.
Z- 8-201)

Signature of Class Member or Survivor Date

Name or Social Security Number of First Petitioner:



[Document No. ORAUT-TKBS-0012-5 [Revision No. 01 __| Effective Date: 10/01/2007 | Page 4tof74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 7 of 36

A.2.6 VOLUME OF 24-HOUR SAMPLE

Excretion rates and concentrations vary greatly within a 24-hr period and between individuals. The
models developed for calculating intakes from in vitro analyses are based on 24-hr samples. The
daily urine excretion rate is 1,400 mL/24-hr sample; for feces, this value is 135 g/24-hr sample.

Often, the volume analyzed (V,) was less than the reference excretion rate from standard man. When
this was observed, the calculated activity in the sample was normalized to the reference man volume
to compensate. This adjustment increased the MDA and, therefore, would be favorable to claimants.

A3 Caiculation Method for MDAs for Various Periods Between 1945 and 1988

Once the MDAs were calculated for individual samples, they were segregated by radionuclide and
imported into Excel spreadsheets. Using Excel, the values were plotted for visual inspection. Figures
A-1 through A-18 at the end of this attachment show plots of data the was obtained and calculated for
the following radioisotopes: (urine) “*Am, “Cu, "Cs, gross alpha, gross beta, °H, “Pm, Pu,
pu, "Pu, rare earths, *Sr, “°U, ”°°U, and (feces) ““"Am, gross alpha, 2°Pu, “Th. In the first plot,
the MDA values were plotted against the date of their analysis. This provided a visual representation
of individual MDAs over time. The density of the available data and a gross approximation of typical
sample MDAs for a given period were quickly observed using this format.

Two bar charts were created for each radionuclide. One shows the annual average MDAs between
1945 and 1988. The other shows the number of analytical results recovered for each year in the
same period. These two charts provide additional information on gross trends exhibited by sample
MDAs, and on the temporal distribution of the data.

These visual aids were used to establish general periods during which sample MDAs were similar.
The sample MDAs in these periods were grouped as similar populations. Using Excel's statistical
functions, the following statistical parameters for each group of samples were determined:

The population size (N)

The 50" percentile value of the population,

The population's arithmetic mean

The arithmetic standard deviation of the population
The 95% confidence level on the mean

The 95% confidence levels on the mean for each radionuclide are the recommended MDAs for that
radionuclide during the period evaluated.

A4 Additional Considerations

The recovered data indicate the analytical MDAs tended to remain fairly consistent for years. Abrupt
changes in the MDAs were identified for groups of radionuclides during specific years. Following the
change, the MDAs remained generally consistent in the several succeeding years. This “step-wise”
pattern allowed MDAs from several years to be grouped as one MDA for a specified period. Table 5-9
highlights these groupings by enclosing similar MDAs in a box.



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 _| Effective Date: 10/01/2007 | Page 8of74__|

5.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the
National Institute for Occupational Safety and Health (NIOSH) but are rather general working
documents that provide historic background information and guidance to assist in the preparation of
dose reconstructions for particular sites or categories of sites. They will be revised in the event
additional relevant information is obtained about the affected site(s). These documents may be used
to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of
buildings that served a specific purpose at a site. !t does not necessarily connote an “atomic weapons
employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees
Occupational Iliness Compensation Program Act [EEOICPA; 42 U.S.C. § 73841(5) and (12)].
EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon
which such building, structure, or premise is located ... in which operations are, or have been,
conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises,
grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” (42 U.S.C.

§ 73841(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted
above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE
facility encompassed by EEQICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines
eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for
members of the Special Exposure Cohort). The compensation decision for cancer claimants is based
on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C.
§ 7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in
the performance of duty for purposes of the compensation program if, and only if, the cancer ... was
at least as likely as not related to employment at the facility [where the employee worked], as
determined in accordance with the POC [probability of causation‘] guidelines established under
subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation
guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees
with a covered cancer or restrict the “duty” to nuclear weapons work.

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures,
premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982
(42 U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 73841(12)].
While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the
section of EEOICPA that deals with the compensation decision for covered employees with cancer
[.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an
exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation
exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including
tadiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and
external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts
are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose
reconstruction. NIOSH, however, does not consider the following exposures to be occupationally
derived:

« Radiation from naturally occurring radon present in conventional structures
« Radiation from diagnostic X-rays received in the treatment of work-related injuries

' The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC.



LDocument No. ORAUT-OTIB-0034 | Revision No. 00 _| Effective Date: 12/13/2005 | Page3of29 |

TABLE OF CONTENTS

SECTION TITLE PAGE
1.0
2.0
3.0

3.1 Selected Bioassay Data

3.2 Analysis
4.0 Intake Modeling... 7

41 Assump'
4.2 Bioassay Fitting .
4.3 Radionuclides and Material Types
4.3.1 Strontium-90.
4.3.2 Uranium-234.
4.3.3. Plutonium-239..
4.3.4 Americium-241.
4.3.5 Additional Radionuclides

5.0 Assignment of Intakes and Doses
5.1 Contribution from Additional Radionuclides
5.2 Dose Assignment

References



[Document No. ORAUT-OTIB-0034 | Revision No.00__| Effective Date. 12/13/2005 | Page9of29 |

4.3.3

4.3.4

4.3.5

calculated GSDs for Type S ?“U. Plots of the results of the individual fits to the
84th-percentile values for all three solubility types are not shown here since
they were largely in agreement.

Plutonium-239

For this analysis, the analyzed material types for Pu are assumed to be
types M and S. Although the bioassay results are for all alpha-emitting isotopes
of plutonium, the results were assumed to represent the concentration of “Pu
alone,. Figures A-18 through A-21 show the fits for Type M material
corresponding to each of the subsets of years across the intake period used to
develop the composite figure. Figure A-22 shows the composite of the subsets
of selected years used to fit the 50th-percentile intake values for type M 7*°Pu.
Table A-10 summarizes the intake periods and corresponding intake rates for
the 50th- and 84th-percentile values, and the GSDs for type M*°°Pu. The
GSDs were determined by dividing the 84th-percentile intake rates by the 50th-
percentile intake rates. Again, the same intake periods were applied to the
84th-percentile values for Type M *°Pu because the values followed a similar
pattern; results of the individual fits are not shown here since they were largely
in agreement.

Figure A-23 shows the fit for Type S material based on a chronic intake for the
entire set of years fitted to the bioassay data for the last three years (1986
through 1988). Table A-11 summarizes the intake period and corresponding
intake rate for the 50th- and 84th-percentile values, and the GSDs for type S
28°Py. Again, the GSDs were determined by dividing the 84th-percentile intake
rates by the 50th-percentile intake rates. The same intake periods were applied
to the 84th-percentile values for Type S 7*°Pu because the values followed a
similar pattern; results of the fit are not shown here since they were largely in
agreement.

Americium-241

ICRP 68 (1995) assigns all forms of americium to Type M. Figures A-24 and A-
25 show the fits corresponding to each of the subsets of years across the intake
period used to develop the composite figure. Figure A-26 shows the composite
of the subsets of selected years used to fit the 50th-percentile intake values for
Type M**tAm. Table A-12 summarizes the intake periods and corresponding
intake rates for the 50th- and 84th-percentile values, and the GSDs for Type M
41am. Again, the GSDs were determined by dividing the 84th-percentile intake
rates by the 50th-percentile intake rates. The same intake periods were applied
to the 84th-percentile values for Type M “‘Am because the values followed a
similar pattern; results of the individual fits are not shown here since they were
largely in agreement.

Additional Radionuclides

It is recognized that hundreds of different radionuclides were present at ORNL
at some point during its operations. However, bioassay data for additional
tadionuclides beyond *°Sr, 7°°U, °Pu, and *‘Am were deemed to be of little
use for coworker estimations, largely because there were too few
measurements to be statistically reliable for intake estimation. For this analysis,
three additional radionuclides, both important to internal dosimetry and present
in the reported air monitoring data, were considered. These are Ru, '°’Cs,
and “Ce.



[Document No. ORAUT-TKBS-0012-5 |__Revision No. 01 _| Effective Date: 10/01/2007 | Page63of74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 29 of 36

Number of Sample Results Recovered per Year
Between 1945 and 1990 for U-233 in Urine
450 ene

goo |

350

3

Number of Samples (N)

Figure A-13. MDA data for U-233 in urine (part 3 of 3).

1 EH
+ Sample MDA
sl Estimated MDA For the Period
=~ Soin Percentile
EH 4
5 EO
i
&
g 1E-1 |
<
1E2
1E3
1945 1950 1955 1960 1965 1970 1975 1980. 1985 1990

Year

Visual Comparison of Data to Estimated U-238 MDAs in Urine
Figure A-14. MDA data for U-238 in urine (part 1 of 3).



{Document No. ORAUT-OTIB-0018___] Revision No. 07 Effective Date: 08/09/2005 |" Page 350f35 |

ATTACHMENT C
SITE APPLICABILITY

This TIB is applicable to all DOE sites with the limitations listed in the TIB and exceptions listed below.
Table 4 contains three inventory lists of radionuclides. Dose reconstructors should use these lists as
follows:

¢ The alpha list should apply alone to sites or facilities that primarily handled plutonium.
¢ The uranium list should be applied alone to sites or facilities that primarily handled uranium.

* The alpha plus the beta list should be applied to reactor sites, national laboratories, and any
other sites.

Some large sites contained facilities that performed different functions. For example, the Hanford Site
is a reactor site but also had fuel fabrication facilities. The fabrication facilities normally controlled air
concentrations to uranium limits. Therefore, the uranium fist should be applied to individuals working
in those facilities.

Some sites operated with particular nuclides in campaigns. Once the campaign ended, the nuclide
was no longer routinely handled. For example, the Fernald site produced thorium products during
several campaigns but did not routinely handle it between these campaigns. Therefore, it is
permissible to remove thorium from the list of possible radionuclides during times when the individual

was not exposed fo thorium. This can be done in other situations when there is clear evidence that
the individual was not exposed to a particular radionuclide.

Exception

Until 1962, the Nevada Test Site performed atmospheric testing of nuclear weapons. These tests had
the potential to create high airborne concentrations for relatively short periods. While it is possible
that the continuous assumption made in this TIB will overestimate these intermittent exposures,
further evaluation is necessary to verify this. Therefore, this TIB is not currently applicable to
exposures at the Nevada Test Site prior to 1963.



[Document No. ORAUT-OTIB-0018

[ Revision No. 01

Effective Date. 08/09/2005 [ Page 24 of 35_|

Table 7-1 (Continued). Whole-body count (WBC) results below which this
TIB method is still considered an overestimate.

Absorption Time since ‘Organ of WBC resukmust
Radionuclide type last WBC (d)* concem be less than (nCi):

Ce-T44 Mors T80 ‘Adrenals — “86
‘inary bladder oO
Brain 69
Breast 60
Gall bladder 136.
Heart w all 79
Kidneys 80
Tver 5
Muscle 75
‘Ovaries 97
Pancreas 86
Testes 78
Thyrod 82
Red bone marrow 20
Bone surface 22
Stomach 46
SI 20
ULI 6
[an] Ey
in 58
Speen BS
Thymus 78
Uterus 95
Er 6
Lung 4
Colon 3
ophagus: 78
Ce-144 MorS 365 Adrenals 103
Urinary bladder. 108
Brain 84
Breast 72
Gall bladder T60
fart w all 9
Kidneys. oF
Liver 6
Muscie 90
Ovaries 410
increas TOS
Testes oF
yrowd oo
Red bone marrow 24
Bone surface 27
Stomach $3
SI 22
UCT 6
[aay 6
Skin 70
Spleen 100
Thymus 96
Uterus 110
ET 7
Tung a
jon 6
Esophagus 95

15%, The choice of days 1s meant to represent general rheasurement frequencies, Such os quarterly,

semiannually, or annually. {tis recognized that workers did not get whole-body counts at intervals that
were exact to the day


AECU-817 51

percentage of such results continued to deviate considerably from this, we could only conclude that
our procedure of selection is really biased or that our population is, perhaps, not distributed in ac-
cordance with the above law.

Now consider a population distributed by the above law, and suppose N individuals are drawn
and the average taken

Xp 4 Xgere Xy
ees

Statistical theory shows that the values of M obtained when this procedure is repeated indefinitely
again follows a normal distribution law with the same mean as for the parent population, but with a
standard deviation given byo/¥N .

If N is large enough,o/¥N is very small and since

Probability (= -<mem +s) = 0,683

we can accept M (note that M is determined from the sample) as an approximate value for m and we
shall not be in error by more than the small quantity /VN (except in $1.7 per cent of the cases when
we make such an estimate).

if we would like to be correct more than about a third of the time, we can use the fact that

20 20
Pp - 22 \ 209,
robability (m WW <M<m +3) 0.954

and now assert that M approximates m with error less than 20/¥N. We can expect this statement to
be wrong in about 5 per cent of the cases, Hf we use the probable error, we can assert that

_ 0.81450. 0.67450

M<m
Ww WN

Accordingly we should expect that the average M of the sample will deviate from the true mean m
by more than 0.6745¢/ YN as often as not.

The above discussion tells us nothing very precise, since not only the true value of m but also
the value of g would, in practice, be generally unknown. There is, however, good reason to believe
that statistical phenomena due to long-lived sources of radiation follow a somewhat different law,
the Poisson distribution. The Poisson distribution depends only on a single parameter m which is
again the mean. If m is not too small, the Poisson distribution is well approximated by a normal
distribution having the same mean m and with standard deviation o= Ym, This fact enables us to
regard M as, in a sense, giving us the approximate value of the mean and the standard deviation
needed in the earlier statements.

More precisely, suppose a radiation field has an average rate of m particles per unit of time
across the region of the counter. We ignore ail deficiencies of the counter and deal here only with
statistical fluctuations due to the randomness of the radiation field itself. The number of counts of
our ‘‘perfect’’ counter per unit time then constitute a population which is Poisson distributed with
mean m, However, we immediately replace this by a normal distribution with mean m and standard
deviation v= ¥m. If m is not assumed to be too small, the agreement will be good in general. Then
if a count is taken for N units of time (N repetitions of the sampling process) we can assert that for
the average rate M computed by M = total count/N

Probability (m — ve <Mcm+ ve)- 0.683

Probability (= ) 205



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 _| Effective Date: 10/01/2007 |_ Page 67 of 74_|

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 33 of 36

Annual Average MDAs for Gross Alpha in Feces
Calculated from Recovered Data Between 1945 and 1990
12 — —o

MDA (dpm/24 hy
2

2

02+

Figure A-16. MDA data for gross alpha in feces (part 2 of 3).

Number of Sample Results Recovered per Year
Between 1945 and 1990 for Gross Alpha in Feces
140 a

Number of Samples (N)

Figure A-16. MDA data for gross alpha in feces (part 3 of 3).



AECU-817 9
cu —

and negiect the comparatively small mass of the electron, the mass of the atom as a whole becomes
equal to the number of particles in the nucleus. Thus the atomic weight of the normal helium atom

is 4, since the helium nucleus contains 2 protons and 2 neutrons, The number of protons in the nu-
cleus, which is equal to the number of orbital electrons in the atom, determines the so-called ‘“atom-
ic number,’’ which is usually designated by ‘'Z.’? One fact of interest which has been observed is
that quite a range of elements have an atomic weight ‘‘A,’’ which is approximately equal to twice the
atomic number.

Niels Bohr made 2 great contribution to our knowledge of orbital electron structure outside the
nucleus, Bohr’s work was an initial application of Planck’s quantum hypothesis. Planck’s theory was.
that energy cannot be divided continuously until arbitrarily small amounts of energy are obtained,
but, instead, energy comes in small units which cannot be further subdivided.

Following Rutherford’s work it had been assumed that electrons could occupy any point external
to the nucleus. This implies that the electron could have any total energy less than a maximum
characteristic of the nucleus in question. The quantum hypothesis, of course, prohibits this and states
that the electron can have only certain definite energies. It is to be understood that this is a negative
energy, since work would have to be done on the electron to move it away from the attraction of the
nucleus,

Electrons, which are at a given distance from the nucleus, settle into orbits about the nucleus.
for which the electrical attraction is balanced by the centrifugal force acting on the electron, The
older ideas would have permitted any ciruclar orbit to exist. Bohr’s theory stated that only specific
orbits are permissible (see Fig. 2.5). A generalization of Bohr’s idea made by the German physicist,
Sommerfeld, actually allows several elliptical orbits for each of the circular Bohr orbits. In the e)-
liptical orbit the electron has the same total energy, that is, kinetic energy plus potential energy, a5
does the electron in the equivalent circular orbit.

As we have said, each of the Bohr orbits plus their associated elliptical orbits correspond to a
certain total energy for the electron, An electron which reduces its total energy is known to radiate
the difference in energy in the form of electromagnetic waves. If for some reason an electron is
forced to move from a stable orbit close to the nucleus to one further out, it will then, ina short
time, return to its original position. To move the electron outward, an external force is required.
This does work upon the particle, giving the additional energy. In returning to its original position,
in a lower energy state, the added energy 1s lost. This appears in the form of light or X rays. Since
each orbit has a definite energy associated with it, each of the electron jumps will cause the radia-
tion of a photon of a definite energy. The frequency or ‘color’ of the electromagnetic radiation is
determined from the energy of the transition by the expression E = hy, where E is the energy differ-
ence between the initial and final orbits, h is Planck’s constant, and p is the frequency.

The chemical properties of a given atom are entirely determined by the electron structure and
are independent of the nucleus. There are preferred numbers of electrons in each of the orbits about
the nucleus. Two electrons will fill the first shell, eight the second, and eighteen the third, etc. An
outer shell which is nearly filled, but still incomplete, seeks to add electrons. An outer shell which
has only one or two electrons seeks to give those electrons away. This is the basis of simple chem-
ical binding in the formation of a molecule (see Fig. 2.6). The atom which desires to give electrons
shares them with an atom which is seeking to complete a shell, Electrons which are available for
such sharing are known as ‘‘valence electrons.”

Electrons can be added to or subtracted from normal atoms in other ways; that is, an electron
may be knocked out of an orbit, and the atom left in a net positive electrical state, or an atom may
complete a shell from a supply of free electrons and become electrically negative. Such charged
atoms are known as “tions.” Ionization is frequently found where gases have been partly bombarded
with radiations or in solutions where chemical molecules have dissociated, with the shared electrons
going entirely into nonmetal atoms and leaving the metal atom in a positive state. Such charged atoms
will migrate in an electrical field owing to the electrica} attraction and repulsion and can be collected
as in an electrolytic cell or ionization ehamber.

tt is obviously quite possible for an atom to retain the same total number of electrons and protons
and therefore retain the same chemical characteristics while the mass of the nucleus is changed by



[Document No. ORAUT-TKBS-0012-5 | _RevisionNo. 01__| Effective Date: 10/01/2007 | Page 350f74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 1 of 36
TABLE OF CONTENTS

SECTION TITLE PAGE
Ad Recovery of Historical Data .

A.1.1 Data Description...

A.1.2 Data Classification

A.1.3 Synopsis of Recovered Data .
A.2 Calculation Method for Sample MDAs

A.2.1 Standard Deviation of Background Counts 40

A.2.2. Background Count Time. 40

A.2.3. Instrument Efficiency ...

A.2.4 Yield

A.2.5 Volume of Aliquot Analyzed

A.2.6 Volume of 24-Hour Sample ...

A.3 Calculation Method for MDAs for Various Periods Between 1945 and 1988 ..0........ cee 41
A4 — Additional Considerations, ............ccccceccseceececteeeceesneresrsnesesneserresesseaeseseseessesencersssesenseseenrennae 41
LIST OF TABLES

TABLE TITLE
A-1 Data used to determine historical MDAs for radionuclides in urine and feces.

A-2

MDAs calculated from recovered data for radionuclides


UNIT 7

(Since Unit 7 dealt primarily with the work of the laboratory, as mentioned in the Preface, it is
not included.)

80 AECU-817



16 AECU-817

Impurities that have appreciable capture cross section must be excluded If sustained reaction
is to take place. The presence of nonfissioning u238 makes it necessary to be conservative with neu~
trons if natural urantum is to be made self-sustaining in a chain reaction. Uranium tsotopically en-
riched in U235 can be made to react with remarkably less mass, inferior moderators, and with ap-
preciable foreign materials present. The fission products that form, however, limit the reactivity
of an assembly as it operates. Some of these newly generated elements have relatively high capture
cross sections and, as they build up, tend to poison the reaction by robbing it of neutrons,

In order to maintain a nuclear chain reaction indefinitely we must not only replenish the ‘‘fuel’’
but remove the ‘‘ashes,’””

The ability of any assembly of fissionable material to maintain a chain reaction may be expressed
as its reactivity represented by its ‘‘k’’ factor. The k factor is the ratio

Number of fissions resulting _
Number of fissions supplying neutrons

Neutrons from a source may be multiplied by an assembly with k < 1, but when k = 1, a self-main-
tained chain reaction exists,

A pile is controlled by inserting poisons of high capture cross section such as boron or cadmium
so that the reactivity may be held to 1 or reduced to less than 1 to stop the reaction (k = 1 at any
steady power operating level, k > 1 paves the way for an increasing rate of reaction).

Reactivity has a temperature coefficient which we hope will protect pile operations against a
catastrophe, although we do not depend on it. How temperature could affect nuclear reactions may
be postulated by considering that the increased kinetic activity of moderator atoms at elevated tem-
peratures relative to that of the massive fissionable atoms hinders the moderator’s function of slow-
ing down neutrons by recoil.

Control of a pile is feasible because an appreciable fraction of the fission neutrons is released
after a momentary delay, Should the k of an assembly be raised to a point where reaction without
the delayed neutrons would take place, control would be lost. We can imagine a critical assembly
abundant enough in neutrons to sustain fission captures without the moderation and thrifty conser-
vation of neutrons described above.

A pile is useful because it has an abundance of neutrons. Incidental to the reaction there is a
considerable neutron flux over a generous range of energies. This is useful in bombardment of other
atoms to produce transmutations, It made practical the production of plutonium by converting 238
to U289 which decays in two rather rapid stages to Pu259, similar transmutations are the heart of
our current radioisotope industry.

The nuclear chain reaction releases energy which is dissipated in absorbing materials and thus
converted to heat. Present proposals to utilize this atomic energy postulate the use of heat engines,
In the future we may realize more direct application of the kinetic energy imparted to the particles
resulting from fission.

2.5 RADIOACTIVITY DEFINED
By W. H. Ray

The discovery of radioactivity in 1896 by the Frenchman, Becquerel, is a classic example of a
finding incidental to an objective investigation which is of greater significance than the objective.
In seeking a materia} that would fluorescent readily when excited wth X rays (discovered the pre-
vious year). Becquerel was checking some salts of uranium which he had previously observed would
fluoresce under ultraviolet irradiation. Pausing in his work, he stowed the salts in the same drawer
with some photographic material. When he later developed the emulsions which had been wrapped
in light tight paper enclosures, he was surprised to find that they had been affected, Thinking this
penetrating radiation resulted from previous excitation of the salts, Becquerel ran a series of ex-
periments to check the response quantitatively.. He not only found that salts prepared in darkness had
the same property, but that this radioactive characteristic was associated always with uranium.



(Document No. ORAUT-TKBS-0012-5 | _Revision No. 01_| Effective Date: 10/01/2007 [ Page 61 of 74 ]

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 27 of 36

Annual Average MDAs for Sr-90 in Urine
Calculated from Recovered Data Between 1945 and 1990
70 :

60

50

MBDA (dpmi24h)

204

Figure A-12. MDA data for Sr-90 in urine (part 2 of 3).

Number of Sample Results Recovered per Year
Between 1945 and 1990 for Sr-90 in Urine
1400 7

1200

1000

g

8

Number of Samples (N)

8

200

Figure A-12. MDA data for Sr-90 in urine (part 3 of 3).



[Document No, ORAUT-TKBS-0012-5 | Revision No. 01 _| Effective Date: 10/01/2007 | Page 57 of74_ |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 23 of 36

Number of Sample Results Recovered per Year
Between 1945 and 1990 for Pu-239 in Urine
1400

1200 {—___________

1000

800

Number of Szmpler (N)

400

200

Figure A-9. MDA data for Pu-239 in urine (part 3 of 3).

1B+2

+ Sample MDA
MDA of 9.2.(95%CL on Mean)
Percentile of 6.7 . i

1 E414

LENO f-—-—

‘Activity (Spmv/24 h snaaple)

1945, 1950 1955 1960 1965 1970 1975 1980 1985 1990
Year

Visual Comparison of Data to Estimated Pu-241 MDAs in Urine

Figure A-10. MDA data for Pu-241 in urine (part 1 of 3).



{Document No. ORAUT-OTIB-0078 | Revision No. 01 [Effective Date: 08/09/2005 | Page 100f 35 _|

4.0 INHALATION INTAKE
41 INTAKE QUANTITY
The inhalation intake is determined by

Intake = (air conc.)(breathing rate)(exposure period) (1)
From Equation 1, intakes in Bq/yr can be determined by

Intake (Ba/yr) = (air conc. pCi/mL)(1.2 m*/hr)(2,000 hr/yr)(10° mL/m*)(3.7 x 10* Bq/yCi)
or
Intake (Bo/yr) = (8.9 x 10")(air conc. uCifmL) (2)

where the air concentration is obtained from Table 4-1 or from site-specific limiting air concentrations
(in pCi/mL) when applicable.

Table 4-1. Default limiting air concentrations.

Air concentration Associated daily

Exposure type (pCi/mL) intake (pCi)
Alpha-emitting radionuclides except for uranium facilities 2E-12 13.2
(includes Hanford from 1949, see Attachment A)
Beta-emitting radionuclides except for uranium facilities 2E-9 7.32E4
{includes Hanford from 1949, see Attachment A)
Uranium including recycled uranium (includes Hanford from 6E-10 3.96E3
1949, see Attachment A, and Los Alamos from 1951) _
ORNL alpha-emitting radionuclides except for uranium 3E-11 1.98E2
facilities, 1944-1952
ORNL beta-emitting radionuclides, 1944-1952 TE-7 6.60E5
Hanford alpha-emitting radionuclides except uranium facilities, 4E-1T 2.64E2
1946-1949
Hanford beta-emitting radionuclides, 1946-1949. 1E-8 6.60E4
Los Alamos alpha-emitting radionuclides except uranium 3.2E-11
facilities or Po-210 facilities, 1948-52
Los Alamos alpha-emitting radionuclides except uranium 4E-12
facilities or Po-210 facilities, 1953-66
Los Alamos beta-emitting radionuclides, 1948-52 1E-7
Los Alamos beta-emitting radionuclides, 1953-66 3E-9
Los Alamos uranium facilities, 1948-50 1.8E-10
Los Alamos Po-210 only*, 1948-53 (exposure ended in1953) 6.3E-10
Mound Po-210 only*, 1952-71, (exposure ended in 1971) NA 5.88E3
Rocky Flats apha-emitting radionuclides in Pu chemistry 4E-12 26.4
operations or default for the site, 1953-1994 (2E-12 uCi/ml was
used in the metal operations).

a. Los Alamos and Mound had special air concentration limits or urinalysis limits for Po-210. The doses associated w ith
this intake will be larger than for general alpha-ervitting radionuclides at 13.2 pCi/d for years of exposure but will be less
for years after exposure ends. The dose reconstructor w ill have to run both cases and determine w hich option gives the
largest probability of causation.

The intake per calendar day in pCi is
Intake (pCifcal. d) = (air conc. yCi/mL)(1.2 mMr)(2,000 hréyr)(10® mL/m?)(10® pCisuCiy/365 cal. d/yr



AECU-817

ON- REGION
CHAMBER
Region

a

PULSE SIZE

VOLTS
Fig, 4.1 — Pulse Size

Dead Time and Recovery Time

4

After a counter has fired, the electrons are drawn very rapidly to the positive electrode (usually

the central wire). The positive ions are drawn much more slowly toward the wall. As a result, the

positive ions remain in the volume of the counter as a sheath, more or less well defined. This sheath

reduces the electric field between itself and the positive wire to such a value that ionization by col-
lision is impossible. Therefore, if a particle enters the counter during the time when this condition
prevails, it will not be recorded, The counter is ‘dead’? and remains in this state until the positive
{on sheath has moved out to a certain well-defined radius. At this critical radius pulses begin to
occur, but they are not of the normal height. The time required for the sheath to move the rest of

the distance to the counter wall has been named the ‘‘recovery time,’’ Each of these times is of the
order of 100 sec.

Quenching

When the positive ion sheath reaches the wall of the counter, it might have enough energy to
produce photons as a result of the positive charge being neutralized with a negative charge residing
on the wall, Furthermore, electrons might be ejected out into the counter volume, These photons
or electrons, or both, would then start the ionization process again. Then there would prevail the
wnhappy circumstances of obtaining several counts for a single initial ionizing event. This condition
may be corrected by different methods, The notable ones will be mentioned in the following.

Resistance Quenching

Wa high-value resistor is inserted in series with the counter and high voltage, a potential drop
will occur across that resistor when current begins to flow. Consequently the potential across the

tube will be smaller by the same amount. In this case the positive ions will not have so much energy

on reaching the wall as previously. If the value of the resistor is made very large, the ions will not
have enough energy to eject the bothersome photons or electrons. Thus the discharge ceases,

COUNTS PER MINUTE

VOLTS
Fig. 4.2-—Counts per minute



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01

| Effective Date: 10/01/2007 |

Page 14 of 74

Table 5-5. New division codes (MMES 1995).

Code | DIVNUM Division Code | DIVNUM Division
AC 01 Analytical Chemistry HS 08 Health & Safety RSRH
AS 70 Administrative Services Ic 09 instrmt & Controls
AT 07 Applied Technology IF Isotopes
Bi 02 Biology IR 43 Info. Resource Org.
BS 95 Business Systems is 72 isotopes Division
CH 04 Chemistry LP. 26 Lab Prot.
CM 20 Central Management MC 11 Metals and Ceramics
cs 63 Computing and Telecom oc 62. Controller Office
cT 03 Chemical Technology OP. 27, 32 Operations

EA 93 ESA Os 22 Operational Safety

EC 35 Environmental Compliance. PC. 64 Procurement

EH 36 Env. & Health Prot. PE 21 Plant & Equipment

EN 15 Energy PH 13 Physics
EP 12 Eng. Physics & Math PU 73 Publications
ER 29 Employee Relations QA 24 Quality
ES 42 Env. Sciences RE 87 Env. Restoration
[aa 16 Eng. Technology RP 14 Robotics
EX 90 Executive Offices RR 06 Research Reactors
FE 19 Fusion Energy. RU Rust
FM 37 Finance and Materiais $s 18 Solid State
FR 14 Fuel Recycle TR 60 Treasurer Office
GE 69 ORNE Engineering Vt Visitor
GR 71 Graphics WM. 27 Waste Management
HE 23 Health 4

Table 5-6. EOP urinalysis abbreviations and codes (Mani 1983; MMES 1995).

Alphanumeric code Numeric code Alphanumeric code Numeric code
CMO Cm-244 000 | Other PAQ Pa-230 alpha 016 | T-204
coo Co-60 001 $-35 PA3 Pa-233 beta 017 Np-237
cso Cesium B (Cs-137) | 002 Co-60 PH2 P-32 018 {| Ag-110m
cs7 Cs-137 003 Pb-210 PM7 Pm-147
EU Eu-154 004 Na-24 Poo Po-210
FPO, Fission products. PUO, Pu-239 alpha
FP (Cs-137) 005 | Zr-95/Nb-95 Pu
Fuo Total rare earths () |006 | Tc-99 Put Pu-241
GAO, Gross alpha Pug Pu-239
GA (Pu-239) 007 | As-74 RAO Ra-226
GBO, Gross beta (Sr-90) RUS Ru-106
GB 008 Br-82/Br-83 SRO Sr-90
GDO Gd-153 009 Fe-59 SR5 Sr-85
GGO Gross gamma SR9 Sr-89

(Cs-137) 010) | Mn-54 TAO Ta-180
GUO Gross alpha TH ‘Th-232

(Pu-239) O14 1-134 TPO Trans plutonium
HY3__[H-3 012 [Cs-132 alpha (Am-241)
25 1-125 013 Gross beta TRE Total rare earths ()
134 1-134 014 | Ba-140 URO U-234 alpha
NPO Np-237 O15 {Sb-125

The ORNL database contains values for isotopic activities during times when isotope-specific
analyses were not possible or routinely performed. in earlier years, the element of concern was



[Document No. ORAUT-OTIB-0034_[ Revision No. 00 _| Effective Date: 12/13/2005 | Page 230f29 |

* annual bioassay
value (50th percentile)
© notused in fit
fit

Figure A-13. Assumed
Type S.

'U intake, 1959 to 1961, 50th-percentile results,

* annual bioassay
value (50th percentile)
* not used in fit
fit

Figure A-14. Assumed ““U intake, 1962 to 1963, 50th-percentile results,
Type S.

* annual bioassay
value (50th percentile)
© not used in fit
fit

Figure A-15. Assumed
Type S.

U intake, 1964 to 1982, 50th-percentile results,



12 AECU-817

tal particle in the nucleus is the ‘‘neutron,’’ whose discovery was announced by James Chadwick of
England as the ultimate result of a set of experiments which had been done in many parts of Europe.

When free neutrons were originally observed, it was thought that they were very energetic high
penetrating gamma rays. Experiments finally showed them to be particles almost exactly equal in
mass to the proton, but with no electrical charge. Protons and neutrons taken together in definite
number and frequently varying proportions make up all the nuclei of known chemical elements. There
are, however, some arrangements in proportions of these fundamenta) particles which produce nuclei
that are more stable and longer lasting. It is these nuclei which are found in nature,

At first sight, it might seem that the mass of any given nucleus ought to be some integral mul-
tiple of the proton mass. However, this is not the case, since in being bound to the nucleus as con-
stituent, the particle loses a small amount of its total mass proportional to the binding energy which
holds the particle in the nucleus.

According to Einstein’s statement of the equivalence of mass and energy

E = mc?

The energy loss experienced by a free particle when it becomes part of the nucleus appears with a
change in mass equal to the energy change divided by the quantity c2, where c is the velocity of light.
If the actual mass of a nucleus is subtracted from the mass of an equal number of free protons, a
difference obtained is called the ‘‘mass defect’’ and is proportional to the total binding energy hold-
ing the nucleus together. If the mass defect is divided by the total number of particles in the nucleus,
one obtains a quantity, called the “‘packing fraction,” which is proportional to the binding energy per
particle in the nucleus (the approximate curve shape in a packing fraction is shown in Fig. 2.9). Nuclei
whose packing fractions have the lowest possible values are, in general, most stable.

It might be well to add a word about the atomic mass unit. This unit is not based on the mass of
a free neutron or proton. It is based instead on the atomic weight of the most abundant isotope of
oxygen divided by 16. Since, from physical experiments, the mass of the oxygen atom is very slightly
less than the mass of 16 individual protons, it may be assumed that there are 16 fundamental parti-
cles in an oxygen nucleus, and the atomic weight divided by 16 will represent an average mass of
each individual particle within the nucleus. The number obtained in this way will be smaller than
the mass of the free proton, because each of the particles in the oxygen nucleus has lost a small
fraction of its mass in the form of binding energy.

It is possible for a nucleus to exist in several different states analogous to the various states
in which the electrons around the nucleus may be arranged. There is no picture of orbits which can
apply here. However, experimenta) sources state that 2 nucleus may change from one energy state
to another, with the excess energy being emitted as an extremely high-energy electromagnetic photon
called a ‘‘gamma ray.”’

2.3. TRANSMUTATION AND NUCLEAR REACTIONS
By R. S. Thackeray

In the last two lectures a great mass of material was presented hurriedly, In order to crystal-
lize agreement on the fundamental points on which great emphasis is to be placed, we shall briefly
review some of the points covered,

The characteristics of the fundamental particles about which we have talked are listed in
Table 2.1,

The portion of this lecture which was devoted to review also covered a mechanism by which the
orbital electrons can radiate photons whose energy is characteristic of the difference in energy be-
tween stable electric states. Suppose that a heavy object, on roller bearings to eliminate friction,
could be pushed to the top of a hill. One can imagine that such an object would be difficult to move
up the incline and that its velocity would never become very great. If the object’s velocity was al-
ways close to zero, it would never have a great momentum and could not do appreciable work on any



44 AECU-817
now

Ege
a- 0

AQ

It is obvious from the last expression that if one increases the proportional amplification, one
reduces the amplifier gain needed and thereby avoids the attendant difficulties of high-gain ampli-
fiers, It should be noticed that a reduction of input capacity also increases the pulse voltage output.
Practical amplifiers are usually followed by discriminator circuits which reject low-level disturb-
ances and produce uniform pulses for feeding rate meters or sealers,

Proportional counters have the advantage of speed over Geiger counters and are practically
limited by the speed of the amplifiers one can construct.

Some gases used for proportional counters and reasons for their use follow:

Air is used in proportional alpha survey counters to avoid the inconvenience of a sealed tube.
These counters can detect alpha path lengths of 1 mm. However, air has a steep proportional char-
acteristic, and air counters require well-regulated power supplies, Since peta ion densities may be
of the order of 20 per centimeter as compared to 30,000 per centimeter for alphas, beta proportional
counters require considerably more gas amplification than alpha counters, This makes air unfeasi-
ble for use in a beta proportional counter. Methane and argon~carbon dioxide (98 per cent A, 2 per
cent COg) are used for beta proportional counters and quantitative alpha proportional counters* be-
cause of their long proportiona) gas characteristic (long with respect to voltage).

Counter construction and operation hints: Precautions must be taken to minimize spurious dis-
charges in the counter and its associated circuit. These precautions involve the use-of good cables
and connectors, well-insulated counters, filtered and regulated high-voltage supplies, and smooth,
clean counter-electrode surfaces.

‘The use of some thin membranes as windows on alpha air proportional counters (nyton or rubber)
has now become standard practice. The window is of considerable value in avoiding discharges due
to acid fumes and moisture in the air.

Rate meters may sometimes be of value to a surveyor, but since there is no way of knowing
how the contamination is embedded in a surface, a quantitative reading is usually of little value.

Colloquium

Question: How about the counting rate as a function of voltage?

Answer: With the laboratory-type counter, high counting accuracy requires precautions in the
amplifier discriminator circuit to minimize multiple pulses and background pulses, If the gas am~
plification is too high for a given amplifier, the following bad effects result from overdriving the
amplifier: multiple pulsing and differentiation difficulties. There is a plateau associated with the
proportional counter, and it occurs when the gas amplification is sufficient to allow one to count all
but the smallest pulses produced in the tube. The proportional beta counter circuit is much more
difficult than the alpha counter because the spread of pulse heights is considerably greater. The
spread of pulses for alpha is 25 to 50, and it is of the order of 500 to 1000 for betas.

Question: Is a plateau similar to that of the plateau in the Geiger region somewhat independent
of the voltage?

Answer: Yes if the amplifier has sufficient gain and does not produce multiple pulses due to
differentation difficulties or dead spots due to some tube being driven positive.

Question: What about acid in the presence of the open window? And how about beta tolerance?

Answer: You can get a spurious count from it, Beta tolerance of an alpha proportional counter
1s of the order of 107

Question: What about probing too fast?

Methane is not used in a gas-flushed counter where inflammablility is a problem, such as in
some types of survey work,



‘ ‘ ORAU TEAM
we ey Dede Mocller Miw . Dose Reconstruction

aS Project for NIOSH

Oak Ridge Associated Universities ! Dade Moeller & Associates | MJW Corporation

Page 1 of 74.
Document Title: Document Number: ORAUT-TKBS-0012-5
Oak Ridge National Laboratory — Revision: ot
Occupational Internal Dose Effective Date: 10/01/2007
Type of Document TBD
Supersedes: Revision 00
Subject Expert(s): Thomas La Bone
Site Expert(s): Mike Bollenbacher, Kenneth Fleming, and Elizabeth M. Brackett
Approval: Signature on File Approval Date: 09/13/2007
Robert E. Bums, Jr., Document Owner —_ —
Concurrence: Signature on File Concurrence Date: 09/13/2007
John M. Byme, Task 3 Manager
Concurrence: Signature on File. Concurrence Date: 99/13/2007
Edward F. Maher, Task 5 Manager
Concurrence: Signature on File Concurrence Date: 09/27/2007
Kate Kimpan, Project Director
Approval: Brant A. Ulsh Signature on File for Approval Date: 10/01/2007
James W. Neton, Associate Director for Science
1 New 0 Total Rewrite &) Revision (1 Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR
REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.



FETE A MTT

(Document No, ORAUT-TKBS-0012-5 | Revision No. 01__| Effective Date: 10/01/2007 |

Page 7 of 74

Naf/Ge
nCi
NCRP
NIOSH
ORNL

pCi
POC

SBC
TBD
U.S.C.

WBC
wk

uci
§

sodium iodide/germanium

nanocurie

National Council on Radiation Protection and Measurements
National Institute for Occupational Safety and Health

Oak Ridge National Laboratory

picocurie
probability of causation

scanning bed counter
technical basis document
United States Code

Whole-Body Counting (Facility)
week

microcurie

section or sections



UNIT 8. SOME PRACTICAL PROBLEMS

8.1 GUARDING HAZARDOUS AREAS AND CONTAMINATION CONTROL
By G. C. Cain*

All hazardous areas should be promptly identified, marked, or guarded as soon as hazards occur
to keep personnel from being over exposed and to keep other areas from becoming contaminated.
That brings up the question of what a hazardous area ts. From the point of view of radiation pro-
tection, any area is to be considered hazardous if the level of radiation or the degree of contamina-
tion is such that, over long periods of time, any person working in the area might be expected to
receive radiation exposures above the established maximum permissible levels. A wide-awake sur-
vey will, in many cases, anticipate a hazardous area or contamination, even before it occurs, from
experience and knowledge of the operations involved.

The maximum permissible levels for an occupied area are as follows: beta and gamma radia-
tiont of 12.5 mrep/hr; alpha-emitting contamination of 3000 dis/min in an area of 150 cm2, or the
minimum amount of alpha material which can be detected by use of the ‘‘smear’’ technique; a ther~
mal neutron flux of 1600 neutrons/cm? sec; a fast neutron flux of 200 neutrons/em® sec; and atmos-
pheric contamination of 10° .c/cc for beta~emitting materials and 3 x 107! yc/ce for alpha-emit-
ting materials. If two or more types of radiation are present, the combination shall not produce an
exposure to any part of the body biologically equivalent to more than 12.5 mrep/hr? of beta or gamma
radiation.

Radioactive materials may present special hazards in case of fire or other accidents. For the
protection of firemen, areas containing radioactive materials which may become hazardous In case
of fire are marked by use of an 8-in, magneta plaque. Firefighting procedures require the use of
masks in certain designated areas until it has been determined by authorized personne} that no sig-
nificant inhalation hazard exists. As an emergency procedure, firemen may work in levels of radia-
ton as high as 20 r/hr and may be permitted to receive single exposures as high as 5 r.

All the maximum permissible levels should be considered just so—as maximum permissible
levels. They are not tolerable unless unavoidable, Our job is to try to see that people work in areas
as free of radiation and radioactive contamination as possible.

The majority of radioactive materials originate in the pile building. They are usually sent from
there to the other buildings. Some conditions under which guarding requirements arise are (1) active
samples from the pile, (2) active products from other areas, i.e., isotopes, (3) transfer and storage
of smaller active samples in the laboratories, (4) storage and use of sources, (5) transferring hot
waste to the burial ground (may need patrol escort), and (6) shipments to X-10 from other sites of
active wastes.

Ways and means of guarding or identifying these areas bring on a lot of discussion in determining
just how areas should be marked off. There are small radiation-danger tags for individual small
samples, contaminated tools, and equipment, We can use signs of a temporary or permanent nature.
If it is temporary, then by all means describe the level at which the sign is placed and also the max-
imum level at the source of radiation. Those signs should be brought up to date each day. There are

*Notes taken by C. Scott; edited by G. C. Cain,
TAt the International Conference on Radiation Protection held at Chalk River, September 29-30,
1949, a maximum permissible rate of exposure of 0.3 rep/week was recommended,

AECU-81T 81



[Document No. ORAUT-OTIB-0016 | RevisionNo.01 __| Effective Date: 08/09/2005 |__Page 13 of 35_|

These intakes would be applied to the same years for which inhalation intakes occurred, using the
same radionuclide. Use the f, associated with the absorption type from Annex F in ICRP Publication
68 (ICRP 1994b).

6.0 DOSE DISTRIBUTION

Assuming an intake at the limiting air concentration for every minute of employment is probably an
overestimate. However, there are sources of uncertainty. For instance,

e The air sample resutts might not have been representative of the air breathed. Ifa worker was
close to a localized source, the air concentration the worker breathed might have been higher
than the air collected by a sampler several to tens of feet away. Offsetting this problem
somewhat is the likelihood that the worker breathed larger particles than the air sampler
collected. The time spent near a localized source would have been less than full time.

enforcement of the limiting air concentrations at some facilities for some periods might have
been lax.

e Prior to widespread use of alarming continuous air monitors (CAMs), exposure to air
concentrations above limiting air concentrations for periods of days to a few weeks was
possible. For instance, a filter exchanged weekly with delayed counting for radon decay could
easily result in a 2-week lag between the change in air concentration and change in access
control. Even with the use of CAMs, radon progeny interference was a problem with alpha
monitoring in many facilities such that immediate recognition of an air concentration slightly
greater than the plutonium limit was difficult

e Forbeta emitters, the counting efficiency for converting counts to total activity is dependent on
the beta energy. If the beta energy of the radionuclide on the filter is less than the beta energy
of the calibration source, the total activity would have been underestimated. This is offset
somewhat by the generally lower dose per unit intake from the lower energy beta emitters.

For alpha emitters, self-absorption of the alphas due to high dust loading might not have been
compensated for.

e The volume of air sampled might not have been measured accurately; for instance, it was

common to use flow rate meters and the average of the on/off flow rate to estimate the volume
of air rather than using total air volume meters.

© Operational logistics might have temporarily compromised the air sampling in some areas,
such as pump failures, leaking filters, filters destroyed by harsh chemicals, or filters obviously
contaminated by a source other than-the air.

These problems arguably could have enabled intakes exceeding the air concentration limits for some
periods. Because of the sensitivity to these uncertainties of the method used in this TIB to determine
dose, the dose uncertainty distribution is lognormal with the geometric mean as determined by
the equations in this document with a geometric standard deviation of 3.

Remember that exposure to tritium, radioiodines, radon, or ‘4C has to be accounted for separately
from these intake/dose estimations.


ORAU TEAM
/ Dose Reconstruction
Dade Moeller
wee ay Project for NIOSH

Oak Ridge Associated Universities | Dade Moeller & Associates | MJW Corporation

Page 1 of 29

Document Title: Document Number: | ORAUT-OTIB-0034
Revision: 00

Internal Dosimetry Coworker Data for X-10
Effective Date: 12/13/2005

Type of Document: OTIB

Supersedes: None

Subject Expert: William E. Kennedy, Jr.

Document Owner

Approval: Signature on File Approval Date: 12/07/2005
Judson L. Kenoyer, Task 3 Manager

Concurrence: _ Signature on File Concurrence Date: _ 12/07/2005
Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 12/13/2005
James W. Neton, Associate Director for Science

K New [1 Total Rewrite 1 Revision ( Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE REPLACE THE PRIOR
REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.



[Document No. ORAUT-TKBS-0012-5 | RevisionNo. 01 __| Effective Date: 10/01/2007 | Page 62 0f74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 28 of 36

1E+2

E+
: 1E+0
:
<

1E-]

1E-2

1945 1950 1955 1960 1965 1970 1975 1980, 1985 1990

Year

Visual Comparison of Data to Estimated U-233 MDAs in Urine

Figure A-13. MDA data for U-233 in urine (part 1 of 3).

Annual Average MDAs for U-233 in Urine
Calculated from Recovered Data Between 1945 and 1990

10 nme

MDA (dpmi24 hy

Figure A-13. MDA data for U-233 in urine (part 2 of 3).



UNIT 3. APPLICATION OF PHYSICS TO HEALTH PHYSICS

3.1 THE ROENTGEN AND ITS DERIVATIVES
By Ross S. Thackeray

The first purpose of this discussion is to make clear that the roentgen is not a unit of radiation.
It does not describe directly either the number of particles in the beam or their energy. A physicist
has little use for this unit since it tells nothing about the rays themselves or what they are likely
to do.

The roentgen is a unit of radiation exposure based on the effect of X or y radiation on the air
through which it passes. It is also important to note that the unit applies only to X and y radiations
in air. Any attempt to use the unit for other radiations or for dosages within a medium other than
air implies an extension of the present definition.

The roentgen is currently defined as ‘‘that quantity of K or y radiations such that the associated
corpuscular emission (secondary electrons) per 0.001293 g of air (which is equal to 1 cc at 0°C and
760 mm Hg pressure) produces in air ions carrying 1 esu of quantity of electricity of either sign.”

We can consider the various parts of this definition in the light of what we know radiation can
do to the material which absorbs it. X or y photons of low energy will separate electrons from atoms
by the process of photoelectric absorption. In this way the photon disappears, and a high-speed elec-
tron is ejected from the atom. This process has caused one ionization, and the quantity of electric
charge separated by this process is too small to measure, even when repeated many times in a beam
of radiation. However, an amplification of the amount of charge separated occurs because each
photoelectron will cause between 10,000 and 100,000 ionizations along a short track. If the multiple
ionization by secondary electrons is repeated frequently as in a beam, there is soon a charge avail-
able which is easily measured with an electroscope.

As the energy of the radiation is increased, other processes begin to absorb more of the pho-
tons. These processes, called ‘Compton scattering’’ and ‘‘pair production,’’ also give rise to
secondary electrons which cause large numbers of ionizations for each ionization caused directly
by the photons.

From this it is easy to see that the roentgen unit considers the ionizations caused only by sec-
ondary particles ejected from some known volume of air, Their ionized tracks, however, may go
outside of the known volume, and it is important that their total ionization be collected wherever it
occurs.

Many units have been proposed for comparing exposures within a material other than air and
for measuring dosages of radiations other than X or y. Some of these have been based on the equiv-
alence of the biological effect of a given amount of the other radiations to 1 r of X or y. This equiv~
alence, however, is not useful, since the effects are too different to compare. The failure of the bio-
logical equivalence led to consideration of some sort of physical equivalence.

One roentgen of X or y rays corresponds to the absorption of 83 ergs of energy by each gram
of bombarded air. Dr. H. M. Parker has based his unit, the rep (roentgen equivalent physical), on
this fact, The rep is defined as that amount of any ionizing radiation from which tissue will absorb
energy to the extent of 83 ergs/g,

It will be noticed that Parker has actually used the energy absorption per gram of air per roent-
gen for the absorption in tissue. Tissue will absorb energy to the extent of approximately 93 ergs/g/
incident roentgen, and therefore the rep actually embodies about a 12 per cent factor of safety.

22 AECU-817



64 AECU-827
Appendix B: Example of NBS Certificate
National Bureau of Standards Certificate for One Specimen of Radium Salt

NBS No.
Submitted By

Description. The material is contained tn a metal tube mm long and __mm in external diameter.
The specimen and container weight . The Bureau is informed that the material is a compound
of radium. After the measurements had been completed, the specimen with a card bearing the num-
ber of this certificate was enclosed in a package suitably inscribed and secured by a seal of this

Bureau.

THIS CERTIFIES that the specimen described in the preceding paragraph has been compared with the
Radium Standard of this Bureau and has been found to have a gamma radiation equivalent to that from

Xxx.xx milligrams of radium

in radioactive equilibrium and contained in a tube of Thurtngian glass 0.27 mm thick.* The uncertainty
in this value does not exceed 0.7 per cent. Observations extending over 13 days indicate that the
radium contained in the specimen is in radioactive equilibrium.

For the Director,

by
Chief, Radioactivity Section,
Division of Optics.
Test No.
Date;
*NOTE: The Bureau is informed that the tube is of platinum and has walls 0.5 mm thick, On this
basts if the radioactive material were contained in a glass tube of the kind specified above,

the gamma radiation from the specimen when in equilibrium would be approximately 4.3
per cent greater than the value named in the body of the certificate.

The observation upon which this certificate is based do not serve to distinguish between radium prep-
arations and those containing mesothorium and radiothorium.

5.9 INSTRUMENT CARE AND REPAIR
By Davis, Gatner, Ryan, and Hamrick*

Cost of Instrument

The following list indicates the cost of some of our more commonly used instruments:

*Notes taken by C. Scott; edited by D. M. Davis.



Special Exposure Cohort Petition U.S. Department of Health and Human Services

under the Energy Employees Occupational Centers for Disease Control and Prevention
{iness Compensation Act National Institute for Occupational Safety and Health
OMB Number: 0920-0639 Expires: 09/20/2013

Special Exposure Cohort Petition — Form B Appendix — Petitioner 3

Use this Appendix for Petitioner 3.

This appendix form is to be used as needed. Petitioner 3, or his or her representative, should complete the
Parts applicable to him or her.

Refer to the General Instructions on completing petitioner information for Parts A, B, or C.

If you need more space to provide additional information, use the continuation page provided at the end of
the form and attach the completed continuation page(s) to Form B.

Except for signatures, please PRINT all information clearly and neatly on the form.

Q An Energy Employee (current or former), Start at C

If you are: |Q A Survivor (of a former Energy Employee), Start at B
Start at A

Representative Information — Complete Section A if you are authorized by an Employee or
Survivor(s) to petition on behalf of a class.

O A Representative (of a current or former Energy Employee),

A. Are you a contact person for an organization? 0 Yes (Go to A.2) Q No (Go to A.3)
A2 Organization Information:

Name of Organization

Position of Contact Person
A3 Name of Petition Representative:

Mr/Mrs./Ms. First Name Middle Initial Last Name

A4_ Address:
Street Apt # P.O. Box
City State Zip Code

A5 Telephone Number: ( ) -
A6 Email Address:

A.7 OQ Check the box at left to indicate you have attached to the back of this form written authorization to
petition by the survivor(s) or employee(s) indicated in Parts B or C of this form. An authorization

form for this purpose is provided.

Name or Social Security Number of First Petitioner:



Special Exposure Cohort Petition U.S. Department of Health and Human Services

under the Energy Employees Occupational Centers for Disease Control and Prevention
lliness Compensation Act National Institute for Occupational Safety and Health
OMB Number: 0920-0639 —-_ Expires: 09/20/2013

Special Exposure Cohort Petition — Form 8
Proposed Definition of Employee Class Covered by Petition — Complete Section E.
E14 Name of DOE or AWE Facility: _©2k Ridge National Laboratory (X-10)

E.2 Locations at the Facility relevant to this petition:

Page 4 of 7

All areas and buildings within the perimeter of the X-10 plant.

E.3 — List job titles and/or job duties of employees included in the class. In addition, you can list by
name any individuats other than petitioners identified on this form who you believe should be
included in this class:

All contractor employees, subcontractor employees, and Atornic Energy Commission employees who were monitored or should,

hhave been monitored for any of the various radtonuclides and fission products present at the X-10 plant during the proposed time period

E4 Employment Dates relevant to this petition:

Start January 1943 End December 1952
Start End
Start End
E.5 Is the petition based on one or more unmonito! unrecorded, or inadequately monitored or
recorded exposure incidents?: Yes v No

If yes, provide the date(s) of the incident(s) and a complete description (attach additional pages
as necessary):

Name or Social Security Number of First Petitioner:



(Docent No. ORAUT-OTIB-0034__|

Revision No. 00

[Effective Date: 12/13/2005 [_ Page 25 of 29

Table A-9. Type S TMU intake periods and rates.

annual bioassay
value (50th percentile)
not used in fit

fit

Figure A-18. Assumed Pu intake, 1951 to 1952, 50th-percentile results, Type M.

Type SU intake rate (dpmiday) |] GSD Type S$ ““U intake rate (dpm/day) | GSD
Year | 50th percentile | 84th percentile Year | 50th percentile | 84th percentile
1951 850.7, 2048 2.41 1970 23.7 71.203 3.00
1952 850.7 2048 2.41 4971 237 71.203 3.00
1953 850.7 2048 2.41 1972 23.7 71.203 3.00
1954 850.7 2048 2.41 1973 23.7 71.203 3.00
1955 850.7 2048 2.41 1974 23.7 71.203 3.00
1956 675.7 2048 3.03 1975 23.7 71,203 3.00
1957 247.7 1,917 7.74 1976 23.7 71,203 3.00
1958 247.7 1,917 7.74 1977 23.7 71.203 3.00
1959 509 1,917 3.77 1978 23.7 71.203 3.00
1960 509 2,125 4.17 1979 23.7 71.203 3.00
1961 509 2,125 417 1980 23.7 71.203 3.00
1962 235.3 1,422 6.04 1981 23.7 71.203 3.00
1963 235.3 1,422 6.04 1982 23.7 71.203 3.00
1964 23.7 71.203 3.00 1983 15.25 71.203 4.87
1965 23.7 71.203 3.00 1984 15.25 31.58 2.07
1966 23.7 71,203 3.00 1985 15.25 31.58 2.07
1967 23.7 71,203 3.00 1986 15.25 31.58 2.07
1968 23.7 74.203 3.00 1987 15.25 31.58 2.07
1969 23.7 71.203 3.00


AECU-817 19
N=AeTMAt
Putting in boundary conditions No = N at t = 6, A is found equal to Ng, from which
N =Nge*t
or using I for the amount of radioactive material present instead of N atoms

I= fget

If the mean free path of a radiation in a material is 1 and a very thin slice Ax is taken from a
slab x thick, then the probability of interaction (and removal) in one slice Ax is Ax/1 with the prob-
ability of transmission,

tT

Wf there are n slices in the x thickness of material, Ax = x/n and the probability of transmission
through n slices is
x\n
( . x)

Carrying this to a limit as n ~ = and Ax — 0, we get the over-all probability as; e-*/1 which, using
u in place of 1/1 and an incident intensity Ig, gives

T= Ige"2%

Repeated practice in using the simple graphical methods of solution is urged.

2.7 FUNDAMENTAL PARTICLES AND UNITS OF MEASUREMENT
(Activity Evaluation)
By Ross S. Thackeray

In previous lectures we have covered the fundamental particles and their arrangement in atomic
structure, Others have spoken of the way in which atoms disintegrate in natural activity, We shall
now discuss radiations from active atoms, and we shall find that all the particulate radiations are
actually free fundamental particles. There are a number of types of radiation which we shall take
up in order.

Alpha particles (or stripped helium atoms)
Neutrons.

Protons

Beta rays (or electrons)

Gamma rays (or x rays)

The alpha particles are helium nuclei consisting of two protons and two neutrons bound tightly
together. The external electrons, normally associated with helium atoms, are stripped away in a
process called ‘‘double ionization."

Neutrons can occur as radiation and have the same properties as when bound in nuclei, that is,
they have no charge, and they have a mass equal to 1 amu.

Protons or stripped hydrogen atoms also appear as radiation having one plus charge and a mass
of l amu.

Beta rays are nothing more than rapidly moving electrons freed from their orbits and streaming
through space,



[Document No. ORAUT-OTIB-0018 | Revision No.01 | Effective Date. 08/09/2005 [_ Page 7 of 35 ‘|

and DOE Order 5480.11 (DOE 1988)]. Since December 1993, 10 CFR Part 835 has established
limits on air concentrations. Table 2-1 lists the history of workplace air concentration limits applicable
at DOE sites for a number of important particulate radionuclides.

Table 2-1. Limiting air concentrations for selected particulates from 1953 to present.

Limiting air concentration (uG/mL)
AEC Manual | DOE Order
Radionuclide* |_NBS 1953 NBS 1959 | Chapter 0524 5480.11° 10 CFR 835
Ipha emitters
Po-210 7E-11 2E-10 2E-10 3E-10 3E-10
Th-230 NL 26-12 2E-12 3E-12 3E-12
Th-232 NL 2E-12" 3E-1T 5E-13 BE-13
U natural sol q.7E-17 7E-11 7E-1T NL NL
U natural insol T.7E-17 6E-17 6E-1T NU NL
234 sd (D) NL 6E-10 6E-10 SE-10 SE-10
U-234 insol (Y) NL TE-10 1E-10 2E-11 2E-11
-238 sd (D) RL 7E-11 7E-17 6E-10 6E-10
U-238 insol (Y) NL TE-10 TE-10 2E-11 2E-1T
p-237 RC 4E-12 4E-12 2E-12 2E-12
Pu-238 AL 2E-T2 2E-12 SE-12” SE-12
'u-239 2E-12 2E-12 2E-12 2E-12 2E-T2
Am-2471 3E-11 6E-12 6E-12 2E-12 2E-12
m-244 NC 9E-12 9E-12 4E-12 4E-12
ta emitters
In-54 RT 4E3 4ES 3E-7 3E-7
58 AC SE-B SEB 3E-7 3E-7
0-60 TES 9E-9 9E-9 TES TE-8
[Fe-59 6E-7 SEB BE-9 TE-7 TE-7
2n-65 2E6 6ES 6E8 1E-7 TE-7
Sr-90 2E-10 3E-10 TE-9 2E-9 2E-9 —
Y-9T 4E-8 3E8 3E-8 5ES SE-8
[b-95° 4E-7 TE-7 TE-7 5E-7 5E-7
95 NC 3E8 SEB 6ES 6E-8
| Tc-99 NC 6ES GES 3E-7 3E-7
u-103 NL SES 8E-8 3E-7__ 3E-7
1-106 3E-8 6E-9 8E-8 5E-9 5E-9
Cs-134 NL TE-8 TE-8 4E8 4E-8
8-137 2E-7 TE-8 TE-B 7E8 7E-8
| Ba-140 6E-8 4E-8 4E-3 6E-7 6E-7
ta-140 6E-8 TE-7 tE-7 SE-7 SE-7
Ce-147 NL 2E-7 2E-7 SE-7 2E-7
Ce-144 7E-9 6E-9 6E-9 6E-9 6E-9
Pm-147 2E-7 TE-7 bE CES 6E-8
Eu-T62 NL 3E-7 2E8 TE-8 TE-8
Eu-T54 6E-9 4E-9 7E-9 8E-9 8E-9
Eu-155 NL 7ES 7ES 4ES 4E3—

a. Most limiting form (e.g., Soluble, insoluble, inhalation class D,W,Y) is listed, with the exception of
uranium, for which the soluble formw as based on chemical toxicity

From 1968 version

From 1988 version (effective January 1983)

NL — not listed.

Printed as 2E-12 wCi/mL, but a footnote stated that continued use of 3E-11 uCi/mL was recommended
until further investigati
f. Based on special definitions for the curie that added disintegrations from selected progeny.

oaoc



a AECU-817

5.2 PROBABLE ERROR AND CONFIDENCE INTERVALS
By W. S. Snyder

The basic problem is that of selecting a sample from a population and from a study of the sam-
ple inferring some information concerning the parent population.

A sample, even when selected in a manner as free from bias as we can determine, may still be
unrepresentative of the parent population. If a ball is drawn at random from an urn containing 99
red balls and i black ball, we may get the black ball, and if the experiment is repeated 10 times, we
may get the black ball on every draw. The theory of probability only asserts that the eventuality is
exceedingly unlikely. However, the same theory also asserts that if the experiment of making the
10 draws is repeated often enough, the eventuality of 10 blacks wil! certainly occur, and even ina
certain percentage of the cases,

Probability and statistics are principally concerned with making precise statements about the
results of a successton of trials and generally cannot assert anything very precise about a single
trial, In the above example we assumed the population known. Generally we have to make more or
less plavsible assumptions about our population, and then we study our sample to substantiate or
invalidate our assumption. We might begin by assuming our urn to contain red and black balls in
equal proportions. But after 10 draws all yielding black we would certainly have serious doubts
concerning this hypothesis, The hypothesis is not vigorously disproved; it is rather exceedingly
unlikely.

One of the commonest assumptions concerning a population is that it is normally distributed.
This means that the population when classified by a numerical measure of the attribute being studied
follows the pattern of the Gaussian or normal error curve, This curve is defined by an equation

_ bh ete- mi?
Y* Varo 00

containing two parameters, m (the mean) and g (the standard deviation), To say the population is
distributed in accordance with this law is to assert that the fractional part of the population lying
between any two limits a and b is equal to the area under the curve between the same limits. The
following tabulation lists some of the intervals most often used [see diagram (the shaded area is
68.3% of total area)}.

Area
m + 0,675¢ 0.50
mio 0.683
mt 2o 0,954
mi 3e 0,997 some

Suppose an individual X is selected from the population in a random manner, 1.e., so a8 not to
favor any one individual of the population over another, We say the probability that X would lie be~
tween the values m +o is 0.683, and we abbreviate this statement thus

Probability (m-—o < X < m +0) = 0.683
Similarly
Probability (m ~ 20 < X < m + 20) = 0,95
Nevertheless, if an individual is drawn, he may not lie in the interval m + 20, and the above

statement does not assert that he will. It simply asserts that this latter event is relatively untikely,
that in a large number of such trials in only about 5 per cent of the cases should this happen. If the



AECU-817 a

you happen to be called upon to resurvey the same object or area —or if another surveyor, not
knowing the results of your survey, is called on to survey it.

Another important factor in obtaining accuracy is to allow sufficient warm-up time before at-
tempting to make a survey with an instrument, The recommended warm-up time for most of the
instruments ranges from 30 sec up. In the Health Physics Instrument Manual compiled by D. M.
Davis, his calibration procedure calls for a 5-min warm-up time before the calibration procedure
is started. Obviously you cannot postpone every survey you are required to make for 5 min while
you allow the instrument to warm up. You can allow as much warm-up time as possible, however,
and remember to always allow at least 30 sec tor those instruments which require a warm-up pe-
riod, A procedure which I have found generally satisfactory is to take the 30-sec time in the office
and make a test setting to see if the instrument appears to be working. Then, before leaving to make
the survey, I put the sensitivity switch back on the ‘‘zero’”’ position and run the microamperes up to
around § while walking to the place where the survey is to be made. This allows me to add time to
the warm-up period without delaying the actual survey too much.

Let us go back now and consider for a few minutes the other factor governing your choice of an
instrument for making a survey, the type of radiation to be surveyed. By type of radiation Iam re-
ferring to whether the material to be surveyed is an alpha, beta and/or gamma, or neutron emitter.
This information may come either from your own knowledge of operations and materials in your
areas or from the person who requests the survey. If this information is not available, it will be
necessary for you to make a scanning survey first to determine the type of instrument needed to
actually measure the intensity of the radiation, Once you have identified the radiation which you wish
to survey, your choice of an instrument to use should be governed bv the points given you by Kuna
and Patterson. To briefly reconsider these points:

1, For low-level beta or gamma radiation the Victoreen or Lauritsen electroscope serves best,
with the electroscope being the more accurate and the Victoreen being the more sensitive of the two.

2. For intermediate-level beta or gamma radiation in the range of 5 mr/hr to 10 r/hr the Cutie
Pie is the most commonly chosen. I should like to point out that in the range up to I r/hr the electro-
scope is still usable and more accurate than the Cutie Pie.

3. For high-intensity beta or gamma in the range from 10 r/hr up to 100 r/hr the Fish pole
probe is the only choice available.

4, For surveying of alpha emitters, either the Zeus or the Zeuto may be used, The Zeus is
preferable because it permits you to determine whether the activity is purely alpha or a mixture,
while the Zeuto does not, even though it is sensitive to the beta-gamma radiation.

5. For surveying thermal neutron emitters it is necessary to use two electroscapes, One scope
is the ordinary Lauritsen type, and the other is of the same type, but it has a thin coating of boron
on the inside of the chamber. This boron, under neutron bombardment, emits alpha particles. These
particles produce ionization in the chamber in addition to that produced by gamma rays from the
source or beam being surveyed. The difference in the readings on these two scopes is due to the ad-
ditional ionization produced by the alpha particles and, through calibration, can be used to measure
the thermal neutron flux.

6. For fast neutron surveys we have only the Chang and Eng, which has so many limitations that
for general survey purposes the instrument is nearly useless.*

5.7 RADIATION MEASUREMENTS WITH FILM
By John H. Robersont

The Photographic Plate

Photographic film is composed of two essential parts, (1) emulsion and (2) the base or support
for the emulsion. The active ingredient in a photographic emulsion is pretty well standardized, con-

*More satisfactory instruments for fast-neutron surveys, using a proportional counter, have
been developed in this division during the fall of 1949.
tNotes taken by C. Scott; edited by J. H. Roberson,



07-18-11P01:15 RCVD

July 13, 2011

SEC Petition

Division of Compensation Analysis and Support
NIOSH

4676 Columbia Parkway, MS-C-47

Cincinnati, OH 45226

Re: Employee:
Employee SS#:
Petitioner:
Survivor:

To Whom It May Concern:

Enclosed please find a Special Exposure Cohort (SEC) Petition for the Oak Ridge
National Laboratory (X-10) concerning January |, 1943 through December 31, 1952. I have also
enclosed a Petitioner Authorization Form, signed by 7 which authorizes

as the petitioner for
this SEC Petition.

I have additionally enclosed continuation pages, with additional information concerning
section F.1, and attachments referenced in section F.1 of this petition.

I trust that you will ensure this petition is filed. Thank you for your attention to this
matter.

Very truly yours—

Enclosures

go


AECU-817 85

IV. Chemical- and mechanical-filter respirator
1, Combination of chemical and mechanical removal of contaminant

Important considerations in the selection of any type respirator would include: low resistance
to breathing, definite degree of protection under well-defined conditions, comfort in wearing, light-
weight, and littie interference with vision.

Masks should be intended primarily for emergency use, or for use in conditions not applicable
to control by other means, but have definite limitations in their use which should be carefully ob-
served.

Respirators at the laboratory have been chosen for use in those situations where the inhalation
of radioactive materials above prescribed maximum permissible values is a possibility. These
values at the present time are: 1.0 x 10-8 uc/ce of air (for long-lived 8,» emitters) and 3.0 x 197}
pe/cc of air (for long-l'ved a emitters).

The Oak Ridge National Laboratory Fire Department has standardized on two masks: the Chemox,
which is a self-contained oxygen unit, for use inside buildings and the Army assualt mask for use
outside buildings.

The Health Physics and Safety Departments both function in an advisory capacity as to the type
of respiratory equ‘pment necessary for protection against the various hazards encountered about
the plant area. The Health Physics Division recommends the protection necessary in those situations
where the inhalation of radioactive materials is a possibility, while the Safety Department recom-
mends the proper type of protection for these operations not involving the use of radioactive mate-
rials, Protective equipment chosen for use at the laboratory has been selected in most cases to
furnish protection against both chemical contaminant and radioactive materials.

Situations in the field requiring respiratory protection are varied and depend on the particular
area involved. In all instances where air contamination is above masking tolerance levels, masking
or evacuation of personnel is mandatory. One disadvantage confronting the surveyor is the length
‘of time necessary in running and counting an air sample, using the precipitron, to determine the ac-
tivity level. It is dependent on the surveyor’s good judgment and knowledge of operations in his area,
therefore, as to whether to mask or evacuate personnel in the interim while the actual activity level
is being determined. Discretion is necessary also in the use of protective equipment where its use
might interfere with a man’s safety. For example, a job requiring but 1 or 2 min to complete ina
high-radiation zone where air contamination is a possibility might be better accomplished without
the use of the mask, if its use would affect the workers efficiency to the extent of requiring a longer
period of time in the high-radiation zone to accomplish the work. In like circumstances, if the use
of a mask interferes with a man’s movements, increasing the hazard of slipping or falling, its use
might not be justified,

Respirators and masks should be stored in a clean dry location. Sterilization after use is usually
done by the health physics surveyor, using 70 per cent ethyl alcohol. Contaminated masks are also
decontaminated by the surveyor, Laundry decontamination facilities are available for masks exten-
sively contaminated,

The number and types of masks and respirators should be held to the minimum necessary to
furnish protection needed 10 the various areas, These needs are best determined by the senior sur-
veyors in the respective areas, Army assault masks wil! furnish protection against most types of
dusts, fumes, smokes, and mists, Where oxygen lack is suspected, or when absolute protection is
needed, the Chemox, Scott Air Pack, or an airline respirator should be utilized. For chipping, grind-
ing, or abrading solid materials, one of the lightweight respirators, approved for dusts or fumes,
may be used.



[Document No. GRAUT-OTIB-0018 | Revision No. 01 Effective Date: 08/09/2005 | Page 28 of 35_|

REFERENCES

AEC (U.S. Atomic Energy Commission), 1968, Manual, Chapter 0524, “Standards for Radiation
Protection,” AEC 0524-01, November 8.

DOE (U.S. Department of Energy), 1988, “Radiation Protection for Occupational Workers,” Order
5480.11, Washington D.C., December 21.

ICRP (international Commission on Radiological Protection), 1959, Report of Committee Il on
Permissible Dose for Internal Radiation, Publication 2, Pergamon Press, London.

ICRP (International Commission on Radiological Protection), 1994a, Human Respiratory Tract Model
for Radiological Protection, Publication 66, Pergamon Press, London.

ICRP (International Commission on Radiological Protection), 1994b, Dose Coefficients for Intakes of
Radionuclides by Workers, Publication 68, Pergamon Press, London.

NBS (National Bureau of Standards), 1953, Maximum Permissible Amounts of Radioisotopes in the
Human Body and Maximum Permissible Concentrations in Air and Water, NBS Handbook 52
(also referred to as NCRP Publication 11), Washington, D.C.

NBS (National Bureau of Standards), 1959, Maximum Permissible Body Burdens and Maximum
Permissible Concentrations in Air and Water for Occupational Exposure, NBS Handbook 69
(also referred to as NCRP Publication 22), Washington, D.C.

Office of Compensation Analysis and Support, OCAS-TIB-009, Technical Information Bulletin:
Estimation of ingestion Intakes, Rev 0, April 13, 2004.

ORAU (Oak Ridge Associated Universities), 2004, Technical Information Bulletin — Maximum Internal
Dose Estimates for Certain DOE Complex Claims, ORAUT-OTIB-0002, Rev. 1, Oak Ridge,
Tennessee.



(Document No. ORAUT-OTIB-0034 | Revision No. 00 __| Effective Date: 12/13/2005 | Page 6 of 29 ]

1.0 PURPOSE

Technical Information Bulletins (TIBs) are general working documents that provide guidance
concerning the preparation of dose reconstructions at particular sites or categories of sites. They will
be revised in the event additional! relevant information is obtained. TIBs may be used to assist the
National Institute for Occupational Safety and Health in the completion of individual dose
reconstructions.

In this document, the word "facility" is used as a general term for an area, building, or group of
buildings that served a specific purpose at a site. It does not necessarily connote an "atomic weapons
employer facility” or a "Department of Energy facility” as defined in the Energy Employees
Occupational Illness Compensation Program Act of 2000 [42 U.S.C. § 73841(5) and (12)].

There are instances of energy employees who, for a variety of reasons, were not monitored for
internal exposure during the course of their employment at a U.S. Department of Energy (DOE)
facility, or whose records of such monitoring are incomplete or unavailable. In such cases, data from
coworkers may be used to approximate an individual's possible exposure. The purpose of this
document is to provide the details of the calculation and assignment of intakes based on coworker
data from Oak Ridge National Laboratory (ORNL) for the purpose of estimating unmonitored
exposures or where records of monitoring are incomplete or unavailable, whether for discrete periods
or for the entire period of employment.

2.0 OVERVIEW

Analysis of Coworker Bioassay Data for Interal Dose Assignment (ORAU 2004a) describes the
general process used for analyzing bioassay data for assigning doses to individuals based on
coworker results.

Bioassay results for ORNL were obtained from the Oak Ridge Institute for Science and Education
(ORISE) Center for Epidemiologic Research (CER) Dosimetry Database, which contains urinalysis
tecords from the ORNL site for the period 1951 to 1988. ORISE obtained this database from ORNL
for the purpose of conducting an epidemiology study of site workers. The database results are in
units of disintegrations per minute (dpm)/24 hours. Because of the varied operations at the different
ORNL facilities over time with the potential for exposure from numerous different radionuclides, the
database contains urinalysis data for numerous radionuclides. These data are stored using Electronic
Data Processing (EDP) codes, as documented in Section 5.1.5.2 of the X-10 Technical Basis
Document (TBD) (ORAU 2004b). In summary, data were stored under 64 different EDP codes and
included measurements for five radioisotopes of uranium, four radioisotopes of plutonium, seven other
transuranic radionuclides, numerous fission and activation products, and gross alpha and gross beta
measurements. The majority of the EDP codes contained fewer than 100 data entries. In reviewing
these data, it was determined that only a few of the EDP codes contained enough entries to allow
statistical evaluation for purposes of dose reconstruction. These EDP codes were: SR and SRO for
°°Sr (12,893 entries from 1951 to 1988), URO and UR for total uranium (11,434 entries from 1951 to
1988), PUO, PU9, PU, GAO, and GUO for plutonium alpha-emitters (15,476 entries from 1951 to
1988), and TPO and TP for transplutonium radionuclides, primarily ‘Am (5,670 entries from 1968 to
1987). Further review of these data indicated that, except for follow-up for accidental exposures, they
were collected on an annual sampling basis; therefore, the analysis that follows considers chronic
exposures to estimate annual intakes. A statistical analysis of these data was performed in
accordance with ORAU (2004a). The resultant values were input to the Integrated Modules for
Bioassay Analysis (MBA) Expert OCAS-Edition computer program, and a fit to the data for each of
the four radionuclides was performed to obtain intake rates for assigning dose distributions.



[Document No. ORAUT-O11IB-0016 | Revision No. 01 _] Effective Date: 08/09/2005 | Page 50f35_]

ACRONYMS AND ABBREVIATIONS


-—

Special Exposure Cohort Petition U.S. Department of Health and Human Services
under the Energy Employees Occupational Centers for Disease Control and Prevention
liness Compensation Act National (nstitute for Occupational Safety and Health

OMB Number: 0920-0639 Expires: 09/20/2013

Special Exposure Cohort Petition — Form B Page 3 of 7
Labor Organization Informati Complete Section D ONLY if you are a labor organization
D.1_— Labor Organization Information:

Narne of Organization

Position of Contact Person
D.2 Name of Petition Representative:

0.3 Address of Petition Representative:

Street Apt # P.O. Box
City State Zip Code
D.4 Telephone Number of Petition Representative: _) :

D.5 Email Address of Petition Representative:

D.6 Period during which labor organization represented employees covered by this petition
{please attach documentation): Start End

D.7 Identity of other labor organizations that may represent or have represented this class of
employees (if known):

Name or Social Security Number of First Petitioner: ,



{Document No. ORAUT-TKBS-0012-5 | Revision No. 01 _| Effective Date: 10/01/2007 | Page 2 of 74

PUBLICATION RECORD

EFFECTIVE
DATE

REVISION
NUMBER

DESCRIPTION

06/29/2004

00

New technical basis document for the Oak Ridge National Laboratory
— Occupational Internal Dose. First approved issue. Initiated by
Robert E. Burns, Jr.

05/30/2006

00 PC-1

Page change revision to update required language in the Introduction
on page 8 in Section 5.1. Adds Section 5.4 on pages 28 and 29 on
using air monitoring data to assign best estimates of internal dose for
unmonitored ORNL workers during the 1944-1947 period. Makes
reference changes on pages 30 and 32 in the Reference Section.
Adds Attachment B, on pages 67, 68, 69, and 70, which gives the
technical justification for this protocol. No changes occurred as a
result of internal formal review. Incorporates NIOSH formal review
comments. No sections were deleted. This revision results in no
change to the assigned dose and no PER is required. Training
required: As determined by the Task Manager. Initiated by Elizabeth
M. Brackett and Thomas R. La Bone.

Approval:

Signature on File 05/22/2006
John M. Byme, Document Owner

Signature on File 05/22/2006
John M. Bye, Task 3 Manager

Keith McCartney Signature on File for 05/22/2006
Edward F. Maher, Task 5 Manager

Signature on File 05/22/2006
Kate Kimpan, Project Director

Signature on File 05/30/2006
James W. Neton, Associate Director for Science

10/01/2007

01

Approved Revision 01 initiated to include Attributions and
Annotations, Purpose and Scope sections. Revised Section 5.4.
Made editorial changes, No further changes occurred as a result of
formal intemal review. incorporates formal NIOSH review
comments. Training required: As determined by the Task Manager.
Initiated by Robert E. Burns, Jr.



[Document No. ORAUT-TKBS-0012-5 | _Revision No. 01 __| Effective Date. 10/01/2007 | Page 22 0f74 |

facility was still being used to verify if intakes had taken piace rather than for routine monitoring of site
personnel. A new Health Physics Division Report (Morgan, Snyder, and Davis 1967) begun in 1966
included data on numbers of in vivo counts and basic statistics for in vivo monitoring. They included
numbers of individuals whose results exceeded the U.S. Atomic Energy Commission (AEC) reporting
level (50% of the permissible body burden averaged over the year) or some other specified amount.
Table 5-13 summarizes this information. NOTE: The use of annual averages could be misleading in
cases involving short-lived or rapidly cleared materials. Mani (1983, Table 15) provided similar data,
but he reported all in vivo counts that exceeded specified levels rather than annual averages. Table
5-14 lists this information.

Table 5-13. Qualitative information concerning radioactive
material detected utilizing the WBC facility from 1966 to 1983.
Number of persons exceeding the permissible body
Year burden (based upon in vivo measurements)
1966° | 0 persons exceeded 50% of permissible body burden
1967" | 0 persons exceeded 50% of permissible body burden
1968* |'0 persons exceeded 50% of permissible body burden
1969" | 0 persons exceeded 50% of permissible body burden
1970" | 0 persons exceeded 25% of permissible body burden
1971 | 0 persons exceeded 25% of permissible body burden
1972° | 0 persons exceeded 25% of permissible body burden
1973" | 0 persons exceeded 25% of permissible body burden
1974” | 2 (employees inhaled Cm-244 believed to range from 20
(15 pCi) to 40% (30 pCi) of organ (lung) burden)
1975” | 4 (1 - Cm-244 of ~50%, 1 — U-238 of <15%, 1—Zn-65
and Co-57 of <15%, and 1 ~ Co-60 of <15% of lung
burden)
1976 | 6 (all six appear to have had detectable amounts of Cm-
244 of <15% of the lung burden)
1977" | 20 (8-— Pu and “Am of <25%, 8 — “"Cs, "Co, Gd,
and Cd of <15%, and 4 - | and "Te <25% of the
jung burden)
1978" | 6 (1- Pu-241 and Am-241 of ~70% (lung), 1 = "1-131
minor amount that could not be quantified (thyroid), and
4—-Co-60 of <25% (lung)
1979" | 0 persons exceeded 25% of permissible tung burden
1980" | 0 persons exceeded 10% of permissible lung burden
1981" | 0 persons exceeded 10% of permissible lung burden
1982” | 0 persons exceeded 10% of permissible lung burden
1983” | 0 persons exceeded 10% of permissible lung burden
a. The AEC permissible level was 50% of the permissible body burden
from 1966 through 1969.
b. From 1970, ORNL indicated in reports that in vivo measurements did
not exceed specified amounts.



t
j

86 AECU-817

8.4 SHIPMENT OF RADIOACTIVE MATERIAL’
By T. H. J. Burnett

The material covered was substantially identical to that contained in an article by Dr. Karl Z.
Morgan on this subject, appearing in the Journal of Applied Physics, July 1948.1 Copies of reprints
of this article were distributed to the students.

In addition, the effects of bremsstrahlen were cited as means whereby the shipment of strong
beta emitters could be accompanied by asséciated X or gamma rays in the energy range such as to
have a k factor greater than 1 and possibly induce film damage. Bremsstrahlen was explained as a
term of German origin meaning literally ‘‘braking rays’ and constituting the radiation of photons
accompanying acceleration changes of the beta electrons as they are deflected by interaction with
matter,

An experimental evaluation of the effective k factor proper to use in shipping isotopes emitting
gamma rays of more than one energy was described, and the results of these tests were given. No
case was found where the k factor exceeded 2, this result being due to the effectiveness of the shield-
ing in cutting out the low energies when sufficient to handle the higher energies also present.

A brief discussion of causes for intensification of film readings behind the badge shield was
given, illustrating the difficulties inherent in interpreting the open-window reading in some cases.
The conclusion emphasized again the importance of the professional investigation made by the health
physics surveyor in all indicated cases of possible exposure,

8.5 SHIELDING
By R. S. Thackeray

The use of shielding in work involving radioactivity is required to limit personnel exposure to
the established tolerances. Shielding may be broken down, in general, into the following categories:

1. Permanent shielding

2. Portable or flexible shielding

3. Close shielding

Gamma-ray shielding is done with heavy masses of material which have a high density. The
radiation is never reduced to zero by such shielding, but it can be reduced to safe levels, Permanent
shielding against gamma rays usually takes the form of thick walls of concrete. Sometimes space
requirements dictate a special high-density concrete, but normally a large thickness of ordinary
material will suffice.

Portable shielding is achieved through the use of specially designed containers of heavy metal
such as our source pots.

Flexible shielding is erected around hot samples, using lead bricks. Gamma rays are, in gen-
eral, too penetrating to allow the use of close shielding.

There is seldom any need to make permanent or portable shields for alpha, beta, and protons.
Close shielding, that is, shielding built closely around the vessel containing the source, is most use-
ful with these radiations. In this way all the radiation can be cut out through the use of a suitable
thickness of plastic, because each of the radiations mentioned has a definite maximum range corre-
sponding to its energy. Close shielding permits one to handle beakers, test tubes, etc., ina more
or less normal fashion.

*A report on the physical, biological, and administrative problems associated with the trans-
portation of radioactive substances is under preparation by Robley D. Evans for the Subcommittee
on Shipment of Radioactive Substances, Committee on Nuclear Science, Division of Mathematical
and Physical Sciences, National Research Council. It is expected that this report will be issued in
1950. The report gives an excellent and comprehensive treatment of the materials indicated by the
title and will be of interest to all persons handling radioactive materials.

tSome changes and extensions of shipping regulations have been made since publication of the
article referred to above, and other modifications are in progress.



UNIT 2. FUNDAMENTAL PHYSICS

2.1 ATOMIC STRUCTURE
By Ross 8. Thackeray*

An atom is far smaller than anything which can be observed with direct means such as a micro-
scope. As the result, knowledge of atomic structure is limited to a series of models which seem to
account for the results of indirect experiments, Even though large numbers of models have been
Proposed and then discarded as the result of further experiments, we shall proceed as though we had
complete, definite knowledge.

‘The first concepts of atomic structure pictured an atom as a hard, solid sphere somewhat like
a ball bearing. This picture was used for a few years aod seemed to work reasonably well until
Rutherford performed his experiment (Fig. 2.1) with alpha particles at the Cavendish laboratory in
Cambridge, Engiand. This experiment consisted in the bombardment of a thin gold foil with alpha
particles. The foil was so thin that most of the alpha particles passed through. However, it was
found that in penetrating the foil these particles were directed from their original path by a process
which came to be known as “scattering” (the assumed picture of scattering behavior from per cent
distribution is shown in Fig. 2.2). The scattered alpha rays were detected with scintillation screens,
and the angular distribution of the scattered rays was determined. From these data Rutherford drew
the conclusion that the foil was not made up of closely packed solid spheres, but instead was made
up of spheres with small positively charged dense centers surrounded by a large volume of virtually
empty space.

Some idea was known of the distance between centers of atoms in thin gold film, since X-ray
experiments had already given that information. The size of the positive nucleus, demonstrated by
the alpha-ray experiments, was very much smaller than the distance between the atoms (Fig. 2.2).

More information about the extranuclear structure of the atom was obtained by such experiments
as foll bombardments, using high-speed electrons, It was found that an incident electron could eject
one or more other electrons from the target material (see Fig, 2.3). Knowing that an atom as a
whole had to be electrically neutral, Rutherford then proposed that the nucleus was surrounded by a
field of electrons which were very much Hghier than the nucleus and carried a negative charge.

Prior to the discovery of the neutron the fundamental atomic particles known were only two in
number, To fit the observed atomic mass and charge data, it was proposed that the nucleus contained
heavy, positively charged particles called ‘‘protons’’ and that there were approximately twice as
Tmany protons in the nucleus as there were electrons in the external orbits (see Fig. 2.4). Since the
atom is electrically neutral, it was also assumed that there were sufficient electrons contained in
the nucleus to make the sum of the negative charges in the atom equal the total positive charge.

Experiments carrted on at the Curie Institute in Paris and in various European universities led
to the discovery of a particle with no electric charge and with mass equal to that of the proton, This
particle, now called the ‘‘neutron,’”’ removed the necessity for assuming that electrons exist in the
nucleus. It is now known that the nucleus contains a number of protons equal to the number of orbital
electrons, and the rest of the nuclear mass consists of neutrons. Since all atoms are assumed to be
made up of various combinations of these three fundamental particles, we now have available a sys-
tematic method for their classification. If we assume a unit of mass equal to the weight of the proton

*Notes taken by C. Scott, edited by R. S. Thackeray.

AECU-817



(Document No. ORAUT-TKBS-0012-5 | Revision No. 01__| Effective Date: 10/01/2007 | Page 240f74 |

Table 5-15 summarizes the maximum activity measured through in vivo monitoring for various
nuclides for the period from 1961 to 1966.

Table 5-15. Maximum measured in vivo activity (nCi) from 1961 to

1966.

Isotope 1961 1962 | 1963 | 1964 | 1965 1966
Na-24 140
Sc-46 4
Cr-51 320 42_|TraceTM_
Co-56 15
Co-57 65
Co-58 20 38_| Trace
Fe-59 40 15
Co-60 80 | 13 2 220 5 12
Co-64 <5
Zn-65 40 3 16_| Trace
Se-75 250
Sr-90 40 | 4,500 800 626
Zr-95/Nb-95 30_| 14 14 47 18 39
Ru-106/Rh-106 131 47 200 30
Sb-125 162 6 18
1-134 12_ {200 28° 73 54 6
Cs-137, 440_ | 360 570 310 104 92
Ce-144/Pr-144 30 <7 75 50
Eu-155, Trace
Au-198
Hg-203 800 0.0005 169 22
Ra-226 <6 <0.5°
Pa-233 2,800
U (enriched) Trace Trace

a. Trace is not defined within any of the annual reports, but is provided here to
give a qualitative “feel” for the amount present.

b. Thyroid.

¢. micrograms.

The results for a given in vivo analysis contained the following information: name, badge number (or
equivalent), division code, HP area, date, analysis code, and results code. Table 5-8 lists result
codes for the period prior to 1978. The reason for the analysis was given. The ORNL Division codes
are the same as those listed in Tables 5-4 and 5-5. [NOTE: The ICRP Publication 2 methodology of
reporting percentages of body or lung burdens (ICRP 1959) was used at ORNL through 1983. AEC
Manual Chapter 0502 required “an evaluation of the radiation exposure status of an employee when
monitoring techniques indicated that a body burden equaled or exceeded 50% of a maximum
permissible limit.” A phone discussion with the current internal dosimetry staff (McLaughlin 2004)
indicated that the staff that was in place in the 1960s through 1980s probably conducted dose
assessments when necessary and that documentation would exist for these assessments. Since
1989, each in vivo measurement showing a positive result has been assessed, including making
estimates of intake and dose. The ORNL internal Dosimetry Program TBD (McLaughlin 2002) states
that dose estimates of less than 1 mrem are reported as zero. The overwhelming number of in vivo
measurements obtained at the facility indicate no elevated activity (McLaughlin 2004).

Mani (1983) indicated that a Na!-CsI phoswich detector was installed and operational in 1967, though
Brown, Patterson, and Abee (1971) state, ‘The several detectors used for measuring gamma
radiation from the human body are all Nai (Tl) crystals.” The 1974 annual report of the Health Physics



66 AECU-617

5. Needle goes below zero
Cause
a, Tube burned out
6, Indicates reading when no radiation is present
Causes
a, Contamination of instrument
b. Light getting into chamber
c. Mechanical balance of meter
7. Reading is less than known radiation
Causes
a. Instrument grounded at wrong place
b, Chamber and box giving signals of opposite polarities
c. Needle on meter sticks
8, Change of zero set in changing position of instrument from horizontal to vertical
Cause
a. Mechanical balance of meter
9, Needle drifting
Cause
a, Batteries getting low
Walkie Poppy. The following is a list of some possible malfunctions of the Walkie Poppy with
probable causes:
1. Majority of troubles are due to batteries
2. High background noise
Causes
a. Leakage in high voltage
b, Bad bias cells of amplified section supply
3. No noise or pops
Causes
a. Tube burned out in amplifier section
b. High-voltage failure
4, Intermittent operation
Cause
a. Bad cable connection
NOTE: The filament batteries must be changed quite frequently. High voltage lags behind the
adjustment. Operating point is very critical.
Victoreen 263. The following is a list of some possible malfunctions of the Victoreen 263 with
probable causes: .
1, Voltage is up but no signal or pops
Cause
a, Bad G-M tube
2, No background at all
Cause
a. Burned-out tube
3. High background outside radiation field
Causes
a. Low batteries
b, Bad tube
c, Contamination on instrument
4, Intermittent operations
Cause
a. Bad cable connections



mee

[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 __| Effective Date: 10/01/2007 | Page 100f74 |

Table 5-1. Radioanalytical results between 1947 and 1988 [2].

Nuclide Sample size Nuclide Sample size Nuclide {| Sample size
Am-241 5,670 1-131 41 S-35 10
Am-243 12 Mn-54 2 Sb-125 4
As-74 7 Mo-99 1 Sm-151 11
BG 2 Na-24 3 Sr-85 41
Bk-249 14 Nb-95 3 Sr-89 37
Br-82 2 Np-237, 55 Sr-90 12,893
Br-83 2 P-32 166 Tc-99 20
C-14 11 Pa-231 55 Th-232 1,125
Ca-45 4 Pa-233 16 TI-201 1
Ce-144 37 Pa-234 1 Tl-204 1
Cf-249 3 Pb-210 2 Tm-170 6
Cf-252 14 Pm-147 80 U-232 4
Cl-36 1 Po-210 66 U-233 829
Cm-242 12 Pu-238 65 U-235 3
Cm-244 299 Pu-239 15,476 U-238 11,434
Co-60 83 Pu-241 112 U-239 11
Cs-134 1 Pu-242 41 Y-88 5
Cs-137 3,561 Ra-226 333 Y-90 31
Fe-59 9 Ra-228 4 2n-65 7
Gross alpha 4,875 Rare earths 1,098 2r-95 20
Gross beta 324 Ru-103 1
H-3 2,070 Ru-106 65

'sotope-specific analyses for in vitro samples did not become routine until 1989. Prior to that time,
chemical methods were used to separate radioelements as well as practicable, and the materials
were assayed in terms of total activity. The activity measured would later be assigned to a
predominant nuclide. Thus, a result from the early years might indicate Sr, when in reality it includes
“Sr. The same is true for early plutonium results and results for transuranic materials. Thus,
“associated” radionuclides are inherently included in such results. Process knowledge of
radionuclides present in various work areas was used to assign nuclides to sample results. The
present ORNL internal dosimetry program uses a limited number of radionuclides for screening
purposes. Positive results are followed up with additional bioassays.

5.1.4 Solubility Classes

Internal dosimetrists at ORNL provided a list of assumed intake modes and clearance class
information for 73 radionuclides. These assumptions, given in Table 5-2, are those used by ORNL
when re-evaluating historical bioassay results. They are provided as a reference for dose
reconstructors, with the caveat that the assumed solubility classes are not known to be based on any
specific studies. In general, they merely reflect conservative choices for dose assessment in terms of
committed effective dose equivalent. Dose reconstructors therefore should not assume that these
classes, which are in terms of the system promulgated in (ICRP 1966), represent favorable to
claimant choices for tissue-specific dose evaluations. Dose reconstructors should instead make a
favorable to claimant choice for the radionuclide and tissue(s) of interest using the system of solubility
classes described in ICRP Publication 66 (ICRP 1994) [3].



AECU-817 61

5.8 CALIBRATION TECHNIQUE
By D..M. Davis*

Sources Used in Calibration

Gamma sources Alpha sources
0.96 mg Ra 4,000 dis/min
23 mg Ra 18,000 dis/min
27 mg Ra 20,000 dis/min
48 mg Ra 42,000 dis/min
100 mg Ra 82,000 dis/min
1 curie Co 189,000 dis/min

200,000 dis/min
300,000 dis/min

The 0.96-mg Ra source is used in the calibration of monitrons, The 27-mg Ra source is used
in the calibration of electroscopes and other fiber inatruments of that nature. The 23- and 48-mg
Ra sources are used in the calibration of film. The Co source is used in the calibration of the Cutie
Pie, Zeus, and other ionization-type instruments. The 100-mg Ra source is often used in checking
our calibration setup, The alpha sources are used for calibration of instruments which are designed
for measuring alpha.

Calibration Cycle

Electroscopes, Lauritsen and L & W, are calibrated at 4-week intervals. Ionization-type instru-
ments are calibrated at 2~week intervals, For various reasons the calibration of the instruments
may vary from time to time.

With a calibration cycle as listed above, the instruments should not vary more than 10 per cent
from the calibration chart. Fiber instruments vary less from the calibration than do ionization-type
instruments,

Calibration Theory

The calibration of an instrument is to set up a standard to be used for measuring radiation in-
tensities in an unknown field. By use of an instrument that is properly calibrated, we are simply
saying that the radiation measured in the field is the same as that measured at a known distance
from a known source. In other words, if by using a 50-mg Ra source we determine that at a given dis-
tance we are getting 100 mr/hr and the meter reads 10 amp at this position, then in the field when
the meter reads 10 pamp we can assume that we are in a field of 100 mr/hr,

Beta Calibration

We are often confronted with the question, ‘‘How do you calibrate for beta?’’ We do not at the
present time calibrate for beta, There are many problems involved jn the calibration of beta, such
as various energy ranges, etc., which make the problem much more complicated than gamma calt-
bration, It is recognized that beta calibration is desirable, and at the present time people in the
Research Section are working on a method for the calibration of beta, We do have, on most of the
instruments, what we call a ‘‘window.’’ This window, however, does not give us an accurate meas-
urement of beta in mrep, but only the amount of ionization, The efficiency of the instrument for beta
radiation varies with the energy of the beta radiation.

One statement might be made as to the effect of contamination of instruments. Contamination
on an instrument wil) always result in an erroneous reading. The fiber instruments will show a

*Notes taken by C. Scott; edited by D. M. Davis.



Special Exposure Cohort Petition U.S. Department of Health and Human Services

under the Energy Employees Occupational Centers for Disease Control and Prevention
iliness Compensation Act Nationat Institute for Occupationa! Safety and Health

OMB Number, 0920-0639 Expires: 09/20/2013

Special Exposure Cohort Petition — Form B dix — Continuation Page

Continuation

~ Photocopy and comp

necessary

calibrating their detection instrumentation, states "We do not at the present time

calibrate for beta. There are many problems involved in the calibration of beta,
such as various eneray ranges, etc., which make the problem much more
complicated than gamma calibration. It is recognized that beta calibration is
desirable, and at the present time people in the Research Section are working on a
method for calibration of beta." This proves that accurate measurements of the
beta emitting radionuclides was impossible in 1949, and any measurements of beta
must be considered inaccurate. Page 69 begins the lecture entitled "PERSONNEL
MONITORING PROCEDURES" and states “Effective personnel monitoring involves
measurements of both external and internal radiation, External radiation is
monitored at present by pocket meters (ionization chambers) and film meters.
Internal radiation presents two problems in that we are faced with both radiation
and toxic effects. Internal radiation is measured by a study of body fluids. (this
method is discussed in Lecture 6.3)."_On page 74 we find lecture 6.3 entitled
"MEASUREMENTS ON INTERNAL EXPOSURE.” This particular lecture is very
short and consists of only five paragraphs. Paragraph five states "Since the
organization of a group for body fluids analysis at Clinton Laboratories in July 1945,
the main concern has been the development of procedures of chemical isolation

of radioactive elements, followed by the perfecting of these procedures to a point
where they were usable. The development of instruments which are reliable has

been one of the major problems. The quantity of any radioactive material to be
found in either urine or feces is exceedingly small. Therefore the counting
instruments must be dependable over a long period of time,” Although itis not
specifically stated, it is clear that ORNL had no reliable means of analyzing
biological samples in 1949. The lack of biological data prohibited ORNL from

accurately assessing exposures, and this is a prevailing message throughout
this lecture series.

This leads us back to determine the accuracy of ORAUT-OTIB-0018, Rev. 1, which
allows NIOSH dose reconstructors to deviate from the stated applicible dates of
1953 the present for this technical information bulletin, and apply this TIB to ORNL
from 1944-1952 for both alpha and beta emitting radionuclides. This TIB is used to
assign internal dose when air sampling was the only method for determining dose.
\t basically assumes that since ORNL had a program in place to limit exposures
by means of a rigorous air sampling scheme, internal doses can be reconstructed
from these limiting concentrations. Although ORNL did perform air sampling, and
did report data in regard to limiting exposures, AECU-817 does reflect that ORNL
was still struggling with instrumentation, calibration, and lack of any biological

Name or Social Security Number of First Petitioner: _

Continuation pg. 3


Page 16 of 74 |

10/01/2007 |

| Effective Date:

Revision No. 01

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[Document No. ORAUT-TKBS-0012-5|

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AECU-817 31

tx 1.64% 1 = 1.15 109 =

-9W 0.7, 1.64 x 10°8W
OE Tt * £
‘Under the condition of equilibrium assumed at the beginning of this section, this is also the permis-
sible rate of accretion of the radioisotope in the tissue.

To determine the permissible concentration of the radioisotope in the atmosphere, we must know
the volume of air breathed per day and the fraction of the inhaled radioisotope which finds its way to
and is retained by the critical tissue.

Let F equal the fraction of radioisotope taken into the body (in this case, into the lungs) that
reaches and is retained by the tissue (subject, of course, to reduction at the rate of {C per day), V
equal the volume of contamination air breathed per day, and Q equal the curies of radioisotope per
unit volume of air. Then the rate at which the radioisotope is increasing in the critical tissue, owing
to inhalation, is

Sa sFQv

Since this is equal to tie rate of decrease
2 -9 W
FQV = 1,15 x 10 TE

or, the permissible concentration of a beta-gamma-—emitting radioisotope in the atmosphere is given by

. -9 _W curies
Q=115x10" Fyre iit volame

(The unit of volume in which the concentration is expressed will be the same as that used to meas-
ure V.)

AS a numerical example, let us compute the permissible concentration of P33 in air, breathed
8 hr/day, 7 days per week, For this isotope the critical tissue is bone, and the following values for
the constants involved in the equation are used by K. Z, Morgan:

W = 12,000 ¢

F = 0.09

V = 4.3.x 108 cm8/8-hr day
T = i2 days

E = 0,5 Mev

Substituting these values into the equation above,

1.15 x 1079 x 1.2 x 104 3
Q= Top asx 108 x 19x 0,5 Tes/om

= 54x 107)? curie/em

If the air is to be breathed 24 hr/day, obviously the permissible concentration will be only one-third
as much, # one wishes to assume a 40-hr week, he may multiply the above result by 7/5.

It is, of course, assumed in the above example that the tissue in question is exposed to negligible
radiation from other sources. The assumption of uniform distribution in the tissue should be care-
fully examined in all cases.



[Document No. ORAUT-OTIB-0018

[Revision No. 01

[Effective Date: 08/09/2005 | Page 17 0f 35_|

Table 7-1 (Continued). Whole-body count (WBC) results below which this
TIB method is still considered an overestimate.

Absorption Time since Organ of WEC resukmust
Radionuclide type last WBC (d)* concem be less than (nCi):

Co-60 Mors i) ‘Adrenals 10
Tnary bladder’ 2t
Brain 25
Breast 5
Gall bladder 23
Heartw alt 4
Kidneys is)
Liver 36
Muscle 13
‘Ovaries 14
Pancreas 11
Testes 8
yroid by
Red bone marrow as
Bone surface 408
‘Stomach it
Si 13
ULI 15
UT 25
‘Skin uci
Spleen J
Thymus: 6
Uerus 20
ET 4
Tung Tt
Colon 20
Esophagus 6
Co-60 MorS 180 Adrenals 412
Urinary bladder 27
Brain 31
Breast 8
Gall bladder 30
Heart wall 7
Kidneys 9
Liver 45
Muscle 7
Ovaries 18
“Pancreas 4
Testes 35
‘Thyroid 19
Red bone marrow 51
Bone surface 138
Stomach 14
SI kid
ULT TS
TET 3T
‘Skin. 29
Spleen 14
Thymus 9
Uterus. 25
ET [J
Tung 1S
TON, 25
Esophagus. 9



AECU-817 13

Filing Board
4th Digit
0}1(2(3 1415 [6 17/8 /9
O]W!
1
» (2 x
B13
a4 z
3 6
6
7 iY
8
e]

4321
A. For four (4) digit Nos. (10]0{0]
B. Place films in groups of 100
C. W = location for film in 0 to 100 range
X = location for film in 3200 to 3299 range
Y = location for film in 5700 to 5799 range
Z = location for film in 8400 to 8499 range

Appendix A: Plastite-film Ring Meter

A plastic-film ring meter has been developed at the Oak Ridge National Laboratory in collabo-
ration with certain commercial plastic manufacturers. It is intended to be used as a beta and/or
gamma monitoring device for measurement of radiation received by specific parts of the body such
as hands,

This plastic-film ring meter is made of a special highly opaque plastic. It is thus highly adapt-
able for all use with film, resisting even out-of-door exposure to strong sunlight for reasonable
lengths of time. This meter may be used either as a ring for the finger, pinned to various parts of
the clothing near the radiation sources to be monitored, or used in any general manner where local
radiation exposure is to be checked. Because of its small size, convenience, and compactness, one
is able to obtain a more accurate radiation-field distribution of intensity readings.

The film ring meter utilizes a cadmium filter above and below the film. This cadmium filter
consists of a 9/16-in, washer with a 3/16-in. hole, the metal being 1 mm thick, The hole in the
cadmium washer acts as an open window for determining the entry of soft radiation onto the film
and radiation quality comparative measurements.

Film disks 37/64 in, in diameter are cut out with a flat punch of our own design and inserted
between the cadmium washers, The film used is tue more sensitive film of the Du Pont type 552
pack, After the meter is loaded with film, the plastic cap which is 11/16 in, in diameter and has a
thin top 0.02 in. thick is then sealed in place, using a solvent.

The ring composition is molded ‘‘tenite’’ with adequate pigmentation and flow specifications.
The solvent which we have found most satisfactory consists of a mixtur . of 40 per cent ‘‘cellosolve””
and 60 per cent acetone, The 0,02-in. plastic cap appears to have a stopping power equivalent to
approximately 25 mg/ ‘cm? of aluminum by experimental tests and thus should pass all beta radiation
above 0.34 Mev.

Rings are unloaded after use by prying off the cover, The ring itself is disposable. The cadmium
filter may be reused,

Adaptability features of the ring include the possibility of resizing by means of heating the band
which is open at the bottom and reforming while warm to the appropriate size desired. Likewise,
this band may be removed from the body of the ring, and the ring may be used as a small button,



Document No. ORAUT-OTIB-0018 Revision No. 01 | Effective Date 08/09/2005 |_Page 310f 35_|

ATTACHMENT B
VALIDATION OF THE BASIC ASSUMPTION IN THIS 1B

The basis for this TIB is that weapons sites with a radiological control program controlled exposures to
airborne radioactive contaminates to less than an applicable limit. The limit varied with time and
isotope of concem. This TIB acknowedges that exposures fo concentrations greater than these limits
did occur. However, the assumption is that, when integrated over time, the average exposure to any
individual was less than that which would occur if someone were exposed to the limit continuously.
The assumption is tested in the following analysis. Note that this analysis does not change the
assumption but merely tests the validity of it.

B.1 URANIUM

Uranium urinalysis data from the Y-12 site was used to test the validity of the assumption. The Y-12
site processed uranium extensively. The data consisted of hundreds of urine samples each month
from 1952 through 1985. Because this TIB provides mean vatues that are to be used ina lognormal
distribution, the median result was used in the comparison. Figure B-1 plots these median sample
results as well as the uranium result predicted by this TIB for an inhalation of type M uranium.

g
3
S

median results

&

~meee Long-term OTIB-18 result

dpmiday urine
we

—— 2-year OTIB-18 result

a
Vi82 6/29/57 12/2062 GIGS «= 12/2/73 «S257. 1414/84 B/TI90
Date
Figure B-1. Y-12 versus OTIB-18 uranium (type M).

The TIB result is considerably higher than the actual Y-12 median urine sample result. The 2-year
result begins 1 month after the start of the chronic intake and is approximately 10 times higher than
the highest actual median result. This verifies that for Y-12 uranium intakes, either the assumed

intake is an overestimate or the tung absorption of the uranium is considerably slower than type M.

Figure B-2 shows the same information using an absorption type S assumption for the TIB values.
This graph indicates that for a long-term exposure, it can be shown that this TIB overestimates actual
results even if absorption type S is assumed. The shortterm exposure, on the other hand, indicates
that it could take as long as 3 years to reach the level of the highest median urinalysis result. This
implies that this TIB approach might not be an overestimate for type S exposures of duration that are

less than 3 years.
B.2 PLUTONIUM

The Rocky Flats Plant worked extensively with plutonium. Plutonium urine sampkes from Rocky Flats
were evaluated from 1953 through 1969, For every year, the median sample was recorded as zero.



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01__| Effective Date: 10/01/2007 | Page 660f74_|

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 32 of 36

Number of Sample Results Recovered per Year
Between 1945 and 1990 for Am-241 in Feces

25

T

Number of Samples (N)

1945,

Figure A-15. MDA data for Am-241 in feces (part 3 of 3).

1E+H
‘Sample MDA.
MDA of 0.6 (959%CL on Mean)

Ze E -
4
2
a
é
=
&
g
El

1E2 4

1945 195019551960 «196519701975 1980 19851990
Year
Visual Comparison of Data to Estimated Gross Alpha MDAs in Feces

Figure A-16. MDA data for gross alpha in feces (part 1 of 3).



[Document No. ORAUT-TKBS-0012-8 | Revision No. 01__| Effective Date. 10/01/2007 L_ Page 73 of 74 |

ATTACHMENT B
ASSESSMENT OF INTERNAL DOSE AT ORNL: 1944 TO 1947
Page 3 of 4

100 3
z ]
[5 q
8 3 /
4 °
=
=
o
a _—_-______
o
£
=
i)
=
= J
a

0.01 T T T
-1 0 1 2 3
Z Score

Figure B-3. Piutonium urine sample data for April 25-May 30, 1945. The
geometric mean = 0.54 pCi/day, the GSD = 6.0, and 7 = 0.80.

the study showed little detectable activity. Two separate lognormal fits were performed on these data,
before and after the implementation of stricter controls. Figures B-2 and B-3 show the outcomes.

The geometric mean before April 25, 1945, is 3.1 pCi/day, while after this date it drops to

0.54 pCi/day. The arithmetic average of the results from April 25 through the end of the study is about
25% of the average during the previous weeks, and the average from the final four weeks is about 3%
of the earlier weeks. These comparisons indicate that the concerns about the collection system
before April 25 may have been justified.

In addition to the urine samples, 1483 alpha air-monitoring results were available for 1944 through
1947 at ORNL. The two largest values, both many orders of magnitude larger than the tolerance level
in use at the time, were excluded from the analysis because one appeared to be a miscalculation
(based on the recorded alpha activity) and the other was collected inside of a hot cell. Results were fit
to a lognormal distribution (Figure B-4). The geometric mean is 5% of the tolerance level applied in
ORAUT (2005a), and the 95th percentile is equal to the toferance level. Notes indicating that assault
masks had been worn were present for just under 10% of the samples; many of the samples were
collected inside of hoods.

Given this information, ORAUT (2005a) appears to be a reasonable upper bound because the 95th
percentile of the air-monitoring results coincides with the value applied in that document. Assault
masks were used in many situations, and many of the results exceed the activity to which the
individual would have been exposed because samples were collected inside of hoods. There is
conservatism built into ORAUT (2005a) by the inclusion of alpha emitters with large dose coefficients
that were likely not present or present only in very small quantities.

|


[Document No. ORAUT-ONIB-0018

[| Revision No. 01

[Effective Date: 08/09/2005 [ Page 16 of 35_|

Table 7-1 (Continued). Whole-body count (WBC) results below which this
TIB method is still considered an overestimate.

‘Absorption Time since Organ of WEC resukmust
Radionuclide type last WBC (d)* concem be tess than (nCi):

Cs-T37 F 365 Adrenals 24
Urinary bladder 2
Brain 19
Breast 19
Gallbladder 16
Heartwall 22
Kidneys 2
Tver 106
Muscle 21
‘Ovaries 22
Pancreas 22
Testes 20
Thyroid 21
Red bone marrow 80
Bone surface 150
Stomach 27
Si 22
ULI 36
tu 85
‘Skin 18
Spleen 2T
‘Thymus. 21
Uterus 22
ET 48
Lung 440
Colon 60
Esophagus 2T
Cs-137 F 730 Adrenals 2
Urinary bladder 2
Brain 2
Breast 2
Gall bladder 4
Feartw all 2
Kidneys 2
Liver 10
Muscle 2
Ovaries 2
Pancreas 2
‘Testes 2
ThyroKd 2
Red bone marrow 8
Bone surface 15
Stomach 2
St 2
ULT a
Tr g
Skin 2
‘Spleen 2
Thymus 2
Uterus 2
ET EJ
Tung 43
Colon 6
2



[Effective Date. 10/01/2007 | Page 520f74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 18 of 36

Revision No. 01

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[Document No. ORAUT-TKBS-0012-5 |

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52 AECU-817

Probability (= -2 ye <M<me2 ve) = 0,954, etc.

Obviously if N is large, the per cent of error

in 68.3 per cent of the cases and

in 95 per cent of such cases, ete.
For example, if the count is taken for 100 sec, with a total count of 22,500 counts, then

and we can assert that m differs from 225 by only

m
2 Vioo

with 95 per cent confidence. Taking M as the approximate value of m, we get an error of only

225
2 Too 73

and can feel reasonably sure that m lies between 222 and 228 ct/sec, i.e., m will not lie between these
limits in only about 5 per cent of the cases where such tests are made. We can use instead

im
3 wets

and assert that m lies between 220,5 and 229.5, with some assurance that this statement will be wrong
in only 0.3 per cent of such cases,

The probability for various other intervals about the mean can be tmmediately derived from ta-
bles available in most texts on statistics, What degree of accuracy should one demand? Should one
demand 0.3 per cent confidence or use the probable error and be satisfied to be right as often as
wrong? Requirements will obviously vary with the risk incurred by being in error or by the diffi-
culty encountered in taking 2 large value for N. Moreover, other factors are usually involved. There
is obviously no point in prolonging the sampling procedure to get an answer which appears to be more
exact than the limits of accuracy of the counter or of lower order than the decay of the source during
the procedure. It would be quite hazardous for a mathematician not familiar with the type of counter
to be used or with the field situation where the count is taken to attempt to set down any precise rules
for practical work,

A much more detailed study of counting techniques is given by Jarrett in Statistical Methods
Used in the Measurement of Radioactivity, MonP-126, For further reading on the fundamentals of
the statistics involved, I recommend Hurd, Statistical Methods (Lecture Notes); Hoel, Introduction
to Mathematical Statistics, Wiley, 1947; and Rider, Statistical Methods, Wiley, 1939; ail of which are
in the Health Physics Library or in the Training School Library.



[Document No. ORAUT-TKBS-0012-5 [Revision No. 01 _| Effective Date: 10/01/2007 | Page 48 of 74_|

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 14 of 36

Number of Sample Results Recovered per Year
Between 1945 and 1990 for Cs-137 in Urine

3

‘Number of Samples (N)
&

Figure A-3. MDA data for Cs-137 in urine (part 3 of 3).

18+]

(E+

VE1

Activity (dpm/24 h sample)

1E2
1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Visual Comparison of Data to Estimated Gross Alpha MDAs in Urine

Figure A-4, MDA data for gross alpha in urine (part 1 of 3).



(Document No. ORAUT-TKBS-0012-5 |

Revision No. 01

| Effective Date: 10/01/2007 | Page 29 of 74

Germanium array and phoswich

Monitoring type Applicable period Frequency
Lung counting |1976-1994 Monitoring frequency was not effectively implemented.
(cont'd.) §.25- x 1/16-in. Nal - 2- Cst This phoswich detector was used to monitor for low energy
iphoswich photons.
1980-1987 Monitoring frequency was not effectively implemented.

Detectors used together.

1992-present
combination germanium and Nal

Fixed Bed Counter is SBC in fixed position. Used only to
measure fission and activation products (e.g., °’°Co and
4Ce) immediately after assimilation.

1994—present
Germanium system

Varies from annual to semiannual depending on exposure
potential set by Internal Dosimetrist.

1994~present
§.25- x 1/16-in. Nal - 2- Cs!

Phoswich detector can be used, but germanium detection
system noted above is routinely used.

phoswich
1960-1961 INo set frequency. (Period of facility development —limited
|4- x 4-in. Nal (TI) routine human counting, typically incidents)
1961-1962 INo set frequency (Development activities to change to
Whole-body 8- x 4-in. Nal (Tl) larger detector — typically incidents)
counting 1963-1976 Table 5-12 Selection criteria used in 1965 for WBC
8- x 4-in. Nal (Tl)

1992—present
combination germanium and Nal

SBC. Monitoring frequency implemented in 1994 to
include annual and semiannual counts depending on
exposure potential set by Internal Dosimetrist.

1963-2? Incident counting and research. May have been used for
3- x 3-in. Nal (TI) frequent monitoring of workers potentially exposed to
Thyroid counter iodine, but this information was not located.
(and other 1994-present Typically, site research personne! would be placed on this
organs) germanium (uses one of three program and monitored on semiannual frequency, as well
germanium detectors used for as persons involved in known or potential iodine intake.
lung counting)
5.4 USING AIR MONITORING DATA TO ASSIGN BEST ESTIMATES OF DOSE

Information regarding internal dose at Oak Ridge National Laboratory (ORNL) from 1943 through

1950 is somewhat limited, with no bioassay results appearing in the DOE claimant files. However,
there are fairly extensive air-monitoring records in the 1944 through 1947 time frame that are suitable
for assigning best estimates of dose. The general protocol presented in OTIB-0018, “internal Dose
Overestimates for Facilities with Air Sampling Programs’ is used with the following modifications:

1) The approach applies only to best estimates of doses received in the 1/1/1944 to 12/31/1950
time period at ORNL.

2) For the alpha-emitter category, the radionuclides are limited to plutonium and uranium.

3) Acconstant distribution is used in IREP for the doses rather than a lognormal distribution.
The technical justification for this approach for 1944 through 1947 is given in Attachment B.
Air monitoring data are not available for 1948 through 1950. A fit to the coworker data for 1951 and
1952 indicate an intake rate of 200 pCi/day of Type S **°Pu for 1948 through 1950 [6] whereas
application of the OT!IB-0018 protocol to the same period gives an intake rate of about 198 pCi/day”

? See Table 4-1 in OTIB-0018 (ORAUT 2005a).



AECU-817 13

obstruction which it might strike. All the effort that went into moving this heavy load would disappear
in the raising of the body to a greater height, However, if the body were allowed to descend under the
force of gravity in a frictionless slide down the hill, it would acquire a great velocity and would do
work against any obstruction in its path.

Table 2.1

Name of Electrical
particle charge, esu Mass, g Mass, amu
Proton 44.8 x 19710 1.67 x 10-24 1
Neutron 0 1.67 x 10°24 1

- “10 -28 —
Electron 4.8 x 10 9.1 x 10 Te40

10 -28 a.

Positron +4.8 x 10 9.1 x 10 7840
Neutrino 0 ~0 ~0

A similar situation prevails in the orbital structure of the atom. Hf an electron in an inner shell
is lifted to an outer shell by energy supplied by a photon or another electron, that energy disappears
in the work of moving the electron against the attraction of the nucleus. If, however, the electron
goes back to its original position, the energy used to raise it to an outer orbit becomes available and
is radiated as an electromagnetic photon of energy (see Fig. 2.10 for an interaction diagram for pho-
tons).

Other concepts which were reviewed were the definition of atomic mass, atomic number, and
isotopes.

There was a brief discussion of nuclear reactions at the end of this period. Various methods of
obtaining high-speed subatomic particles were introduced, but the student is referred to reference
texts and literature for the mechanisms and operating characteristics of these methods.

‘The nucleus is held together by amazingly great forces, and it requires very energetic bom-
bardment to disrupt the nucleus. The tonization potential for the various electrons of a normal atom
requires energy on the order of 10 to 15 ev. To cause the same sort of disturbance in the nucleus,
it would require energy on the order of several million electron volts.

There are a variety of common nuclear reactions, but the most frequently used in work with
piles are those in which neutrons are absorbed in an atomic nucleus and some secondary particles
are emitted. One can characterize such exchanges as n,a; n,2n; n,y;and n,p. This symbolism means
that a neutron is absorbed by the target nucleus and that alpha particles, neutrons, gamma rays, or
protons are ejected. In any case in which the ejected particle carries a charge, it changes the atom
from one element to another. This process is called ‘‘transmutation.’’

Another concept which is frequently met with in our work is that of cross section, A cross sec-
tion can be given for a single process or a combination of processes, and it signifies the relative
probability of the occurrence of the process in question. Thus if we write sigma (c) as the symbol
for the cross section, a high value of Og indicates that the material with this cross section will
scatter neutrons from their orbital paths very efficiently. Similarly, a low value for a, indicates
that few of the neutrons entering a thin slab of the appropriate material will be absorbed in the ma-
terial and that most of them will escape. 0, corresponds to the combined probability of scattering
and absorption and indicates the total loss of neutrons from a beam due to the insertion of an ele-
mentary thickness of material.



aed as

RT ia

[Document No. ORAUT-OTIB-0018

[ Revision No. 01

[Effective Date: 08/09/2005 [Page 27 0f 35_|

Table 7-2 (Continued). Urinary excretion results below which this TIB method is still
considered an overestimate.

Radionuclide or | Nature of | Time since last Excretion rate
‘analysis type _| oxposure_| urinalysis (d)_| Organ of concern pera peu
Etemental U cont- Tacility, WA, assumes Re ne Marow 5,300 Kilt}
natural U, type | chronic exposure
F for at least one
year. Sampling
during work days.
Bone surface C510 78
Stomach 7,280 318
Si 7.280 om
TEI 7240 Ey
cor 7320 42
Skin T, 200 920
Spleen T2390 Eys]
THYMUS 7,290 O24
Uterus 7250 920
ET 555,000 382000
Tung 827,000 597000
Colon 7300 26
sophagus 7200 O24
TEU Muliply results for natural Uby 0.031
FeV Muliply results for natural Uby 0.01
EemenelU Uracil, Sampling on third [Adrenals 376 258
natural U, day of no Urinary Badder 351 250
type F exposure.eg., [Brain 374 267
Monday moming [Breast 375 28
Gall bladder EG] 267
Elemental Ucont. TU facility, ‘Sampling on Heart wall 376 Zoo
natural U, thinl day of [Ridneys 70 50
type F exposure Ther BBS. B32
Muscle 374 267
varies TO T700
ancreas 375 266
Tesies 7,570 TT20
Thyra 374 267
"Red bone marrow T5360 TT20
one surface 2,000 TA50
fomach 373 266
T 370 264
vay 360 257
ToT 382 273
1 37S 267
pleen 375 268
“TAYMUS 375 ist:
Uterus 374 267
T T55,000 T1000
ang 240,000 T7100
Ton 376 259
Esophagus 375 268
TEU Multiply results for natural U by 0.31
FEU Muliply results for natural Uy 0.01
Elemental 0 AiphaTadlity, | Because Intales at alpha facies are consdewd acute for overesiimales, any
any absorption | detecton of U above natural background will have to be evaluated separately from
type OTB-018.



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 _| Effective Date: 10/01/2007 | Page 31 of 74 _ |

alt inhalation dose reconstructions under EEOICPA, as this represents use of the best
available science. Use of older solubility class information or assumptions to maximize
committed dose could result in organ dose assessments that would not be favorable to the
claimant.

{4] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007.
These codes are from worker dosimetry records.

[5] Burns, Robert E. Shonka Research Associates. Senior Health Physicist, April 2007.
The selection of 7°°Pu as the default radionuclide, when the true isotopic mix is not known from
measurement, is made to ensure assessments favorable to claimants.

[6] LaBone, Thomas R. ORAU Team. Deputy Principal Internal Dosimetrist. September 2007.
The urinary excretion in 1951 (0.13 dpm/day on 7/1/1951) and 1952 (0.11 dpm/day on
71111952) from OTIB-0034 (ORAUT 2005b) were modeled with a chronic intake of Type S
299 u from 1/1/1948 to 12/31/1952. The calculated intake rate was 200 dpmiday.

{7] Brackett, Elizabeth M. ORAU Team. Principal internal Dosimetrist, and LaBone, Thomas R.
ORAU Team. Deputy Principal Interna! Dosimetrist. October 2006.
The changes to Section 5.4 were initiated by Brackett and LaBone and the calculations were
performed by LaBone.



AECU-817 25

The early signs of damage due to external radiation are lassitude, fatigue, and variation in the
number of white blood cells. The red blood cell decrease is a late effect of overexposure, leading
to fatal anemia.

Internal radiation refers to radiation received by the ingestion or inhalation of radioactive sub-
stances, The dose of internal radiations is measured by the concentration of radioactivity in the
excreted body fluids, and in the case of inhalation it is measured by the concentration in the expired
air. The late effects of continued inhalation of excessive amounts of radon have been realized by the
high incidence of lung cancer among workers in the uranium mines of Bohemia,

There have been several attempts to calculate the lung dosage, but unfortunately the validity of
these results cannot be checked because the reports on the Schneeberg miners did not include the
site of origin of the cancer. However, on the basis of these calculations, the American Bureau of
Standards recommended a permissible tolerance concentration of radon as 19-14 curie/ce of air.

There are other radioactive gases, such as radon, mesothorium, xenon, argon, krypton, and
jodine, which must be considered hazardous, The tolerance value for radioactive iodine in the at-
mosphere is based on the selective absorption and deposition of fodine in the thyroid gland, Per-
migsitle atmospheric concentration has been tentatively placed at 1.0 x 10~ s eurte/ce over 24 br
per day. In the case of the noble gases, argon, krypton, and xenon, the hazard is due primarily to
beta and gamma radiation from gas outside the body.

Internal radiation from ingestion and deposit of radioactive material in the body was first en-
countered among radium-dial painters, Although only a few micrograms were retained in the body,
this proved to be a lethal dose. The continued bombardment of the bone marrow led to fatal anemia
or production of malignant bone tumors. The actual path of radioactive materials in the body is very
complex, and, a8 yet, no conclusive estimates can be made, The experience with the radium potson-
ing in the luminous-dial industry has led the Bureau of Standards to establish 0.1 ug of radium as
the limit which can be deposited in the body without resulting in later damage,

The skin is composed of two layers of tissue, the epidermis which consists of the protective
layer and the dermis which carries the nutrient vessels, accessory organs, and nerve supply. The
thickness of the epidermis varies over the body, but in general it ranges from 0.07 to 0.12 mm. On
the palms of the hands it averages 0.8 mm, while on the soles of the feet it is as thick as 1.4mm,
The hands can take a greater dose than the whole body tolerance.

The results of radiation are produced in the dermis as damage to the capillaries, venules, and.
arterioles supplying the skin, The mechanism is thought to be identical with the erythema produced
by ultraviolet irradtation of sunlight,

The thickness of the epidermis is of importance in relation to the penetrability of alpha and low-
energy beta radiation. The alpha particles of uranium have a range in tissue of about 0.5 mm and
therefore do not produce any skin injury, When the energy of the alpha particles ts greater, such as
those from a cyclotron, the penetration is sufficiently great to produce erythema.

It must be remembered that erythema results as an injury from relatively acute doses of radia~
tion, As the time over which the dose is administered is lengthened, the tolerance increases.

The reproductive organs may sustain damage either to the germ plasm or to the cells which
carry the germ plasm. To obtain a permanent sterilization of the female ovary requires 400 to 600 r
delivered*within the ovary. Sterilization in the male is produced by 800 to 1000 r in the testes. How-
ever, sterility in the male is reversible, and although there may be temporary sterility at lower
levels, fertility is usually regained after a period of time,

In 1927 Muller demonstrated that the mutation rate of the fruit fly could be accelerated by ex-
posure to X rays. Radiation-induced mutations have been found to have characteristics which bear
on the practical consideration of radiogenetic changes possibly associated with occupational exposure.
The most important of these is that there is a linear relationship between dose and increase in mu-
tation rate, There is no threshold effect—the cumulation of exposure is therefore additive, The
magnitude of the effect is independent of the wavelength and dosage rate of exposure. These genetic
studies of the fruit fly have led to considerable discussion concerning their application to man in
relation to the tolerance dose. Actually, very little is known about genetics in man, and still less
is known about the effects which might be produced by exposure to radiation, As yet there is no



94 AECU-817

4, For wounds: rinse quickly under running water followed by a soap and brush wash if possible.
Accompany employee to Medical where further hand decontamination will be attempted only under
their advice and supervision.

Decontamination Kits Contain.

1, Cleanser
a. Zephonite (powdered soap)
b. Lanokleen (powdered soap)
c. Liquid soap
d. Old Dutch Cleanser

2. Four scrubbing brushes

3. Two 1-lb citric acid jars

4. Two 1-lb KMnO, jars

5. Two bars of soap

6. One bottle of NaHSOg

7, Lanolin

8. One pair of surgical gloves

Keep kit up to full strength by replacing used articles from stores as soon as they are depleted,

8.7 THE DESIGN AND OPERATION OF RADIOCHEMICAL LABORATORIES*
By W. H. Ray

Responsible Operation Essential

In contrast to a chemical works, a laboratory cannot be made foolproof by its deSign or equip-
‘ment, In addition to facilitating design and beneficent equipment, safety in a laboratory requires the
elimination of fools and foolish operation, The prerequisite is education, but some persons must
expect to be excluded from laboratory work if they exhibit inaptitude at protecting lives. Perhaps
licensing may evolve, if legal activity is incited by misfortunes in laboratories dealing with espe-
cially hazardous materials and processes. This can be forestailed if appropriate selection and re-
striction is exercised by managements responsible for laboratory operation.

Safety in a radiochemical laboratory requires that each individual permitted to control any
radioactive material shall be competent and responsible to prevent exposure to any person of radia-
tion dosage in excess of the maximum permissible rate (0.1 rem/day), This responsibility extends
beyond the radiation produced by sources assigned to the responsible person, and in some cases
may mean that no additional exposure to radiation from these sources can be permitted. Control
over all radioactive material is to be maintained until it is transferred to other competent and re-
sponsible individuals or until an approved preparation for disposal has been compounded.

Numerous rules and devices for assistance in keeping exposures to radiation less than the
maximum permissible may be proferred, but they cannot make a laboratory safe if those entrusted
with radioactive materials do not evidence responsibility for the associated hazards. The paramount
rule is to prevent the exposure of any part of the body to irradiation in excess of 0,1 rem/24 hr. In-
telligence exercised by the persons responsible will find many ways to achieve this, whether the
hazard is exposure to penetrating radiations or the intake of radioactive contaminants that may be
retained in the anatomy.

The following examples indicate some of the malpractices in a radiochemical laboratory which
can be guarded against without special knowledge.

1, Use of food product containers to hold active materials

2. Use of refrigerators for foods and radioactive samples jointly

3. Eating in radiochamical laboratories

4, Smoking while handling radioactive materials, especially while wearing protective gloves

*The following material, previously written by Mr. Ray for another purpose, was issued in lieu
of lecture notes.



30 AECU-817

Permissible Concentration of a Radioisotope in the Atmosphere for Constant Exposure over a Long
Period af Time

If the daily atmospheric concentration of a particular radioisotope is maintained at a value which
is roughly constant for a period of time long compared to its effective half-life in the critical tissue
of the body, the rate of loss of the isotope from the body will equal the rate at which it is taken in
from the atmosphere. We use this relationship as follows:

‘The rate at which the atoms of a radioisotope disintegrate is proportional to the number present.
After chemical elements are fixed in body tissues, the rate of elimination also is generally approxi-
mately proportional to the amount present. The rate of radioactive decay of an isotope may be de-
scribed by giving the fraction of its atoms disintegrating per unit of time or by giving the length of
time for half of any quantity of the isotope to disintegrate. The former ts called the “disintegration
constant’’ of the radioisotope, and the latter is called the ‘‘radioactive half-life.’”

Let T, equal the radioactive half-life, in days, and f, equal the fractional decrease per day

0.7
fee

and T, equal the elimination half-life, in days, and f, equal the fraction eliminated per day

0,7

fe= Te

As a result of these two processes, the effective fractional decrease per day is
fat, +f,

This daily decrease gives rise to an effective half-life of the radioisotope in the tissue,

Since

2 a orT= Tr Te
TT, *Te Tr+Te

if C is the number of curies of the radioisotope in the tissue, the rate of reduction is

ac 0.1C
a Crp

If we take the permissible dosage rate for a beta-gamma—emitting radioisotope to be 0.1 rep/day,
the number of curies which will give this dosage rate is, from the preceding section,

= OW __ 2 1.64 x 10° &
6.1 x 107E, E

The rate at which this quantity of radioisotope in the tissue is being reduced by a combination of
radioactive decay and physiological elimination is



[Document No. ORAUT-TKBS-0012-5 | RevisionNo.01__| Effective Date: 10/01/2007 | Page 150f74 |

Table 5-7. Fecal analysis codes (Mani
1983; MMES 1995).

Alphanumeric code
GFO | Pu-239 gross alpha (includes Th)
PFO | Pu-239
RFO | Rare earths ()
SFO _|Sr-90
SF9 | Sr-89
TFO | Trans plutonium (Am-241)
UFO [U-234
OFO | Other
Table 5-8. Result codes for whole-body counting [4].
Before 1971 1971-1978
Code Description* Code Description*
1 Normal human spectrum 0 = <15% MPOB
2 Less than 10% MPBB 2 = <25% MPOB
3__|Less than 25% MPBB 4 |= <50% MPOB
4 Less than 50% MPBB 6 = <100% MPOB
5 Greater than 50% MPBB 8 >100% MPOB
N__| Insignificant & indeterminate
S__| Significant & indeterminate

*MPBB = maximum permissible body burden; MPOB = maximum permissible organ
burden.

extracted chemically from the biological sample and the total radioactivity of the element in the extract
was measured, At some point after the extraction and sample count, the total sample activity was
attributed to a specific radionuclide. Many of the isotopic assignments were based on process
knowledge.

To reflect the performance of the instrumentation and analytical methods more accurately during the
period prior to 1990, Table 5-9 assigns MDAs for some isotopes to their corresponding radioelement
rather than the specific radionuclide. For example, routine separation of the alpha emitters “°Pu and
288Pu did not occur until alpha spectrometric analyses became routine in 1989, but the recovered
database reports both separately. They have been combined in this document as “Plutonium.”
NOTE: “Pu is reported separately because it is a beta emitter that can be assayed separately from
the alpha-emitting isotopes of plutonium.

5.2 IN VITRO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND
REPORTING PRACTICES

§.2.1 In Vitro Minimum Detectable Activities
5.2.1.1 | Minimum Detectable Activities between 1947 and 1989

Table 5-9 lists historical MDAs for radionuclides of concern in urine and feces. These MDAs were
calculated from analytical records recovered from the in vitro sample databases supporting the ORNL
historic workforce dose assessment project (MMES 1995). Attachment A contains details of data
recovery and additional information on specific radionuclides. Blank MDA entries in the early years in
Table 5-9 indicate that no analytical results for that radionuclide were recovered from that year. This
could be because the records have not been located, or because the analysis was not performed.



[L_ Document No. ORAUT-OTIB-0034 | Revision No. 00 __| Effective Date: 12/13/2005 | Page 130f29 |

. When a “Sr intake is assigned, the worker should also be assigned intakes of ‘Ru,
*87Cs, and “Ce, consistent with the average isotopic ratios developed from the air
monitoring data given in Table 5-7, using the °°Sr intake as the basis. For example, if a
worker were exposed in 1965 to Type F Sr, the intake (from Table 5-1) would be 15.5
dpm/d. Intake of ‘Ru would also be assigned at a value of 103.8 dpm/d (i.e., 15.5
dpm/d of ®°Sr times 6.7, the average ratio of ‘Ru to “Sr from Table 5-7), along with a
34.1 dpm/d intake of *"’Cs and 170.5 dpm/d intake of ““Ce. Although Table 5-7 is
based on data from 1975 through 1984, the average ratios calculated should be used
for all years.

The lognormal distribution is selected in the Interactive RadioEpidemiological Program
(IREP), with the calculated dose entered as Parameter 1 and the associated

GSD as Parameter 2. The GSD is associated with the intake, so it is applied to

all annual doses determined from the intake period. The GSD for ®°Sr should be
assigned when intakes of ‘Ru, '*’Cs, and/or “*Ce are included.

Table 5-7. Evaluation of the contribution of additional radionuclides.

Reported “Sr | Reported Ru | Ratio: | Reported ""Cs| Ratio: | Reported "Ce | Ratio:
concentration | concentration |**Ruto| concentration | *’Cs to | concentration | “Ce to
Year (10° uCirmL) | (10% pCiimL) | “Sr | (10 yciimt) | “sr (10° wCi/mL) Sr

1975" 0.78 67 86 1.4 18 11 14
1976 0.77 0.88 14 0.55 07 24 3.4
1977 1.30 1 85 1.60 1.2 21 16
1978 0.81 97 12 2.14 26 16 19
19797 0.15 1.56 10 0.67 45 18 12
19807 0.08 0.49 6.1 0.37 46 0.98 12
1981° 0.46 4.40 9.6 1.2 26 4 24
1982" 0.15 0.40 27 0.24 16 0.03 0.2
1983* 0.08 0.12 15 0.10 12 NA NIA
1984" 0.07 NIA NA____|0.07 4.0 NA NA
Peak ratio : : 12 - 5 = 24
Ratio range : - 4.1- 12 [- 07-46 |- 0.2 - 24
Average ratio | - - 6.7 - 2.2 : 1

a. References: 1975 — (Union Carbide 1975); 1976 — (Union Carbide 1976); 1977 — (Union Carbide 1977), 1978 — (Union Carbide 1978);
1979 — (Union Carbide 1979); 1980 — (Union Carbide 1980); 1981 — (Union Carbide 1981); 1982 — (Union Carbide 1982); 1983 —
(Martin Marietta 1983), and 1984 — (Martin Marietta 1984).



26 AECU-817

evidence that the present generations of radiation workers have produced offspring which differ from
those of the general population. If experimental findings in the lower organisms are accepted as
valid for man, then one can only escape some degree of radiogenetic effect by avoidance of all radi-
ation exposure.

Recent animal experiments have indicated that long-continued radiation shortens the over-all
life span of the animals. On the basis of this consideration, plans are now being formulated for re-
ducing the tolerance dose to 0.3 r/week. The fact that radiation is used extensively in medicine must
also be considered, and it is suggested that radiation workers inform the doctor of the conditions
under which he is working.

‘The estimation of the dose received by the body is normally founded on the reading at one region.
There are certain conditions under which such a reading may be unreliable. In the case of directional
radiation, for example, where radiation is aimed at a specific part of the body, the whole body will
not receive the same dose. This dose may be large enough so that if the whole body were subjected
to it, the damage would be intolerable, yet the area to which radiation is directed may not be affected.

For emergency cases, certain tolerances have been defined to provide for permissible exposure
above that of the daily tolerance dose. In the case of fire, for example, individual exposure will be
limited to a total dose of 5 r tq be acquired at intensities not exceeding 20 r/hr. If the intensity is
10 r/hr, the fireman will be permitted to stay in the area for 1/2 hr. If the intensity is only 2.5 r/hr,
he will have an exposure time of 2 hr.

For cases of extreme emergency, not generally anticipated in industrial operations, a man
twenty-one years of age, not previously subjected to significant radiation exposure, may receive a
single whole-body exposure of 25 rep provided he receives no other significant exposure before the
age of forty-five. For persons over forty-five, a single whole-body exposure of 50 rp may be con-
sidered permissible provided no other significant exposures are received in a lifetime, If a person
received a single whole-body exposure of 25 rep before reaching forty-five, he would be permitted
a single whole-body exposure of only 25 rep after the age of forty-five.

The permissible tolerance doses for a 24-hr day used in practice at ORNL at present are listed
in Table 3.1.

Table 3.1
Type of radiation* Roentgen Rem Rep
X-ray 0.1 0.21 0.1
Gamma 0.1 0.1 OL
Beta 0.1 0.1
Fast neutron 0.1 0,02
Thermal neutron 0,2 0.02 ~0,1
Alpha 0.1 0.01

*Considered from standpoint of internal effects.

The maximum permissible level for plutonium in the atmosphere is tentatively set at 5x 19-10
ue/ce or 3x 19-12 pe/ce for an 8-hr working day, 6 days per week, for a one-year period. The ob-
ject is the prevention of the accumulation of a total of more than 1 yg of plutonium in the body during
a person’s life.

The air-contamination tolerance level for the more hazardous and common beta-gamma radio-
isotopes, such as iodine, strontium, barium, etc., is considered to be about 10-7 pe/ce of air,

Working-time calculations are based on the period over which the dose is administered, If a
man has worked in an area where the intensity is 0.1 r/hr, he will be limited to 1 hr of work in that
area and will not be permitted to be exposed to any further radiation over the 24-hr period. If the
intensity of the area is 0,01 r/hr, the man may be continually exposed over an 8-hr period or even
10 hr if necessary.



{Document No. ORAUT-TKBS-0012-5 | Revision No. 01 | Effective Date: 10/01/2007 | Page 26 of 74_]

Sides of the bed. As described below, it can also be used as a backup chest counter, if necessary,
due to its 3-to-50% efficient germanium detectors (Watts et al. 1995). The detectors are in a solid
center shield. The shield is designed so there is a minimum of 4 in. of lead around each detector.

During this time, the 5.25- by 0.0625-in. Nal - 2-in. Cs! phoswich detector was used to conduct chest
counts. Some time between 1977 and 1994, another phoswich detector with similar dimensions was
procured and used until three 70% efficient germanium detectors were installed in the counting room
in 1994 (Hembree 2004; McLaughlin 2004). A chair counting geometry was reinitiated at that time,
with the individual positioned with his/her back reclined 45 degrees from the vertical. The three
germanium detectors are positioned with two detectors over the right lung (one over the upper portion
and the other over the lower portion) and one over the left lung. The third detector is over the upper
Portion of the left lung.

In addition to the above, there are three other detector arrangements in the WBC facility. A
germanium thyroid counter can be set up using one of the detectors used for lung counting. The
individual is in the same chair used for tung counting, but the chair is inclined at a 25-degree angle.
Other organ-specific geometries can be used as necessary (skull, liver, hand, etc.). The 5.25- by
0.0625-in. Na! - 2-in. Csi phoswich and the SBC (in a fixed-bed position) are backup systems that
could be used for lung counts if needed.

Until 1994, field health physicists selected individuals for in vivo monitoring. At that time, the intemal
dosimetry group became responsible for the selection of personnel for in vivo monitoring with
guidance from field health physicists. Berger (2003) indicated that the in vivo monitoring program
may not have been effective for monitoring ORNL personnel until the late 1980s because of reliance
on field health physicists to determine the individuals to be monitored.

5.3.3 Cesium Counting Artifact

Many ORNL workers had measured body burdens from intakes of *’Cs from nonoccupational
sources (e.g., fallout and consumption of local venison). At present, ORNL uses a value of 20 nCi
*7Cs in a whole-body count for consumers of venison as the decision level to follow up or conduct a
dose assessment (ORNL 2002). However, the Laboratory has never used it to negate an individual's
occupational dose that was received on the site (McLaughlin 2004).

Fallout affects everyone in North America, and body burdens of '"’Cs measurable in whole-body
counters were common in the 1960s and 1970s. NCRP Report 94 (NCRP 1988) provides mean body
burdens of *’Cs for the United States for the years most likely to produce interference with
occupational whole-body count results. Table 5-16 lists these values.

§.3.4 Minimum Detectable Activity

As stated above, the thin (5-in.-diameter by 0.0625 in.) Nal (Tl) crystal array used in 1963 to detect
low-energy photons emitted primarily from transuranics could detect approximately 40 nCi of “Pu, if
there had been a preexposure chest count. If there was no preexposure count, the detection
capability was approximately 80 nCi (Morgan, Snyder, and Struxness 1963).

Table 5-17 lists MDAs for the HPGe array of six individual detectors in the aluminum block (used
around early- to mid-1980s).



[Document No. ORAUT-TKBS-0012-5 | Revision No.01__| Effective Date: 10/01/2007 | Page 200f74 |

Counting facility utilized a 4- by 4-in. Nal (TI) crystal, but it is not clear what the counting geometry

was. There is a photograph of an individual laying on a nylon-strapped, aluminum beach chair, but it

was not clear whether that was the geometry initially employed in the facility. The Argonne Chair
counting geometry was set up for whole-body counting on February 27, 1961 (Morgan, Snyder, and
Struxness 1961). The tubular steel chair was tilted so the individual’s body was in a V position with

the detector placed approximately 50 cm directly over the hips (Morgan, Snyder, and Struxness 1961, |
p. 223).

Before construction of the WBC facility, health physicists realized that there could be problems with

siting a low-level radiation counting facility in the plant environment, with radioactive effluents from
operations being generated immediately adjacent to the facility. Thus, ORNL designed and installed a
recirculating air treatment system that pumped air from inside the counting room through charcoal

traps, cooling coils, and heaters to remove radon, odors, and excess moisture (Brown, Patterson, and |
Abee 1971; Morgan, Snyder, and Struxness 1961). A slight positive pressure was maintained in the
counting room by using cylinders of “aged” breathing air to make up for leakage from the system.

To improve the detection sensitivity, additional layers of material were applied to the floor in the WBC.
facility counting room over the years. Morgan, Snyder, and Struxness (1963) indicates that |
0.04-in.-thick layers of tin and cadmium were laid over the interior surface of the lead, only on the

floor, with a 0.01-in.-thick layer of copper over the tin and cadmium. The original vinyl tile was placed
over the copper. These materials were installed to reduce background radiation emitted from the lead
shield as a graduating shield to minimize the contribution of low-energy X-rays. In the mid-1990s, the
vinyl tile surface was removed and a 0.03-in.-thick {ayer of stainiess steel was placed directly over the
other metal layers. Site personnel indicated that the stainless-steel layer was primarily for aesthetic
purposes to cover the oxidized copper layer. With the addition of frictional surfaces for slip reduction,

this was the latest form of the floor. Table 5-11 lists the construction history of the counting room.

Table 5-11. Construction history of the counting room.

Installation date Material Thickness (in.)
1959 (entire facility) | Pre-WW II steel 14
1960 (entire facility) | Special low-radioactivity lead 0.125
1963 (floor only) Tin 0.04
1963 (floor only) Cadmium 0.04
1963 (floor only) Copper 0.01
Mid-1990s Stainless steel 0.03

Subsequent improvements to the shielded counting room, detectors, and counting geometries are
described below. The shielded room was used to conduct all in vivo measurements at ORNL until
1992, when a Canberra scanning bed counter (SBC) was installed in Room 16 of Building 2008.

A discussion with Berger (Berger 2003), who was responsible for the facility from the mid-1970s until
the early 1980s, indicated that although the facility was in operation, a formal in vivo monitoring
program was not in place until the late 1980s. Prior to that time, the WBC facility was used almost
exclusively to either confirm known or suspected intakes of radioisotopes or for research purposes.
Selection of individuals for counting was performed by field health physics personnel based on
expected contaminants of concern and the probability of exposure until the early 1990s, when the
selection of individuals for in vivo monitoring became the responsibility of the Internal Dosimetry staff.



([Document No. ORAUT-TKBS-0012-5 | Revision No. 01 _| Effective Date: 10/01/2007 | Page 27 of74 |

Table 5-16. Mean body burdens of *’Cs from fallout in the
United States.

Year | Body burden (nCi) Year {| Body burden (nCi)
1953 0.27 1966 9.7
1954 14 1987, 5.6
1955 2.2 1968 3.5
1956 4.3 1969 2.7
1957 5.1 1970 2.7
1958 6.5 1971 27
1959 8.1 1972 27
1960 6.8 1973 27
1961 46 1974 16
1962 6 1975 1.1
1963 1 1976 16
1964 19 1977 1.1
1965 16

Table 5-17. MDA values for six-detector
HPGe array (Berger and Lane 1981).

Nuclide Organ MDA (nCi)
Pu-239 Lung 21.5
Am-241 Lung 14
4-125 Thyroid 0.04
Cs-137 Lung 16.4
Gd-153 Lung 0.87
Eu-152 Lung 0.71
U-233 Lung 1.42
Co-57 Lung 0.94

Tables 5-18, 5-19, and 5-20 list MDA values for the ORNL SBC, germanium thyroid counter, and fixed
bed counter, respectively (Watts et al. 1995). MDA values are calculated for each individual
measurement made using the germanium lung counter, SBC, and fixed bed counter. Figure 5-1 is a
plot of MDA values for different nuclides for the ORNL germanium lung counter as a function of chest
wall thickness. Figure 5-2 shows MDA values for °Pu and "Am for the ORNL phoswich lung
counter as a function of chest wail thickness. Table 5-21 lists the frequencies of in vivo counting since
1960 and, for the various detectors, a summary of the best estimates of the applicable periods of use.

Table 5-18. MDA values for ORNL scanning bed counter (Nal/Ge) (Watts 1995).

tsotope | MDA (nCi) tsotope | MDA (nCi) lsotope MDA (nCi)
Cr-51 24/30 Min-54 2.1/2.4 Co-57 4.9/5.9
Fe-59 3.9/4.5 Co-60 22.9 Zn-65 4.3/6.0
Rb-88 12/15 Y-88 1.8/9 Nb-95, 2112.7
Zr-95 3.7/4.1 Tc-99m 4.415.3 Ru-103 2.6/2.9
Ru-106 21/30 Cd-409 NG/224* Sn-113 3.4/4.0
Sb-124 2.2/3.1 £131 2.7/3.1 Ba-133 3.7/4.5
£133 2.7/2.9 Cs-134 2.2/3.0 Cs-137 2.4/2.8
Ce-139 4.3/4.6 Ba-140 9.2/9.1 La-140 2.3/2.4
Ce-141 7718.4 Ce-144 36/43 Eu-152 8.5/9.7
Gd-153 21/21 Eu-154 6.1/7.2 Pb-212 6.3/7.6
Bi-214 5.8/9.8 U-235 5.9/6.1

*NG = not given



AECU-817 55

5.4 SURVEY INSTRUMENT USAGE AND SELECTION (1)
By E, J. Kuna

This lecture, primarily, will be a résumé of my experiences with instruments, and as such, it
will deal with a practical application of an instrament’s characteristics to the surveyor’s job. It
must, of necessity, be a biased and prejudiced viewpoint, with theory subordinated for actual working
practices,

The various and sundry instruments manufactured and developed at Oak Ridge National Labora-
tory are used to discover, differentiate, and measure the various types of radiation. Thus, for quali-
tative work the Walkie Poppie or the Walkie Talkie serves a useful purpose. For quantitative work
the Zeus, Zeuto, or the Cutie Pie fills the requirements. In simplier terms, ‘‘first you must find it
and then you must measure it.’’

‘The degree of radiation encountered limits one’s choice of an instrument. Obviously an instru-
ment such as the Cutie Pie with a maximum range of 10 r/hr could not be used 1n cell work where
radiations up to 100 r/hr are encountered. This example would require a Fish Pole Probe which
has an additional advantage of a 6-ft handle to cut down the exposure possibilities of the surveyor
(inverse square law).

‘The radioactive tolerance of the area limits the choice of an instrument, Obviously a Walkie
‘Palkie or Victoreen 263 would not be used in a high-background area. These would do well ina low-
background area such as cold chemistry laboratories.

Finally, the degree of accuracy required of low-radiation readings would also limit the choice
of an instrument. Such a requirement is required of isotope shipments where the electroscope is
best suited to fill the need,

All instruments require practically the same technique of operation, First, of course, is the
warm-up period, followed by the zero adjustment or setting. A few instruments also require the
correct voltage setting, such as, for example, the Walkte Poppy. In use, the factors of distance and
type of radiation encountered are noted along with the degree of radiation. At this point your own
expertence, knowledge, and observation guide you in making a judgment.

Factors of environment, light, heat, wind, and dampness affect all instruments to a degree. The
Cutie Pie stands up well under this test, while the Zeus and Zeuto are very poor actors. Maneuver-
ability and weight of the Cutie Pie puts it first again in comparison with the Zeus and Victoreen 263.
‘The accuracy of these four instruments, the Cutie Pie, the Zeus, the Zeuto, and the Victoreen 263,
is approximately the same. The calibration is made to within an error of +10 per cent,

‘The Cutie Pie, Zeus, and Zeuto all have ion chambers that pick up the ionization current and
send it through a selected resistance, The voltage drop across the resistance is then measured in
microamperes. The Victoreen 263 is different in that it uses a G-M tube instead of an ionization
chamber.

Summary

1. The Cutie Pie, a gamma-calibrated beta-indicating instrument with a 5 mr/br to 10 r/hr
range, an accuracy of 10 per cent, highly maneuverable, highly resistant to environmental factors.
‘This has a window for differentiating between beta and gamma.

2. The Zeuto, an alpha-measuring instrument with a reported accuracy of 10 per cent, highly
affected by environmental factors.

8, The Zeus, an alpha- and gamma-calibrated beta-indicating instrument between the ranges of
5 mr/hr to 5 r/hr. The accuracy is reported to be 10 per cent. It is highly affected by environmental
factors and low maneuverability.

4, Victoreen 263, a gamma-calibrated beta-indicating G-~M-tube instrument with three ranges
from 12 to 20 mr/hr, It is best used as a qualitative instrument. It is slow reacting and heavy, and
it has low maneuverability.

Remember that accuracy is a relative term useful for laboratory conditions. Do not compare
two different instruments since their degree of sensitivity varies, Educate the survey requester to



AECU-817 65

Cutie Pie $200.00 a-c Poppy $450.00
Zeus 300.00 Beckman 300.00
Electroscope 90.00 Victoreen 263 230.00
Zeuto 240.00 Monitron 500.00
Scaler 350.00 —500.00 Dosimeter 45.00
Landsverk 200.00
G-M tube 10.00-50,00

The prices of the instruments may vary slightly between different companies, but the prices
given are essentially correct.

Care of Instrument

The manuals accompanying most instruments state that they are light, portable, rugged, easy
to operate, etc. The instruments are only rugged, however, when used in proper survey techniques.
Abuse or maltreatment of any of our portable survey instruments will result in improper operation
or faulty readings. Some of the ways in which the instruments may be abused or mistreated are as
follows: (1) leave batteries on, (2) contamination, (3) jars or rapid movement of instrument, (4)
dropping or knocking the instrument over, and (5) tampering with the instrument in the field.

We should like to stress that surveyors should not attempt repair work on the instruments in
the field, When an instrument is in need of repair, the surveyor should call the Calibration Building
and have the instrument picked up for repair. Where possible, a replacement will be issued at the
time the instrument is picked up.

Repair of Instrument

There are several definite types and varieties of instruments, each having its own peculiarities
as to malfunctions and methods of repair. We shall discuss only three instruments, which will be
three definite types of instruments: the Cutle Pie, the Walkie Poppy, and the Victoreen 263.

Cutie Pie. The following is a list of some possible malfunctions of the Cutie Pie with probable
causes:

1, Will not zero

Causes
a. 71/2-volt batteries do not have same voltage
b. Resistors on both sides of zero potentiometer not balanced
¢, 22 1/2-volt battery down
d, Filament battery down
2, Needle fluctuates
Causes
a. Insulator in back of ionization chamber dusty or dirty
b, A bad connection anywhere
c, A high-value resistor open
d. Resistor dirty or tube dirty
e. Burned wax
3. Reads off scale
Causes
a. Stud mounting ton chamber grounded to case
b. Corroded or dirty connection
c, One of resistors bad
d, Bad switch on potentiometer
4, Will not zero on sensitivity scales
Causes
a, High-value resistors bad
b, Tube base may be bad
c. Dirty insulator at end of ion chamber



2 AECU-817

Gamma and X rays are electromagnetic photons of very high energy. They are completely
identical, and the difference in name only signifies that gamma rays arise from the changes in the
nuclear energy levels, while X rays are caused by similar shifts in electron energies.

It may happen that a gamma ray making its way through the orbital shells of the atom will strike
an electron and eject it as a fast-moving particle. The energy of the gamma photons will be given to
the electron as kinetic energy which will be equal to the energy of the photon minus the ionization
potential. This process is called ‘‘internal conversion.”

‘The process of beta decay liberates a high-speed electron, but the mechanism is something en-
tirely different, In a beta-active atom a neutron breaks up and gives a proton plus an electron. The
electron is given considerable energy and moves rapidly away from the nucleus. There is a very
slight change, approximately 1/2000 amu, in the mass of the nucleus so that the weight, in atomic
mass units, of the nucleus remains essentially unchanged. However, the number of neutrons has de-
creased by one, and one proton has been added. This raises the atomic number of the atom 1 unit
and completely changes its chemical characteristics.

We can pause here to consider the difference in experimental measurements on interna) con-
version electrons and discuss beta particles. Although the two radiations are completely identical
on a particle for particle basis, we find that the beta particles from a given isotope are emitted with
a large spread of energies. On the other hand, the internal conversions of electrons all have the en-
ergy of the gamma ray which liberated them. Any given isotope will emit only a very few definite
energies of gamma radiation, and therefore, ali the internal conversion electrons will have similar
discrete energies.

Alpha decay is somewhat similar and causes a loss of 4 amu in the nuclear mass. It decreases
the atomic number of the material by two units, thus causing a transmutation.

Free neutrons and protons are usually found as the results of neutron bombardment of some
target material.

In attempting to specify the strength of a given source of radiation, we usually speak in units of
curies. The curie was originally defined as the rate of disintegration of the radon in equilibrium
with 1 g of radium. This, however, applied only to a radon source. To generalize this unit, it is now
arbitrarily defined as 3.7 x 10/0 disintegrations/sec. If it is desired to express the strength of any
source in terms of this unit, divide the measured number of disintegrations per second for the source
by 3.7x 1010, since the curie is a very large unit of activity, we frequently use submultiples. The
millicurie (me) is ¢/1000 or, similarly, 10° c, and the microcurie (yc) 15 ¢/1,000,000 or 10-6 c.

2.8 INTERACTION OF RADIATION WITH MATTER
By W. H. Ray

Radiations are absorbed by matter by processes which impart energy to the absorber. The
cases of concern to Health Physics are those in which ionization is involved in the process.

Beta particles dissipate their kinetic energy by knocking orbital electrons from their respective
atoms, An average of 32 ev is required for the formation of each ion pair. When the beta particle
has lost its kinetic energy, it penetrates no further,

Alpha particles produce fons similarly, but because of their relatively lower velocity and greater
charge they are able to produce more ions per unit length of path.

Photons impart their energy to respective electrons in the absorbing medium which in turn pro-
duce ionization in a manner similar to that of beta particles. There are three methods by which a
photon may impart its energy to an electron:

1. Photoelectric process
2. Compton scattering
3. Pair production

Each method has a probability of occurring as the photon proceeds through the absorbing medium.
Consequently there is a probability that some will penetrate any absorber thickness which may be
established. The probability depends on the orbital electron density encountered and the energy of the
photon.



(Document No. ORAUT-OTIB-0034 | Revision No. 00 _| Effective Date: 12/13/2005 | Page4of29 |

LIST OF TABLES
TABLE TITLE PAGE

5-1 Combined *’Sr Type F intake periods and rates
5-2. Combined *°Sr Type S intake periods and rates
5-3. Annual ?*U Type F, M, and S intake periods and rate:
5-4 Annual Pu Type M intakes and rates
5-5 Annual Pu Type S intakes and rates.
5-6 Combined *'Am Type M intake periods and rates
5-7 Evaluation of the contribution of additional radionuclides
A-1 Statistical summary of annual ORNL strontiumX-10 °°Sr 24-hour urinary excretion

tates, 1951-1988
A-2 Statistical summary of annual ORNL uranium 24-hour urinary excretion rates,

1951-1987

A-4 Statistical summary of annual ORNL transplutonium 24-hour urinary excretion

rates, 1968-1987...
A-5 Type F °°Sr intake periods and rates
A-6 Type S “Sr intake periods and rates
A-7 Type FU intake periods and rates.
A-8 Type MU intake periods and rates
A-9 Type S TMU intake periods and rates
A-10 Type M TM®Pu intake periods and rates
A-11_ Type S **°Pu intake period and rates
A-12 Type MTM‘Am intake periods and rates



a AECU-817

(b) For a first trial, neglect the lower energy radiation which is absorbed much more rapidly
than the other. At 1 ft the dosage rate due to the 1.5-Mev gamma is 164 r/hr, and at 1.5 ft it is 73
r/hr. Four TVL’s will reduce this to 7.3 mr/hr, and an additional half-value layer will reduce it to
3.7 mr/hr, which is sufficiently close to 5 mr/hr for our purpose. For 1.5-Mev radiation through
lead, this is equivalent to 4.3 TVL’s times 1.6 in. per TVL, or 6.9 in.

In checking on the effect of the neglected 0.5-Mev gammas, one finds from the table that, for
0.5-Mev gammas, 6.9 in. of lead is approximately 11 TVL’s, from which it follows that their contri-
bution is practically zero. However, in many cases the contribution of the softer radiation will not
be negligible, and the estimate of shield thickness must be adjusted to take it into account.

Limitations of the Exponential Law in Shielding from Gamma Radiation. In the above examples
it was assumed that the number of quanta lost from the beam in traversing the shield was the number
reacting with the material of the shield. This is true only for very narrow beams and for very thin
shields. If the beam is wide, quanta from other parts of the beam are scattered toward the position
of interest; and, if the shield is thick, significant numbers of quanta scattered out of the beam may
reach the point of interest through repeated scattering. The magnitude of these effects depends upon
the geometry involved and cannot be simply predicted, but for a thick shield the emergent radiation
may be several times that predicted by the simple exponentia) law, using the absorption coefficient
applicable to narrow beamis and thin shields. In practice one may include the uncertainty involved
in an appropriate safety factor. Measurements by J. H. Roberson, on the attenuation of a wide beam
of a hard gamma radiation in lead, indicated that, for the geometry used

R/Rg = eH (1 + 0.2x)

where x is measured in centimeters. While this equation cannot be considered applicable to other
conditions than those under which the measurements were made, it is given here to give some indi-
cation of the nature and magnitude of the error involved in the simple exponential law on which the
use of tenth-value layers is based,

8.6 DECONTAMINATION OF PERSONNEL AND EQUIPMENT
By H. A. McClearen

Part I
L Common contaminants
A. Uranium
B. Plutonium
C. Fission products
1, Ruthenium 1. Krypton
2. Cerium 8. Barium
3. Strontium 9, Iodine
4. Columbium 10. Tellurium
5. Zirconium 11, Rare earths
6, Xenon

D. The degree of decontamination required depends upon two factors
1, Permissible personal exposure after decontamination
2. Probability of the spread of contamination

E. The decontamination procedure depends upon
1, Nature of the material to be decontaminated
2. Chemical composition of the contaminant

II. Decontamination reagents

A. Water and steam
1, Cold and hot water
2. Steam sparging (high temperatures and agitations)



AECU-817 a

8.8 SURVEY TECHNIQUES AND REFINEMENTS
By T. H, J. Burnett and Associates

The purpose of this discussion is the furtherance of a high professional standing, especially in
the opinions of the groups associated with us at this laboratory,

It ts valuable and desirable to exercise whatever reasonable additional care may be required
so that those cases in which an apparent disagreement or contradiction is thought to be seen between
the readings and/or recommendations of different individual health physicists will be reduced to a
minimum,

The following six points are those frequently involved in such situations of potential criticism:

1, Continuity of effort involves situations from which might arise apparent conflicts of opinion
on working times recommended for permissible exposure. On the same job, when it is continuing
over several work periods, different health physics representatives may give different working times,
and this is often noticed and wondered about, particularly by workmen with more meager scientific
backgrounds,

This situation ts best avoided by checking the log as soon as you come on duty to see the range
of readings on such continuing jobs. Second, the prior safety work permits issued for specific jobs
should always be checked. Thus, when a different working time seems more proper, it can be given
with a casual comment on the probable reasons for its difference.

It is not hereby suggested that efforts should be strained to get the precise readings of your
Predecessor. However, knowing their values gives a basis for inquiry into the reasons, peculiar to
each specific job, for different radiation levels and may indicate the need of more frequent survey
coverage than had been anticipated.

A brief discussion of the features of the Safety Work Permit was given by A. D. Warden, fol-
lowed by an outline of a proposed new procedure for individual exposure records for ‘‘hot’’ jobs by
7. C, Hart,

Before resuming, it was stressed that the responsibility lies strictly with the supervisor to
stipulate whether or not a radiation hazard should be anticipated, Health physics will exercise due
Professional care and concern lest any radiation hazards be overlooked.

2. Nonuniformity of permissible exposure is to be expected under certain circumstances. The
maximum permissible daily exposure to an employee working routinely in a laboratory with a rela-
tively high background may well be set lower than that of a maintenance employee called in for an
occasional repair job. Similar discrepancies arise when part of the work crew has previously worked
on a “‘hot”’ job, getting some fraction of their maximum permissible exposure for the day. These
individuals, then, would have working times proportionately shortened in comparison to those of their
fellows,

tn such circumstances it is well for the health physics representative to consider the advisability
of a casual word of simple explanation to the foreman in charge of the maintenance crew to allay any
untoward misgivings.

3. Variation between instruments is difficult to forestall, since if one survey is made with an
instrument having the allowable variation of approximately 10 per cent ‘‘high”’ and a survey is made
of the identical conditions with another instrument having a variation of 10 per cent ‘‘low,’’ a net
difference in these two surveys of 22 per cent will result, In every case then, it is easily seen that
the surveys made should be carried out with the maximum of care so as to utilize the inherent accu~
racy of these instruments to the full, A minor variation of 10 per cent can be casually cited as due to
unavoidable instrument variation, When possible, it is desirable to designate on the safety work per-
mit the number of the instrument used so a recheck with the same instrument can be made if desir-
able and also as an aid to discovering instrument malfunctioning.

4, Background change is the most frequent, important, and significant cause for variation in
recommended working times. This may take the form of an increase in working time, such as would
accompany successful progressive decontamination efforts, or the increase of shielding (in the case
of covering a hot pipe line with dirt) due to the furtherance of the work or the use of temporary lead
shielding.



{Document No. ORAUT-TKBS-0012-5 | RevisionNo. 01 _| Effective Date: 10/01/2007 | Page210f74 |

§.3.2 Detectors, Geometries, and Techniques

As described above, the initial in vivo counter at ORNL utilized a 4- by 4-in. Nal (Tl) crystal and a tilted
chair counting arrangement (Morgan, Snyder, and Struxness 1961). In July 1961, counting activities
in the WBC facility were suspended to modify the detector arrangement with the installation of an 8-
by 4-in. Nal (T!) crystal to replace the smaller detector to increase efficiency and reduce counting
times. In 1962, several thin (5-in.-diameter by 0.0625-in.) Nal (TI) crystats were installed in the facility
to quantify low-energy photons (e.g., mainly isotopes of plutonium). Also in that year, calibration
studies were conducted using an arc-shaped geometry. This geometry was expected to result in less
variation in counting efficiency than the chair geometry. (The arc geometry placed an individual laying
in an arced position with the anterior portion of the body facing the detector at a distance of about 1

m, so each portion of the body was approximately the same distance from the detector (Mani 1983).

Morgan, Snyder, and Struxness (1963) indicated that the thin crystal detectors could see
approximately 40 nCi of *°Pu, if there was a preexposure chest count. if there was no preexposure
count, the detection capability was approximately 80 nCi. In 1963, a SBC replaced the chair
geometry using the 8- by 4-in. Nal (Tl) crystal for whole-body counting. The bed and individual were
moved under the stationary detector. This geometry was used to determine roughly the part of the
body in which the gamma-emitting radioisotopes were located. The detection efficiency was
approximately equivalent to that of the chair geometry (Morgan, Snyder, and Struxness 1963).
Morgan, Snyder, and Struxness (1963) noted that the computer output provided a “gross” spectrum
as well as having the ability to “strip the “°K and "Cs background counts.” Several spectra observed
during that period indicated a large portion of the “net” spectra with negative values. Morgan, Snyder,
and Struxness (1964) noted that in March 1963 “written weekly reports of in vivo counting activities
and results was begun.” (The research to generate this document did not find any of these reports).

Morgan, Snyder, and Struxness (1965) reported that “Baseline counts on essentially every person
with a potential for future exposure was completed in May 1965.” By the third week of May 1965, a
Prioritizing system for selecting individuals for whole-body counting was initiated. Table 5-12
describes this system. Although several located documents stated that baseline and specified
monitoring frequencies were utilized to make in vivo measurements, Berger (2003) and McLaughlin
(2004) indicated that a full in vivo monitoring program did not exist at ORNL until approximately 1994,
when site internal dosimetrists became responsible for identifying personnel for counting. Prior to that
time, the area health physicists selected individuals for in vivo monitoring. The area health physicists
were responsible for determining radioisotopes to which the worker could have been exposed and
counting frequencies. This fed to inconsistent approaches to the selection of individuals for
monitoring.

Table 5-12. 1965 selection criteria for whole-body counting.

Priority Selection criteria
4 Persons suspected of having sustained exposure
2 Persons being recounted as follow-up to initial elevated in vivo results
| 3 Persons who work directly with radioactive material (once every 3 mo)
4

Persons who work in areas where radioactive materials are handled, but do not work directly with
the material (once every 6 mo)

5 New hires or other persons requiring a baseline count and a limited number of persons prior to
L_ termination

Morgan, Snyder, and Struxness (1966) indicated that the “job of determining which employees should
be counted, the frequency with which they should be recounted, and preparing the necessary
schedule cards and lists is requiring more time than had been anticipated.” For this reason, the WBC



Document No. ORAUT-OTIB-0034 [Revision No. 00__| Effective Date: 12/13/2005 | Page2of29 |

PUBLICATION RECORD
EFFECTIVE | REVISION
DATE NUMBER DESCRIPTION
12/13/2005 | 00 New technical information bulletin for assignment of X-10 internal

doses based on coworker bioassay data. First approved issue.
Training required: As determined by the Task Manager. Initiated by
William E. Kennedy, Jr.



{Document No. ORAUT-TKBS-0012-5 | Revision No. 01 __| Effective Date: 10/01/2007 | Page 440f74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 10 of 36

1E+l —

1TE+O

El

‘Activity (4pnv24h sample)

1E-2

+ Sample MDA

1E-3
1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Visual Comparison of Data to Estimated Am-241 MDAs in Urine

Figure A-1. MDA data for americium-241 in urine (part 1 of 3).

Annual Average MDAs for Am-241 in Urine
Calculated from Recovered Data Between 1945 and 1990

MDA (dpm?24 h)

Figure A-1. MDA data for americium-241 in urine (part 2 of 3).



[Document No. ORAUT-TKBS-0012-5 | _ Revision No. 01__| Effective Date: 10/01/2007 | Page 330f74 |

Brown, P. E., G. R. Patterson, and H. H. Abee, 1971, The ORNL Whole Body Counter, CF-71-1-71,
Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Cox, A. C., 1944, Progress Report - First Nine Months Operation of the Pilot Plant, CN-1879, October
9.

Dixon, E., 2004, personal conversation discussing the historical records and their retention, April
2004.

Evans, R.D., 1955, The Atomic Nucleus. McGraw-Hill, New York, New York.

Hembree, W. R., 2004, personal conversation discussing the operation of the WBC facility during the
late 1980s to present, January 2004.

Henley, L., 2003. Personal communications with Mike Bollenbacher, Knoxville TN, October 15.

ICRP (International Commission on Radiological Protection), 1959, Report of Committee !I on
Permissible Dose for internal Radiation, |CRP Publication 2, Pergamon Press, London,
England.

ICRP (International Commission on Radiological Protection), 1966, “Deposition and Retention Models
for Internal Dosimetry of the Human Respiratory Tract,” CRP Committee II, Task Group on
Lung Dynamics, Health Physics, Volume 12.

ICRP (International Commission on Radiological Protection), 1975, Report of the Task Group on
Reference Man, Publication 23, Pergamon Press, Oxford, England.

ICRP (international Commission on Radiological Protection), 1994, Human Respiratory Tract Model
for Radiological Protection, Publication 66, Pergamon Press, Oxford, England.

Mani, K. V., 1983, Assessment of Oak Ridge National Laboratory Internal Dosimetry Data for DOE
Health and Mortality Study, master’s thesis, 1983.

McLaughlin, D. A., 2002, Oak Ridge National Laboratory Internal Dosimetry Program Technical Basis
Document, Revision 4, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

McLaughlin, D. A., 2004, personal conversation discussing the operation of the WBC facility, January
2004.

MMES (Martin Marietta Energy Systems, Inc.), 1995, Documentation of the dBase IV In vitro Sample
Databases Supporting the ORNL Historic Work Force Dose Assessment Project.

Morgan, K. Z., 1961, Health Physics Division Annual Progress Report for Period Ending July 31,
1961, ORNL-3189, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Morgan, K. Z., W. S. Snyder, and E. G. Struxness, 1963, Health Physics Division Annual Progress
Report for Period Ending July 31, 1963, ORNL-3492, Oak Ridge National Laboratory, Oak
Ridge, Tennessee.

Morgan, K. Z., W. S. Snyder, and E. G. Struxness, 1964, Health Physics Division Annual Progress
Report for Period Ending July 31, 1964, ORNL-3697, Oak Ridge National Laboratory, Oak
Ridge, Tennessee.



[Document No. ORAUT-OTIB-0034 | _RevisionNo. 00 __| Effective Date: 12/13/2005 [ Page 26 0f29 |

* annual bioassay
value (50th percentile)
not used in fit

fit

Figure A-19. Assumed “Pu intake, 1953 to 1968, 50th-percentile results,
Type M.

+ annual bioassay
value (50th percentile)
© notused in fit
fit

Figure A-20. Assumed “Pu intake, 1969 to 1984, 50th-percentile results,
Type M.

* annual bioassay
value (50th percentile)
© notused in fit
fit

Figure A-21. Assumed Pu intake, 1985 to 1988, 50th-percentile results,
Type M.


eo

Dose Reconstruction
Project for NIOSH

Oak Ridge Associated Universities 1 Dade Moeller & Associates | MJW Corporation

Page 1 of 35

Type of Document: OTIB

Document Title: Document Number:  ORAUT-OTIB-0018
Revision: 01

Internal Dose Overestimates for Facilities with

Air Sampling Programs Effective Date: 08/09/2005

Supersedes: Revision 00

Subject Experts: Elizabeth M. Brackett and Donald E. Bihl

Document Owner

Approval: Signature on File Approval Date: 08/05/2005
Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 08/04/2005
Richard &. Toohey, Project Director as

Approval: Signature on File Approval Date: 08/09/2005
James W. Neton, Associate Director for Science TT

1 New [1 Total Rewrite & Revision 0 Page Change

DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE REPLACE THE PRIOR REVISION.

PLEASE DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.


Special Exposure Cohort Petition
under the Energy Employees Occupational
MIness Compensation Act

U.S. Department of Health and Human Services
Centers for Disease Control.and Prevention
National Institute for Occupational Safety and Health

OMB Number: 0920-0639 Expires: 09/20/2013
Pi Zot7

Special Exposure Cohort Petition — Form B
Survivor Information — Complete Section B if you are a Surviv

r representing a Survivor.

8.1. Name of Survivor:
Mr/Mrs./Ms. First Name Middle Initial Last Name

B.2 Social Security Number of Survivor:

8.3 Address of Survivor:
Street Aot# P.O. Box
City State Zip Code

8.4 Telephone Number of Survivor: , d

8.5 Email Address of Survivor:

B.6 Relationship to Employee: _Ispouse [_Json/Daughter [Parent

[| [__]Grandchila

ployee Information — Complete Section C UNLESS you are a labor organization.

C.1. Name of Emplovee:
Mr/Mrs./Ms, First Name ~ Middle Initial Last Name
C.2 Former Name of Employee (e.g., maiden name/tegal name change/other):
Mr/Mrs./Ms. First Name Middle Initial Last Name
C.3 Social Security Number of Employee:
C.4 Address of Employee (if living):
Street Apt # P.O. Box
City State Zip Code
C.5 Telephone Number of Employee: ) :
C.6 Email Address of Employee:
C.7 Employment Information Related to Petition:
C.7a Employee Number (if known):
C.7b Dates of Employment: Start 1945 End 1975
C.7c Employer Name: EL. DuPont (for Univ. of Chicago), Monsanto Chemical Company, Union Carbide
C.7d Work Site Location: Oak Ridge National Laboratory (X-10)
C.7e Supervisor's Name:

Name or Social Security Number of First Petitioner: _



[Document No. ORAUT-TKBS-0012-5 [Revision No. 01 _| Effective Date. 10/01/2007 | Page 600f74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 26 of 36

Number of Sample Results Recovered per Year
Between 1945 and 1990 for Rare Earths in Urine

‘Number of Samples (N)
o § 8 &
953 os
|
| |
973 kam
974 “pmai

Figure A-11. MDA data for rare earths in urine (part 3 of 3).

1EH3 po

LEH

E+

‘Activity (dpm/24 h ample)

1 EW

1E-}

1945 1950 1955 1960 1965 1970 1975 1980 1985

Year

Visual Comparison of Data to Estimated Sr-90 MDAs in Urine

1990

Figure A-12. MDA data for Sr-90 in urine (part 1 of 3).



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01__| Effective Date: 10/01/2007 | Page 64 of 74 ]

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 30 of 36

Annual Average MDAs for U-238 in Urine
Calculated from Recovered Data Between 1945 and 1990

3

MDA (dpm/24 h)

Figure A-14. MDA data for U-238 in urine (part 2 of 3).

Number of Sample Results Recovered per Year
Between 1945 and 1990 for U-238 in Urine
1000

3

3

3

3

3

Number of Samples (N)
gs

3

Figure A-14. MDA data for U-238 in urine (part 3 of 3).



[L_ Document No. ORAUT-OTIB-0034 | Revision No, 00 | Effective Date: 12/13/2005 | Page 20of29 |

* annual bioassay
value (50th percentile)
* not used in fit
fit

Figure A-8. Assumed ~“U intake, 1951 to 1987, 50th-percentile results, Type F.

Table A-7. Type F ““U intake periods and rates.

Type F TMU intake rate (dpm/day) Type F TM“U intake rate (dpmiday)

Year | 50th percentile | 84th percentile | GSD Year | 50th percentile | 84th percentile | GSD
1951 11.19 28.28 2.53 1970 0.413 1.272 3.08
1952 11.19 28.28 2.53 1971 0.413 1.272 3.08
1953 11.19 28.28 2.53 1972 0.413 1.272 3.08
1954 11.19 28.28 2.53 1973 0.413 1.272 3.08
1955 11.19 28.28 2.53 1974 0.413 1.272 3.08
1956 4.213 28.28 6.71 1975 0.413 1.272 3.08
1957 1.942 17.97 9.25 1976 0.413 1.272 3.08
1958 1.942 17.97 9.25 1977 0.413 1.272 3.08
1959 5.171 17.97 3.48 1978 0.413 1.272 3.08
1960 5.174 17.97 3.48 1979 0.413 1.272 3.08
1961 5.171 17.97 3.48 1980 0.413 1.272 3.08
1962 1.881 12.05 6.41 1981 0.413 1.272 3.08
1963 1.881 12.05 6.41 1982 0.413 1.272 3.08
1964 0.413 1.272 3.08 1983 0.413 1.272 3.08
1965 0.413 1.272 3.08 1984 0.413 1.272 3.08
1966 0.413 1.272 3.08 1985 0.0957 0.227 2.37
1967 0.413 1.272 3.08 1986 0.0957 0.227 2.37
1968 0.413 1.272 3.08 1987 0.0957 0.227 2.37
1969 0.413 1.272 3.08



AECU-817 83

in addition to the protective clothing normally worn by the personnel in the areas that have been
discussed, extra items of wearing apparel such as overshoes, shoe covers, gloves, and caps are kept
in limited numbers for messy emergency jobs.

1, The overshoes are not satisfactory at present. '

2. Overshoes are used in liquid decontamination jobs.

3. Shoe covers are used in areas where walking surfaces are dry.

4, The canvas gloves are used normally where objects to be handled are dry.

5. Rubber gloves of all types are used in liquids.

6. Gloves and cloth caps are expendable and are seldom usable after being laundered,

Lab coats are frequently worn by supervisors in the various areas, A large number of super-
visors fee] that they do not work directly with the hot material enough to warrant the wearing of
protective clothing, The lab dresses and slack-coveralls are worn by the female employees in the
various laboratories.

Protective Clothing for Maintenance Men

The maintenance personnel are not furnished with free issue shoes or coveralls. It is up to the
Health Physics Division to recommend on work orders the wearing of shoe covers or overshoes.
When the maintenance man does get his persona] shoes contaminated over the tolerance value, the
company will issue him a pair of safety shoes, and we shall confiscate or store his own shoes, Ifa
hot job is to be performed, the Area Supervisor in the area where the job is to be performed will
issue the maintenance men coveralls, gloves, overshoes, caps, etc. The health physics men probe
the maintenance men’s personal clothing carefully because that is principally what they wear at work.

Protective Clothing: Spreader of Contamination

Although we use protective clothing to keep the radioactive materials from our skin, it is one of
the greatest spreaders of contamination that we have. Shoes track the contamination all over the
plant. The gloves we wear can contaminate hand rails, telephones, chairs, etc. Some areas have
containers for grossly contaminated clothing and keep this separate from mildly contaminated cloth-
ing. This helps protect laundry personnel.

Protective Clothing Program at Present

Our protective clothing program at present is only about 33 1/3 per cent efficient for the fol-
lowing reasons:

1. There are not enough regulations concerning the necessity of wearing protective clothing

2. Enforcement of what rules we have is very lax

3. Some supervision professes only weak interest in the protective-clothing program

How the Protective-clothing Program Can be Remedied

The protective-clothing program can be remedied by zoning as follows: (1) cold areas, (2) cool
areas, (3} warm areas, and (4) hot areas. The clothing requirements should be set up for these areas,
and then supervision should be allowed to enforce the rules with common sense and a little psychology.

Laundry in Relation to Protective Clothing

Clothing Stores personnel mark all garments prior to issue with the station number followed by
the badge number with nonwashable ink. This is the means of identifying each garment.

When protective clothing is picked up, it is delivered to the laundry receiving station. There
the clothing is listed by station and badge number and passed on, in hampers, to the Health Physics
Division checking station.

Health physics technicians make a prewash check — beta-gamma always and alpha in some in-
stances. Eighty-five per cent of the clothes pass this prewash check, are sent to the laundry for a



AECU-817 43

Fig. 4.4

Figure 4.4* isa gas characteristic curve, where Qo is a function of the high voltage across the
chamber.

When the chamber is operated at voltage V, the charge collected, Qo, is relatively independent
of the chamber voltage, and all the ions, Q,, produced by a particle are collected (Q, = Qz). The
pulse voltage output

Typical values of V range around 0.001 volt. The operation of a chamber in the region of V, requires
a high-gain amplifier (approximately 300,000) which introduces the probability of stray counts due to
mechanical shock and electrical disturbances.

Raising the voltage to V2 causes the chamber to reach the proportional operating region. This
is usually a high voltage for a parallel-plate counter, but the proportional region can be reached.
The counter is in the proportional region when the field intensity is so high that the electrons pro-
duced by radiation receive enough energy to produce ion pairs,

To obtain a high field intensity requires a lot of voltage on a parallel-plate chamber, but a cyl-
inder with a coaxial fine wire requires much less. This is the usual type.

Proportional Ptocess. A particle forms ion pairs in the gas. Electrons are attracted toward
the center wire, and as they get close to the center wire the attraction increases, and the velocity
between collisions increases until the electrons acquire enough energy to produce ion pairs.

Qg, the collected charge, is now different from Q), the charge induced by a particle. The ratio
Q2/Q; is called the gas amplification factor ‘‘A.’’ Now

Q2
Ve but Q, = AQy

A

c

Practical gas amplification factors may be 500 or more. The gain of an amplifier is defined as
the ratio of the output voltage to the input voltage or

but

— a.
‘Reducing chamber pressure moves curve to left.



AECU-817 2

tron, in the ranges in which we are most interested, may be considered to be proportional to the
fourth powers of the atomic numbers of the atoms to which electrons are attached. Certain of the
body tissues, particularly the skin and the bones, contain elements of higher atomic number which re-
sult in higher absorption of energy from soft gamma or X radiation than occurs in either air or soft
tissue,

Perhaps the variation of energy per gram per roentgen of dosage should be considered as a de-
fect of the roentgen as a unit of dosage. In any event, when considering dosage from radioisotopes
distributed through body tissues, it is more convenient and, so far as is known, just as significant
to measure dosages in terms of energy absorbed per gram of tissue. To afford a rough comparision
between these two methods of expressing dosage, the term “roentgen equivalent physical,’ custom-
arily referred to as the rep, has been widely used. As commonly used, 1 rep corresponds to the
absorption in tissue of the same quantity of energy (83.8 ergs/g) as would be absorbed in air from
a dosage of 1 r. (Since this lecture was given the rep has been defined in terms of a larger energy
unit. The new rep corresponds more nearly to the energy absorption in tissue when the energy ab-
sorption in a small air cavity in the tissue is 83.8 ergs/g. At ORNL the rep is defined in terms of
an energy absorption of 95 ergs/g of tissue.)

It should be obvious from the preceding discussion that a tissue dosage of 1 rep is generally
only roughly equivalent to a tissue dosage of 1 r. However, a practical approach to the problem
is to assume as a working basis that tissue tolerance can be expressed in terms of energy per
gram. (This assumption is as well grounded as the assumption that tolerance for penetrating radia-
tion can be expressed in roentgens.) It is customarily assumed, also, that an exposure of 0.1 rep/24-
hr day constitutes a tolerance exposure. These assumptions are made in subsequent discussion,

One curie of a radioisotope is defined as a quantity of the isotope such that it disintegrates at
the rate of 3.7 x 1010 atoms/sec. This is equivalent to a rate of disintegration of 3.2 x 1015 atoms/24-
hr day. If the average energy per disintegration is E Mev, the energy liberated in a 24-hr day by 1
curie of the radioisotope is 3.2 x 10!5 & Mey. since 1 Mev = 1.6 x 10-6 erg, this is equivalent to
5.1 x 109 & ergs.

Relationship between Concentration of a Radioisotope in Tissue and Dosage Rate

For simplicity, let us consider a tissue of uniform composition, in which the atoms of a radio-
isotope are uniformly distributed, and of which any linear dimension is large compared to distance
in the tissue required for the radiations from the atoms of the isotope to be effectively absorbed. In
such a case, except near an edge of the tissue, any small portion of the tissue will absorb just as
much energy as the radioactive atoms in that portion radiate (see Appendix B). If the radioisotope
is uniformly distributed through the tissue, in most cases this relationship gives a good approxima-
tion to the rate of absorption of energy in the tissue. If the tissue has a small linear dimension, the
above assumption leads to a computed value for the absorption of radiation which is larger than the
actual value,

From the above relationship, the dosage rate resulting from a radioisotope uniformly dispersed
through a tissue is computed as follows:

Let C = number of curies of the radioisotope in the tissue

E = energy per disintegration, measured in Mev
W = weight of tissue, measured in grams

R = dosage, measured in rep

t = time, measured‘in days

Then the total energy radiated per day measured in Mev is 3.2 x 1015 CE, and the energy per
gram per day is 3.2 x 1015 CE/w.

Since 5.23 Mew/g is required to give a dosage of 1 rep, the dosage rate in rep per day is

R _ 3.2x 1015 CE/W Mev/g/day
t ~ 5.23 x 10 Mev/¢/rep

6.1 x 107 CE/W rep/day



[Document No. GRAUT-OTIB-0018

[ Revision No. 01

[Effective Date: 08/09/2005 | Page 19 of 35 _|

Table 7-1(Continued). Whole-body count (WBC) results below which this TIB
method is still considered an overestimate.

Absorption Time since ‘Organ oF WEC resulfmust
Radionuclide type last WBC (d)* concern be less than (nCi):
Mn-54 ForM 90 Adrenals 24
Urinary bladder 40
Brain 42
Breast 23
Gall bladder 35
Heart w all 20
Kidneys 29
Liver 47
Muscle Ei)
Ovaries 31
Pancreas 30
Testes 65
Thyroid BT
Red bone marrow 70
Bone surface 113
Stomach 39
Ss! 42
ULT 45
ttl 100
‘Skin aT
‘Spleen 39
“Thy mus 26
Uterus 43
ET 15
Tung TSO
Colon a
Esophagus 25
Mn-54 ForM 180 Adrenals 8
Urinary bladder’ 17
Brain 14
Breast 16
Gallbladder 12
Heart wall 13
Kidneys 10
Liver 16
Muscle’ 14
Ovaries 11
Pancreas 10
Testes 22
Thyroid 18
Red bone marrow 24
Bone surface 39
‘Stomach 14
si 11
ULT 15
tll 35
Skin 16
Spleen 16
“Thy mus 17
Uterus 1S
eu 6
Tung 117
Colon 24
Esophagus 17



18 AECU-817

when necessary, severe contamination could be kept to a minimum, and it might not pay to have the
investment in laundry equipment. Cheap and effective expendable garments may be devised that are
to be worn only once.

One way of salvaging garments is to put those garments with short-lived radioactive contamina -
tion in storage so that the activity may decay. It is debatable whether such a storage procedure is
economical, It does not seem feasible to hold garments for more than 6 months. At one time shoes
that were confiscated were kept for 6 months or more, but, invariably, when the shoes were safe to
give back, they were not fit to wear or the employee had terminated.

Agarment’s life is about 15 to 25 washes if the cases of contamination are relatively mild.

There ig the additional cost of checking, pickup and delivery, and segregation. Garments are
now assigned to individuals, Originally they were assigned to operating areas. Protective garments
should be written off the inventory when issued. Here at Oak Ridge where we have an ‘‘academic
atmosphere,’ the clothes must be pressed. This adds to the cost.

Fifteen to 20 per cent of the clothes actually have a decontamination wash. Contaminated clothes
are given three washes before casting them out if they fail to respond.

There is also a problem of laundry waste and laundry fumes. Garments which the laundry has
been unable to decontaminate satisfactorily are buried. Storage and checking add to the cost con-
siderably and may not be economicai.

Monitoring Techniques. Hanford uses an adaptation of the fourfold counter, This is for gamma
and hard betas. Alpha contamination is best checked by Poppy. Depending on the thoroughness of
inspection, it requires from 4 to 30 min to check a pair of coveralls with Poppy.

6.6 HEALTH PHYSICS PUBLIC RELATIONS*
By W. H. Ray

IL. Health Physics objective: reduce radiation injury to the individual and to the race to a minimum
IL, Means of attaining objective
A, Education
1, Intelligent people when informed can do much to take care of themselves
2, Many can be made to feel 2 moral responsibility for prevention of overexposure to others
3. Propaganda can develop beneficial protective habits through mob psychology
B. Guidance
i. Surveyor surveillance
2. Professional consultation
3. Procedures and standards
C. Coercion
1, Rules and laws
2. Line organization and directives
Il. Techniques of applying means
A. Salesmanship
1, Character
a, Absolute honesty
b. Dependability
c. Keep confidences
2. Magnanimity
a. Serve others
b. Do more than is expected
¢. Do not ‘‘pass the buck.’’ Feel responsible
3. Personality
a. Alertness

*This is substantially the same lecture given in a previous Health Physics Survey Course by
W. H. Ray, November 24, 1947.



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01

[Effective Date: 10/01/2007 | Page 70 of 74

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 36 of 36

Annual Average MDAs for Th-232 in Feces
Calculated from Recovered Data Between 1945 and 1990

MDA (dpmi24 hy

Figure A-18. MDA data for Th-232 in feces (part 2 of 3).

800

Number of Samples (N)

Number of Sample Results Recovered per Year
Between 1945 and 199@ for Th-232 in Feces

Figure A-18. MDA data for Th-232 in feces (part 3 of 3).



AECU-817

project, assuming that so many persons had been killed with 2 Ib of radium (this was all the radium
that had been produced at that time). They asked how many persons would be expected to die as a
result of pile operations with the equivalent of millions of pounds of radium. Fortunately it was de-
cided to go ahead with the Plutonium Projects, but to put forth an unprecedented effort for radiation
protection in addition, .

Development in pile research and operation of piles has developed simultaneously with the field
of Health Physics. Initially it was thought that Health Physics would be a medical problem, and, as
a result, we reported to the medica] group here when Clinton Laboratories began operations. Later
it was found that if we did our part appropriately, there would be little need for medical advisors on
problems of damage.

One by one the Health Physics Divisions throughout the Plutonium Projects have obtained ma-
turity and have become full-size divisions. This has been true primarily at the Argonne National
Laboratory, Hanford Works, and at Oak Ridge National Laboratory (ORNL). At the University of
Chicago the first Health Physics Division was organized under the direction of E. O. Woolan, who
is now Assistant Director of the Physics Division here. J. C. Hart, R. R. Coveyou, a few others,
and myself joined that group in 1943. H. M, Parker and C. Gamertefelder had joined it in the latter
part of 1942. During the fall of 1943, Parker, Coveyou, Gamertsfelder, Hart, and myself came to
ORNL (then Clinton Laboratories), and we set up a Health Physics Division and began a training
program for personnel who were scheduled to move out to Hanford. it was our opinion at that time
that in a Health Physics organization all that was needed was a few senior men to serve in an ad-
visory and consultation capacity, and all the survey work would be done by Operations. We have had
to change that philosophy.

Radiation is actually no more dangerous than any other hazard so long as safety rules are com-
plied with. The only difference is that we do not have a physical sense which tells us when we are
receiving an exposure.

We were made to realize from our early experiences that operations and research personnel
often became so impressed with the work they were doing that they were inclined at times to ignore
certain hazards, Research personnel who thought they knew all the answers were sometimes the
worst, because they had the attitude that it was their necks if they wanted to take chances.

With the help of the Manhattan District and the University of Chicago, our local contractor, we
developed a new philosophy in that we actually did do a great deal of the service work out in the field.
In addition to pocket meters and film badges (our only service function then), it was decided to train
a group of Health Physics surveyors in the use of instruments and interpretation of instrument read-
ings and see that radiation protection measures were enforced. A month or so after operations got
under way at ORNL, this new philosophy was developed, and it has been held ever since.

During the past year our philosophy has changed again somewhat. We have been convinced that
we have been doing too much of the scrub-up type of work for operations and research and that when
they make a mess, the best way to impress on them that they should be more careful is to let them
clean it up themselyes, At Chicago, during the past two years, they have had a special clean-up squad
to wash up radiation messes. Here and at Hanford we feel that a chemist or top-notch operations man
should do his own clean-up work, and therefore spills do not occur very often. Therefore we have
encouraged the research personnel and operations to do these routine jobs wherever possible, and
we have been pleased to have their full cooperation, If equipment is to be decontaminated, we show
them how to use the instruments, We hope to have the research and operations personnel so trained
and equipped with instruments in the near future that they can do their own clean-up work, This is
not doing Health Physics out of 2 job, since it gives us more time for other things that are more im-
portant to be done.

The problems presented by such an operation as we have here are somewhat different from the
problems of the hospital radiologist. Here we not only have alpha, beta, gamma, and X rays, but we
also have fast neutrons and thermal neutrons. In addition to these particles, we have the recoil fis-
sion elements and the neutrinos. We also have epithermal neutrons. However, we do not have much
knowledge of the biological damage associated with neutrinos or epitherma] neutrons,

Dr. Morgan then outlined the principal objectives of the Health Physics students:



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 __| Effective Date: 10/01/2007 | Page 390f74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 5 of 36
Table A-1. Data used to determine historical MDAs for radionuclides in urine and feces.
: Nuclide_} Fecal | Urinalysis | Row totals Nuclide | Fecal | Urinalysis | Row totals
: | Totals 1,859 $9,288 61,147 Pb-210 2 2
Alpha 1 3 4 Pm-147 16 64 80
Am-241 130 5,540 5,670 Po-210 E} 57 66
Am-243 6 6 12 Pu-238 4 61 65
As-74 7 rd Pu-239 124 15,352 15,476
| BG 2 2 Pu-241 1 1114 112
Bk-249 2 12 14 Pu-242 41 41
Br-82 2 2 Ra-226 4 332 333.
Br-83 2 2 Ra-228 4 1
C-14 14 11 RE 32 698 737
Ca-45 4 4 Ru-103 4 1
Ce-144 4 33 37 Ru-106 65 65
Cf-249 2 41 3 $-35 10 10
Cf-252 3 1 14 Sb-125 1 1
Cl-36 41 1 Sm-154 1 1
Cm-242 41 14 12 Sr-85 1 1
Cm-244 26 273 299 Sr-89 37 37
Co-60 2 81 83 Sr-90 142 12,751 12,893
Cs-134 1 4 T 1 1
Cs-137 4 3,557 3,561 Te-99 19 19
Fe-59 1 8 9 Th-232 1,079 46 4,125
GA 69 1,875 1,944 T2014 1 1
G-alpha 6 6 TI-204 1 4
GAO 154 2,771 2,925 Tm-170 4 5 6
GB 278 278 TPU 1 1
GBO 46 46 TRE 2 197 199
H3 2,070 2,070 TREB 4 1
1434 4 a TRE-B 4 4
Mn-54 2 2 U-232 1 1
Mo-89 1 7 U-233 3 826 620
Na-24 3 3 U-235 3 3
No-95 3 3 U-238 26 11,409 11,434
Np-237 1 54 55 U-239 ch 11
P-32 166 166 Y+RE 157 157
Pa-231 1 54 55 Y-88 5 5
Pa-233 3 13 16 Y-90 31 31
Pa-234 1 4 Zn-65 7 7
21-95 1 19 20

To use this equation one must know, or be able to calculate, the standard deviation of background
counts, the duration of the count, the yield of any chemical extractions used, the efficiency of the
detector used, and the amount of the sample analyzed. The method of recovering this information
from the historical database is discussed in the following sections.



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 __| Effective Date: 10/01/2007 [ Page 47 of 74 ]

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 13 of 36

1 E38

: + MDA (apm/24 hy
‘ Estimated MDA For the Period
: [+ = 50th Percentile

1E+2 4

Activity (dpmn/24 bs sample)

E+ 4

1E+O
1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Visual Comparison of Data to Estimated Cs-137 MDAs in Urine

Figure A-3. MDA data for Cs-137 in urine (part 1 of 3).

Annual Average MDAs for Cs-137 in Urine
Calculated from Recovered Data Between 1945 and 1990
350

250 +

MDA (dpm/24h)

504

Figure A-3. MDA data for Cs-137 in urine (part 2 of 3).



AECU-817

UNIT 4. RADIATION-DETECTION METHODS 38

4.1 Yonization Mechanism
By R. 8. Thackeray
4.2 Yonization Chambers
By F. J. Davis
4.3 Gelger-Mueller Tubes and Scaling Circuits
By G. 8. Hurst
4.4 The Proportional Counter
By R. A. Dandl

UNIT 5. RADIATION-DETECTION TECHNIQUES 46

5.1 Counting Techniques
By P. E. Brown
5.2 Probable Error and Confidence Intervals
By W. 8. Snyder
5.8 The Pocket Meter and Projection Minometer
By 1 8. Abee
5.4 Survey Instrument Usage and Selection (I}
By E. J. Kuna
5.5 Survey Instrument Usage and Selection (1)
By G. R. Patterson
5.6 Survey Instrument Usage and Selection (III)
By R. L, Clark, E. L. Sharp, and P. E. Brown
5.7 Radiation Measurements with Film
By J. H. Roberson
5.8 Calibration Technique
By D. M. Davis
5.9 Instrument Care and Repair
By D. M. Davis and Staff
5.10 The Photographic Emulsion
By M. M, Shapiro

UNIT 6. APPLIED HEALTH PHYSICS 69

6.1 Personnel Monitoring Procedures
By J.C. Hart
6.2 Exposure Reporting and Personnel Monitoring Records
(Demonstration of Personnel Monitoring Techniques)
By J. C, Hart
6.3 Measurements on Internal Exposure
By L. B. Farabee
6.4 Waste-disposal Theory and Practice
By R. J. Morton
8.5 Decontamination Lauuiry
By W. H. Ray
6.6 Health Physics Public Relations
By W. H. Ray



82 AECU-817

many cases where signs are left for weeks and months, and personnel begin to pay little or no atten-
tion to them. If they are kept up to date, personnel will respect them. Rope barriers, i.e., rope off
an area and put tags on it. In all cases you want to rope the area or barricade off at or under the
TMmaximum permissible level. In the case of a permanent hazard, put up a permanent barrier to keep
people out of the area. You might lock existing doors and place signs on them and give supervision
the responsibility for locking or unlocking the doors. In an extreme emergency you can call out a
patrolman, and he will stand at the door and keep people away from it. You might in some cases
paint lines on floors to distinguish between contaminated areas and areas free of contamination. This
will aid materially in preventing the spread of contamination,

Contamination Control. This particular part of the surveyor’s work in the field takes up more
of his time than any other one thing, even radiation control, The best means to control contamination
is to prevent it in the first place. If you prevent it, you have it licked right then. In many cases, and
in the majority of cases, if you know your area, you will know, in general, in advance when to expect
contamination and can guard against it before it occurs. In some cases contamination cannot be pre-
vented, but it can be kept at a minimum, Try to arrange consultations with the person who is handling
the area even before he attempts the job. In many cases he will come to you. The jobs that cause
you the most trouble are the little ones. The big ones are carefully thought out, and, as a result,
fewer accidents happen. On the small jobs that normally occur in that area, personnel usually will
not come to you and ask. It is up to you to find them.

Other than preventing contamination before it occurs, perhaps the next best way to control it is
to conduct thoroughly regular routine surveys of the building. St is a monotonous job, but usually you
come upon something unexpected, Many hazards have been found in this way.

T believe it is safe to say that in most areas the protective clothing rules and regulations are
good enough if they are enforced, In most areas protective clothing is available. Friskers, in many
cases, have helped to control contamination. Instruments help control contamination. Monitrons and
integrons, when showing a high level of activity, have quite often been the only indication of a spill.
If the employees would take showers and clean up before leaving the area, it would help to keep con-
tamination localized, Anything going to the shops, stores, or outside the area should be checked be-
fore leaving.

‘We have had cases where a job has been carried out without a single exposure. This is a result
of the Operation’s Superintendent calling meetings to plan the work and following through on all
recommendations submitted by the Health Physics Division, This saves exposures, contamination,
and cleanup.

In order to complete the job of guarding hazardous areas and controlling contamination, you
must first know your particular area, Another requirement is to know the men you work with, The
better you know the men, the better you can get your particular job done. Know the rules that apply
to your area, Above all, do not undersell your job.

8,2 FULFILLING CLOTHING NEEDS
By A. D. Warden and J. T, Sutherland

E. L. Sharp gave a brief discussion of protective clothing in another lecture, I shall attempt to
go into the protective clothing needs a little more in detail.

The words “‘protective’’ for clothing in this particular case is a misnomer. The words ‘‘expend-
able clothing’ might be a better term to use. We may get a small amount of protection from soft
betas, but we cannot expect clothing to stop very penetrating radiation.

Let us look at protective clothing in this light. ‘It is much easier to peel off clothing than it is
to peel off the skin.” Protective clothing may be easily replaced. Some types of contamination are
extremely hard to remove from the skin. We have had a few cases of hand contamination where it
was necessary to scrub the hands with various solutions until the hands were practically raw, and
yet we could not remove all of the contamination.



[Document No. ORAUT-OTIB-0078 _[ Revision No. 01 | Effective Date: 08/09/2005 | Page Sof 35_]

Most sites performed total alpha and total beta counting on air samples, using either a time delay or
mathematical correction to account for radon progeny. Radionuclide identification was either not
performed or was performed occasionally as a check on basic assumptions from knowledge of the
facility source terms and processes. Trigger levels and prompt decisions for changing access to an
area, usually by placing a “respiratory protection required” restriction for the area, were usually based
on total alpha or total beta air concentrations. Because the radionuclide was not identified for every
sample, the limit for the most restrictive, plausible radionuclide was offen used to establish the trigger
level. For facilities handling transuranics, the most restrictive air concentration for the alpha
measurements was based on plutonium; for facilities with potential exposure to fission or activation
products, the most restrictive air concentration for beta measurements was based on %¢y; at uranium
facilities, the limits for uranium were generally used.

The validity of the assumption that workers’ intakes, excepting accidental intakes, were controlled
such that intakes at limiting air concentrations for a chronic period are reasonable median values is
discussed in Appendix B. Appendix B uses worker bioassay results.

3.0 APPLICABILITY, LIMITATIONS, AND ASSUMPTIONS

3.1 APPLICABILITY

This TIB applies to:

«Sites or facilities that rigorously sampled particulate air concentrations in areas of risk and
controlled exposure to intakes according to the measured concentrations.

« Employment between 1953 and the present. (See first bullet in Section 3.2 for an exception.)

¢ Claims for which itis likely that the covered employee had no significant intakes of particulate
tadioactive material. (See Section 7.0 for a discussion on insignificant intakes in the context of
this TIB.)

¢ Intakes of particulate radioactive material only.

* All organs except respiratory tract organs for monitored workers and the thyroid; however, it
can apply to the thyroid for uranium facilities or plutonium-only facilities (i.e., where there is no
chance of exposure to radioiodines).

3.2 LIMITATIONS
This TIB has the following limitations:

* Itdoes not apply to employment prior to 1953; however, if the limiting air concentrations for the
site of exposure are known for years prior to 1953, those concentrations can be used in the
manner discussed in Sections 4.0 and 5.0. In addition, see Attachment A. It does not apply
to Nevada Test Site outdoor exposures prior to 1963.

 {t applies only to particulate radioactive material. These particulate intakes are in addition to
any intakes of °H, radioiodines, ‘C, or radon/thoron and their progeny, as applicable.



[Document No. ORAUT-OTIB-0018 | Revision No.01 _| Effective Date: 08/09/2005 |_Page 11 of 35 ]

or

Intake (pCi/cal. d) = (6.6 x 10"4)(air conc.) (3)
For most sites intakes will be assigned for alpha and beta intakes. Only the uranium intake should be
assigned for facilities with exposure to only uranium (or natural thorium), with the exception of years
when exposure was to recycled uranium. For recycled uranium, add intakes as listed in Table 4-2.

Table 4-2. Intakes of contaminants in recycled uranium
as fraction of uranium intake.

Activity fraction of contaminant
Radionuclide {e.g. pCi X/pCi U)

Pu-238 0.06

Np-237 0.005

Tc-99 0.4

Th-232 0.02

Ru-106 0.04

4.2 CHOICE OF RADIONUCLIDE

Because total alpha and total beta activity were measured on the air filters, the measurements
represented all the particulate activity being breathed (as opposed to bioassay measurements that
usually measure only certain radionuclides such that unmeasured radionuclides have to be accounted
for in addition to the measured radionuclides). To be most accurate, the fractions of the total activity
(alpha or beta) would have to be assigned to each radionuclide in the mixture in the air breathed.

Those fractions were generally not determined on a regular basis and would have varied among sites,
facilities, processes, or even specific work tasks.

Rather than estimating radionuclide fractions, this TIB overestimates the internal dose by assigning
100% of the intake to the single radionuclide that produces the largest dose per unit intake to the
organ of concem. In addition, organ dose depends on the absorption type of the radionuclides. The
radionuclide and absorption type combination that produces this “largest dose” can differ by organ,
time of exposure, and time after end of exposure. Because the annual organ dose is the dose of
interest, and the annual dose varies from year to year for a given radionuclide, the largest dose
contributor could change over the years. For example, for a 10-year chronic intake of beta emitters,
‘34Cs type F delivers the largest dose to the urinary bladder during years 1 through 10. However, if
the date of diagnosis is sometime after the intake period, the assumption of a '°Ru type F intake
during the 10 years of exposure yields the largest dose for years 11 through 14, as does “Sr type F
for all subsequent years. Rather than determining which radionuclide and absorption type
combination produces the largest dose, an additional claimant-favorable assumption was made.
Each year between the start of exposure and date of diagnosis is evaluated individually and the
largest dose is assigned. For the previous example, it would be assumed that the intake was
comprised of 100% type F ‘Cs when assigning years 1 through 10. For years 11 through 14, it
would be assumed that the intake had been entirely made up of type F ‘®Ru, and that it was °°Sr for
years 15 through 65.

Table 4-3 lists the inventory of radionuclides and absorption types from which the combination
producing the largest dose can be obtained. This is the default list at present

Notes: Ac-227 is a progeny of **'Pa in the °U decay chain. tt was probably not present in significant
quantities unless 7°'Pa or 7“’Ac was purposely concentrated; therefore, it was not included in the



[Document No. ORAUT-OTIB-0018 __| Revision No. 07 Effective Date: 08/09/2005 [ Page 150f35 |

Table 7-1 Whole-body count (WBC) results below which this TIB method is
still considered an overestimate.

‘Absorption ‘Time since ‘Organ oF WEC resukmust
Radionuclide type last WBC (d)* concem be less than (nCi):
Cs-T37 F Ww Adrenals Kis
Urinary bladder. 30
Brain 28
Breast 27
Gall bladder 68
Heart w all 31
Kidneys oT
Liver T50
Muscle 30
Ovaries 32
Pancreas 32.
Testes 29
Thyroid 30
Red bone marrow TS
Bone surface 220
Stomach 30
Si 31
ULI 52
peal 30
kin 5
‘Spleen ST
‘Thymus 30
Uterus. 32
Er 69
Lung 640
Colon 87
Esophagus. 30
Cs-137 F 180 Adrenals 39
Urinary bladder 34
Brain 32
Breast 30
‘Gall bladder [si3j
Feartw all 35
Kidneys: 35
Liver 170
Muscle 33
Ovaries 36
Pancreas 36
Testes 33
Tryrod Ku
Red bone marrow 130
Bone surface 250
Stomach 34
SI 36
ULT 38
PCLT 140
Skin’ 29
Spleen 35
Thymus: 34
Uterus 36
ET 78
Tung 20
Tolon os
Esophagus 34



AECU-817 ll

wet te te |

Ee Fy" "ae ae

(en)

~E2DEHS

Fig. 2.7

BEAM OF GOLLIMATING SLOT
euccrnone, | ATTRAGTING SLOT

See Se VAPOR WHERE
f, ee (ONS FORMED
ie) Pi
IP aceerenarne

ATTRACTING SLOT’
6 VOLTAGE

MAGNET OF
KNOWN FLUX

DEFLECTED,
1ONS

Fig. 2.8

PACKING FRACTION

INGREASING ATOMIC MASS.

Fig. 2.9



[Document No. ORAUT-OTIB-0018

[Revision No. 01

| Effective Date: 08/09/2005 [| Page 21 of 35 |

Table 7-1 (Continued). Whole-body count (WBC) results below which this
TIB method is still considered an overestimate.

‘Absorption Time since Organ oF WEC resuk must
Radionuclide type last WBC (d)* concem be less than (nCi):
Nb-95 Mors 90 ‘Esophagus 14
Nb-95 Mors T80 Adrenals 4
Urinary bladder Ex
Brain 12
| Breast 4
Galil bladder 8
art w all Kj
Kidneys 4
Liver 13
Muscle 6
Ovaries 3
Pancreas 5
Testes 16
“Thyroid S
Red bone marrow 14
‘Bone surface 12
Stomach 4
SI 3
ULT 3
jaa 4
KIN Ey
‘Spleen 5
Thymus 4
Uterus 7
ET 1
Tung 3
Colon 3
Esophagus. 4
Nb-95 MorS 365 Adrenals 0.7
Urinary bladder. 0.2
Brain 03
Breast OT
Gallbladder 02
Heart w all 01
Kidneys 0.4
Liver 03
Muscle 0.2
Ovaries OT
Pancreas OT
Testes 04
Thyroid 0.2
Red bone marrow 0.4
Bone surface 0.3
Stomach 0.1
ST OT
ULT OT
Lu OT
Skin’ 0.2
Spleen 07
Thymus: 0.1
Uferus 02
ET 00
Tung OT
Colon 01
Esophagus 07



[Document No. ORAUT-TKBS-0012-5 |_RevisionNo.01_| Effective Date: 10/01/2007 | Page 420f74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 8 of 36

Some MDAs changed frequently, creating a pattern of similar MDAs in short adjacent periods. When
this occurred, the largest estimated MDA from that period was selected to be the representative MDA
for the entire period. The tritium 7H) MDA listed between 1968 and 1981 in Table 5-9 is an example

of this adjustment.

Most recovered data contained isotope-specific entries in the NUCLIDE field. However, the ability to
differentiate routinely between isotopes of the same radioelement did not always exist before 1989.
Many of the isotopic assignments were based on process knowledge. Therefore, during the period
prior to 1989, the MDA for some isotopes are reported as the MDA for the radioelement instead of the
fadionuclide. For example, the alpha emitters **Pu and *°Pu are reported separately in the
database. They have been combined in this TBD as plutonium. NOTE: **Pu is reported separately
because it is a beta emitter that can be measured separately from the alpha emitting isotopes of
plutonium. If the recovered data did not extend to 1989, the MDA calculated from the last known data
was extrapolated forward (see Table 5-9).



92 AECU-817

a. Scrap metal
b. Machinery
ec. Clothing
d, Furniture
e. Mobile equipment
V. Contamination control, decontamination, of equipment in low activity level areas (area outside
Restricted Area)
A. Frequent surveys in all maintenance shops
B. Cafeteria surveys
1, Building smeared and probed
2. People wearing protective clothing are forbidden to eat in the main cafeteria

Part 0
1. Personnel decontamination, general
A. All personnel should be checked following all jobs that involve work with radioactive materials
B. Friskers in all buildings and at all area gates should be kept in good working order so that
employees who are unknowingly contaminated may be detained
C. Probe personnel quickly to determine the extent of contamination
D. If the employee is generally contaminated have him remove his clothing and take an immediate
shower; If hands are badly contaminated have gross contamination removed before the shower
is taken
Tl, Hand contamination
A. Study the Hand Decontamination Procedure enclosed in the Health Physics Hand Decontamina-
tion Kit
B. Employees whose hands are unsuccessfully cleaned should be sent to Medical
Il. Body contamination
A. Health Physics Hand Decontamination Kit contains the necessary reagents
B. Hair contamination is the main problem
IV. Medical correlation with Health Physics
A. Contaminated wounds
B. Hand contamination not reduced
C. Ingested active materials
1. Feces
2, Urine
D. Blood-count checks made on employee receiving high dosage rate

Appendix A: Hand-Out Material Suggested Reagents; Their Use for Alpha, Beta, Gamma Decontami-
nation

Material Treatment

Stainless steel HNO3 washes, varying concentration and temperatures, followed by
NagCO3

Mild steel 10% citric acid, 5% Zepco soap powder, using stiff scrub brushes, fol-
lowed by rinse

Lead HCl preferable to HNO3, plenty of rinse water, hoods desirable to re-
move fumes

Brass Acetone wipes with kleenex to emery cloth, ammonium citrate, trisodium
phosphate, 6N citric acid

Aluminum 10% HNOg with vigorous brushing, synthetic detergents

Glass Small glassware, immersed 60 to 70% HNOg and soaked for several

days, followed by sulfuric ~chromic acid cleaning solution, aqua regia,
10% HNO3, with 1% HF



(Document No. ORAUT-TKBS-0012-5 [Revision No. 01_| Effective Date: 10/01/2007 L Page 50f74__|

LIST OF FIGURES

FIGURE TITLE PAGE

1 Current MDA values for ORNL germanium lung counter versus chest-wall thickness ............ 28
-2 Current MDA values for ORNL phoswich lung counter versus chest-wall thickness for
241m and 7*Pu seseneeece

A-1— MDA data for am
A-2. MDA data for curium-244 in urine.
A-3 MDA data for Cs-137 in urine...

A-4 MDA data for gross alpha in urine
A-5 MDA data for gross beta in urine.
A-6 — MDA data for H-3 in urin
A-7_— MDA data for Pm-147 in urine.
A-8 MDA data for Pu-238 in urine
A-9 MDA data for Pu-239 in urine
A-10 MDA data for Pu-241 in urine
A-11 MDA data for rare earths in urine
A-12 MDA data for Sr-90 in urine..
A-13 MDA data for U-233 in urine.
A-14 MDA data for U-238 in urine.
A-15 MDA data for Am-241 in feces
A-16 MDA data for gross alpha in feces .
A-17 MDA data for Pu-239 in feces .
A-18 MDA data for Th-232 in feces .
8-1 Plutonium in urine samples from 1945 .
B-2 Plutonium urine sample data through April 18, 1945.
B-3 Plutonium urine sample data for April 25-May 30, 1945
B-4 = Alpha air-monitoring results for 1944-1947........



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01__| Effective Date: 10/01/2007 | Page74o0f74 |

ATTACHMENT B
ASSESSMENT OF INTERNAL DOSE AT ORNL: 1944 TO 1947
Page 4 of 4

Because it is not known who wore the assault masks and where specific individuals worked, it is
difficult to postulate a smaller airborne concentration. Urine sample results are somewhat ambiguous
because activity was detected in them but they were likely contaminated during collection or analysis.

1.0E-06
1.0E-07
1.0E-08
1.0E-09 +
1.0E-10 +——
4.0E-11 4
4.0E-12
1.0E-13
1.0E-14 T + T — T

-2 A 0 1 2 3 4

Z Score

Figure B-4. Alpha air-monitoring results for 1944-1947. The geometric mean =
1.4 x 10°? CifmL, the GSD = 6.35, and 7 = 0.97.

Air Sample Result (Ci/mL)

As a best estimate for individuals likely to have a potential for exposure, ORAUT (2005a) values will
be used but the nuclides will be limited to those likely to be present in significant quantities (plutonium
and uranium isotopes), thus eliminating unnecessary conservatism. Due to the extent and quality of
the alpha air-monitoring program, this approach is considered to be applicable to beta/gamma-
emitting radionuclides at ORNL during this period.



[Document No. ORAUT-TKBS-0012-5 | _RevisionNo.01 | Effective Date: 10/01/2007 | Page 68of74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 34 of 36
1E+1
“+ Sample MDA.
Est. MDA of 0.4 (95%CL on Mean)
‘0th Percentile of 0.1 a
2
i 1E4 —_ Se ° —
4 :
2
1E2 -}——-—
1B3
1945 195019551960 1965 = 197019751980 «s198S~— +4990

Year

Visual Comparison of Data to Estimated Pu-239 MDA in Feces

Figure A-17. MDA data for Pu-239 in feces (part 1 of 3).

Annual Average MDAs for Pu-239 in Feces
Calculated from Recovered Data Between 1945 and 1990

MDA (dpm/24 h)

Figure A-17. MDA data for Pu-239 in feces (part 2 of 3).



[Document No. ORAUT-TKBS-0012-5 [Revision No. 01 | Effective Date: 10/01/2007 | Page 36 of 74_ |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 2 of 36
LIST OF FIGURES
FIGURE TITLE PAGE

A-1 MDA data for americium-241 in urine
A-2 MDA data for curium-244 in urin
A-3 MDA data for Cs-137 in urin
A-4 MDA data for gross alpha in
A-5 — MDA data for gross beta in urine
A-6 = MDA data for H-3 in urine...
A-7 MDA data for Pm-147 in urine
A-8 MDA data for Pu-238 in urin
A-9 MDA data for Pu-239 in urine
A-10 MDA data for Pu-241 in urin
A-11 MDA data for rare earths in urine
A-12_ MDA data for Sr-90 in urine.
A-13 MDA data for U-233 in urine.
A-14 MDA data for U-238 in urine...
A-15 MDA data for Am-241 in feces
A-16 MDA data for gross alpha in feces
A-17 MDA data for Pu-239 in feces
A-18 MDA data for Th-232 in feces

AA Recovery of Historical Data

A large amount of raw bioassay data exists at the ORNL complex. A portion of these data was
transcribed into an electronic database hosted on ORNL equipment and maintained by ORNL
personnel.

As part of this dose reconstruction effort, this database was queried for historical urinalysis records on
October 30, 2003. A supplemental query was performed on December 12, 2003. These queries
requested information on the samples and the instrumentation used in the analysis. No personal
information or individual dose results were requested or received.

These database queries generated 66,204 records from the period between 1945 and 1988. These
records were compiled into a single database. A data validation and comment field was added to
each record. Then the records were examined and classified according to usability.

A114 DATA DESCRIPTION

The database queries retrieved 17 parameters from the ORNL internal dosimetry database:

« NUCLIDE —A unique abbreviated identifier for specific nuclides
SAMPLEID — A unique identifier for individual samples
e SAMPLETYPE - A one-character identifier designating the type of sample (urine, fecal, etc.)



[Document No. ORAUT-OTIB-0034 | Revision No. 00 | Effective Date: 12/13/2005 | Page 11of29 |

results of this evaluation are summarized in Table 5-7, which shows the reported
radionuclide concentrations, the ratio of the concentrations of the three selected
radionuclides to °°Sr, the peak concentration ratio, range of the concentration ratios,
and the average concentration ratio.

5.2 Dose Assignment

In most cases, doses to be assigned to individuals potentially exposed on a routine
basis are calculated from the 50th-percentile intake rates; the material type resulting in
the largest probability of causation (which is determined by the Department of Labor) is
selected. A comparison of the intake rates shows:

. For Sr, the calculated intake rates for Type S material are one to two orders of
magnitude higher than the intake rates of Type F material for all intake periods.
However, because the Type S material remains in the lungs for an extended period
while the Type F material is transferred to the systemic organs, it is necessary to
compare the annual doses on a case-by-case basis to determine which will deliver the
larger dose to the organ of interest.

Table 5-3. Annual ?**U Type F, M, and S intake periods and rates.

*U Type F intake rate, dpmid °U Type M intake rate, dpmid =U Type S intake rate, dpmid
50th 50th 50th

Year percentile GSD percentile GSD percentile GSD
1951 41.19 3.08 47.03 3.33 850.7 3.0

1952 41.19 3.08 47.03 3.33 850.7 3.0

1953 11.19 3.08 47.03 3.33 850.7 3.00
1954 11.19 3.08 47.03 3.33 850.7 3.00
1955 11.19 3.08 47.03 3.33 850.7 3.00
1956 4.213 6.71 44.45 9.70 675.7 3.00
1957 1.942 9.25 7.49 9.70 247.7 7.74
1958 1.942 9.25 7.49 9.70 247.7 7.74
1959 5.174 3.48 21.81 3.33 509 4.17
1960 5.171 3.48 21.84 3.33 509 4417
1961 5.171 3.48 21.84 3.33 509 417
1962 1.881 6.71 6.56 7.24 235.3 6.04
1963 4.881 671 6.56 7.24 235.3 6.04
1964 0.413 3.08 1.641 3.33 23.7 3.00
1965 0.413 3.08 4.641 3.33 23.7 3.00
1966 0.413 3.08 4.641 3.33 23.7 3.00
1967 0.413 3.08 1.641 3.33 23.7 3.00
1968 0.413 3.08 1.641 3.33 237 3.00
1969 0.413 3.08 1.641 3.33 23.7 3.00
1970 0.413 3.08 1.641 3.33 23.7 3.00
1974 0.413 3.08 4.641 3.33 23.7 3.00
1972 0.413 3.08 4.641 3.33 23.7 3.00
1973 0.413 3.08 1.641 3.33 23.7 3.00
1974 0.413 3.08 1.641 3.33 237 3.00
1975 0.413 3.08 1.641 3.33 237 3.00
1976 0.413 3.08 1,641 3.33 23.7 3.00
1977 0.413 3.08 1,641 3.33 23.7 3.00
1978 0.413 3.08 1.641 3.33 23.7 3.00
1979 0.413 3.08 1.641 3.33 237 3.00
1980 0.413 3.08 1.641 3.33 23.7 3.00
1981 0.413 3.08 1.641 3.33 23.7 3.00
1982 0.413 3.08 1.641 3.33 23,7 3.00
1983 0.413 3.08 1.641 3.33 15.25 467



[Document No. ORAUT-OTIB-0018

[Revision No. 01

[Effective Date: 08/09/2005 | Page 23 of 35_|

Table 7-1 (Continued). Whole-body count (WBC) results below which this
TIB method is still considered an overestimate.

Absorption Time since Organ of WES resuk must
Radionuclide type last WBC (d)* concem be fess than (nCi):

Ca-T4T Mors TEU Esophagus 27

Ce-T4T Mors 365 Adrenals’ TS
Urinary bladder 2.4
Brain 27
Breast 09
Gallbladder 1.8
Heart w all O7
Kidneys T7
Liver 0.3
Muscle 17
Ovaries 09
Pancreas. 1.4
Testes 33
Thyroid x3
Red bone marrow 4.4
Bone surface 02
Stomach 0.3
Si 0.2
ULT <OT
To <O.T
‘Skin 27
Spleen 1S
Thymus 0.9
Uterus. 20
ET <010
Tung <OT
Colon <OT
Esophagus. 0.9

Ca-144 MorS 90 Adrenals 56
Urinary bladder 59
Brain 46
Breast EY
Gall bladder EE]
Heart w alll 52
Kidneys 53
Liver 3
Muscle 49
Ovaries 60
‘Pancreas 56
Testes by
Thyroid 54
Red bone marrow 13
Bone surface 18
Stomach 30
ST TS
ULT rs
Lu 4
Skin. 38
Spleen 55
Thymus 32
Uterus 65
ET Ey
Tung 3
Colon 4
Esophagus 52



AECU-817 35

One other factor that is involved, when a solid mass is ground up into fine material, is that the
small particles adsorb a layer of air. This condition makes it very difficult to wet the particles. For
these particulates to become wetted, the liquid must overcome the gaseous surface area that the
small particle has adsorbed. The finer particulates in the extremely fine particle range have the
power of Brownian movement. This movement causes, in many cases, a phenomenon called floccula-
tion, in which the smaller particles, by collision, conglomerate to form masses of larger size and
settle much more rapidly than fine particles. Flocculation seems to depend on the material. Gen-
erally it requires experimentation to determine whether or not a given material will flocculate. Dust
formed from substances such as quartz or minerals will not generally flocculate. Certain fumes or
smoke such as zinc oxide fume and tobacco smoke will flocculate extremely well.

The fact that a relatively small amount of material will give rise to a large amount of particulate
matter is of importance to health physicists due to the fact that when spills occur which involve radio-
active materials, the dispersal of that resulting material must be prevented. A method that has been
carried out, not in regard to particular spills or localized areas of activity, is vacuum cleaning,
which has been used only to the extent of removing dust accumulated over a period of time. It might
be well in the future that vacuum cleaning should be carried out more often and in the vicinity of a
spill, particularly after it has dried out,

Ventilation requirements in regard to the nonactive materials have consisted in installing venti-
lation equipment in order to keep the concentration of these materials down to certain allowable
levels, These have been determined over a period of years by industrial hygienists and have worked
out, for the most part, extremely well. Unfortunately past experience has indicated that ventilation
requirements for the purpose of merely keeping within a certain level are not entirely satisfactory
when radioactive materials are used. In the ventilation of an industrial process it is a relatively
easy matter to capture the contamination and dispose of it. This is not true in the case of radio-
active materials, The problem of ventilation merely leads to the problems of what we are going to
do with the material when we do capture it and how to capture the material we are actually interested
in, in the first place.

The distances traveled by small particles are fantastic at times. Some rough approximations
made, assuming particles being discharged from a 200-ft stack, indicate that if a number of factors
are neglected, such as terrain and turbulence, a 1-4 particle ejected in a 10 mph wind might possibly
travel as far as Memphis or Washington, D.C.

For larger particles the distance traveled is less; a particle 100 4 in size with the same winds
prevailing, would probably land within 271 ft from the stack.

‘The effects of the various particles in the body are varied, and disabilities therefrom may be
divided into four classifications:

1, Fibrosis from such dusts as silica and asbestos,

2. Toxic reactions from such dusts as those of lead, arsenic, manganese, phosphorus, and cad-
mium,

3. Metal fume fever caused by inhalation of such freshly formed metallic oxides as those of zinc
and magnesium. It appears that a man working in conditions where he is exposed to those metals
builds up a resistance during the week and loses it during the week end, The symptoms are high
fever and a feeling similar to coming down with a good case of grippe.

4, Allergies and fungus growths due to dusts of organic or bacterial origin. In many cases the
dusts themselves do not produce the growth, but bacteria adhering to those particles caused the
trouble.

It is possible that a fifth classification might be added, i.e., radioactive dusts or particulates,
but it is necessary to be extremely conservative since so little is actually known about the action of
these particles.

In view of the dust conditions to which we are normally exposed and dust conditions to which
men in industry are exposed, the efficiency of the body’s protective mechanisms, especially in the
lungs, is indicated by the fact that during an individual's lifetime the amount of dust actually inhaled
and retained is small. The filtering agency of the nostrils, the mucous membrane of the nose and
throat, and the lymphatic system of drainage all serve to protect an individual against the inhalation
and retention of normal dust concentrations.



AECU-817 i

Table 8,2, Continued

Comparative thickness to give equivalent absorption Table of specific gravities

Energy,
Mev Pb Steel «= AL HQ
0.2 1 1.7 29 W Aluminum 2.7
0.5 1 2,8 V4 19 Limestone 2.7-3.0
1.0 1 15 4.2 10 Concrete 2.7
15 1 14 Ad 10 Iron (gray cast) 71
2.0 1 1.5 4.3 uN Steel 7.6
2.5 1 1.6 4.3 i Lead 11.2
Tenth-value layers*

Energy, Concretet

Mev Lead Steel or aluminum Water

ih. cm In. cm In, Cm ih. Cm

0.2 0,09 0.23 0,69 18 2.97 6.8 6.5 17

0.5 0.60 1,7 45 3.8 4.1 10.5 10 25

1.0 1,4 3.4 21 5.3 5.7 14 14 34

15 1.6 4.1 2.3 5.8 6.5 17 16 40

2.0 1.8 45 2.7 6.9 7.6 19 19 48

2.5 1.9 4.9 3.0 16 8.4 21 22 55

*One half-value layer = 0.30 tenth-value layer,
tSee note in preceding paragraph on use of cement blocks.

Shape and Thickness of Shields. Example 1, below, illustrates the application of the above
values to a spherical shield about a geometrically small source. If the shield is flat and relatively
close to the worker or object being shielded, an additional thickness of perhaps 1 half-value layer
will be required to reduce the radiation level to the desired value.

Examples,

1, It is desired te reduce the dosage rate at a certain position from 750 r/hr to less than toler-
ance (i.e., to less than 12.5 mr/hr). The energy of the radiation is thought to be 0.5 Mev. What
thickness of spherical lead shield will be required?

Solution: Express the existing and maximum desired radiation in the same units of measure-
ment; then determine by what power of 10 the first must be divided to reduce it to a value less than
the second. In this case 750 r/hr = 750,000 mr/hr, Division by powers of 10 is most easily accom-~
plished by shifting the decimal point. To reduce 750,000 to less than 12.5 is accomplished by shifting
the decimal point 5 places to 7.50000. Then § TVL's will reduce the radiation to 7.5 mr/hr, From
the table one finds that, for 0.5 Mev, 1 TVL of lead has a thickness of 0.60 in.; therefore the thick-
ness of lead required is approximately 3 in.

2, Twenty curies of radioisotope emit one 1.5-Mev gamma and one 0.5-Mev gamma per disinte-
gration. (a) What is the dosage rate at 15 ft? (b) What thickness of spherical lead shield will reduce
the radiation to 5 mr/hr at 18 in.?

Solution: (a) At 1 ft the dosage rate due to the respective gamma energies is:

0,5-Mev gamma, 3.1 r/hr/curie x 20 curies = 62 r/hr

1.5-Mev gamma, 8.2 r/hr/curie x 20 curies = 164 r/hr

At 15 ft the dosage rates are 1/225 times as large. Total dosage rate at 15 ft, 0.276 r/hr plus
0.729 r/hr equals 1 r/hr.



AECU-817 53

5.3 THE POCKET METER AND PROJECTION MINOMETER
By H. H. Abee

The pocket meter used at Oak Ridge National Laboratory is a pencil-type ionization chamber.
This meter is produced by the Victoreen Instrument Company. It consists of a tenite barre) with
aluminum fittings at each end, an aluminum rod for a center electrode, polystyrene insulators, and
an aluminum cap. The tenite barrel is molded to the aluminum fittings. A plug fitting closes one
end of the barrel. The other end of the barrel has a tube fitting which leaves it open for insertion
of the center electrode assembly, The barre} contains a graphite-impregnated paper liner to make
it conducting and has a wall thickness of approximately 2.5 mm, The center electrode is an alumi-
num rod 1/16 in, in diameter. The electrode is insulated from the barrel by polystyrene insulators.
One insulator, slightly larger in diameter than the electrode and molded to one end, insulates the
electrode from the closed end of the barrel, The second insulator, whose diameter is such that it
fits the inside of the barrel, is molded to the electrode about 1/4 in, from the other end This leaves
a portion of the electrode protruding for making contact during charging and reading. The third
insulator is a small bead molded around the center of the electrode to prevent accidental contact of
the electrade with the wall of the barrel. The aluminum cap plugs into the open end of the barrel,
Friction between a hard-rubber ring around the cap and the inner wall of the open aluminum fitting
holds the cap in place.

‘The pocket meter was designed to measure gamma radiation. The walls of the meter are made
of a substance which has the same effective atomic number as air, consequently the chamber may
be any size, If the wall substance did not approach the effective atomic number of air, the calibra-
tion of the meters would be energy dependent. The wall thickness of the meter is sufficient to stop
beta particles of less than 1 Mev energy. However, beta radiation capable of penetrating the wall
will be recorded. As far as the meter is concerned, gamma rays and penetrating beta rays are not
distinguishable in any way.

The pocket meter 1s charged to 150 volts by means of a projection minometer which will be
discussed later, When exposed to gamma radiation, the air in the chamber ionizes and allows part
of the charge to leak off. The amount of discharge or voltage drop ts proportional to the radiation
exposure. Exposure to 100 mr of gamma radiation will give a voltage drop of approximately 40 volts.
Exposures between 0 and 300 mr are measured accurately by the pocket meter.

Pocket meters are calibrated with radium gamma rays, which consist of mixed energies ranging
from about 0.3 to 2.2 Mev, with an average quantum energy of 0.8 Mev. They measure equally well
gamma rays of all energies above 0.2 Mev. Between 0,08 and 0.2 Mev, their readings are about 10
per cent higher, which causes the wearer’s record to show an exposure slightly larger than he actu-
ally received,

Each employee who works in the restricted area wears two pocket meters daily. The two meters
should be worn side by side so they will have approximately the same geometry with respect to any
incident radiation, The purpose of wearing two meters instead of one is to decrease the probability
of getting an insignificant exposure record. The meters are collected and read daily. After reading,
the lower of the two pocket meter readings is recorded as the significant reading, since any factor
which tends to give a faulty reading would cause a meter to discharge more than the amount caused
by the radtation exposure.

Faulty readings may be caused by any of the following reasons: (1) insulator leakage caused by
dust particles or moisture on the insulator, a scored insulator, or a cracked insulator, (2) dropping
the meter, or (3) misuse of the meter such as beating it on the edge of a table or taking the cap off
and touching the electrode with the finger or some other object,

If a meter in routine use indicates an exposure of 30 mr or more, it is considered to be a leak-
ing meter until proved otherwise by simple testing procedures. Should the meter fail to meet the
requirements of the testing procedures, it is laid aside for repair.

In the case of insulator leakage where the insulator {s not damaged, there are several things
that can be done to repair the meter. First, the insulator may be blown with a stream of dry air to
remove dust particles and surface moisture. Should this fail to stop the leakage, the insulator may



AECU-817

all operations at various levels. You are directing the radiation-protection procedure on a job, and
yet on this job you probably find people working who are assigned to it from a number of divisions
or departments. I am sure you have found out already that you do not go up to a plumber and tell him
to get away from what he is doing before he has obtained a slight overexposure. In practically all
cases we work with supervision and try to offer advice and recommendations to the supervisor in
the area in which we work, and he passes these recommendations on down to the people under him,

If damage may result to the plumber from what he is doing, then try to remove him from the area

in which he is working as quickly and quietly as possible. Otherwise, go to the supervisor. If no co-
operation is available, contact your own supervisor, and he will see that the case is taken care of.
Do not cut channels except in serious emergencies where life is endangered. Offer advice where and
when it will do the most good and in a manner that will be accepted, This is a challenge to each indi-
vidual, Some are born with a good personality and a knack of getting along with people, but others
must develop it.

1. Have a General Knowledge of the Total Laboratory Operation, with a Working Knowledge
of the Various Operating Techniques in Your Particular Assigned Area. This is also a matter of
psychology, in that you must be able to create opportunities to learn of the operation without taking
up someone’s time or aggravating him.

Previously I pointed out the fact that if you become a good Health Physicist here, you are not
necessarily limited to ORNL or to Oak Ridge, There is a great need for Health Physicists in many
walks of life and in all countries of the world, Wherever you have people working with radioactive
materials or instruments that protect from radiation, you find a need for Health Physicists. The
greatest need has been, up to now, on the AEC projects. Altogether, we have trained in our division
roughly 150 persons in Health Physics, some for periods of one year or more, and others for months,
weeks, or days. We have sent trained personnel to Mound, Hanford, Los Alamos, and general AEC
operations. Many opportunities in the AEC operations still exist, and these operations are quite
anxious to get additional personnel, Our division is conducting at this time a Health Physics training
program for 10 persons from other laboratories,

As time goes on, the need for Health Physicists and/or Radiological Physicists in hospitals be-
comes more urgent. There is going to be a cancer research program at our local hospital, and a
project is starting in Knoxville which will need Health Physicists on the staff, Various universities
are taking this job seriously Purdue, Chicago, M.LT., and others are sending men here for train-
ing, and they are going back to set up their own nuclei of organizations. The steel industry, using
c** and radioactive iron, is anxious to get started in tracer work and to obtain Health Physicists,
The petroleum industry has indicated that it would like to send people here for training. The Army,
Navy, and Air Force need senior and junior Health Physicists The Public Health Service will need
quite a number of persons trained in the field of Health Physics, Some senior men from the Public
Health Service are in training here now. TVA has a senior man here in training. The transportation
industry will be anxious to get a few senior men trained in Health Physics because they are respon-
sible for seeing that all shipments of radioactive material are made safely. Insurance companies
are interested and have inquired about the possibility of organizing a school for insurance specialists.
An effort is already under way to give safety engineers some elementary but fundamental training in
Health Physics. Many research organizations are interested in securing Health Physicists.

1,2 HEALTH PHYSICS ORGANIZATION AT OAK RIDGE NATIONAL LABORATORY
By J.C Hart*

Mr. Hart’s lecture was primarily concerned with explaining the organization of the AEC and the
laboratory to the students, with the aid of diagrams and charts which he had previously put on the
blackboard, He spent some time in explaining the various job titles and specific differences in re
search and operations.

*Notes taken by C Scott; edited by J. C. Hart.



AECU-817 7

2, Leak Test. After the vibration test, meters are recharged and placed on an 8- to 16-hr
stand test.. Meters reading 20 mr or less are declared good meters and returned to service. Meters
reading 25 mr or more are sent to the repair shop,

NOTE: The passing 20 mr or less value varies from time to time, depending upon the needs of
the laboratory. The value has never been higher than 30 mr nor less than 10,

A Simple Laboratory Method for Calibrating Film
A. Formula

KX mg Ra
ae
where K = 7.8 r/hr for 1 mm Pt = 8.4 r/hr for 1/2 mm Pt

mg Ra = milligram of Radium
d* = distance in cm
§ = source
B. Method (assumed)
1, 160 mg of radium in a 1-mm platinum capsule
2, 24-hr exposure

mr/hr =

3. Table
Exposure, mr Distance, cm
100 432.4
250 273.5
500 193.5
150 157.8
1000 136.8
2000 96.7

Film Marking Equipment

A. Brass Plate Shield
1. Capacity, 54 film meters
2, Dimension, 18 by 18 in,
3, Thickness, 1/4 in,
B, X-ray Marker
1, X-ray—-General Electric dental X ray (price approximately $750)
2. Settings
a. kv, 35 to 46
b. ma, 5 to 6
ec, Time of exposure, 3 sec
d, Distance from tube to brass plate, 30 in,
3. Box shield
a. Inside dimension, 20 by 20 by 50 in.
b, Wood, 9/4 to 1 3/4 in.
¢, Lead lining, 1/8 in. sealed at all joints

Film-changing Equipment

A. Film slide and opener
B. Contamination checking counter of ‘“‘over’’ and ‘‘under’’ G-M tube beta-gamma Poppy



C.2

C.3
C4

C5
Cé

C7
C.7a

C.7b
C.7¢
C.7d

C.7e

Special Exposure Cohort Petition U.S. Department of Health and Human Services

under the Energy Emptoyees Occupational Centers for Disease Control and Prevention
\iness Compensation Act National Institute for Occupational Safety and Health
OMB Number: 0920-0639 Expires: 09/20/2013

Special Exposure Cohort Petition — Form B Appendix — Petitioner 3
B Survivor Information complete Section B if you are a Survivor or representing a Survivor.
B.1 Name of Survivor:

Mr./Mrs./Ms. First Name Middle Initial Last Name
B.2 Social Security Number of Survivor:
8.3 Address of Survivor:

Street Apt # P.O. Box

City State Zip Code
B.4 Telephone Number of Survivor: )
B.5 Email Address of Survivor:
B.6 Relationship to Employee: Q Spouse Q Son/Daughter Q Parent

Q Grandparent Q Grandchild

Employee Information — Complete Section C.

Name of Employee:

Mr/Mrs./Ms. First Name Middle Initial Last Name
Former Name of Employee (e.g., maiden name/legal name change/other):

Mr./Mrs./Ms. First Name Middle Initial Last Name
Social Security Number of Employee:
Address of Employee (if living):

Street Apt# P.O. Box

City State Zip Code
Telephone Number of Employee: )

Email Address of Employee:

Employment information Related to Petition:
Employee Number (if known):

Dates of Employment: Start End
Employer Name:
Work Site Location:

Supervisor's Name:

Name or Social Security Number of First Petitioner:



$4 AECU-817

then be washed with absolute alcohol to clean the surface of any remaining foreign matter and mois-
ture, If this does not stop the leakage, the insulator may then be dipped in a solution of carbon tetra~
chloride and polystyrene which will give the insulator a coating of polystyrene.

Where the insulator of a meter is scored or cracked or when the above procedures fail to repair
the meter, a new center electrode assembly should be installed,

This present meter is a vast improvement over the original pocket meter used here. The orig-
inal meter had a bakelite barrel which was coated on the inside with graphite. Flaking off of the
graphite in many cases caused the meters to give faulty readings. Instead of a plug-type cap, a
threaded cap was used, Continual screwing of the cap into the meter wore off fine aluminum parti-
cles which collected on the insulators and caused leakage. This cap was also more cumbersome to
use operationally, The center electrode assembly was different. The large insulator was not molded
to the electrode, but was held in place by screwing it up against a ledge on the electrode with a small
nut, There was no center-bead insulator on the electrode; consequently the meter was more frequent-
ly discharged by accidental contact of the electrode with the wall of the barrel. As a comparison,
here are some operational figures showing the performance of the two meters, In December, 1945,
19,812 meters were read, Of these, 603 meters were completely discharged. In December, 1947,
44,141 meters were read, Of these, 70 meters were completely discharged, Hf it is assumed that
none of the readings are due to radiation, in the first case the expected discharged or off-scale pairs
would be

603_\? '
(ees) x 9,906 = 8.9 pairs

In the second case the expected off-scale pairs would be

7_\?
win) * 22,118 = 0.05 pair

The Projection Minometer

The projection minometer is a quartz-fiber string electrometer. It contains a power supply for
charging and a lens system for projecting the string image on a screen. The lens system has a
magnification of 40x, The screen is a scale calibrated from 0 to 300 mr, and full-scale deflection
of the image is 76 mm, When the minometer is discharged, the image of the fiber should ‘be at the
extreme right of the scale on a position marked Z. Upon charging, the deflection in millimeters of
the fiber to the left is determined by the charging voltage and the tension on the fiber. The tension
on the quartz fiber is adjusted so the 150 volts of charge on the instrument will give a full-scale de-
flection of 76 mm to the left, This position is the zero point of the scale. Should two or more tminom-
eters be required in an operation, it is imperative that the tension on the fiber of the instruments
be adjusted so that a given charging voltage will give full-scale deflection on each instrument.

When a pocket meter is inserted inte the minometer to be read, the two act as condensers in
parallel. Following the principles of condensers in parallel, these two instruments share their charge
in proportion to their respective capacities,

Summary

‘The pocket meter is a very good instrument for getting a daily dosage record of personnel work-
ing around radiation. The minometer is a part of the pocket meter, even though it is separated in

usage.



we AECU-817

sisting of a chemical compound of silver and bromine. The AgBr has two characteristics which make
it desirable for photographic emulsion. One is that when struck by light, the nature of the crystal
changes very slightly. Subsequently the AgBr at the place where the light or radiation struck the
emulsion can be changed chemically and retained in permanent form. Second, AgBr is insoluble in
water, which means that it can be washed and treated with various chemicals without danger of
losing it.

The AgBr is mixed in a gelatin carrier, which has very conventent properties for photographic
use, It is transparent, and when immersed in water it will swell to several times the original size,
permitting chemical solutions to go throughout the gelatin and get to the AgBr.

The base gives support to the gelatin, holding it in shape when the plate is immersed in water.
The gelatin swells in one direction only and to 8 or 10 times the original thickness, When dried it
comes back to the original position.

It is hard to describe the effect of radiation on photographic emulsion, Historically we find that
early experimentors discovered that some colors of light would activate a silver halide crystal.
Other colors would not affect it. Ordinary AgBr emulsion will respond to blue and green light and
also to radiation in the near ultraviolet. Toward the red end of the spectrum the photographic effect
falls rapidly to zero.

When light strikes AgBr it produces a change in the energy of electrons in the crystal. The
thing that is important is that this changed crystal can be reduced by certain chemical agents so that
the Ag in the AgBr crystal becomes metallic. This is the process known as developing.

After developing the image, the next thing to do is clean out from the gelatin that part of the
AgBr not struck by light. The AgBr is soluble in certain chemical solutions, one of which is sodium
thiosulphate, commonly known as ‘‘hypo.’’ After development and ‘‘fixing’’ with hypo, the photo-
graphic image remains in the gelatin in the form of finely divided metallic silver, which appears
black in color,

‘The speed of an emulsion as used in ordinary photography depends on the amount of light re-
quired to produce a given blackening in the developed plate. Speed is achieved in emulsion by in-
creasing the size of the AgBr grain. It takes a certain number of light quanta to activate a grain.
Apparently the number of quanta required is almost the same for big crystals as for little crystals.
In order to get the maximum amount of silver in the developed image, the largest permissible grain
size is used,

Dosage Measurement with Film

The heaith physicist tries to make dosage measurements in terms of roentgens. The roentgen
is based on the ionization produced in air. In order to measure roentgens with photographic film,
it is necessary to interpret the effect of radiation which was absorbed in the film into ionization that
the radiation would produce in air. Since film and air do not absorb radiations of all wavelengths in
the same proportion, the film is said to be ‘‘energy dependent’ (i.e., wavelength dependent) when
used as a dosimeter. Therefore the response of films depends on ratio of absorption of or gamma
rays in film to the absorption that the same rays would have had in air, The absorption coefficient
in air is almost a constant in the whole X-ray region. Ag and Br are heavy elements, and their
photoelectric effect is large in the soft X-ray region of the spectrum, Because of this photoelectric
effect, the absorption in the film emulsion is not the same as it is in air, which is composed of light
elements.

Dr. Roberson then went on to explain some of the work Joe Deal had done in cooperation with
the National Bureau of Standards,* and discussed the personnel monitoring film meters to some
extent.

*Deal, Roberson, and Day, Roentgen Ray Calibration of Photographic Film Exposure Meters,
Am, J. Roentgenol, Radium Therapy, 59: 5 (1948).



[[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 _| Effective Date: 10/01/2007 | Page 38 of 74_|

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 4 of 36

the same analyte (e.g. GA, GAO, and G_ALPHA are all gross alpha measurements). Table A-1 lists
the recorded analyses performed by radionuclide.

A2 Calculation Method for Sample MDAs

Equations A-1 and A-2 show the derivation of the estimated MDA, which was calculated from 61 165
analytical records. MDAs were caiculated using the method in Brodsky (1992):

MDA= 3+4.65 sors) (At)
r, (nin) xi x24 same)
dis
where
MDA = Minimum Detectable Activity (dpm/24 h sample)
SD, = Standard deviation of the total background counts (cnts/min)
Th = Duration of the background count (min)
k = Calibration factor (cnts*24 h sample/dis)
k ents x 24h sample = InstrEf cnts x Yieldx V, (mL) (A-2)
dis dis Vv mi
'\ 24hsample
where:
InstrEff = Instrument efficiency (cnts/dis)
Yieldi = Fractional chemical recovery
Va = Volume of the aliquot analyzed (mL for urine, g for feces)
Vr = Volume of the sample submitted, or the volume of a standard 24-hr sample from

reference man (1,400 mL/24-hr sample for urine, 135 g/24-hr sample for feces)
(ICRP Publication 23 (ICRP 1975] default values for 24-hr voids), whichever is
larger



This page intentionally left blank.


88 AECU-817

Dosage Rate from a Radium Source. For a small radium source, within a 1 mm platinum shield,
the dosage rate, measured in roentgens per hour at 1 ft, is given by

Ry = 0,0084M

where M is mass of radium measured in milligrams. (The dosage rate at 1 ft from an unshielded
1-mg source is about two-thirds of 8-hour tolerance.)

Dosage Rate as a Function of Distance from a Point Source, The above relations give the dosage
rate at 1 ft from an unshielded point source, If shielding and Scattering are negligible, the dosage
rate at any other distance is found by dividing the value at 1 ft by the square distance measured in
feet,

R =R,/D2

Shielding of a Point Source of Monoenergetic Gamma Radiation. For monoenergetic sources,
shielding problems are simplified by the fact that, in most cases, the fractional reduction in the dos-
age rate produced by a layer of shielding material is independent of the amount of shielding that has
already been used, To use this property to the best advantage, we define and make use of the concept
of a tenth-value layer (TVL) of shielding material. A TVL is the thickness of material which will re-
duce the dosage rate of the radiation passing through it to one-tenth of its incident value. It follows
that two tenth-value layers will reduce the dosage rate by a factor of 1/100; three TVL’s by a factor
of 1/1000; six TVL’s by a factor of 10-8, etc,

It is sometimes useful to know the thickness of a half-value layer; i.e., the thickness necessary
to reduce the dosage rate of the radiation passing through it to one-half of its incident value. One
half-value layer is equivalent to 0.30 tenth-value layer.

One of the parts of Table 8.2 gives the thicknesses of tenth-value layers of different materials
as a function of the energy per quantum of gamma radiation. From the preceding discussion, it will
be apparent that a TVL of one material will be equivalent to a TVL of any other material for the
gamma energy for which they are given. The values given for concrete are for poured concrete with
limestone aggregate. They are satisfactory for cement blocks of the same weight per unit volume,
but for the many cement blocks of lesser density, proportionally greater thicknesses are required,

‘Table 8.2— Absorption of Gamma Radiation

Mass absorption coefficients, Linear absorption coefficients,

Energy em?/, (‘4 per cm
Mev A Ph Fe Al H,O Pb Steel Al H2O
0.2 0.062 0,90 0.17 0.126 «0,14 10 1.30 0.34 0.14

0.5 0,025 0.15 0.080 0.080 0.090 “7 0.61 0.22 0.090
1.0 0.0123. 0.060 «0.058 90.058 (0.67 0.67 0.44 0,16 0.067
1.5 00,0082, 0.050» 0.052 0.052 (0.057 0.56 0.40 0.14 0.057
2.0 0.0062 0.046 0.043 0.043 0.048 0.51 0.33 0,12 0.048
2.5 0.0049 0,042, 0.040 «0.040 0.042 0.47 0.30 «0.11 (0.042



58 AECU~-817

‘These probes are calibrated for hard gamma from 5 r/nr to as high as 100 r/hr. They aiso
indicate soft gamma and beta, but are not calibrated for these. They are reasonably accurate and
handy to use.

They are somewhat limited in their use by intention of design. They are not accurate below 5
v/hr, and each small dial division is equal to quite a number of roentgens or milliroentgens, as the
case may be. This makes atcurate reading difficult, especially since it is a little hard to hold the
chamber steady at a distance and read the dial at the same time. Usually, however, at the intensity
of radiation for which these probes are used, extreme accuracy is not vital, Persons are not likely
to be entering fields of guch high-intensity radiation.

All in all, where high-intensity radiations are likely to be encountered, Fish pole probes are
very useful and necessary instruments for survey work. My own experience has been that they are
good instruments for their intended use and not so subject to failings as are many of the other in-
struments on which we must now rely.

5.6 SURVEY INSTRUMENT USAGE AND SELECTION (I)
By R, L. Clark, E, L. Sharp, and P, E. Brown*

Instruments covered in the past two lectures were: Cutie Pie, Victoreen 263, Zeus, Zeuto,
electroscope, Poppy, and Fish pole probe.

‘The past two lectures have given you descriptive material and key points in the selection and
usage of the instruments listed above in making radiation surveys in the field. Unfortunately I was
unable to attend these lectures, so a portion of the material presented today is likely to be repeti-
tious, In addition, most of you have been making surveys with these same instruments for some time
now. We hope, however, to be able to present some new ideas, or at least re-emphasize those old
ideas which will be beneficial to you and the people you work with in your career as a health physics
surveyor,

In selecting an instrument with which a survey is to be made, the two most important considera-
tions which should govern your choice are: (1) type of radiation to be surveyed and (2) type of survey
desired. The answer to the second question wili generally be supplied by the person requesting the
survey, The two types of surveys, of course, are qualitative or quantitative. Generally speaking,
however, it will save you a trip back to the office if you take some quantitative instrument along,
even when a qualitative survey is requested. About the only time you will be able to make a purely
qualitative survey for beta-gamma activity is when the results are negative —that is, your survey
with a scanning instrument fails to detect any activity above the natural background. H you do detect
activity on the object or in the area surveyed, the next question put to you will most likely be, “‘How
much?” Realizing this contingency and being prepared to take care of it will, in addition to saving
steps for you, cast a favorable reflection both upon you as an individual and on the whole group, Once
you realize that our official function here is to serve in an advisory capacity, you will understand
why each surveyor is, in addition, a public-relations man for the division, For this reason T should
like to emphasize the necessity for accuracy in making surveys.

It has been pointed out in previous lectures that the accuracy of the instruments you are called
upon to use is decidedly Mmited, Accepting that fact, let that be the only limitation placed on the
accuracy of your surveys. A little practice with a yardstick will help you in making accurate esti-
mates of the distances involved in making surveys. You already know that none of the instruments
now available for survey work will permit you to determine the radiation level at the surface of a
container, on a floor, or, in other words, give the information most generally desired, namely, a
‘contact’ reading. You can do the next best thing and give a reading at a distance, provided you are
able to accurately estimate distances,

‘This ability is not only valuable to the person for whom you survey, but it is an aid in obtaining
what ts called ‘reproducibility of results.’’ This factor of reproducibility is much to be desired if

*Text by R. L. Clark.



[Document No. ORAUT-OTIB-0018 __| Revision No. 01 __| Effective Date: 08/09/2005 [Page 340f35 |

25.00
= 2000
3
3 ee
g 1500 —#— Expected Type F (median)
3 | a Expected 95th percentia
5 10.00 | | ae Median Results
4 16.7 pCiltay
<= one -
7 500

0.00 E =

4000 6000 8000 19000 12000
Days after start of chronic intake

Figure B-5. Hanford Sr-90 expected urine versus actual results.

that the values of this TIB are an overestimate for the respiratory tract. However, there are several
important issues to point out.

» While comparing the recording limit to the TIB mean cannot show the TIB to be an
overestimate, continuous exposures at these levels would be reportable in many of the long-

term workers. Levels below the reporting limit were seen in a large percent of the sampled
population, much larger than 50%.

« Long-term exposure to these levels requires that an employee would be working into the
modern era. In more modern eras, chest counts, whole-body counts, and more sensitive

urinalysis would indicate if the average worker were exposed to these levels for a long
duration.

e The data analyzed was obviously for monitored workers. The monitored population is

probably the population that radiological contro! personne! believed to have the highest
exposure potential.

Taken together, this implies that while it might not be possible to demonstrate that the average
radiological worker was exposed to a lower concentration of type S material, it can be seen that it is
very possible, Therefore this TIB should not be considered an overestimate for respiratory tract dose
unless other information, such as chest counts or whole-body counts, is available for the individual.

However, the TIB assumes a continuous exposure every working minute of every day. For individuals
who were only intermittently exposed, a considerably higher air concentration would have had to be
routinely present to equal the same intake. Intakes considerably higher than those listed in this TIB
would have been detectable in the urine results of those present more frequently in those areas. This

TIB should, therefore, be considered an overestimate for intermittently exposed workers regardless of
the organ of concern.



84 AECU-817

normal laundry wash, and are sent back to the appropriate laundry station. The 15 per cent that fails
to pass is sent to the laundry for a decontamination wash and then sent back for another check.

if a garment has not passed after three decontamination washes, it is taken out of service. This
is one way a person may expect to lose protective clothing. Less than 1 per cent of clothing handled
is lost for this reason, When clothing is found to be free of a harmful amount of contamination, Mr.
Taylor may decide that some garments are beyond repair and take them out of service. This is a
second way a person may expect to lose his protective clothing.

A list is made up weekly of clothing taken out of service because of contamination, and a list is
made up daily of clothing taken out of service because they are beyond repair. Copies of these lists
are sent to Clothing Stores.

Clothing Stores personnel, treating each type of list in the same manner, do the following:

1, Credit the man’s account with the garment listed

2. Check stock for a replacement garment of the right size
. Debits the man’s account with the replacement garment

4. Marks the garment with the appropriate numbers

5, Sends the garment to Mr. Taylor at the laundry for distribution

f the garment of correct size is not in stock, the clothing stores personnel sends notification to
Mr, Taylor and to the employee concerned saying that a certain garment, or garments has been
taken out of service because it is contaminated or because it is beyond repair, his account has been
credited with same, and this notification slip will serve to draw this clothing when it is in stock.

Other ways an employee may lose his protective clothing are:

i. Another employee takes his clothing

2. Changing laundry stations and neglecting to have station numbers changed on clothing

3, Numbers misread by someone handling the clothing

4. Employee is deprived of clothing longer when more than one wash is necessary for decon-
tamination

Trial storage is in effect. Because of the small number of garments Involved and the low value
of these garments, the expected saving is smali and necessitates a very economical procedure, A
hutment, 15 by 15 ft, is being used. Ten bins have been constructed at very little cost. There will
be a 20-week storage with checks made at the end of 4, 10, and 20 weeks. The use of colored tags
for marking garments individually will permit using a single bin for more than one weeks collection.
Mildewing may present a problem. It will take at least several weeks to determine whether sucha
storage system will be economically worth while,

e

8.3 WEAR AND CARE OF MASKS
By H. 5. McAiduff

For many years workers in the dusty or smoky trades attempted to relieve the discomfort of
working under these conditions by wearing cloths or handkerchiefs over the nose and mouth, but it
was not until 1919 that the systematic development of modern respiratory protective equipment began.

A respirator or mask is a device for rendering inhaled air respirable, and these may be classi-
fied as follows:

L Supplied-air respirators
A. Self-contained type
1, Oxygen-breathing apparatus
B. Hose type
1. Air-line respirator
IL. Air-purifying respirators
1, Air is purified by drawing through chemicals which remove the contaminant by chemical re-
action
I. Mechanical-filter respirator
1. Removal of dust, fumes, or mists by suitable filters



UNIT 6. APPLIED HEALTH PHYSICS

6.1 PERSONNEL MONITORING PROCEDURES
By J.C. Hart*

The personnel-monitoring program is designed for the three main purposes of providing for (1)
the prevention of unnecessary exposures to personnel, (2) the establishment of permanent iegal and
Clinical records, and (3) the statistical aid to other methods of measuring radiation.

Personnel meters currently in use indicate the extent of exposure to radiation. Tissue damage
may be approximated from the readings if certain facts are known regarding the type of radiation,
the energy involved, and habits of the users. The health physics groups throughout the various proj-
ects have made records available to individuals at the request of the individual.

Effective personne] monitoring involves measurements of both external and internal radiation.
Externa) radiation is monitored at present by pocket meters {ionization chambers) and film meters.
Internal radiation presents two problems in that we are faced with both radiation and toxic effects.
Internal radiation is measured by a study of body fluids (this method is discussed in Lecture 8.3).
Pocket meters used for measuring external radiation are currently worn in pairs, Their primary
purpose is to give a quantitative measurement of radiation exposures on a daily basis. Film meters
are not practical for measuring small doses and/or giving a quick check, Film meters do measure
larger quantities of radiation and have the added feature of providing a per manent record. The ef-
fective pocket-meter range is from 0 to 300 mr. The effective film-meter range is from 0 to 20 r.
A densitometer capable of measuring densities up to 4 or 5 would increase the range of the film
meter by a factor of approximately 2 or 3. The film meter is impractical to process on a daily basis
because of economics and the length of time required.

The film meter tends to show a threshold effect. The emulsion currently in use (Du Pont 552)
Picks up teaves of radium gamma between 7 and 14 mr, A variation in densitometer readings intro~
duces a slight error. Also, not all films give the same density over their surface. These three ef-
fects prevent us from interpreting accurate film readings below 30 mr on a corresponding radium-
gamma calibration. Fortunately the pocket meter fills in this gap, providing readings to 5 mr.

It is desirable to integrate readings between pocket- and film-meter weekly totals, The problem
which presents itself as the most difficult in integrating the film meters and pocket meters begins
with the definition of the heaith physics week. This is increasingly difficult where the laboratory
works shifts. The weekly pocket-meter total must be computed through Saturday, since all records
culminate. It follows from this that the week must end with the beginning of the first shift following
the initial film servicing, Thus the health physics week begins with the 4 to 12 shift on Saturday at
4:01 P.M.

Every available film meter is processed Saturday morning. This would include the meters for
all persons who worked the 12 to 8 shift Saturday morning and those who will be working the 4 to 12
shift Saturday night, The only meters not processed on this initial servicing are those being used
by the 8 to'4 shift, These meters will be processed on the 4 to 12 shift routine. From this it is seen
easily that the 4 to 12 shift workers on Saturday use films which will not be processed until the fol-
lowing week, Thus we are correct in assuming that the health physics week begins at 4:01 P.M. on
Saturday,

——*Notes taken i
‘Notes taken by C. Scott; edited by J. C. Hart.

AECU-817



[Document No. ORAUT-OTIB-0018

[ Revision No. 01

[Effective Date: 08/09/2005 [

Page 26 of 35 |

Table 7-2 (Continued). Urinary excretion results below which this TIB method is still

considered an overestimate.

Radionuclide or Nature of Time since last Excretion rate
analysis type exposure urinalysis (d)_]| Organ of concern pera POUL
Radiosronbum or | Type F_ 365 waste
$r-90 cont. 162 116
‘Ovaries 200 743
Pancreas 20Z Tat
Tesies TST 15
Thyroid 173 Ta
Red bone marrow 7 5
Bone surface 9 6
stomach 159 114
j 166 119
UI 71 57
To 65 Lg
Shr T2T 86
Spleen TE THO
Thymus 174 124
Uiers’ 203 145
ET 107 76
Tung 2,424 T7531
Colon 66 Eid
Esophagus 174 12a
730 Adrenals 7g 36
Uninary bladder 2 20
Bran 55 39
Breast 51 36
Gall bladder 7 84
Heart wall 73 $2
Kidneys 70 50
Liver ST TST
Muscle 6 EL
‘Ovaries 77 EE
Pancreas 78 56
Testes 62 44
Thyroid 67 48
Red bone marrow 3 2
Bone surface 3 2
aaiosronsum or | Type F 730 Stomach BT a
‘Sr-90 cont. oT 6F 46
DLT Zz Li]
CLT 25 1s
Kin 47 3
Spleen 70 50
Thymus 67 48
Uterus 78 56
ET aT Ys)
Tung VIZ BooTM
ToTon 2 Te _
Esophagus or a
paid bgt
Elemental U U facility, NA, assumes. Adrenals 7,300
natural U, type | chronic exposure [Unnary bladder 7,200 86s"
F for atleast one
year. Sampling Brain 1,290 920
[Note change in during work days. [Bread 7,290 925
units} Gall badder 720 bye
Heart wall TO w27
Rigneys a0 17
Tier 3,050 278
Muscle 7,290 922
‘Ovaries 5,320, 3800
Pancreas 7,290 924
Testes 5,410 3870
THyrod T2590 U2T



[Document No. ORAUT-OTIB-0078 | Revision No.01 _| Effective Date: 08/09/2005 | Page 29 of 35_]

ATTACHMENT A
AIR CONCENTRATION LIMITS PRIOR TO 1953 OR OTHER LIMITS

Air concentration limits established for a site prior to 1953 can be used to extend the applicability of
this TIB to those earlier years.

Because Hanford Site tolerance levels were used to control intakes, they fit the criteria of air
concentration limit. The alpha and beta particular air concentration limit at Hanford from at least 1949
to 1952 were 1 x 107? uCi/mL for alpha in nonuranium facilities, 1 x 10° sCi/mL for beta, and 1.5
10° jg U/mL (which converts to 1.1 * 10!° yCi/mL) for uranium facilities (Patterson 1949). These
concentrations are less than the concentrations listed in Table 4-1; thus, the use of Table 4-1 values
is a plausible overestimate. Therefore, this TIB approach can be used unmodified for Hanford claims
from 1949 to the present. Higher tolerance levels of 1 x 10° yCi/mL for beta-emitters and 4 x 10"
yCi/mL for plutonium were in place at least by October 1945 (Cantril 1945).

Oak Ridge National Laboratory (ORNL) used the concept of tolerance levels for air concentrations
prior to 1953. By July 1944, ORNL had established an air concentration limit of 5 x 10° pg/mL

(3 x 10" uCi‘mL) for plutonium based on alpha counting (Parker 1944a). A limit for beta-emitters in
air (based on tolerance in the thyroid for *°'l even though they might not have been collecting iodine
properly on ther filters) was established at least by 1944 at 1 x 10° yCi/mL (Parker 1944b).

Los Alamos also established tolerance levels for air concentrations for specific buildings as early as
1948. The tolerance levels varied by building, apparently based on the predominant radionuclide
used in the building. For beta-emitters the limit was stated as 1 x 10°” uCi/mi in one reference (LANL
1948). The air concentration limit for alpha-emitters other than uranium or *"°Po was 70 dpmim? (3.2
x 107" uCi/ml) for 1948 through 1952 (LANL 1947, 1950, 1952). This was actually the highest limit as
many buildings had tower limits. For uranium buildings the highest air concentration was 400 dpmim?®
(1.8 x 107° UCi/ml) reported in both late 1947 and 1950 so assumed to apply throughout the period
1948-1950 (LANL 1947, 1950). For “°Po, which was a source of exposure at Los Alamos only for
1948-53 (ORAU 2004a), the air concentration limit was the same throughout the period at 1400
dpnvm’ (6.3 x 10° pCi) (LANL 1947,1950, 1954). The concentration limits for alpha-emitters and
beta-emitters changed in 1953. The highest air concentration limit for a plutonium facility was 9
dpm/m? (4 x 10°” sCi/ml) (LANL 1954) which continued to show in reports in 1961 and 1963 (LANL
1969). Fission products were not in heavy use at Los Alamos; the only limiting air concentration for
beta-emitters found after 1953 (actually referred to as fission products) was listed at 6700 dpm/m? (3 x
410° pCi/ml) in a monthly report in 1961(LANL 1969). It was assumed that this limit applied from 1953
to 1966. The limits for uranium in reports during and after 1951 (LANL 1952, 1969) were less than the
NCRP value in Table 4-2 so the NCRP value was assumed to apply, in other words not adjustment
from the default value in Table 4-2 was needed.

The Rocky Flats site used two air concentration limits for plutonium operations throughout its
operational period: 2 x 101? uCi/ml was used for plutonium metal operations and 4 x 10°? \Ci/ml was
used for plutonium chemistry operations. These values were used through 1994 (ORAU 2004b).
REFERENCES

Cantril, S. T., 1945, “Tolerance Limits,” Letter to file, HW 4.159 or 7-2602, Hanford Works, Richland,
Washington, October 18.

LANL (Los Alamos National Laboratory), 1947, Group CMR-12 Monthly Reports for 1947.



AECU-817 37

hr day, The maximum concentration of a radioisotope in water which could be permitted under this
condition is found from an equation given in Lecture 3.4.

fe)
= -9 W Cc _1,64x 10
Cc =1,64x 10 E or, Ge

In this equation C/W is the isotope concentration in curies per gram, and E is the average energy
of radiation, measured in Mev.

The concentration represented by the above equation will give a radiation dosage rate of 01
rep/day to a smail body such as a fish egg, but to a hippopotamus the surface dosage would be only
approximately 0.5 rep/day, and the internal dosage would decrease with distance from the surface.
It is of interest to note that for some radioisotopes in water the potential radiation hazard due to
ingestion greatly exceeds the submersion hazard, while for others, due to very low absorption from
the digestive system, the reverse is true.

Possible Effects of Physical and Biological Processes

Using the ORNL liquid-waste —disposal system as an example, it is generally possible to so
control the concentration of the waters discharged from White Oak Lake that they fall below the per-
missible concentrations for drinking water and submersion given above. There have been a few oc-
casions on which these concentrations have been exceeded for short periods of time. Since these
concentrations are very greatly diluted in the Clinch River, they are felt to represent no hazard to
persons (and fish) downstream, However, in many instances nature has shown remarkable ability
to concentrate chemical elements by both biological and physical processes, Many plants and lower
animals greatly concentrate certain chemical elements. Such concentrations are sometimes more
pronounced if the elements are present in very low quantity, and especially if the organism has a
diet deficiency in these or similar chemical elements, conditions which are likely to hold in the case
of radioisotopes of some of the rarer elements, The most familiar example of concentration at
ORNL is by algae which grow in the Settling Basin and frequently accumulate sufficient activity that
a small piece will produce a significant intensity of radiation at a distance of many feet.

Experiments have shown that the clay in this area adsorbs many of our radioactive wastes very
strongly, In fact, it is felt that the deposition of radioisotopes in the bed of White Oak Creek and
Lake is largely due to this property of the clay which enables it, in settling out, to carry the radio-
isotopes with it. More effective physical concentration is accomplished by the resin columns that
have developed for the almost quantitative removal by ion exchange of both positive and negative
ions from solution, The same principle is used in many water-softener plants. The health physicist
looks with suspicion upon any uncontrolled concentration of radioactive material, and in this case
his misgivings are intensified by the knowledge that such concentrations are frequently sensitive to
the degree of acidity of the solution. A solution may permit concentration over a long period of time;
then a slight change in pH may release the accumulated radioisotope in a very short time, giving
rise to undesirable concentrations,

It is said by some chemists that it is possible for some of the more hazardous radioisotopes, if
dumped into a drain pipe, to be almost completely plated or perhaps adsorbed to the interior of the
pipe, waiting for some unsuspecting plumber to come along and receive a serious overdosage.

At the present time our knowledge of the extent to which concentrations may take place in nature
and the conditions under which they may represent serious hazards is too limited for us to evaluate
with any certainty the hazards involved in any method of waste disposal.



AECU-817 33

Permissible Concentrations of Alpha-emitting Materials

Where it is believed that the behavior of alpha~emitting radioisotopes in the body is sufficiently
well known to use the above method, the calculation is performed in the same manner except that
the effect of the alpha radiation is considered to be 10° times as great, per Mev, as that of beta and
gamma radiation. This factor may be introduced into the above equations by multiplying the alpha
contribution to E by 10. (There is some consideration of using a higher factor than 10, but this has
not been accepted.)

Factors Determining Permissible Concentrations

From the above equations it is apparent that permissible concentrations are lowered by small
mass of critical tissue, high retention in the tissue, long effective half-life, and high energy of dis-
integration, In general, permissible concentrations for materials deposited in the bones are low
because of the comparatively long effective half-life. It may be observed that, if the elimination
half-life and the radioactive half-life are greatly different, the shorter of the two is the principal
factor in determining the effective half-life. Thus the effective half-life of 45-day Sr®9 in bone is
43 days, while that of 25-year Sr® is 197 days,

Practice at Oak Ridge National Laboratory

It Is generally impractical to consider individual isotopes in setting maximum permissible
limits of concentration in the atmosphere, since the isotopic concentration of the radioactive con-
tamination cannot generally be readily determined. It is the practice, therefore, to set the permis-
sible limits for beta-gamma sufficiently low to take care of the worst cases to be anticipated.

In the case of alpha contamination the frequent occurrence of comparatively large quantities of
short-life alpha emitters make it desirable to determine the percentage of long life activity before
assessing the hazard, At the present time decay determinations are frequently made for this pur-
pose, but a more rapid method is desirable, The Geiger -Nutall relationship between the range and
the rate of decay affords a means for rapid discrimination between long and short half-life alpha
emitters by use of a range analyzer.

Noble Gases

Noble gases which do not decay Into radioactive materials are less hazardous from the point of
view of inhalation than from external radiation,
Particulate Atmospheric Contamination

Hazards from particulate contamination which is highly soluble in the lungs and body fluids may
be assessed by the methods given above.

At the present time there is insufficient biological information to assess the hazards due to in-
haled insoluble particulate radioactive materials.

Appendix A

Experimental results indicate that the energy absorbed per ion pair produced by X and gamma
radiation in air is approximately 32.5 ev. For a dosage of 1 r, the energy absorbed per gram of air
would be

Lesu _ 2,08 109 ion pairs 32.5 ev _ 13 7 .
0.00129 ¢ x eau * jon pair * 5.23 x 1019 ev/g = 5.23 x 107 Mev/g = 84 ergs/g

*The use of the value of 20 for the relative biological effectiveness of alpha radiation was recom~
mended by the Chalk River Conference on Radiation Protection, September 29-30, 1949,



Work performed under Contract No.

Originally entitled:
"fealth Physics Applied Training Course."

Issuance of this document does not constitute authority
for declassification of classified material of the same ar
similar content and title by the same authors.

This report has been reproduced with minimm alteratia
directly fran manuscript provided the Technical Informatim
Service in an effort to expedite availability of the informa-
tion contained herein.

Reproduction of this information is encouraged by the
United States Atomic Energy Cammission. Arrangements for
your republication of thie document in whole or in part
should be made with the author and the organization he
represents.

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Operated by
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Reprinted - AEC, Oak Ridge, Tenn, -A23502
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Reprinted - AEC, Oak Ridge, Tenn. Whi272



(Document No. ORAUT-TKBS-0012-5 | Revision No. 01_| Effective Date. 10/01/2007 | Page 460f74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 12 of 36

06

Os

MDA (dpm/24 hy

2

a

Annual Average MDAs for Cm-244 in Urine
Calculated from Recovered Data Between 1945 and 1990

Figure A-2. MDA data for curium-244 in urine (part 2 of 3).

‘Number of Samples (N)

g

|
|

‘
|
|
|
|

|
i

69 $e ——

50 J

40 |

Number of Sample Results Recovered per Year
Between 1945 and 1990 for Cm-244 in Urine

Figure A-2, MDA data for curium-244 in urine (part 3 of 3).



24 AECU-817

recommendations for radiation safety. The reports were similar, but they dealt with protective ma-
terials for use in X-ray and radium laboratories rather than tolerance limits.

In 1925 the International Congress of Radiology in London considered the problem, but no action
was taken until 1928 at the Stockholm Meeting where definite proposals were adopted and an Inter-
natjonal Committee on X-ray and Radium Protection was formed. There was no reference to a toler~
ance dose, but they stated that the known effects to be guarded against were injuries to the superficial
tissues, derangement of internal organs, and changes in the blood.

In 1934 the tolerance dose was suggested as 0.2 r/day. The basis for this figure was established
from the opinions of various individuals, who published their concepts during the period from 1925 to
1932. These opinions were not substantiated by experimental evidence, but merely by observing radi-
ation workers who showed no obvious effects. The daily dose to which they were exposed was caicu-
lated, and a generous safety factor was applied.

As an outgrowth of the International X-ray and Radium Commission, the United States formed
an advisory committee on X-ray and radium protection, which published its first proposals in the
Bureau of Standards Handbook, The first tolerance dose set by this committee was 0,2 r/day, which
was later changed to 0.1 r/day for X and y radiation, This limit is widely used today, although it is
by no means the final word,

There is no legal status of radiation safety recommendations. The complications of such an ar-
rangement were recognized, but the American Advisory Committee feeis that public opinion can be
just as effective as word of law. Moreover, by not having a legal status, recommendations could re-
main flexible and be readily changed to suit changing conditions. Certain states have set up, through
their Departments of Labor and Industry, rules to cover the safety of workers engaged in luminous-
dial painting, but in no case are the recommendations actually Jaw on the statute books.

The difficulties in establishing a tolerance dose are due to varying biological factors. Different
tissues have different reaction to the same dose, Different organs react in different ways. Each
species of animals varies in response to radiation damage, Within the species there is a variation
in strain; however, this is more a matter of degree than of type. This obviously makes it rather dif-
ficult to extrapolate to human applications from animal experiments.

Generally speaking, the less specialized tissues are more vulnerable to radiation than the com-
plex cells. For example, the nervous cells seem fairly radioresistant, while the more primitive
cells of the reproductive or lymphatic systems are early indicators of radiation damage. The rapidly
growing and active cells tend to be more radiosensitive than cells in the resting stage.

Although the actual mechanism whereby the cell is damaged is not known, the result is attributed
to a combination of a direct action on the cell and an indirect action on the environment of the cell.
‘The radiation damage will therefore be dependent on the environmental conditions, If, for example,
the cell has a diminished oxygen supply at the time of irradiation, the effects are less pronounced,
There is experimental evidence that change in acid-base rejationship affecting permeability of mem-
branes can increase radiosensitivity in certain tissues. Physical factors, such as heat, cold, and previ-
ous radiation, also affect the response of some cells.

There are two kinds of effects observed, the threshold and the nonthreshold, In the case of a
threshold effect, the damage is not measurable by present methods until a certain dose is adminis-
tered. The skin and blood-forming organs respond to radiation in this manner. For the nonthreshold
effect there is no initial dose that must be exceeded before an effect is obtained, This is true of the
influence of radiation upon the germ plasm of lower organisms,

The reversibility of radiation effects is important, particularly in occupation exposure. Revers-
ibility implies that the tissue may be returned to its normal state after exposure is discontinued,
This faculty depends on the regenerative properties of the tissue. Some tissues, such as the skin
and the blood-forming elements, are endowed with a special type of repair mechanism, Other tis-
sues, such as the brain and the kidney, do not have this property, and any repair is by the formation
of a scar, which does not take over the function of the origina! tissue which it replaces,

In order for an effect to be reversible, it must not produce injury beyond the normal capacity for
regeneration. The total dose and the totai time affect the powers of regeneration. A certain dose may
be safe if administered over a long enough period of time; however, if given as an acute dose, it may
damage the tissue beyond repair.



28 AECU-817

Technical reasons for limiting radioactive contamination concentration are based upon avoidance
of experimental errors due to activity turning up inappropriately. It has been said that the loss of
1075 of the activity from an experiment will foul the results, This is a function of the total radio-
activity, but all those interested in sensitive radiation detection and measuring equipment do not want
to see radioactive backgrounds increased above the inevitabie concentrations in nature.

3.4 PERMISSIBLE INHALATION CONCENTRATIONS
By Forrest Western

Introduction

In a discussion of permissible inhalation concentrations of radioactive materials, it is convenient
to classify the material as follows:

A, Gaseous material

i, Noble gases

2. Chemically active ga:.-s
B. Particulate materials

1, Insoluble

2, Soluble

In such a classification it is to be borne in mind that there are wide variations in the chemical ac-
tivities of gases and in the solubility or insolubility of particulate materials.

Except for the possible case of noble gases which do not rapidly decay into chemically active
radioactive products and for insoluble particulate materials, the hasic problem in the determination
of permissible concentration may be broken up into two parts: (1) What concentration of a given
radioisotope can be tolerated in a particular tissue or organ of the body? (2) What concentration of
a given radioisotope in the atmosphere will give a tolerance concentration of the isotope in the crit-
ical tissue of the body? It is the purpose of the following discussion to provide a basis for making
quantitative estimates of the answers to these questions, making use of certain simplifying assump-
tions and appropriate biological information.

Under normal circumstances it is believed that dosages to body tissues from beta-gamma —emit-
ting radioisotopes within the tissues should not exceed those permitted from external sources of
radiation when measured in equivalent units,

As defined by the Fifth International Congress of Radiology (Chicago, 1937), I r is ‘‘the quantity
of X or gamma radiation such that the associated corpuscular emission per 0.001293 g of air pro-
duces in air ions carrying 1 esu of quantity of electricity of either sign.’’ From its definition, the
roentgen is a measure of the amount of energy, per unit of mass, absorbed in air from the radiation.
In Appendix A it 1s shown that the exposure if air to 1 r of radiation results in the absorption by the
air of 5.23 x 107 Mev of energy per gram of air, or 83.8 ergs of energy per gram of air.

General Relationships and Units of Measurement

In general, when a quantity of a radioisotope is fixed in the tissues of a particular organ of the
body, the rate of dosage to the tissues of that organ depends on the concentration of the isotope in the
organ, on the energy per disintegration of atoms of the isotope, and, to a lesser extent, on the com-
position of the tissue. Customarily, concentrations are measured in curies per gram, energies of
disintegration are measured in millions of electron volts (Mev), and dosages are measured in rep.

The dependence of energy absorption upon tissue composition arises in two ways. For the more
energetic radiation, loss of energy to the tissue is largely through Compton scattering, for which
the loss per electron of tissue is only slightly dependent upon the composition of the material. Be-
cause of the large hydrogen content of soft tissues, there are approximately 10 per cent more elec-
trons per gram than in air, and therefore the absorption of energetic radiation in soft tissue is about
10 per cent higher than in air. For the less energetic gamma and X radiation, a considerable frac-
tion of the loss of energy to matter is through photoelectric absorption. Roughly, the loss per elec~



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 | Effective Date: 10/01/2007 | Page 72 of 74_]

ATTACHMENT B
ASSESSMENT OF INTERNAL DOSE AT ORNL: 1944 TO 1947 |
Page 2 of 4
_ 1000 3 — —
-
eo 1003
o q .
3 J
3 10 3
2 js
g 43
5 :
& 014 a 4
& ; ,
0.01 r r :
-2 -1 0 1 2 3
Z Score

Figure B-1. Plutonium in urine samples from 1945. The geometric mean = 1.7
pCi/day, the GSD = 5.3, and 7 = 0.79.

_ 1000 y -
z q 1
eo 100
So °
g 3
6 10%
& |
g 13
5 ]
£ 014 7
Fe

0.01 + ~ : 7

3 -2 1 0 1 2 3

Z Score

Figure B-2. Plutonium urine sample data through April 18, 1945. The geometric
mean = 3.1 pCi/day, the GSD = 4.2, and 7 = 0.82.



AECU-817 ST

When ifs limitations are realized and it is used as intended, the Lauritsen electroscope is a
very good instrument,

The Poppy probe and amplifier is a pulse detector intended purely for qualitative detection or
indication. It gives an audible signal indicating each pulse, but it cannot be used for quantitative
measurement, Some of the Poppy amplifiers 1 have seen include a rate meter ostensibly for quan-
titative measurement of the radiation. However, none of those I have seen were calibrated or of any
value,

When used for alpha detection the Poppy probe does not respond to beta or gamma radiation,
and thus, unlike the Zeus and Zeuto, it can be used in the presence of beta~gamma radiation.

There are at least two models of the Poppy amplifier: one is a portable model which is battery-
powered, and another uses an a-c power supply and amplifier,

Unfortunately the Poppy has many limitations. One serious limitation is already evident. The
radiation is not measured,

‘The battery ~powered amplifier is subject to the failings of all battery-operated instruments and
to quite a few additional ones. It is extremely sensitive to weather and humidity changes. It is very
hard to adjust to proper voltage, etc. In short, those I have seen never work. Several modifications
have heen tried, but to date I have not found one that is satisfactory.

‘The a-c—powered poppy is much more reliable and is a reasonably good detector. However, it
does have serious limitations. The a-c Poppy is not very portable. It, too, is frequently very difficult
to adjust to a proper or satisfactory operating voltage without background noise. It is subject to the
line interferences and background noises of the a-c power supply —noises caused by motors, flvo~
rescent lights, etc. Frequent loose connections in the cable and probe also give rise to considerable
noise and flash indications, These noises, which for the most part cannot be told apart from true
indications of alpha, make detection of small quantities of contamination or radiation difficult. The
lower limit of detection of the instrument makes it practically impossible to detect concentrations
of less than 1000 dis/min, In view of the emphasis placed on the need for freedom from alpha con-
tamination, this seems to be a very Serious limitation.

My experience with the Poppy for detecting the soft beta radiation of c)4 is nil, and my knowl-
edge of this application is very limited, I cannot give you any idea from practical experience how
good or bad the instrument is. I hope I shall be able to give you some useful knowledge of its basis
and its possibilities.

‘The range in air of beta particles from C4 is only on the order of 1 ft, Because of their very
Jow energy, these particles, or the large majority of them, are stopped by even what are considered
“thin-window’’ detectors, The very thin nylon window of the Poppy probe, which is only about 2/10,000
in, thick, makes it suitable for detecting these low-energy betas.

Some modifications are needed, however, The amplification-factor characteristics of air which
is normally used in the Poppy probe are such that the voitage setting necessary for beta detection
is too near the point of breakdown or the voltage setting for constant discharge.

The probe is slightly modified to permit a slight flow through the probe of a gas such as an
argon-COg mixture. Other gases also have amplification factors with a rate of change slow enough
to be satisfactory, but the argon-CO2 mixture was selected here for various reasons, Since the
voltage for C14 beta detection is more critical, the power supply of the amplifier is also slightly
modified to give a more stable voltage and decrease the lag in the change of voltage when the knob
is turned.

‘The operation of the instrument is very similar to that of the alpha Poppy. However, when the
operating voltage is properly set for the alpha Poppy there is very little background noise, Since
the C’* Poppy is sensitive to beta and secondary electrons from gamma, there wili be background
noise comparable to that in a G-M tube detector. This makes it difficult to find the proper operating
voltage, but a small beta source will help, This background also makes detection of low levels of
activity rather difficult.

The Fish pole probe is quite similar to the Cutie Pie. Since it is calibrated for and used for the
measurement of higher intensities of radiation, the ionization chamber is on the end of a pole. The
surveyor can, from a safe distance out of the beam or field of radiation, hold the chamber in the
beam or field and read the indication on the meter supported by a shoulder strap.



[Document No. ORAUT-OTIB-0034_|

Revision No. 00__| Effective Date. 12/13/2005 |

Page 17 of 29

]

+ annual bioassay

value (50th percentile)
* not used in fit
___ fit

Table A-5. Type F “Sr intake periods and rates.

Figure A-1. Assumed *’Sr intake, 1951 to 1988, 50th-percentile results, Type F.

"Sr intake rate (dpmiday)
Start date End date 50th percentile | 84th percentile | GSD
1/1/1951 12/31/1953 475.2 4389 9.24
1/1/1954 12/31/1954 80.99 810.4 10.04
1/1/1955 12/31/1960 47.34 197.3 4.17
1/1/1961 12/31/1964 47.34 62.43 1.32
1/1/1965 12/31/1988 15.52 36.88 2.38

Figure A-2. Assumed °

* annual bioassay

value (50th percentile)

not used in fit
fit

"Sr intake, 1951 to 1953, 50th-percentile results, Type S.



{Document No. ORAUT-OTIB-0034 | Revision No. 00__| Effective Date: 12/13/2005 | Page 10 of 29

]

5.0

ASSIGNMENT OF INTAKES AND DOSES

The resulting calculated intake rate information, useful in dose reconstruction for the
radionuclides identified for ORNL, is discussed in this section. For each radionuclide, the
50th- and 84th-percentile intake rates, and the GSDs, are provided in specific tables. The
GSD values have been adjusted from the values given in the tables in Appendix A to allow for
the addition of doses from different intake periods into a single input line for a given year in the
IREP input file, and to ensure that none are less than 3, the value used when assigning
intakes to individuals from person-specific bioassay results.

5.1

Strontium-90. For ®Sr, several intake periods were defined as shown in Table 5-1 for
Type F material and Table 5-2 for Type S material. Note that there were five intake

periods defined for *°Sr Type F intakes and six intake periods defined for “Sr Type S
intakes.

Table 5-1. Combined °°Sr Type F intake periods

and rates.
Sr Type F intake rate,
Date range dpmid

From To 50th percentile GSD
1/1/1951 _| 12/31/1953 475.2 10.0
1/1/1954 | 12/31/1954 80.99 10.0
1/1/1955 _| 12/31/1960 47.34 417
1/1/1961_| 12/31/1964 47.34 3.00
1/1/1965_| 12/31/1988 15.52 3.00
Table 5-2. Combined °°Sr Type S intake periods
and rates.

Sr Type S intake rate,
Date range dpmid

From To 50th percentile GSD
1/A11951_| 12/31/1953 24646 9.53
1/1/1954 | 12/31/1954 7,232 9.53
1/1/1955 | 12/31/1960 2,379 5.73
1/1/1961_| 12/31/1964 2,379 3.00
1/1/1965 _| 12/31/1983 795.0 3.00
1/1/1984 | 12/31/1988 425.5 4.51

Uranium-234. Table 5-3 contains the calculated annual intake rates for the years
1951 through 1987, for Types F, M, and S material.

Plutonium-239. Tables 5-4 and 5-5 contains the calculated annual intake rates for
Pu for Types M and S material respectively for the years 1951 through 1988.

Americium-241. Table 5-6 contains the combined 7‘Am Type M material intake

periods and rates for 1968 through 1988. Note that there were two intake periods
defined.

Contribution From Additional Radionuclides

To account for additional intakes, an evaluation of air monitoring data from the ORNL
perimeter reporting stations for the period 1975 through 1984 was conducted. The
approach was to develop the ratios of the isotopic concentration ratios of other
radionuclides to the concentration ratio of °Sr reported in the air monitoring data. The


Special Exposure Cohort Petition U.S. Department of Health and Human Services
under the Energy Employees Occupational Centers for Disease Control and Prevention
iiiness Compensation Act National Institute for Occupational Safety and Health
; OMB Number. 0920-0639 Expires: 09/20/2013
Special Exposure Cohort Petition — Form B Appendix —- Continuation Pag

Continuation

age — Photocopy and complete as ne

samples to determine the effectiveness of the air sampling program at ORNL.

These lectures reveal that ORNL was in its infancy, concerning extrapolation of air
sampling to damage to biological tissues, instrumentation to determine the accuracy
of their sampling program, and the determination of the particle size of the
contaminants. it is not logical to accept all of the air samples accuracy, and

those air samples to extrapolated to intemal dose, when it is clear that the ORNL
methods were exceptionally crude, by today's standards.

This SEC petition has selected an end date of December 31, 1952, due to the fact
that ORAUT-OTIB-0018 was developed to apply from 1953-the present. Although
an exception to that date of application was specifically made for ORNL, itis
becoming clear that ORNL exception was granted without closely reviewing
ORNL's crude methods for gathering and analyzing air samples. It is very clear
that co-worker internal dose cannot be assigned prior to 1951. It is very clear that
all radionuclides were not addressed in either air sampling or the very limited
number of in vitro samples performed prior to 1953. As previously mentioned

a majority of radionuclides, present at ORNL were not considered in bioassay
sampling until sometime later than 1953. Attachment B, for ORAUT-TKBS-0012-5,
Rev. 1, states that there were "1483 alpha air-monitoring results available for 1944-
1947 at ORNL.” This is apparently the evidence that shows that ORNL had a
rigorous air sampling program in place and allowed ORAUT-OTIB-0018 to
determine internal dose for workers at ORNL. 1483 air samples reflects only

1.02 air samples per day during this time period. NIOSH is basically assigning
internal dose, for ORNL workers, based upon 1483 questionable air samples
performed from 1944-1947, and extrapolating that data to cover the entire period
from 1943-1953, when ORAUT-OTIB-0018 does not cover the numerous
radionuclides for which no sampling was performed, and also specifically does not
cover radioactive gases (such as tritium which was prevalent at ORNL), and

does not cover workers who were likely to have incurred exposures of particulate
radioactive material (page 8 of 35 in ORAUT-OTIB-0018).

Itis not feasible that 1.02 air samples daily (1483 from 1944-1947) could possibly
provide the necessary data to limit everyday exposures to even the sampled
radionuclides, which | assume to be plutonium, uranium, and possibly strontium.
Added to this extremely limited amount of air sampling is the fact that these 1944-
1947 air samples are extrapoiated to cover all years from 1943-1953, then they are
again extrapolated to determine internal dose. Internal co-worker data is also not
available prior to 1951, according to ORAUT-OTIB-0034, so this very limited
dataset of air samples is utilized for any co-worker doses that must be addressed

Name or Social Security Number of First Petitioner:

Continuation pg. 4


[Document No. ORAUT-OTIB-0034 | Revision No. 00 _| Effective Date: 12/15/2005 | Page 12 of29 |

1984 0.413 3.08 1.641 3.33 15.25 3.0
1985 0.0957 3.08 0.294 3.33 15.25 3.0
1986 0.0957 3.08 0.294 3.33 15.25 3.0
1987 0.0957 3.08 0.294 3.33 15.25 3.0

Table 5-4. Annual %°Pu Type M intakes and

rates.
"Bu Type M intake rate, dpm/d

Years 50th percentile GsD
1951-1952 40.75 3.0
1953-1959 10.98 79
1960-1968. 10.98 3.0
1969-1984 7.35 3.0
1985-1988. 1.614 4.2

Table 5-5. Annual ”°Pu Type S intakes and

rates.
“Pu Type S intake rate,
dpmid
0th
Year percentile GSD
7951-
1988 415 5.50
. For “U, the calculated intake rates for Type S material are one to two orders of

magnitude higher than the intake rates of Type F and M material for all intake periods.
Again, because Type S material remains in the lungs while Type F and M material is
transferred to the systemic organs, it is necessary to compare the annual doses on a
case-by-case basis to determine which will deliver the larger dose to the organ of
interest.

. For Pu, The calculated intake rates for type M material should be used for all
systemic organs and intake periods. For non-systemic (respiratory and Gl tracts)
organs, the type S intake rate may only be used as an underestimate. If an
overestimate or best estimate is needed for non-systemic organs, an individualized fit
to the bioassay data, assuming type S material, for the specific work period of the
Energy Employee being evaluated must be performed. Table A-3 provides the
bioassay data to be used to perform the individualized fit. Both the 50th- and 84th-
percentile values must be fit using the same intake dates and/or periods; the 50th-
percentile intakes are used to assign the intake and the 84th-percentile is used to
determine the GSD for each intake. The GSD for each intake will be the ratio of the
84" percentile intake rate to the 50” percentile intake rate. For input into IREP, the
dose from each intake must be determined separately.

. For **!Am, the calculated intake rates for Type M material should be used for all organs
and intake periods as given in Table 5-6.

Table 5-6. Combined 74'Am Type M intake periods and
rates.

Am Type M intake rate, dpmid
Year 50th percentile | 84th percentile GSD

1968-1984 6.673 14.61 3.00

1985-1988 2.207 6.659 3.00



32 AECU-817

Evaluation of Quantities C and E Appearing in the Above Equations

The principal difficulty involved in making estimates of dosage rates due to radioisotopes within
the tissue is that of obtaining reliable values for C and E. It has already been indicated that the above
equation ts valid only if the material is uniformly distributed through the tissue. In many cases the
extent to which this is true is difficult to estimate, Further difficulty arises from inadequate knowl-
edge of rapidity and percentage of absorption into the tissues and of rates of elimination from the
tissues, While the chemical behavior of a radioisotope is the same as that of stable isotopes of the
same chemical element, it must be kept in mind that one cannot always predict quantitatively the
behavior of trace amounts of the element from experiments performed with relatively large amounts.

For many radioisotopes, values of E are only roughly known. Generally the isotopes in which
we are interested disintegrate by the emission of a beta particle, frequently followed by the emission
of one or more gamma quanta, Sometimes, however, the disintegration scheme is quite complex, and,
more frequently than not, it 1s not definitely known. Examples of the more common cases, together
with the method of determining E for each case, are given below.

1, p32; maximum beta energy, 1.69 Mev; no gamma emission, In all beta spectra the energies
per disintegration range from zero to the maximum in such a way that the average energy per dis-
integration is approximately one-third of the maximum energy observed. Sincé there is no gamma
energy to consider in this case

_ 1.60
a)

E 0,56 Mev

2. Na24; maximum beta energy, 1.39 Mev; gamma energies, 2.76 and 1.38 Mev. Measurements
by K. Siegbahn, Phys. Rev., 70: 127 (1946), indicate that the emission of each beta particle is followed
by 1 gamma quantum of each of the energies given above. Assuming all energy to be absorbed in the
tissue

5 = LSP, 2.16 + 1.98 = 4.6 Mev

In this case our assumption that all the energy is absorbed in the tissue gives too high a value for E,
since quanta of these energies are very penetrating.

3. Na2°; maximum beta energies, 3.7 and 2.7 Mev; gamma energy, 1.0 Mev. Measurements
by Bleuber and Zunti, Helv. Phys. Acta, 19: 421 (1946) and 20; 195 (1947), indicate that 55 per cent of
the disintegrations involve the emission of only a beta particle with maximum energy of 3,7 Mev, while
the remaining 45 per cent involve a beta with maximum energy of 2.7 Mev followed by a 1.0 Mev
gamma.

7 0.45 (2 - 1.0) 21.53 Mev

The disintegrations schemes described above are customarily illustrated by diagrams as given in
Fig. 3.1.

32 4 28
wP Na No
Br 139
e
8 5
3 a
13 1
a iw a
lia 72.76
rc 2
rd

10) (2)
Fig. 3.1



Special Exposure Cohort Petition U.S. Department of Health and Human Services

under the Energy Employees Occupational Centers for Disease Control and Prevention
Ilness Compensation Act National Institute for Occupational Safety and Health

; OMB Number: 0920-0639 Expires: 09/20/2013
Special Exposure Cohort Petition — Form B Page 5 of 7

Basis for Proposing that Records and Information are Inadequate for Individual Dose —
Complete Section F.

Complete at feast one of the following entries in this section by checking the appropriate box and providing
the required information related to the selection. You are not required to complete more than one entry.

F.1 v We have attached either documents or statements provided by affidavit that indicate that
radiation exposures and radiation doses potentially incurred by members of the proposed class,
that relate to this petition, were not monitored, either through personal monitoring or through area
monitoring.

(Attach documents and/or affidavits to the back of the petition form.)

Describe as completely as possible, to the extent it might be unclear, how the attached
documentation and/or affidavit(s) indicate that potential radiation exposures were not monitored.
Specifically, concerning all radionuclides, for the Oak Ridge National Laboratory (X-10):

There is little or no in vitro bioassay sampling data available for
reconstructing the internal dose for the years 1943-1953. There are no
in vivo records available for reconstructing the internal dose for the years
1943-1953. Concerning in vitro sampling, the lack of availble samples is
documented for all radionuclides other than plutonium 239, uranium, and
cesium, in the graphs provided in Attachment A in ORAUT-TKBS-0012-5.
These graphs provide the proof that very few urine samples were found
for all other radionuclides present at X-10.

F.2 \/ We have attached either documents or statements provided by affidavit that indicate that
radiation monitoring records for members of the proposed class have been lost, falsified, or
destroyed; or that there is no information regarding monitoring, source, source term, or process
from the site where the employees worked.

(Attach documents and/or affidavits to the back of the petition form.)

Describe as completely as possible, to the extent it might be unclear, how the attached
documentation and/or affidavit(s) indicate that radiation monitoring records for members of the
proposed class have been lost, altered illegally, or destroyed.

Name or Social Security Number of First Petitioner:



[Document No. ORAUT-OTIB-0018_ | Revision No. 01 __| Effective Date: 08/09/2005 [Page 20 of 35_]

Table 7-1 (Continued). Whole-body count (WBC) results below which this
TIB method is still considered an overestimate.

Absorption | Time since Organ oF WEC resukmust
Radionuclide type last WBC (d)* concem be less than (nCi):

Wn-54 Form 365 Adrenals’ 05
Urinary bladder [Xs]
Brain 08
Breast 0.9
Gallbladder 0.6
Heartw ali 0.7
Kidneys OS
Giver 09
Muscle 0.7
Ovaries 0.6
Pancreas 0.6
Testes 1.2
Thyroid TO
Red bone marrow 13
Bone surface 21
Stomach 07
SI 06
ULI 08
To iE]
‘Skin ek)
Spleen OS
Thymus: 09
Uterus: 0.8
ET 03
Lung 74
Colon TS
Esophagus Uk}

Mn-54 ForM 730 Not applicable for all
orgars

Nb-95 MorS 90 Adrenals 18
Urinary bladder 34
Brain 50
Breast 13,
Gallbladder 35
Heart w all qT
Kidneys ©
Liver 53
Muscle 24
‘Ovaries 14
Pancreas 21
Testes 65
Thyroid Kx
Red bone marrow 57
Bone surface $1
Stomach 18
SI 13
ULT iT
pany 15
‘Skin oF
‘Spleen 22
“Thymus. 14
Uterus. 27
ET 3
Tung TZ
Colon ic}



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 [Effective Date: 10/01/2007 | Page9of74 |

5.1.1 Purpose

The purpose of this TBD is to describe internal dosimetry systems and practices at Oak Ridge
National Laboratory (ORNL) from late 1943 to the present. It discusses historical and current
Practices used in the evaluation of internal radiation exposure of monitored and unmonitored workers.
This TBD can serve as a supplement to, or substitute for, individual monitoring data.

§.1.2 Scope

ORNL began operations in early 1943. The startup of the Graphite Reactor and plutonium separation
activities in late 1943 introduced the potential for personne! exposures from intakes of radioactive
material. Laboratory operations involving radioactive materials increased over subsequent years as
ORNL expanded its roles in radionuclide production and development of chemical separations
processes.

Development of methods and techniques for internal monitoring (bioassay) was one of the many
priorities at ORNL in its early years of operation, because such methods simply did not exist.
Although ORNL used air sampling and radiological contamination monitoring programs as qualitative
indicators of internal exposure, urinalyses for various internal contaminants did not begin at the site
until about 1947. (A limited number of in vivo measurements appear to have begun at ORNL in
1959.) ORNL maintained early tolerance levels for airborne contamination based on “product’ (ie.,
?°Pu) concentrations in the air (Cox 1944). [These early tolerance levels for alpha and beta-gamma
contaminants were 3E-11 and 1E-07 \Ci/em’ respectively during the mid 1940s.} {In addition, the
Laboratory later established tolerance levels for materials such as ‘*'! and noble gases.

As written, this TBD is applicable for the period 1947 to the present.

Attributions and annotations, indicated by bracketed callouts and used to identify the source,
justification, or clarification of the associated information, are presented in Section 5.5.

5.1.3 Radionuclides of Concern

Because of the many diverse processes and experiments that took place there, a complete list of
radionuclides encountered at ORNL would be difficult to assemble. Radionuclides likely to produce a
measureable internal dose include uranium, activation products, fission products, and transuranics {1}.
The earliest urine sample results that were provided by ORNL were for the isotopes “Pu and "Sr.
The electronic data that were provided by ORNL for use in estimating isotopic MDAs came from a
project performed in the early to mid-1990s to convert hardcopy data over to a dBase IV database.
Funding ran out on the conversion project in the mid-1990s and the entire set of site data was never
completely converted, but a significant number of results were made available for our use. Table 5-1
lists radionuclides included in in vitro bioassay results provided by ORNL for the period from 1947 to
1988. The results were provided for estimating minimum detectable activities (MDAs) for various
analyses. The source of these data was an electronic database created by ORNL in the early 1990s
from hard-copy bioassay records. The sample size values in Table 5-1 are the numbers of analyses
for that nuclide included in the data ORNL provided. These values do not reflect the total number of
in vitro bioassays performed by ORNL in this period, because not all hard-copy records are in the
database.



[Document No. ORAUT-TKBS-0012-5 | RevisionNo. 01 _| Effective Date: 10/01/2007 | Page 32 of 74_|

REFERENCES

Auxier, J. A., D. M. Davis, and J. E. Turner, 1975, Applied Health Physics and Safety Annual Report
for 1974, ORNL-5055, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Auxier, J. A., 2003. Personal communications with Mike Botlenbacher, Knoxville, Tennessee, October
15.

Berger, C. D., and R. E. Goans, 1981, “A Comparison of the Nal-CsI Phoswich and a Hyperpure
Germanium Array for the Jn Vivo Detection of the Actinides,” Health Physics Journal, volume
40, pp. 535-542.

Berger, C. D., and B. H. Lane, 1981, Performance of a Large Hyperpure Germanium Detector Array
for In Vivo Detection of Low-Energy Photon and X-ray Emitters: Analytical Procedure and
Current Capabilities, ORNL/TM-7925, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Berger, C. D. and B. H. Lane 1982, The Oak Ridge National Laboratory Whole Body Counter:
intemal Operation Procedure Manual, ORNL/TM-8423, Oak Ridge National Laboratory, Oak
Ridge, Tennessee.

Berger, C. D., and B. H. Lane, 1984, Calibration of a Large Hyperpure Germanium Detector Array for
Actinide Lung Counting with a Tissue-Equivalent Torso Phantom, ORNL/TM-8723, Oak Ridge
National Laboratory, Oak Ridge, Tennessee.

Berger, C. D., 2003, personal conversation discussing the operation of the WBC facility during the late
1970s and early 1980s, November.

Berger, C. D., 2004, personal conversation discussing the operation of the WBC facility during the late
1970s and early 1980s, January.

Bloss, T. W., 1944a, Dust Surveys, memorandum to K. Z. Morgan, 1944, Oak Ridge, Tennessee,
August 5.

Bloss, T. W., 1944b, Dust Survey, memorandum to K. Z. Morgan, Oak Ridge, Tennessee, August 15.
Bloss, T. W., 1944c, Dust Survey, memorandum to K. Z. Morgan, Oak Ridge, Tennessee, August 17.
Bloss, T. W., 1944d, Dust Survey, memorandum to K. Z. Morgan, Oak Ridge, Tennessee, August 21.

Bloss, T. W., 1944e, Dust Survey, memorandum to K. Z. Morgan, Oak Ridge, Tennessee,
September 5.

Bloss, T. W., 1944f, Dust Survey, memorandum to K. Z. Morgan, Oak Ridge, Tennessee,
September 14.

Bloss, T. W., 1944g, Dust Survey, memorandum to K. Z. Morgan, Oak Ridge, Tennessee,
September 21.

Brodsky, A., 1992, "Exact calculation of probabilities of false positives and false negatives for low
background counting,” Health Physics, volume 63, number 2, pp.198-204.



[Document No, ORAUT-OTIB-0018 | Revision No.07___| Effective Date: 08/09/2005 L_Page 9 of 35__]

3.3

Intakes of particulate material with a documented particle-size distribution other than 5-ym
activity median aerodynamic diameter (AMAD) must be checked to ensure that the dose to the
organ of concem from this TIB is greater.

If site-specific limiting air concentrations were greater than those listed in Section 3.3, those
concentrations must be used. If the site-specific concentration was applicable to
contamination for a given radionuclide only and exposure was limited to that radionuclide, the
dose reconstructor can apply the intake to that radionuclide only. For instance, Mound and
Los Alamos National Laboratory had special limits for exposure to 2199 and Los Alamos had
special limits for general alpha-emitters and beta-emitters through 1966.

It does not necessarily provide an overestimate to respiratory tract organs for monitored
workers.

It does not provide an overestimate to the thyroid (regardless of monitoring), unless the worker
was employed only at a uranium facility or a plutonium-only facility (i.e., where there was no
chance of exposure to radioiodines).

ASSUMPTIONS

This TIB applies the following assumptions:

All significant intake exposure was covered by an air sampling program, or the radiation
protection program had a valid method to ensure that air concentrations did not exceed limits.
(For instance, airborne contamination might have been controlled by controlling the inventory
of radioactive material in an area and/or the physical form of the material.)

Chronic intakes were for 40 hours per week, or 2,000 hours per year, of particulate radioactive
material with a 5-ym-AMAD particle-size distribution.

The breathing rate was 9.6 m°/workday (ICRP 1994a) or 1.2 m°/hr averaged over an 8-hr day.

For sites or facilities with an exposure to alpha-emitting radionuclides that was principally
transuranics, chronic exposure was at 2 x 10°? yCimL.

For sites or facilities with an exposure to beta-emitting radionuclides, chronic exposure was at
2 10° pCi/mL.

For sites or facilities with an exposure principally to uranium, chronic exposure was at
6 x 107° yCi/mL.

Measured air sample concentrations were not always representative of the air concentration
breathed. However, when applied to the long-term, 40-hour-per-week, chronic-intake
assumption, it is assumed that the air sampling program was sufficient to prevent long-term
intakes from exceeding the limiting air concentrations. See Section 6.0 for additional
consideration of the possible nonrepresentativeness of air sampling.



AECU-817 at

where Co = count taken after 24 hr (tg)
Cy = count taken after 4 hr (t))
At = time tg ~ ty, in hours

X= decay constant of thorium B; A = 0.683

10.6
T = (e~AS4 (values of T for different values of At may be obtained from tables used in
counting room}

Therefore

Example: C, = 400 ct/min, C2 = 200 ct/min, at = 20 hr = 0,271 (from chart)

Cu 200 — (400 x 0.271. Bro iss = 120 ct/min

If tables are not available, use table of exponentials in handbook. These counts, Cy: can be
then converted to microcuries per sample by the following formula;

et/min ct/min x 10-6
He/sample = geo, x (3,7 x 1010 x 60) x 1078 geo. x 2.2
or for beta
beta = 2.17 x et/min x 1078 pe/sample
or for alpha

alpha = 0.91 x ct/min x 1076 pe/sampie

The surveyor then converts microcuries per sample to microcuries per cubic centimeter of air.

B. Absorption curves for beta activity are drawn for samples sent in thus enabling one with the
aid of appropriate tables to determine the kind of activity present. This is done for alpha and beta
activity.

It is conceivable that an emergency might arise when all counting facilities were destroyed or
unusable, With only a Victoreen 263 and a few lead bricks an emergency facility for simple counting
can be devised, With the tube removed from the stainless-steel shield and placed inside a lead cave,
counts may be made on hands, shoes, smears, and even air samples, For the permissible counting
rates given below, the G-M tube should be located 3 1/4 in, from the article checked.

Hands or gloves, 700 ct/min
Shoes outside, 10,000 ct/min
Smears or exposed surfaces, 200 ct/min

If the Victoreen 263 is not calibrated in counts per minute, a conversion of 3100 x mr/hr will
give an approximate counting rate.

Air samples may be counted for beta activity by inserting the tube inside the cylindrical sample
and counting. The tube must, of course, be protected from contamination by a thin paper covering,

The over-all geometry will depend upon the assembly used, but may be in the order of 20 per
cent,



4 AECU-817

1, Make a Study of Various Radiation Problems. One of these problems is shielding. In order
to protect personnel by shielding, it 1s necessary to know something about the nature of radiation and
how to shield one from this radiation, If you are working in the pile building, you know that you have
to put up a certain shield to stop the neutrons and would prefer to use light material such as paraffin
to slow down or stop the neutrons. You might mix paraffin with iron (for fast neutrons) or boron (to
capture thermal neutrons), but then, if you left it at that, you would run into trouble because fast
neutrons react with iron and slow neutrons with hydrogen to produce gamma radiation. It would then
be necessary, in addition, to surround the paraffin mixture with lead or heavy material to stop the
gamma rays which are produced by slowing down and capture of neutrons, Shielding of radiation can
be quite a complex problem.

2. Develop Methods of Preventing or Minimizing Radiation Exposures. Along with the idea of
making the study of radiation, the next step is making a study of how to prevent radiation exposure.
This is our primary objective. Shielding is not the only method of preventing radiation exposures.
For example, using psychological means or ‘‘horse sense’’ in dealing with people may be an impor-
tant element in obtaining cooperation and preventing overexposure to radiation. In preventing radia-
tion exposures you will find quite a challenge. If you develop any new methods of minimizing radia-
tion exposure, bring it to the attention of the supervisor so that he can pass it on to others.

3, Understand the Operation and Interpretation of Radiation-detection Instruments. I do not
mean by this that each of you needs to become an expert in electronics, A few of you may go on in
electronics or some other branch of physics. I hope that quite a few of you will take additional
college work and advance in these fields, A large majority will certainly not become specialists in
electronics, but I do hope that each of you will understand the fundamentals of instryment usage,
how the instruments operate, their limitations, etc. Film meters, for example, have certain inherent
limitations. Understand what they are good for and what they are not good for. Know why a film
meter may give a different reading from a Cutie Pie; if a supervisor wishes to know why these two
meters read differently, give the whole radiation story to the supervisor, and I am sure that you will
gain the complete respect of the people with whom you work.

4, Understand the Rules and Procedures for Radiation Protection. There ts no substitute for
this. I have to study the rules and procedures every day or so, but I suggest, to begin with, that you
obtain a copy of the manual of rules and procedures for working with radioactive material and prac-
tically memorize the rules and procedures for the area in which you are working.

As you acquire a knowledge of all operations relative to Health Physics, you become a senior
man and take on more responsibility. I have had this question asked: Here is a fellow in Health Phys-
ics working side by side with someone else. Both are doing the same type of work, but one is making
30 per cent more than the other. Why is this? The answer is that the person who is receiving 30 per
cent more not only knows how to do this particular job, but he knows how to carry out Health Physics
functions on 10 other operations, He is a much more useful man, although he may be doing just the
same type of work as a less experienced man on a particular job, Look forward to familiarizing
yourself with all the varied Health Physics survey functions. If you have research interests, add to
your education, and there is no reason why you cannot go on and become a research Health Physicist.
If you wish to be a supervisor, make a study of that field. There is plenty of chance for advancement
in Health Physics.

5. Maximum Permissible Levels of Exposure, You should know the maximum permissible
level of exposure for every operation with which you are associated, or know how to find them in a
hurry. It is not enough to know that the present level is 1/10 r/day and the new level is 3/10 r/week,
You want to know the permissible contamination on the hands, when a person should wear gloves,
and when he should wear complete protective clothing or just a lab coat. Never use the word ‘‘toler~
ance.”’ Replace it with “‘maximum permissible exposure.’’ No radiation is tolerable. We all take
Some radiation exposure and can never get away from the natural radiation of the earth, but we should
not tolerate any unnecessary radiation exposure,

6. Become Thoroughly Familiar with Employee Relations and Supervisory Relations. This will
be discussed further in later lectures. Later on you will be able to attend supervisory lectures,
Health Physics occupies a rather delicate position as far as management 1s concerned, We tie into



AECU-817 47

0.7CM
Fr

In the above diagram the plates of the Chamber are 0.7. cm apart and at difference of potential 1250
yolts. The chamber is suspended by rubber supports to prevent jarring.

For beta counting, a counter, with a mica window of approximately 3 mg/cm? thickness, and a
Rigginbotham scaler are used, A recorder and a timer are auxiliary equipment. The tube is mounted
in a vertical lead pig with a series of four shelves below the tube for supporting the sample. A sam-
ple counted on the top shelf has a geometry of approximately 30 per cent and on the bottom shelf
approximately $ per cent, The diagram (Fig. 5.1) shows the position of tube in relation to the four
Shelves.

+ Ate GoM ‘sr
t z 2 SHELF, ~ 50% GEOMETRY
z 31M 20 SHELF, ~ 10% GEOMETRY
gi 30 SHELF, ~ 4% GEOMETRY
4 47H SHELF, ~ 2¥2% GEOMETRY
Fig. 5.4

The tube and sample holder are all enclosed by a vertical lead pig for shielding from background
radiation. The geometry is as indicated in the diagram and is posted on the door of the shield. Con-
version factors are also supplied for converting from other geometrics to 10 per cent, (Smear sam-
ples should not have a large amount of loose activity since this will contaminate the counter.)

Mud, water, fish, etc., samples are counted for beta activity only.

Reliability of Counts

Probable Error. In addition to the instrumental and other errors, causes of statistical varia-
tions, the accuracy of a count depends on the number of counts taken, The greater the number of
counts, the less will be the probable error.

‘The probable error in the determination of the number of counts per minute from a sample is

given by the expression
Y(PE of BG)? + (PE total counts)?

ct/min of sample

As the ratio of the total counting rate to the background counting rate becomes higher, the con-
tribution of the probable error in determination of the background rate contributes less to the prob-
able error in the determination of the net counting rate. For routine counting where the ratio-of the
total counting rate to the background rate is greater than 5, it is our practice to approximate the
determination of the probable error of counting by use of the expression

0.6745 sum of counts

sum of counts



[Document No. ORAUT-OTIB-0034 | Revision No. 00 | Effective Date. 12/13/2005 | Page 210f29 |

* annual bioassay
value (50th percentite)
. not used in fit
fit

Figure A-9. Assumed *"U intake, 1951 to 1987, 50th-percentile results, Type M.

Table A-8. Type MU intake periods and rates.

Type MU intake rate (dpm/day! Type MU intake rate (dpmiday)

Year | 50th percentile | 84th percentile | GSD Year | 50th percentile | 84th percentile | GSD
1951 47.03 117.8 2.50 1970 1.641 4.809 2.93
1952 47.03 117.8 2.50 1971 1.641 4.809 2.93
1953 47.03 117.8 2.50 1972 1.641 4.809 2.93
1954 47.03 117.8 2.50 1973 1.641 4.809 2.93
1955 47.03 117.8 2.50 1974 1.641 4.809 2.93
1956 11.45 117.8 10.29 1975 1.641 4.809 2.93
1957 749 72.66 9.70 1976 1.641 4.809 2.93
1958 7.49 72.66 9.70 1977 1.641 4.809 2.93
1959 21.81 72.66 3.33 1978 1.641 4.809 2.93
1960 21.81 72.66 3.33 1979 1.641 4.809 2.93
1961 21.81 72.66 3.33 1980 1.641 4.809 2.93
1962 6.56 47.49 7.24 1981 1.641 4.809 2.93
1963 6.56 47.49 7.24 1982 1.641 4.809 2.93
1964 1.641 4.809 2.93 1983 1.641 4.809 2.93
1965 1.641 4.809 2.93 1984 1.641 4.809 2.93
1966 1.641 4.809 2.93 1985 0.294 0.659 2.24
1967 1.641 4.809 2.93 1986 0.294 0.659 2.24
1968 1.641 4,809 2.93 1987 0.294 0.659 2.24
1969 1.641 4,809 2.93



a AECU-817

use argon or nitrogen in place of air. Oxygen forms heavy ions by combining with electrons, result-
ing in long pulses of the order of milliseconds, Argon does not combine with the electrons, and since
the electrons have a very high mobility, they are collected very fast, producing puises less than 1-
usec duration, The sharp pulses can be amplified with greater discrimination against noise. The
long pulses are of the same order of frequency as microphonics and 60-cycle disturbances; this
phenomenon results in a poor signal to noise ratio.

For the counting of low alpha activity the chamber must be free of alpha contamination, Normal
brass gives from 2 to 4 alpha ct/hr/100 m2, The best treatment is usually a fresh-turned surface
untouched by any chemicals or fingers. Farabee’s alpha-counting chambers have a background of
about 0.1 ct/min, The chambers used here for radon have a background of about 50 ct/hr. These
chambers have a wall area of about 1500 em?,

Chang and Eng is a special apparatus composed of two ion chambers for the purpose of meas-
uring fast neutrons in the presence of gamma radiation, The two chambers are balanced against
each other so that the electrometer registers the difference of ionization in the two chambers. One
chamber is coated on the inside with about 1 mm of paraffin and filled with butane to a pressure of
about 20 psi. The other chamber is filled with 30 psi of argon, The butane chamber is sensitive to
fast neutrons as well as gamma radiation, while the argon chamber is sensitive to only gamma radi~
ation, The apparatus is then adjusted so that in presence of gammy radiation the same ionization is
produced in each chamber and thereby shows no Indication on the electrometer, in the presence of
fast neutrons, however, only one chamber shows additional ionization which is indicated on the elec-
trometer,

For the measurement of slow neutrons, the ionization chamber is usually coated with a layer of
boron.

4.8 GEIGER-MUELLER TUBES AND SCALING CIRCUITS
By G. S. Hurst

Geiger counters have been widely used for some twenty years as a very sensitive detector of
ionizing particles. The first electrical counter was devised by Geiger and Rutherford in about 1900,
when they were making preliminary investigations of the nature of alpha particles. This counter,
known as a “point counter,’’ consisted merely of a central electrode, the tip of which was drawn to
a very sharp point, and an outer electrode of an unspecified shape. 4 voltage was applied across the
electrodes; thus there was an extremely high electric field at the sharp point.

Geiger and Mueller, in 1928, realized that a much more efficient counter could be made by using
a very fine wire as the central electrode. The electric field, being proportional to Ur, where r is
the distance from the center of the wire, is very intense in the near vicinity of the wire and is uni-
form throughout the length of the counter. Geiger and Mueller further emphasized that the outer
electrode should be of cylindrical geometry in order to realize the highest counting efficiency pos-
sible. This type of counter bears the name of its first users,

Discharge Mechanism

This part of the discussion will apply to a counter of the type just described, filled with argon
(or one of the other inert gases) to a pressure of 6 to 10 cm Hg.

With reference to Fig. 4.1, suppose the pulse size is studied as a function of the voltage across
the counter. The alpha curve shows the history of the pulses produced by alpha particles, while the
beta curve shows the beta-pulse history. In the ionization-chamber region only those electrons
formed by the initial ionizing event are collected on the central wire. When the voltate is increased,
the field becomes strong enough to cause the electrons to produce additional ions by Collision; thus
the pulses become larger, as shown in the proportional region. As the voltage is increased further,
the limited proportional region is traversed, Finally the Geiger region is obtained, where all pulses
are of the same height, regardless of the magnitude of the initia! ionizing event. Notice that the size
of the pulse does depend, however, on the voltage.



DISCLAIMER

Portions of this document may be illegible in
electronic image products. Images are produced
from the best available original document.



14 AECU-81T

2.4 CHAIN REACTION AND THE PILE
By W. H. Ray

The chain letter was given as an example of chain reaction, The fundamental basis of chain
reaction is the release of several units of something, each of which may stimulate or trigger a re-
peat of that same action.

In chemistry, combustion is an example of chain reaction, In the example of a match, there are
chemicat compounds which will unite with each other and liberate heat, but not unless heated. You
scratch the match to start the chain reaction with heat produced by friction. The heat released by
the chemicals uniting then ignites the match stick, and combustion will continue, heat releasing more
heat, until the chain reaction is stopped for lack of fuel or because the heat is carried away faster
than it is generated, such as when you blow relatively cold air past the match to extinguish it.

When a uranium atom is split there result two atoms in addition to the neutrons released. The
sum of the mass of these two atoms plus the neutrons will be less than that of the uranium atom from
which they came. The loss in mass represents energy which we observe as the kinetic energy of the
particles, The majority of atoms do not fission into equal fragments; therefore there results a broad
spectrum of fission-product elements or isotopes (see Fig. 2.11 for curve of the abundance distribu-
tion of fission-product elements).

Nuclear chain reaction may take place where we have atoms (or, rather, nuclei) which may crack
apart and in so doing eject more than one neutron each (on the average there is more than one neu~
tron per fission). Each neutron has a chance of creating an additional fission if it is captured by an
atom that will split open (Fig. 2.12 shows a symbolic chain reaction and neutron multiplication).

There is quite a problem of utilizing the neutrons that are ejected, that is, of getting fissionable
nuelei to capture the neutrons so that they can split open and release more neutrons, In general,
nuclei accept neutrons more readily if they are not going by too fast. Neutrons, without electrical
charge, can come closer to a nucleus than ions, and the longer a neutron remains in the vicinity of
a nucleus, the more likely it is to be captured. If a neutron is going by rapidly, it is less likely to
be accepted.

Neutrons react with nuclei along their trajectory in several ways (Fig. 2.13): (1) pass by nucleus,
(2) elastic collision, (3) inelastic collision (energy loss by neutron), and (4) capture (may fission or
may be radioactive}.

if a neutron recoils from a heavy nucleus, there is not much loss in energy, but if it collides
with a light nucleus, the kinetic energy may be shared. Sometimes the neutron is captured by a nu-
cleus and transmutation results. In special cases a fission results, Usually the new nucleus is un~
stable and adjusts itself by an energy loss that is often accompanied by an ejection of a particle.
Where this is delayed, we call the new nucleus ‘‘radioactive.”’ A rare case would be where a nucleus
accommodates an extra neutron without some further adjustment.

The likelihood that a bombarding particle will have a certain reaction with a target substance
may be represented by what is called the ‘‘cross section’’ of that particular target for that particular
reaction with the specified particle at a specific kinetic energy (see Fig. 2.14).

Particles incident normally on a unit area of target will react (in the specific manner being
considered) as if only a fraction of the area were occupied by target nuclei, The unit of cross sec-
tion is termed a “barn” and is 10-24 em2.

To enhance the capture cross section of fissionable 0235, means for slowing down the neutrons
of high kinetic energy are employed. A materia! called a ‘moderator’ is introduced. Hydrogen is
ideal as far as mass is concerned, but it has an appreciable capture cross section for neutrons.
Deuterium, helium, beryllium, and carbon are acceptable moderators because of their low capture
cross sections, but the energy dissipated per collision decreases with the increasing mass,

If there are too few fissionable nuclei within mutual proximity, not enough fissioning captures
will result relative to the escape of neutrons from the assembly of fissionable atoms to sustain a
chain reaction. This critical mass required to permit a chain reaction can be reduced if means are
instituted to reduce the escape of neutrons. Enclosing the material within a reflector, usually a wall
consisting of moderator free from impurities, will return an appreciable fraction of the neutrons
that otherwise would escape.



[Document No-ORAUT-OTIB-0018 | Revision No. 01 _| Effective Date: 08/09/2005 [Page Sof 35__]

TABLE OF CONTENTS

Section

Publication Record.

Acronyms and Abbreviations

1.0) PULPOSC on... eeceeseecessessecsssecsecsnesecsecanenecnesnseenecnecusecucenssnecansesssasssnsasssnenscsuceneenacnacnensesessaussessgesssed 6

2.0 Background..

3.0 Applicability, Limitations, and ASSUMptiONS ...............s esses eeeseeseaseeneeneeneentenneenennesnerenateatteed 8
3.1 Applicability .
3.2 Limitations.
3.3 Assumption:

4.0 Inhalation Intake..
4.1. Intake Quanti
4.2 Choice of Radionuclid

5.0 Ingestion Intake............secscsssscsesseeseseesescneessenesessesseesnesseesssonsaversusanasensenssenenacsensneseacsnscnstansess 12

6.0 Dose Distribution

7.0 When To Use This TIB ........c.eccccecessssssessesnscecsessseeescssscsesesenseeneeseasesssasiserersnessestesensasisessenenes 14
REPETENOES oo... ccecccccecsesesesseesscecsesescscseeeesecstsrscsesnssseassrsteseenesereeceucacissucieneecessciseneseaeeseseteenenensearetited 28
Attachment

A Air Concentration Limits Prior to 1953 or Other Limits
B Validation of the Basic Assumption in This TIB.
B.1 Uranium.

B.2 Plutonium 31
B.3_ Fission Products.. -33
B.4_ Conclusion. :

Cc Site Applicability.......



[Document No. ORAUT-TKBS-0012-5 [Revision No. 01 __| Effective Date: 10/01/2007 | Page 550f74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 21 of 36

Annual Average MDAs for Pu-238 in Urine
Calculated from Recovered Data Between 1945 and 1990

06

MDA (dpm/24 h)

Figure A-8. MDA data for Pu-238 in urine (part 2 of 3).

Number of Sample Results Recovered per Year
Between 1945 and 1990 for Pu-238 in Urine

a

Number of Samples (N)

Figure A-8. MDA data for Pu-238 in urine (part 3 of 3).



34 AECU-817

Appendix B

For a given uniform concentration of the radioisotope in a body of any given shape, the rate at
which energy is produced in the body by radiation is proportional to the volume of the body, while
the rate at which energy escapes from the body is proportional] to the surface of the body. However,
for bodies of similar shape the volumes are proportional to the cube of any linear dimension, while
the surfaces are proportional to only the square of a linear dimension. Therefore, as the body is
made larger, the fraction of radiated energy which escapes becomes smaller; and, by making the
body sufficiently large, the fraction of the energy which escapes can be reduced to any negligible
value, In such a case essentially all the radiated energy is absorbed in the same volume from which
ft is radiated, Except for portions of the body very near the surface, any one smail element of vol-
ume is as favorably situated relative to radiating material as is any other. Therefore each element
of volume must absorb radiation at the same rate that radiation is emitted within it.

References

1, K. Z, Morgan, Tolerance Concentrations of Radioactive Substances, J. Phys. & Colloid Chem., $1
(No, 4), July 1947.
2. Koval to Morgan, June 22, 1945: Determination of Particulate Air-borne Long-lived Activity.

3.5 CONTAMINATION AND DUST PROPAGATION
By H. J. McAlduff

Before speaking of radioactive dust propagation and contamination, I should like to review very
briefly some of the facts that are common knowledge about nonactive particles or dust.

Humans are living almost constantly in an atmosphere of dust. The times when air is actually
clean and free from various types of particles are relatively uncommon. The atmosphere that we
might expect in this room might run to the order of several hundred thousand dust particles per
cubic feet of air, while other locations might be higher ur lower, depending on localized conditions.

When we speak of particles, or dust particles, they cam be grouped in certain classification.
‘The first is actual dust which consists of nothing more or less than solid material which has been
dispersed to extremely fine sizes, the size classification being considered from the visible range
with the unaided eye to less than 1 4: for solid material. These dusts are usually produced by grind-
ing, crushing, and blasting. The size range of fume particles ranges from 0.2 to 1p. Fumes are
produced by chemical reactions or processes, and examples would be the burning of magnesium and
zinc. Smokes are produced by incomplete combustion of carbonaceous material such as oil and coal,
‘The formation of mists and fogs is usually the result of condensation of H2O vapor on suitable nuclei
such as submicroscopic dust particles, and the size of the droplets depends on the conditions under
which they are formed. Oak Ridge is an excellent locale for smokes and fogs.

General Properties of Particles

Particulates with which we are concerned in respect to size lie between the regton of colloids
and those which can be seen readily with the unaided eye. They are larger than those in colloidal
systems and yet small enough to exhibit some of their properties and also to have properties not
possessed by large masses of the substance from which they are generated.

When solide or Nquids break up into finely divided particles, two important changes take place:
(1) surface area is greatly enlarged, and (2) space occupied by the dispersed material is increased
many times over the volume of the original mass. For instance, 1 cc of material crushed into parti-
cles 1 19 in size = 102 particles with a total surface area of 6 m”. If we assume a dust concentra-
tion of 100 million particles/it3 of air, this 1 cc of material will be dispersed in an air volume of
10,000 tt3. In other words, relatively small amounts of materia! will give large amounts of particu-
lates.



[Document No. ORAUT-TKBS-0012-5 | _Revision No. 01__| Effective Date: 10/01/2007 | Page 54 of 74

]

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 20 of 36

Number of Sample Results Recovered per Year
Between 1945 and 1990 for Pm-147 in Urine

Number of Samples (N)

Figure A-7. MDA data for Pm-147 in urine (part 3 of 3).

Year

Visual Comparison of Data te Estimated Pu-238 MDAs in Urine

1E+i —
‘Sample MDA
Est. MDA (95%CL on Mean)
SOth Percentile
1 E40 -——————~—_~--— |
z
z
g
Sie
=
£
2
DEAD few me
1B3 |
1945 1950-1955 1960-1965 297019751980 «19851990

Figure A-8. MDA data for Pu-238 in urine (part 1 of 3).



AECU-817

UNIT 7 80

UNIT 8. SOME PRACTICAL PROBLEMS 81

8.1
8.2
83
8.4
8.5
8.6
8.7

Guarding Hazardous Areas and Contamination Control
By G. C. Cain
Fulfilling Clothing Needs
By A. D. Warden and J. T. Sutherland
Wear and Care of Masks
By H. J. McAlduff
Shipment of Radioactive Material
By T. H. J. Burnett
Shielding
By R. 8. Thackeray
Decontamination of Personnel and Equipment
By H. A. McClearen
The Design and Operation of Radiochemical Laboratories
By W. BH. Ray
Survey Techniques and Refinements
By T. BH. J. Burnett and Associates

UNIT 9, QUALIFYING AS A HEALTH PHYSICST 101

9.1

(Evaluation)

Discussion of Achievements
By T. H. J. Burnett



L_ Document No. ORAUT-OTIB-0034 | Revision No. 00 _| Effective Date: 12/13/2005 | Page 19 0f29 |

* annual bioassay
value (50th percentile)
© notused in fit
fit

Figure A-6. Assumed “Sr intake, 1983 to 1988, 50th-percentile results, Type S.

« annual bioassay

value (50th percentile)
* — notused in fit
fit

Figure A-7. Predicted strontium excretion rate from independently fit intakes,
1951 to 1988, 50th-percentile, Type S.

Table A-6. Type S *°Sr intake periods and rates.
“Sr intake rate (dpm/day)

Start date | End date | 50th percentile | 84th percentile | GSD
1/11951 12/31/1953 24646 227730 9.24
1/1/1954 12/31/1954 7,232 68924 9.53
AAN955 12/31/1960 2379 13628 5.73
1/1/1961 12/31/1964 2,389 6431 2.69
1/1/1965 12/31/1983 795.0 1,917 2.41
4/1/1984 12/31/1988 425.5 1,917 4.51



AECU-817 67

5.10 THE PHOTOGRAPHIC EMULSION
By M. M. Shapiro*

In the past two years several physicists have applied the photographic-emulsion technique with
remarkable results: Powell of England, Gardner of this country, and Lattes of Brazil. Powell dis-
covered a new kind of meson (mass approximately 300mg}, He further demonstrated that one of
these heavy mesons can decay spontaneously into a lighter meson of mass about 200 times electron
mass, Gardner plus Lattes succeeded in producing mesons with a cyclotron at Berkeley.

The photographic emulsion consists of little grains of silver bromide (AgBr) embedded in a
matrix of gelatin, AgBr is the sensitive materi.

Agt Bré ions

Action of light: When a quantum of light strikes an AgBr grain, some silver ions are changed to free
silver, When the emulsion is developed, the entire grain is reduced to free silver (and, of course,
Brg). X rays also are capable of affecting photographic emulsion, Among the other particles which
can do so are: group A consisting of gamma rays and beta particles (electrons), and group B, con-
sisting of alpha particles, protons, deuterons (2), tritons (H), and fission fragments, All will
blacken the emulsion, but there is a rather significant difference in the manner in which group B
affects the emulsion from the way in which photons or electrons affect it.

If you have a source of alpha particles and expose them on a photographic plate and later develop
the plate and then inspect it under a microscope, you find the alpha particles have left not just black~
ening, but tracks whose lengths are of the order of 20 to 40 y, (1 yu = 1074 cm); 40x 103 mm = 0.04
mm, You must have a microscope to observe this,

Protons make the same kind of track, except that it does not ionize so densely. There are more
gaps in the track. If you look at these with bright-field illumination you will get black spots, and if
you have a dark field, then you will see bright spots on a black background.

Similar tracks are made by deuterons and tritons. Fission fragments produce exceedingly dense
tracks. In almost any emulsion their tonization is much denser. It is possible to see both fission
fragments in the same emulsion with suitable techniques.

A photographic emulsion is much simpler to operate than a cloud chamber, and one can study
the details of tracks left behind by all these particles in a photographic emulsion.

Neutrons can be studied with these emulsions by the secondary effects which they produce. The
blackening produced by X rays and gamma rays is due entirely to secondary effects, Electrons are
jonizing particles because they are charged, and in passing through the emulsion they produce a
blackening. The Eastman Company has succeeded in making film so sensitive that even electrons
can leav® a visible track, This ts quite an achievement,

Principal Advantages of This Technique

1, Continuous sensitivity, (a) The cloud chamber is only sensitive a smail fraction of the time;
(b) rare events: for use with valuable equipment where time is at a premium.

2, Larger volume, (a) The larger volume is from great stopping power of dense emulsion ver-
sus air, Approximately 2000 times stopping power of air (1600 to 2000); (b) convenient, compact
apparatus; (c) the typical thickness of emulsions is as follows: ordinary, approximately 10 2; spe-
cial, approximately 25 to 300 12; and typical, 50 to 100 p,

3. Permanent record

4, Inexpensive technique

5. Simple tn principle (however, care is required to interpret results)

*Notes taken by C. Scott; edited by M. M. Shapiro



[Document No. ORAUT-OTIB-0018

[Revision No. 01

[Effective Date: 08/09/2005 | Page 18 0f 35 _|

Table 7-1 (Continued). Whole-body count (WBC) results below which this
TIB method is still considered an overestimate.

Absorption ‘Time since Organ of WEC resukmust
Radionuclide type last WBC (d)* concem be less than (nCi):

Co-60 Mors S60 Adrenals 13
Urinary bladder 2
Brain 32
Breast 9
Gall bladder 30
Heart w all 8
Kidneys 19
Tver 4%
Muscle 17
Ovaries 18
Pancreas 16
Testes 36
Thyroid 1%
Red bone marrow 60
Bone surface 139
‘Stomach 15
$I 7
ULl 19
CLT 32
Skin 4
Spleen 17
Thymus 10
Uterus: 26
Er 5
Lung 20
Colon 26
ESophagus 0
Co-60 Mor S 730 Adrenals. 6
Urinary bladder 13
Brain 14
Breast 4
ali bladder 14
Heartw all 4
Kidneys. 9
Liver 21
Muscle 8
‘Ovaries 8
Pancreas 7
Testes 17
Thyroid 570
Red bone marrow 27
‘Bone surface 63
Stomach 7
SI 8
ULT go
TU 15
Skin 40
Spleen 7
Thymus 4
Uterus. 12
ET Zz
Tung og
Colon’ TZ
Esophagus. 4



AECU-817 2

Fast neutrons dissipate their energy by collision with nuclet in an absorber, Nuclei of low mass
are accelerated upon effective collision to an extent that they produce ions in a manner similar to
that of alpha particles. Since the energy shared per collision diminishes as the neutron’s kinetic en-
ergy is dissipated, an exponential type of absorption ts characteristic of this process.

Thermal neutrons have no excess of kinetic energy to dissipate. Energy ts released when ther-
mal neutrons are captured by nuclei in the absorbing medium. In addition to photons and/or particles
released in this capture, energy may be released subsequently if the new nucleus is radloactive.



62 AECU-817

greater error than the ionization type in which it is possible to zero out the reading due to contami-
nation, but contaminated instruments should not be used in survey work.

Precalibration Checks

The following checks are made prior to the calibration of an instrument: (1) contamination
check——-this contamination check is a secondary precaution; all instruments should be checked for
contamination before leaving the field, (2) batteries, (3) dial function, and (4) general physical con-
dition.

Appendix A: Data on Sources and Calibration Standards*

{a) Radium Sources as Gamma Radiation Standards. Radium, in equilibrium with its short-
lived decay products, is used as a standard for many source-strength or radiation-intensity meas-
urements. For this purpose the radium is enclosed in a thin platinum filter, usually of either 1/2
or 1 mm thickness.

For small radium sources within platinum filters of thickness greater than 0.3 mm, the source
strength is given by the equation

S$ = 8,98 (1-0.13t)M

where § is the source strength, measured in roentgens per hour at 1 cm, t is the thickness of the
Platinum filter, measured in mm, and M is the mass of radium, measured in milligrams,

For a 1/2-mm platinum filter, 8 = 8.4 M, and for a 1-mm platinum filter, S = 7.8 M.

It is frequently convenient to use the foot as a unit of distance in estimating the radiation hazard
from a radium source, If the source strength is expressed in roentgens per hour at 1 ft, the abave
equations become

Ry = 0.0097 (1 — 0.13t) M
for a 1/2-mm Pt filter,
Ry = 0.0090 M
and for a 1-mm Pt filter,
Ry = 0,0084 M

(b) Radium-Beryllium Neutron Sources. A common type of neutron source utilizes the alpha
particles from radium and its short-lived equilibrium products to produce neutrons by their reaction
with beryllium. Since three of the products of radium are alpha emitters, there are four alpha parti-
cles per disintegration of radium. Measurements on the neutron radiations from various Ra~Be
Sources give results varying from 8,7 x 108 to 14 x 108 neutrons/sec/g of radium.

Radium-beryllium sources must be shielded for both gamma and neutron radiation. Since the
Physiological effects of the two types of radiation are presumably additive, each type of radiation
must be shielded more effectively than if it alone were present.

(c) Procedure for Determination of Source Intensity. (1) Sources Contained in a 0.5-mm Pt
Tube, The radium-element content of the tube shall be obtained by adding to the ‘‘equivalent value’
appearing on the face of the Bureau of Standards Certificate for that source the percentage of this
value indicated in the note at the bottom of the certificate (for a 0.5-mm Pt tube this percentage is
customarily 4.3), The source intensity will then be obtained in milliroentgens per hour at i cm by
multiplying the radium element content, measured in milligrams, by the factor 8400 mr/hr/mg at 1 cm.

*Supplementary study material compiled by F. Western.



(Document No. ORAUT-TKBS-0012-5 | Revision No. 01__| Effective Date: 10/01/2007 | Page 650f74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 31 of 36

[. Sample MDA, ]
BE s!, MDA 0f 0.5 (95%CL on Mean) . i

LEH p=

1 E+0

IE}

‘Activity (dpmn/24 b sample)

1E2
1945 1950 1955 1960 1965 1970 1975 1980 1985 1990
Year

Visual Comparison of Data to Estimated Am-241 MDAs in Feces
: Figure A-15. MDA data for Am-241 in feces (part 1 of 3).

Annual Average MDAs for Am-241 in Feces
Calcutated from Recovered Data Between 1945 and 1990

e

2

2

MDA (dpm/24 h)

2

°

Figure A-15. MDA data for Am-241 in feces (part 2 of 3).



[Document No. ORAUT-OTIB-0018 | Revision No. 01 [Effective Date: 08/09/2005 [ Page 22 of 35

Table 7-1 (Continued). Whole-body count (WBC) results below which this
TIB method is still considered an overestimate.

Absorption Time since Organ of WEC resuk must
Radionuclide type last WBC (d)* concem be less than (nCi}:
Nb-95 — MoS 730 Not applicable for all
organs
Ce-T4T Mors 80 Adrenals T30
Urinary bladder 240
Brain 270
Breast 106
Gallbladder’ 180
Heart w ail 79
Kidneys T70
Twver 29
Muscle 180
Gvaries 107
Pancreas 140
‘Testes 320
“Thyroid 40
Red bone trarrow TOF
Bone surface 20
‘Stomach 37
SI 18
ULI 6
Lu 6
Kin 200
Spleen T8O
Thymus 708
‘Uterus 200
ET 3
Lung 4
Colon 6
Esophagus TOS
Ce-147 MorS 180 Adrenais 36
Urinary bladder 65
Brain 74
Breast 27
| Gall bladder 50
Feartw all 20
Kidneys 46
Liver 8
Muscle 48
Ovaries 27
Pancreas 39
‘estes BF
“Thyroid 67
Red bone marrow 29
Bone surface 6
Stomach 9
SI 5
ULT T4
TT TS
Skin’ 37
‘Spleen 45
‘Thymus: 27
Uterus. 57
ET 0.8
Tung OF
Colon Ta



Page 23 0f 74 _|

[ Effective Date: 10/01/2007 |

Revision No. 01

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[Document No. ORAUT-TKBS-0012-5 | Revision No. 01

[Effective Date: 10/01/2007 | Page 13 of 74_|

Table 5-4, Historical division codes (MMES 1995).

Codes {| DIVNUM Departments Division name
AC. 01 3390 Analytical Chemistry
Br 02 4455 Biology
CH 04 Chemistry
CT 03 3370 Chemical Technology
DO! 20 3200 Directors.

EC, GE 38 3060 Gen. Engr. and Construc.
ED 06 3480 Education

EL 07 3320 Electronuclear

EM 22 Finance and Materials
HE 23 3090 Health

HP. 08 3810, 4193, 3193, 3490 Health Physics:

Ic 09 3341, 3075 Instrumentation and Controls
te 24 Inspection Engineering
iS 25 3369, 3650, 4362, 3360 Isotopes

LP. 26 3094 Laboratory Protection
MA 10 3166, 3516, 3152 Mathematics

Mc 11 3470 Metals and Ceramics
MET Metallurgy

NP 12 3410 Neutron Physics.

OP 28 3639 Operations

PE 21 3016, 3078, 3003, 3004, 3062 | Plant and Equipment
Pu 13 3405 Physics

Py 30 3173 Public Information

PR 29 3107, 3141 Personnel

RC 14 3430, 4430 Reactor Chemistry

RE 16 4435, 3435 Reactor

RP 7 Research Participation
ss 18 3475 Solid State

TH 19 4460 Thermonuclear

Tl 34 3072, 3148 Technical information

§.1.7.2 Radioanalytical Abbreviations and Codes

Paper records contain abbreviation codes for recording analytes. These codes were often based on
the initials of the analysis or the isotopic abbreviation of the radionuclide. Tables 5-6 and 5-7 list the
codes for radionuclides measured in urine and fecal samples, respectively (Mani 1983) (MMES 1995).

5.1.7.3. Whole-Body Counting Results Codes

Table 5-8 lists codes used on some older cards and forms for documenting whole-body counting
results. Many hardcopy records have been consolidated into individual personal records folders.
However, this compilation is incomplete, with records only for employees with last names beginning
with A through G.

Recovered data show that analytical MDAs tend to remain fairly consistent for a number of years.
Abrupt changes in MDAs can be identified for groups of radionuclides during specific years. Following
these changes, the MDAs remain generally consistent in subsequent years. This “step-wise” pattern
allowed MDAs from several years to be grouped to obtain a single, representative MDA. Table A-2 in
Attachment A provides annual averages for periods of time where bioassay data were available.



AECU-817 ss

4.2 IONIZATION CHAMBERS
By F. J. Davis

The ionization chamber is a radiation-detection device in which the ionization produced by the
radiation is not amplified within the chamber. The dimensions are not critical; some have been built
as large as 1 m in diameter and others the size of thimbles. The chamber is essentially a simple
instrument composed of an outer container and an insulated central electrode. A voltage is applied
to collect ions produced within the chamber.

‘The currents produced from the ionization are usually quite small. A typical background is 2
ion pairs/cc/sec due to cosmic rays, 2 ion pairs/cc/sec due to stray gamma radiation, and 2 ion
pairs/ce/sec due to natural alpha contamination of the walls. The current produced by these 6 ion
pairs/cc/sec would then be

6x 4.8 x 10-10

= 19718
3x 109 ror amp/ee

‘The detector may be either a method by which the rate of voltage rise across a condenser whose
capacity may be the chamber itself is measured, or it may be a method by which the voltage is meas- !
ured across a high resistor of the order of from 108 to 10!2 ohms between the collecting electrode
and ground. Since the currents are so small, good insulators such as polystyrene, teflon, or fluoro-
thene are necessary. A guard ring showld be used In most cases. This eliminates the voltage drop
across the insulator between the central collecting electrode and the guard ring, thereby minimizing !
leakage currents. ;

For greater sensitivity, larger volumes or higher pressures are used. We have a chamber in |!
the laboratory which uses a pressure of 1500 psi. The usual gas in ionization chambers is air; how-
ever, chambers for special purposes use argon, butane, nitrogen, etc.

‘The voltages applied to the chamber need to be high enough to collect ions before they cam re-
combine. The usual voltage is of the order of several hundred volts, To determine the voltage to be
used, a curve of intensity of radiation against ion current should be run to determine linearity. If
chamber voltage is not high enough, the curve will not be linear, i.e., the ion currents for high in-
tensities will be too low. This results from the recombination being proportional to the square of
the number of ions produced.

‘The wall of the chamber should be smooth and free of sharp corners. The thickness of the walls
is determined by the type of radiation being measured. For gamma radiation the walls should be
thick to form large numbers of secondary electrons, which produce the ionization measured, but not
thick enough to absorb most of the gamma. The optimum thickness is usually considered as that
thickness which will just absorb the secondary electrons that are produced at the outer surface; this
is of the order of a few millimeters of aluminum. At atmospheric pressure almost all the ionization
is produced by secondary electrons formed in the walls, while at high pressures above 10 atm the
secondaries produced in the gas itself become the main component.

For measurement of beta rays, the thinner the wail, the better to eliminate the absorption in the
walls, Ionization is produced by the beta rays themselves. To get an idea of the thickness of the wall
to be used for measurement of beta rays, the Feather rule is found useful. This rule states that the
thickness in g/cm? which will just absorb all the beta rays of a certain energy is 0.54 E — 0.13, where
E is the energy in Mev. This rule is good for energies above 0.7 Mev. For ! Mev the thickness would
be 0.41 g/cm, and the walls of 2 chamber through which beta rays of this energy are to be measured
should be thin compared to 0.41 g/cm?.

The range of alpha particles from radium is about 10 me/cm2, so that if any window is used for
the measurement of alpha rays it must be thin compared to the range of the particle. In many cases
the alpha source can be placed inside the chamber, thereby eliminating any window absorption.

For the measurement of alpha rays, it is often better to count pulses rather than to measure the
ion current, If the ions are collected rapidly on an electrode, an electronic pulse is obtained which
can be amplified by a high-gain amplifier. lon chambers which which count the alpha particles usually



AECU-817 63

(2) Sources Contained in a 1,0-mm Pt Tube. The radium-element content of the tube shall be
obtained by adding to the ‘equivalent value’ appearing on the face of the Bureau of Standards Cer-
tificate for that source the percentage of this value indicated in the note at the bottom of the certifi-
cate (for a 1.0-mm Pt tube this percentage is customarily 9.4), The source intensity wil] then be
obtained in milliroentgens per hour at 1 cm by multiplying the radium-element content, measured
in milligrams, by the factor 7800 mr/hr/mg at 1 cm,

(d) Gamma Radiation from a Radium Source. The gamma radiation from a radium source is
almost entirely from its equilibrium products, RaB and RaC. The number of gamma quanta from
these products, per radium disintegration, is approximately 2.3, and the effective average energy
of the filtered radiation is approximately 0.8 Mev. The distribution of energy in the gamma radiation
from a radium source is shown approximately in Table 5.1.

Table 5.1 —Gamma Radiation from Radium Source

Gamma energy, Number of quanta per Origin
Mev radium disintegration of gamma
0.18 0.01 Ra
0.24 0.12 RaB
0.29 0.26 RaB
0,35 0.45 RaB
0.61 0,66 RaC
0.77 0.06 RaC
0.93 0,06 RaC
1.12 0.21 RaC
1.23 0.08 RaC
1.38 0.06 RaC
1,76 0,26 RaC
2.20 0.07 Rac

(e) Absorption by Lead of Gamma Radiation from a Radium Source. Since the radiation from
a radium source consists of quanta of many different energies, the radiation transmitted through a
lead shield becomes harder as the shield is made thicker. One cannot, therefore, give an absorption
coefficient or half-value layer that can be used for all thicknesses of shielding.
Table 5,2 uses data from R. D. Evans, J. Ind, Hyg. and Toxicol., 25: 253 (1943), and G. Failla,
Am. J, Roentgenol, Radium Therapy, 54: 574 (1945),

Table 5.2—Absorption of Gamma Radiation from Radium

Fraction of Fraction of

Lead, in.* radiation transmitted* Lead, in.* radiation transmitted*

0 1 3 0.017

0.125 0.8 3.5 0.01

0.25 0.64 4 0,005

0.5 0.43 5 0.0017

1 0.21 5.5 0.001

1.6 0.1 6 0.00058

2 0.058 14 0.0001

*All values are intended to be significant to two figures. Final significant
zeros are dropped for ease in reading.



98 AECU-817

departures from the inverse square law merge into an inverse first-power relationship. If the
source is an area that can be approximated by a circle, the intensity of radiation at a point along a
normal to its center is the product of the intensity at the surface times (1 - cos2 8), where @ is the
angle between the normal to the center of the circle and a line through the point to the circumference.

Combining these factors to attain protection from penetrating radiation involves integration of
laboratory equipment and the manner in which the work to be done is performed.

If, at a point in an infinite volume of the most radioactive solution to be handled, a radiation
dosage rate of 50 mrep/hr cannot be exceeded, solutions of this concentration may be manipulated
up to 2 hr/day with only contamination control. This concentration would be 20 we/liter if the aver-
age energy per disintegration were 1 Mev. (The concentration to yield the same irradiation is in-
versely proportional to the average energy.) For beta emitters the average energy can often be
considered to be one-third of the maximum,

Sources that will not produce more than 10 mr/hr at 4 in. can be manipulated practically with
tongs and forceps. Contact with radioactive materials should generally be avoided, The inverse
Square law intensity increases with amazing rapidity as distance approaches zero, In general, some
form of remote contro} is necessary for manipulating emitters of penetrating radiation regardless
of the extent of shielding. Once this has been mastered, there is little excuse to use shielding that
is too thin, ;

Nelson Garden of Berkeley has discovered that three-fourths of an experimenter’s time (in
radiochemistry) is consumed in setting up apparatus and executing a cold-run dress rehearsal. He
has devised standard jigs to fit in a shield or ‘‘cave”’ well equipped for remote-control manipulations,
‘These jigs are equipped and tested in an ordinary chemical laboratory without radioactivity before
being introduced into the cave for the actual experiment. At the conclusion of an experiment the jig
emerges from the other end of the cave into a decontamination room for salvage or disposal.

Laboratories should be large enough to permit good housekeeping and make manipulations easy.
Rooms should not accommodate more than one experiment or operation at a time. Only men working
on the job should have access to each laboratory cubicle, and adequate shielding should be incorpo-
rated in walls and entry labyrinth to protect all neighbors. Multistory buildings are a disadvantage
because of the need for additional shielding above and below. Shielding waste pipes and the need for
elevators to handle shielded containers also deserve consideration if a multistory building is to be
used for radioactive work.

Where it is necessary to wear gloves for an operation, it has been found helpful to have an as-
sistant to do the bare-handed work in order to minimize glove removal and reduce the temptation to
touch clean things with contaminated gloves. Not more than two men in a laboratory is advocated to
eliminate unnecessary gratuitous irradiation.

Laboratory Design Tailored to Needs

In addition te applying the factors we have discussed to a specific situation, there are other
considerations that should go into outfitting and operating a radiochemical iaboratory. Packing
units and formal rules can be correctly applied to specific installations only,

From the above, the desirability of radiation instrumentation in the laboratory is obvious.
Radiation instruments are jo be described in the next discussion,

The care required for solid, liquid, and gaseous-borne radioactive wastes is also to be discussed
subsequently, Please bear in mind that this is also a part of radiochemistry laboratory planning.



{Document No. ORAUT-TKBS-0012-5. [Revision No. 01 | Effective Date: 10/01/2007 [Page 6 of 74 _]

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AMAD activity median aerodynamic diameter

cm centimeter

cents counts

ents/dis counts per disintegration

ents/min counts per minute

dia diameter

dpm disintegrations per minute

EDP Electronic Data Processing

EEOICPA Energy Employee Occupational IlIness Compensation Program Act of 2000
ft feet

g gram

g/24-hr grams per 24-hour sample (fecal)

GSD geometric standard deviation

HPGe hyper-pure, intrinsic germanium

hr hour

ID identification

ICRP International Commission on Radiological Protection
in. inch

IREP Interactive RadioEpidemiological Program
IVGS in vivo gamma-ray spectrometer

keV kiloelectron-volt, 1,000 electron-volts

L liter

m meter

gm micrometer

MDA minimum detectable activity

me milliliter

mL/24-hr milliliter per 24-hour sample (urine)

min minute

MMES Martin Marrieta Energy Systems

mo month

MPBB maximum permissible body burden
MPOB maximum permissible organ burden
mrem mitlirem

N population size

Nal sodium iodide

Nal-Cst sodium iodide-cesium iodide phoswich detector

Nal(TI) thallium drifted sodium iodide



UNIT 9. QUALIFYING AS A HEALTH PHYSICIST (EVALUATION)

9.1 DISCUSSION OF ACHIEVEMENTS
By T. H. J. Burnett

Here, at the conclusion of this series of meetings, it should*be pointed out that the work pre-
sented so far comprises but a survey or introduction to your field of health physics. Much yet re-
mains to be learned by you, and only with further study and application can that necessary mastery
be achieved of those more advanced topics requisite to your full stature as a health physicist.

You, as developing scientists, should have a fertile interest in all ew progress in science as a
whole, reading currently both in the science sections of the lay press and in the available scientific
journals. The fertility of your interest will evidence itself by the suggested applications of what you
read to our current endeavors. Typical of this might be new instrument developments, new medical
discoveries such as the application of radioactive materials to treatment and/or research, new
atomic-research centers, new methods of producing radioisotopes, catastrophe planning for atomic-
weaponed attack, etc. For much of your further study, many of you will profit by home study, review,
or night-class attendance in such subjects as mathematics, physics, biology, and chemistry. The
latter subjects are locally available in evening courses most of the year.

Review and discussion by groups of your own of the material just covered in this course and
outlined in these notes should be quite valuable to sustain your recently developed skills at their
maximum usefulness.

Being alert to new problems that arise in your respective responsibility areas will enable you
not only to be of greater service value to your associates, but will give you more interest in your
work, Much valuable applied research to field problems can be done with the facilities at your dis-
posal. Of these, to cite only one example, would be differentially shielded film studies of the ex-
posure -energy distributions in typical work areas and operations, permitting better correlation of
exposure monitoring results with prior instrument surveys.

As to the direction your further studies in health physics should take and what subjects ought
to be included, no attempt is herewith made to present any thoroughly weighed program for an ad-
vanced course, However, it may be readily seen with even a little reflection that the following sub-
jects might very appropriately be treated:

1. Additional work on radioactive decay and the growth of radioactive daughter products. This
would include further study of the exponential equations involved, preferably without the handicap of
avoiding calculus.

2. The above will integrate well with a more thorough and basic study of tolerances and toler-
ance concentrations, which involve consideration of both biological elimination and radioactive decay,
considered as providing a combined exponential relationship of time and dosage.

3. More consideration given to instruments and instrumentation will be profitable since it will
develop a further degree of skill in the choice and proper usage of the appropriate instruments, as
well as the improvement of interpretation,

4, The above will profitably be associated with further and less naive study of ionization mecha-
nism, nuclear structure and reactions, beta decay, and kindred topics. Knowledge of the interaction
of radiation with matter, especially tissue, can here be given a better basis.

5. Emphasis can deservedly be given here to more detailed study of the biological aspects of
radiation injury for the different types of radiation, as well as to a survey of the results to date of

AECU-817 101



{HITS OF a)

AECU-817
THOK WALLED
1D *
LEAD “HOWITZER P\,_-scarterto a SCATTERED 128
NARROW BEAM OF RAYS FROM GOLD @ PART, CONCEPT
e Fo PATH OF ATOM
| EVIDENT
ACTION SIZE
rian ound OF + CHARGED
POLONIUM @ FOIL GENTER
RAY SOURCE
Faw”
SCINTILLATION SCREENS TELESCOPE FOR on ‘SCATTERED
FOR VISIBLE a PARTICLES COUNTING FLASHES “Or @ PARTICLE

Fig. 2.1—Rutherford experiment.

(NGIDENT ELECTRON
eS

PA ~

bss ADDITIONAL ELECTRON

fo ce REPULSED
x2

Fig. 2.3—Evidence of electrons
as components of atoms.

ENERGY LEVEL

DIFFERENT
TRANSITIONS OR JUMPS IN
WHICH A PHOTON OF ENERGY
1S EMITTED

Fig. 2.5

Fig. 2.2—Assumed picture of scattering
behavior from % distribution.

HEAVY, DENSE
SMALL. POSITIVE
INNER NUCLEUS
AREA OF DISTRIBUTED
NEGATIVE CHARGES TO

MAKE ATOM NEUTRAL

Fig, 2.4
LITHIUM FLUORINE VALENCE ELECTRONS
ATOMS, NEUTRAL
UNCOMBINED
JONIZATION-————o—|
PLANE

COMBINED, SHARING
ONE ELECTRON

NOW IONS CHARGED
Aer MOBILE IN AN
PELECTRIC FIELD

EXGESS OF EXCESS OF
4 POSITIVE CHARGE {NEGATIVE CHARGE

Fig. 2.6



[Document No. ORAUT-OTIB-0018 __[ Revision No. 01 [Effective Date: 08/09/2005 | Page 20f35 |

[ EFFECTIVE |
DATE

PUBLICATION RECORD

REVISION
NUMBER

DESCRIPTION

11/23/2004

00-A

New technical information bulletin to provide information for internal
dose overestimates for facilities with air sampling programs. Initiated
by Donald E. Bini.

01/06/2005

Incorporates internal review comments. Initiated by Donald E. Bihl.

03/10/2005

Incorporates changes requested by NIOSH, including Atfachments B
and C. Updates RU contaminant levels. Eliminates discussion about
ORAUT-OTIB-0002 related to ORAUT-OTIB-0018. Initiated by
Donald E. Bihl.

03/18/2005

First approved issue. Initiated by Donald E. Bihl

05/05/2005

01-A

Draft revision fo 1) incorporate into Table 4-3 Nb-95 type S and Tc-
99 type F beta choice and replace RU with U-234 alpha choice, 2)
add Table 7-2 which provides excretion values for uranium and
radiostrontium below which the default intakes was still valid, 3) add
special air concentration values for LANL into Table 4-1,4) correct
error in the ingestion intake formula #4. Text was added or modified
in Sections 1.0 through 5.0, 7.0, and Attachment A. No entire
Sections were deleted. Initiated by Donald E. Bihl.

06/10/2005

01-B

No further changes occurred as a result of formal internal review.
Initiated by Elizabeth M. Brackett and Donak E. Bihl.

07/15/2005

01-C

Incorporates NIOSH review comments. Initiated by Donald E. Bini.

08/09/2005

01

Approved issue of Revision 07. Training required: As determined by
the Task Manager. Initiated by Donald E. Bihl.



[Document No. ORAUT-OTIB-0018 __[ Revision No. 07 | Effective Date: 08/09/2005 [Page 25 0f 35 _]

Table 7-2 Urinary excretion results below which this TIB method is still considered an

overestimate.
Radionuclide or | Nature of | Time since last Excretion rato
analysis type exposure urinalysis (d)_| Organ of concern pela pou
Radiosrontum or | Type F 30 Adrenas 70 TOS
$r-80 Uninary badder 95 oe
Brain T90 136
Breas, 174 124
Gall bladder” 407 286
Heart wall 50 179
Kidneys 240 177
Liver oe BSF
Muscle 214 TSS
‘Ovaries 265 180
Pancreas, 266 190,
Testes 273 752
Thyroid 228 163
Red bone mamrow 9 6
Bone surface T zy
‘Stomach 210 TSU
SI ra) T88
LT o4 67
TO El 67
Skin 60 Ti
Spleen 240 77
Thymus 230 164
Ulerus SE T3T
ET aT TOT
Tung 3,200 i]
Colon 8s 63
Esophagus 2350 T64
780 Adrenals 540 386
Urinary bladder 30 T36
Brain 380 277
Bread KG Zi)
Gall badder BU: 373
Heart wal BU 357
Kidneys 480 ELE
Tiver 183: 130
Muscle 427 305
‘Ovaries. 529 378
Pancreas B32 380
Teses 426 BUF
Thyrod a7 325
Red bone marrow 17 T
Bone surface 23 16
‘Stomach 420 300
ST 439 314
Ur 189 135
TL 170 127
ny 319 228
pean a9 KLy3
Ratiosonium or | Type F TE0 TRYAUS Kiny 328
$1-90 cont. Uteris 537, 384
ET 26: 207
Tung 6400 4571
Colon 176 126
Esophagus 459 328
ELI Adrends OF 135
Utinary Bacder 72 BT
‘Brain 144 TOS
Break 73: oF
Gall bladder, 304 217
Heart wat 89. 135.
Kidneys 782 130
Liver 694 496



Special Exposure Cohort Petition U.S. Department of Health and Human Services

under the Energy Employees Occupational Centers for Disease Control! and Prevention
tiiness Compensation Act National Institute for Occupational Safety and Health

| OMB Number: 0920-0639 Expires: 09/20/2013
Special Exposure Cohort Petition — Form B Appendix — Petitioner 2

8

vivor Information — Complete Section B if you are a Survivor or representing a Survivor.
B.1 Name of Survivor:

Mr./Mrs./Ms. First Name Middle Initial Last Name
B.2 Social Security Number of Survivor:
B.3 Address of Survivor:

Street Apt # P.O. Box

City State Zip Code
B.4 Telephone Number of Survivor: )
B.5 Email Address of Survivor:

B.6 _—_—Retationship to Employee: Q Spouse Q Son/Daughter Q Parent
Q Grandparent Q Grandchild

Employee Information — Complete Section C.

C.1 Name of Employee:

Mr/Mrs./Ms. First Name Middie Initial Last Name
C.2 Former Name of Employee (e.g., maiden name/legal name change/other):

Mr/Mrs./Ms, First Name Middle Initial Last Name
C.3 Social Security Number of Employee:
C.4 = Address of Employee (if living):

Street Apt # P.O. Box

City State Zip Code
C.5 Telephone Number of Employee: « ) :
C.6 Email Address of Employee:

C.7_ Employment Information Related to Petition:
C.7a Employee Number (if known):

C.7b Dates of Employment: Start End
C.7¢ Employer Name:
C.7d Work Site Location:

C.7e Supervisor's Name:

Name or Social Security Number of First Petitioner:



{Document No. ORAUT-TKBS-0012-5 | Revision No. 01 __| Effective Date: 10/01/2007 | Page 17 of74 |

5.2.1.2. Minimum Detectable Activities after 1989

MDA values for present (since 1989) samples are available with each sample. Table 5-10 lists
typical, current MDAs for radionuclides of concern in urine and feces.

Table 5-10. MDAs for in vitro samples after
1989 (McLaughlin 2002).

MDA

Isotope (dpm/24-hr sample)
H-3 9,100
C-14 4,480
P-32 1
$r-90 3
Te-99 200
1-134 20
Np-237 0.02
Th-232 0.02
U-232 0.02
U-233 0.02
U-234 0.02
U-235 0.02
U-238 0.02
Pu-238 0.02
Pu-239 0.02
Am-241 0.02
Cm-244 0.02
C252 0.02
Bk-249 26

5.2.2 Counting Methods for in Vitro Samples

Several counting methods have historically been available at ORNL for determining radioactivity in in
vitro samples. The following sections discuss alpha spectrometry, liquid scintillation, zinc sulfide
scintillation, gamma spectrometry, and beta counting using a gas flow proportional counter.

5.2.2.1 Alpha Spectrometry

ORNL uses alpha spectrometry in the analysis of nuclides that decay primarily by alpha emission,
with only very low-energy photons or none at all. A tracer is added to the bioassay sample before
analysis begins to determine the chemical yield of the process. The radioelements are chemically
separated from the sample and electrodeposited on stainless-steel disks. Plutonium and the other
transuranic elements can be analyzed sequentially; uranium analysis requires a separate sample.

5.2.2.2 Liquid Scintillation

Liquid scintillation is used for the analysis of low-energy, pure beta emitters, specifically °H and “C.
One milliliter of a urine sample is mixed with a scintillation cocktail for analysis.



C Document No. ORAUT-OTIB-0034 | RevisionNo.00 | Effective Date: 12/13/2005 | Page 14 of 29 ‘|

REFERENCES

ICRP (International Commission on Radiological Protection), 1959, Report of ICRP Committee II on
Pemnissible Dose for Internal Radiation, Publication 2, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1993, Report of a Task Group of ICRP
Committee II on Dose Coefficients for Intakes of Radionuclides by Workers, Publication 68,
Pergamon Press, Oxford, England.

Martin Marietta, 1983, Environmental Monitoring Report — United States Department of Energy, Oak
Ridge Facilities — Calendar Year 1983, Y/UB-19, Oak Ridge, Tennessee.

Martin Marietta, 1984, Environmental Monitoring Report — United States Department of Energy, Oak
Ridge Facilities - Calendar Year 1984, Y/UB-20, Oak Ridge, Tennessee.

ORAU (Oak Ridge Associated Universities), 2004a, Analysis of Coworker Bioassay Data for Intemal
Dose Assignment, ORAUT-OTIB-0019, Oak Ridge, Tennessee.

ORAU (Oak Ridge Associated Universities), 2004b, Technical Basis Document for the Oak Ridge
National Laboratory — Occupational Intemal Dose, ORAUT-TKBS-0012-5, Oak Ridge,
Tennessee.

Union Carbide, 1975, Environmental Monitoring Report — United States Energy Research and

Development Administration, Oak Ridge Facilities - Calendar Year 1975, YIUB-4, Oak Ridge,
Tennessee.

Union Carbide, 1976, Environmental Monitoring Report — United States Energy Research and

Development Administration, Oak Ridge Facilities - Calendar Year 1976, YIUB-6, Oak Ridge,
Tennessee.

Union Carbide, 1977, Environmental Monitoring Report — United States Department of Energy, Oak
Ridge Facilities - Calendar Year 1977, YIUB-8, Oak Ridge, Tennessee.

Union Carbide, 1978, Environmental Monitoring Report — United States Department of Energy, Oak
Ridge Facilities- Calendar Year 1978, Y/UB-10, Oak Ridge, Tennessee.

Union Carbide, 1979, Environmental Monitoring Report — United States Department of Energy, Oak
Ridge Facilities - Calendar Year 1979, Y/UB-13, Oak Ridge, Tennessee.

Union Carbide, 1980, Environmental Monitoring Report — United States Department of Energy, Oak
Ridge Facilities - Calendar Year 1980, Y/UB-15, Oak Ridge, Tennessee.

Union Carbide, 1981, Environmental Monitoring Report — United States Department of Energy, Oak
Ridge Facilities - Calendar Year 1981, Y/UB-16, Oak Ridge, Tennessee.

Union Carbide, 1982, Environmental Monitoring Report — United States Energy Department of
Energy, Oak Ridge Facilities - Calendar Year 1982, Y/UB-18, Oak Ridge, Tennessee.


[Document No. ORAUT-OTIB-0034 | Revision No.00 | Effective Date. 12/13/2005 | Page 22 of 29 |

* annual bioassay
value (60th percentile)
© not used in fit
fit

Figure A-10. Assumed
Type S.

“U intake, 1951 to 1955, 50th-percentile results,

* annual bioassay
value (50th percentile)
© not used in fit
fit

Figure A-11. Assumed “*"U intake, 1956, 50th-percentile results, Type S.

* annual bioassay
value (50th percentite)
* not used in fit
fit

Figure A-12. Assumed
Type S.

U intake, 1957 to 1958, 50th-percentile results,



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 __| Effective Date: 10/01/2007 | Page 560f74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 22 of 36

LE+ _ 7 ~

Activity (dpm/24 h sample)

1E3
1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Visual Comparison of Data to Estimated Pu-239 MDAs in Urine

Figure A-9. MDA data for Pu-239 in urine (part 1 of 3).

Annual Average MDAs for Pu-239 in Urine
Calculated from Recovered Data Between 1945 and 1990

MDA (dpm/24 h)

Figure A-9. MDA data for Pu-239 in urine (part 2 of 3).



[Document No. ORAUT-OTIB-0034 | Revision No. 00 | Effective Date: 12/13/2005 | Page 28of29 |

Figure A-23. Assumed “*Pu intake, 1951 to 1988, 50th-percentile composite
results, Type S.

Table A-11. Type S °Pu intake period and rates.

"Tam intake rate (dpmiday)

Start date | Stop date { 50th percentile | 84th percentile | GSD
1951 4988 4.45 22.86 §.50

* annual bioassay
value (60th percentile)
. Not used in fit
fit

Figure A-24. Assumed “'Am intake, 1968 to 1984, 50th-percentile results,
Type M.



AECU-817 23

Another equivalent unit, which bears a closer relation to the roentgen, has recently been pro-
posed independently by Dr. K. Z. Morgan and some British investigators. This unit is called the
“acre,” or air cavity roentgen equivalent, by Morgan and simply the ‘J unit’’ by the British. It
makes use of a hypothetical air cavity inserted in the tissue at the point at which it is desired to
know the dosage. One acre corresponds to the absorption of 83 ergs of energy per gram of air in
the assumed cavity. If the cavity is restricted to a very small size, the radiation going through the
cavity, and particularly the secondary electrons, will be equal to what would have penetrated the
tissue or other material which the air replaced. While the unit is designed primarily for biological
work, it is perfectly possible to define a concrete, uranium, steel, or lead acre without loss of
consistency.

With these units, the roentgen, rep, and acre, it is possible to correlate the effects of radiation
on both living and inanimate material, either with calculated internal dosages or with external’ meas-
urements of the exposure. It is also possible to relate the calculated dosage to the externa] measure-
ment if sufficient information about the radiation involved is available.

3.2. PERMISSIBLE RADIATION EXPOSURE
By Lucille Leibowitz

Permissible radiation exposure is limited by the tolerance dose. The first approach to any dis-
cussion of the subject is a strict definition of the terminology.

The term ‘‘dose’’ is referred to as the energy absorbed in the tissue,

The ‘‘dosage rate’’ is the dose per unit time. There was a tendency in the earlier literature to
use the words ‘‘intensity’’ and ‘“dosage rate” interchangeably. This is very confusing, since inten-
sity is measurement of energy flux, while dosage rate is the energy absorption per unit time.

The ‘‘tolerance dose’ is the dose to which the body can be subjected without production of harm~
ful biological effects. The fact that all radiation produces damage cannot be overemphasized, For
this reason it is advisable that persons should strive to keep radiation exposure to a minimum and
think of maximum allowable rather than tolerance amounts of radiation.

Actually we are exposed continually to background radiation due to cosmic rays and the radio-~
active content of the air; however, such low levels are not known to be harmful to man, although it
does produce certain biological changes in small organisms.

The awareness for the need of a tolerance dose arose from the discovery of severe skin derma-
titis among those working with X rays. Becquerel, who carried a tube of radium in his vest pocket
for demonstration purposes, developed a reaction of the underlying skip.

In 1902 Rollins attempted to formulate some idea of a tolerance dose. He suggested that if a
photographic plate is not fogged in 7 min, the radiation was not of harmful intensity. This dose
corresponds to 10 to 20 r/day for soft X rays in modern terminology.

In 1903 damage to the reproductive organs of animals was observed which could be traced to
overexposure to radiation, In 1904 the blood-forming organs were recognized as sensitive indicators
of radiation damage, The death of Madame Curie is attributed to aplastic anemia, which was undoubt-
edly caused by radiation damage to the blood system.

The fact that dermatitis was the first known reaction to overexposure led to the concept of a skin
erythema dose. This dose was given as the threshold for undesirable skin reactions, The skin, how-
ever, varies in sensitivity over the body. It also varies with the individual, so that no satisfactory
limit could be set. In addition, dermatitis is a late reaction of radiation, and the more sensitive cells
could be severely damaged by a dose which would have no effect on the skin,

In 1915 Russ delivered a paper before the British Roentgen Society in which he encouraged more
precaution in the handling of X rays and radioactive materials, His recommendations were ignored
during the war when protection was sacrificed for expediency. After the war repercussions occurred
in the form of death to many prominent radiation workers in this country and in Europe. The unfavor-
able publicity warrented action,

The American Roentgen Ray Society formed a committee in 1920 to recommend protective meas-
ures, At about the same time, the British X-ray and Radium Protection Committees published their



[Document No. ORAUT-TKBS-0012-5 [Revision No. 01 _| Effective Date: 10/01/2007 | Page 58 of 74

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 24 of 36

Annual Average MDAs fos Pu-241 in Urine
Calculated from Recovered Data Between 1945 and 1990

30

MDA (dpm/24h)

Figure A-10. MDA data for Pu-241 in urine (part 2 of 3).

Number of Sample Results Recovered per Year
Between 1945 and 1990 fer Pu-242 in Urine

40

‘Number of Samples (N)

Figure A-10. MDA data for Pu-241 in urine (part 3 of 3).



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01__| Effective Date. 10/01/2007 | Page 19 0f74_|

§.2.3.3 Environmental Uranium

The following paragraphs were taken entirely from the Oak Ridge National Laboratory Internal
Dosimetry Program Technical Basis Document (McLaughlin 2002). They are included as an aid to
dose reconstructors in the interpretation of uranium bioassay results for ORNL workers.

Environmental levels of naturally-occurring uranium are found throughout eastem
Tennessee. The environmental activity levels in the immediate area surrounding Oak
Ridge are sufficiently high such that dietary intake of uranium is detectable with 24-
hour urine samples. A urinary uranium background study was conducted in the mid
1990's using non-occupationally exposed employees to quantify the range of typical
background uranium excretion. Based upon the results of that study, a discrimination
level (set at the 99th percentile level) of 0.14 dpm/day was established for both U-234
and U-238 to differentiate between environmental and occupational exposure to
uranium. A value of 0.25 dpm/day is applied to total uranium results. Plots of the
observed uranium excretion distributions for U-234, U-238, and total uranium are
provided below.

An activity ratio of U-234 to U-238 of approximately 2 to 1 has been observed within
the analyzed local potable water samples. Activity ratios in this range have been
reported for various aquifers (Osmond). Though U-234 and U-238 should be in secular
equilibrium within nature, the observed enrichment of the isotope U-234 is believed to
be caused by several factors which include the direct transfer of U-238 decay products
across a solid/liquid phase boundary by alpha recoil and differences in solubility
between uranium decay chain members. Recognizing this trend, uranium bioassay
results that are less than 0.2 dom/day that do not exhibit a U-234 to U-238 activity ratio
of 2:1 should be considered suspect and investigated.

5.3 IN VIVO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND
REPORTING PRACTICES

5.3.1 Shielded Counting Room

The ORNL Whole-Body Counting (WBC) facility, sometimes referred to as the In Vivo Gamma-Ray
Spectrometer or VGS), is in Building 2008 in the northwest corner of the Main Plant area. Several
Health Physics Division annual progress reports (Morgan, Snyder, and Struxness 1961, 1963, and
1966) indicate that the WBC facility began operation in June 1960 and another report (Brown,
Patterson, and Abee 1971) indicates a May 1960 date; a more recent document (Watts et al. 1995)
indicates several counts occurred earlier. Thus, it is likely that the WBC facility began limited
operations in 1959.

The main counting room has inner dimensions of 10 ft by 10 ft by 10 ft. Its wails consist of four layers
of pre-World War ff steel with a total thickness of 14 in. Some documents called it the iron or big
toom. The first recorded count in the WBC facility occurred on July 13, 1959, and, “was a background
count [conducted] in the corner of the steel room with the door not in place and the roof incomplete”
(Watts et al. 1995). The room was completed on July 24, 1959 (with the exception of a 0.125-in. layer
of “special low-radioactivity lead” added to all inner surfaces in 1960 to reduce background radiation
levels) and was used for a time to conduct background studies of paint and interior samples and to
count biological samples such as milk, grass, and cow thyroids. The first recorded lung count of an
employee occurred on May 19, 1960. Almost immediately after (May 19 to 20, 1960), it was used to
conduct lung counts of three employees involved in an onsite contamination incident. The original



[Document No. ORAUT-TKBS-0012-5 | Revision No.01__| Effective Date: 10/01/2007 | Page 37 0f74 |

A1.2

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 3 of 36

ALIQUOTVOL — The volume or mass of the sample analyzed
SAMPLEVOL — The sample volume or mass collected
COUNTDATE - The date the aliquot was counted

STARTEDATE -— The date sample processing began

STOPDATE -— The date sample processing ended

RECOVERY -— The chemical yield, in percent

EFFICIENCY — The detectors counting efficiency, in percent
EFF_FACTOR - The reciprocal of the detector’s counting efficiency
COUNTMIN ~ The sample count time, in minutes

T_BKG_RATE - The total background count rate, in counts per minute
BKG_MIN — The background count time, in minutes
BKG_RGNT_R — The count rate of any reagent used
BKG_COUNTS ~ The number of background counts

MDA — The minimum detectable activity

DATA CLASSIFICATION

The 66,204 individual records in the database were classified into one of four categories:

Duplicate Records - 272 duplicate records were found. These were marked with a “D” in the
record's validation field.

Usable without Modification - 58,972 of the records collected (89%) were usable without
qualification.

Usable with Modification - Some of the records collected contained input errors that were
easily identified. Examples include a nuclide entry of C-137, instrument efficiencies listed in
fractions instead of percent, and dates from the turn of the century. Other information in the
database enabled correction of some of these records. Such modifications were designated
by placing an “M” in the records validation field and entering the reason for the change in the
comment field. These records comprise 3% (2,193) of the database.

Unusable Records - 4,767 of the records recovered (7%) were determined to be unusable in
their present form. Most of these records are incomplete or contain gross errors. These were
marked with an R in the validation field. An explanation for a record’s rejection was entered in
the comment field.

The resulting database provided all the raw data used to determine the MDAs of radionuclides in
urinalysis calculated in this attachment.

A.1.3

SYNOPSIS OF RECOVERED DATA

The database queries recovered 61,165 usable records. Of these, 59,288 are clearly records of
urinalyses and 1,859 are records of fecal analyses. Special analyses made up the remaining 18
records. Seventy-five different analyte labels are listed, although some are clearly different names for



Special Exposure Cohort Petition U.S. Department of Health and Human Services

under the Energy Employees Occupational Centers for Disease Control and Prevention
lliness Compensation Act National Institute for Occupational Safety and Health

OMB Number: 0920-0639 Expires: 09/20/2013
Special Exposure Cohort Petition — Form B Appendix — Petitioner 2

Use this Appendix for Petitioner 2.

This appendix fom is to be used as needed. Petitioner 2, or his or her representative, should complete the
parts applicable to him or her.

Refer to the General Instructions on completing petitioner information for Parts A, B, or C.

Ifyou need more space to provide additional information, use the continuation page provided at the end of
the form and attach the completed continuation page(s) to Form B.

Except for signatures, please PRINT all information clearly and neatly on the form.

O An Energy Employee (current or former), Start at C

If you are: QA Survivor (of a former Energy Employee), Start at B

A Representative (of a current or former Energy Employee), Start at A

Representative information — Complete Section A if you are authorized by an Employee or
Survivor(s) to petition on behalf of a class

A1 Are you a contact person for an organization? O Yes (Go to A.2) Q No (Go to A.3)
A2 Organization information:

Name of Organization

Position of Contact Person
A3 Name of Petition Representative:

Mr/Mrs./Ms. First Name Middle initial Last Name

A4 Address:
Street Apt # P.O. Box
City State Zip Code

A5 Telephone Number: ; a -

A6 Email Address:

A.7 0 Check the box at left to indicate you have attached to the back of this form written authorization to

petition by the survivor(s) or employee(s) indicated in Parts B or C of this form. An authorization
form for this purpose is provided.

Name or Social Security Number of First Petitioner:



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 __| Effective Date: 10/01/2007 | Page 51of74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 17 of 36

Number of Sample Results Recovered per Year
Between 1945 and 1990 for Gross Beta in Urine

ES

3

&

Number of Samples (N)

Figure A-5. MDA data for gross beta in urine (part 3 of 3).

1Et6
. ‘Sample MDA
. Est. MDA (95%CL on Mean),
* 50th Percentile

Bets. —___.
Hy
a
£
=
2
2 LE +——-—_————.

LEH !

1945 195019551960 196519701975: 1980 1985 «1990

Year

Visual Comparison of Data to Estimated H-3 MDAs in Urine
Figure A-6. MDA data for H-3 in urine (part 1 of 3).



AECU-817 WW

‘The romance of the investigation, the discovery of additional radioactive substances, and the
development of our understanding about the various radiations make stimulating reading. It is rec-
ommended that the student become familiar with these recent activities of modern science. Here
we shall attempt only to review briefly the knowledge resulting from this i:teresting work.

Radioacitvity is spontaneous transmutation. Some nuclei, whether fovnd in nature or artificially
prepared, are unstable and transmute themselves into nuclei of different composition (always with
less energy). In some instances the new arrangement is even less stable than the initial, and several
transmutations may follow successively before a stable nucleus that will remain permanently results.

We know of nothing that will affect the stability of a particular nuclear arrangement except bom-
bardment to transmute it artificially. This transmutation (or disintegration) involving the loss of
energy does not occur gradually, but all at once, for any particular nucleus. Among a number of
identical radioactive nuclei, disintegration events will occur strictly at random; therefore compu-
tations involve a probability that a nucleus or a group of nuclei may disintegraie ina specified in-
terval of time and must be treated by statistical methods, Such mathematics is scheduled to be
treated in a subsequent lecture. The probability that a specific percentage of any number of a par-
ticular kind of nuclei will disintegrate in a unit of time is solely determined by the kind of nuclei
they are. For some kinds, 50 per cent may be transmuted in a fraction of a second, other kinds may
have half-lives as long as 101 years,

‘There are several things which may be given off by a nucleus in radioactive transmutation. En-
ergy may be lost by quanta of electromagnetic radiation. More than 1 quantum may be released per
disintegration, but these have distinct energy values characteristic of the disintegration of the kind
of nucleus transmuting.

Nuclei of elements having atomic numbers greater than 82 may eject the equivalent of a nucleus
of an helium atom, the alpha particle. The kinetic energy of such a particle is characteristic of the
disintegration involved, If quanta of gamma radiation accompany the alpha, the energy 15 relatively
low, The kinetic energy of the alpha particle is several Mev, Such a transmutation lowers the atomic
number of the nucleus by two and reduces the mass by four units.

Gamma quanta are most commonly found accompanying radioactive tranamutations which eject
an electron (or beta particle) from the nucleus. (Some nuclei of atomic number less than 82 emit a
positron rather than an electron.) However, there are a few radioactive atoms that release a beta
and no gamma, Beta transmutations increase and positron transmutations decrease the atomic num-
ber of the nucleus by 1 with slight change in mass, The mass of the unit electrical charge e is equiv-
alent to about 1/2 Mev. The kinetic energy of these ejected electrons has a maximum value charac-
teristic of the nucleus from whence the electron came. However, most identical disintegrations
supply less than this maximum energy, since the distribution of kinetic energy is over a broad spec-
trum extending to zero.

Because of the apparent identity of all daughters of such a transmutation, it 1s believed that
disintegrations imparting less than the characteristic maximum kinetic energy to the ejected unit
charge dissipate the balance of the energy some other way during the disintegration, The neutrino
is postulated to carry away this energy. The ascribed properties of no mass and no charge render
it undetectable. Therefore some investigators claim that it does not exist. At present we are not
concerned regarding its effect on health because of its inability to produce ions.

Protons and/or neutrons are not emitted in radioactive decay. They may be released only during
or immediately subsequent to nuclear bombardment.

Except for potassium and rubidium, radioactive elements occur (to an appreciable extent) in
nature only above atomic number 80 (all elements greater than $2 are radioactive), There are three
long chains of natural radioactive isotopes terminating in stable isotopes of lead, Isotopes in one
chain are never descendants from another chain, Since all transmutations (radioactive) involving a
change of mass are in increments of four atomic mass units, the nuclei of all members of a chain
differ in mass only by integral multiples of the mass of an alpha particle. This suggests that the
helium nucieus is a sort of prefabricated unit in nuclear structure. On this basis, 2 fourth chain of
radioactive descendants might be expected, Nuclear bombardment can effect transmutation from one
chain to another and hence into the chain absent in nature. By this means many of the decay schemes



AECU-817 97

Therefore rubber gloves are adequate protection. Even the dead outer layers of skin on the hands
afford protection from external sources of alpha radiation.

Beta radiation may penetrate several meters of air, and for high energies 1 in. of wood may
not be sufficient to stop it Beta radiation is attenuated by shielding, but since it takes but little to
stop it completely, practice is to use enough for complete stoppage and thus eliminate concern for
quantity measurements. The liquid in which a beta source is in solution provides an appreciable
degree of self-absorption. If the source liquid is not exposed to the operator (mist, films, and drops
included), the walls of the containing vessels may be made thick enough to provide adequate shielding
without difficulty. For low-energy beta emitters, ordinary glassware may sometimes suffice, The
thickness to be used is determined by the maximum energy of the radioactive beta emitter to be
dealt with and not the quantity.

An exception may be considered when great intensities of beta radiation are involved. Beta ra-
diation is composed of high-velocity electrons. When these strike an absorber, some deceleration
is experienced, something like that in an X-ray-tube target, The resulting electremagnetic radiation,
called “bremsstrahlung,”’ is more penetrating than the beta radiation that produced it, and it can be
observed emerging from the shielding where it is generated. Under the most favorable circumstances
this can be only a few per cent of the incident beta, but in these days of high orders of radioactivity,
bremsstrahlung can by itself be a serious health hazard.

This emphasizes the desirability of checking the effectiveness of shielding with radiation instru-
ments for other reasons than the detection of faults.

Gamma rays, unlike beta and alpha, have no maximum range in an absorbing medium. They are
attenuated exponentially, That means that no matter how thick a shield may be, some fraction of the
penetrating electromagnetic radiation will always get through. Therefore the thickness of shielding
depends not only on the energy of the emitted radiation, but also upon its incident intensity and the
intensity that may be permitted to emerge.

The penetrating ability of gamma rays increases with their energy up to about 2 Mev. Above
this energy per photon absorbers become more effective again, particularly those of high atomic
number, For low-energy gamma radiation, absorbers are more effective as the atomic number in-
creases, This is in addition to crowding atoms together to make a dense absorber. Lead is often
used for economic and spatial reasons. Steel is structurally superior. Concrete also has structural
and economic advantages, but it is not so conservative of space, To get the most absorption in a
minimum thickness, uranium is suggested. Gold, too, is better than lead in this regard.

A convenient ‘‘rule of thumb” to remember is that for the penetrating radiation from radium a
half-absorption thickness of lead is approximately 1/2 in. For example a radium source is stored
in a lead pot 4 in, thick, The radiation coming through the wall of the office adjacent the store room
is twice the desired magnitude. An additional 1/2 in. will produce the desired attenuation.

Tables are available to permit a computation of the shielding thickness required if it is known
what the shielding material and density are to be, the intensity and energy spectrum of the incident
radiation, and the intensity desired for the emergent radiation. The first two factors are determined
by the materials to be used. The last factor is based on the protection of persons or the shielding
of radiation-sensitive equipment.

To protect personnel we are guided by the requirement that exposure of any part of the body to
beta or gamma irradiation must be limited to 1/10 roentgen equivalent physical (rep) per 24 hour
day. This may be achieved with the assistance of shielding, but limiting time of exposure and con-
trolling the distance of approach to a source are equally effective means of preventing overexposure.
It is customary to measure radiation in dosage rate. Milliroentgens per hour has been used almost
exclusively on the Plutonium Projects. If exposure is to a radiation dosage rate of 100 mr/hr,
limiting the exposure time to 1 hr per 24-hr day will keep one under the maximum permissible ex-
posure limit.

If 1 hr/day is not enough working time, doubling the distance from the source in the above ex-
ample (providing it is essentially a point source) will increase the permissible working time to 4
hr/day. One must guard against applying the inverse square law to radiation close to extended
sources, As a linear source, such as a pipe, is approached within a distance equivalent to rts length,



(Document No. ORAUT-TKBS-0012-6 | Revision No. 01__| Effective Date. 10/01/2007 | Page 250f74 |

Division (Auxier et al. 1975) stated that a phoswich detector was in place and apparently was installed
between 1971 and 1974, but no other information was located stating when the detector was placed
in the WBC facility. A 9- by 9-in. Nal (TI) crystal was installed in the WBC facility in 1974 (Auxier,
Davis, and Turner 1975) and placed under the chest area of the counting bed. This detector was
used for stationary chest counts. (A statement in Brown, Patterson, and Abee (1971) indicated that a
distortion of the spectrum occurred if the targe detector was used with the bed moving. Thus, it was
Not used for whole-body scans.) The whole-body scans performed during this period used the 8- by
4-in. detector at a distance of 12 in. from the bed surface and an elapsed counting time of 20 minutes.
(if transuranic isotopes were suspected, the thin crystal assembly was used.) Stationary chest counts
used either the 8- by 4-in. or thin crystal assembly in contact with the chest depending on what
contaminants were expected. When the 9- by 9-in. Nal (TI) crystal was used the individual was
positioned either face-up or —down on the bed and the bed was lowered just shy of contact with the
face of the large detector. The time to conduct stationary chest counts typically ranged from 10 to 40
minutes and sometimes longer to ensure “better counting statistics” (Brown, Patterson, and Abee
1971). The arc counting geometry was described in (Brown, Patterson, and Abee 1971) with the 8-
by 4-in. detector positioned 6 ft above the center point of the counting bed. This geometry differs from
that described in Mani’s assessment of ORNL internal dosimetry data (Mani 1983). The counting bed
was placed on the floor with the ends supported to give the proper curvature. This geometry was
intended for use only when the count rate was too high for the scanning or stationary chest counting
methods. (No evidence of the arc counting geometry actually being used was noted in the references
reviewed in the generation of this document.)

'n addition to the Nal (Tl) detectors described above, Brown, Patterson, and Abee (1971) indicated
that three other detectors were in use within the counting room in 1971. A 3- by 3-in. Nal crystal was
used to measure radiation emanating from specific organs (e.g., mainly for the thyroid, kidneys, liver,
and spleen). Because of its smaller size, it could be placed directly over the organ of interest. Two
other Nal crystals were employed in the WBC facility as wound probes: 2- by 2-in. and 1- by 0.03125
in., with the Jatter used to measure low-energy photons less than about 200 keV.

The next bulk sources of information describing instruments and procedures in the WBC facility were
provided by Berger and Goans (1981), Berger and Lane (1981, 1982, 1984), and Berger (2004).
Berger (2004) indicated that use of the phoswich detector began in late 1976 to early 1977. Until that
time, the three thin Nal(Tl) crystal detector assembly was in use for detection of low-energy photons.
The phoswich detector had the following dimensions: 5.25- by 0.0625-in. Nal - 2-in. CsI. Mani (1983)
indicates that in 1978, an 80-cm? HPGe counting array was in the WBC facility, but it was not fully
operational until May 1980 (Berger and Lane 1981). The array consisted of six separate detectors in
an aluminum block in a close-packed rectangular array (e.g., a six-die pattern) designed to cover one
lung (Berger and Lane 1981). The phoswich and HPGe detectors were used together from the early
1980s to obtain lung count data; they were positioned over the right and left lungs, respectively.
Berger and Lane (1984) indicated that there were problems with getting all the HPGe detectors
positioned close to the chest surface due to the construction of the aluminum block. Detectors on the
chest surface were almost a factor of 3 more sensitive than those farther away from the chest. This
lung count geometry was used until approximately 1987, when problems developed with the
HPGefliquid nitrogen feed system (Hembree 2004). At that time, the HPGe detector array was taken
out of service, and the phoswich detector was the only device used for routine chest counts untit
1994.

in 1992, a Nuclear Data SBC was put in service, primarily to measure mixed fission/activation
products ranging in energy between 100 and 2000 keV, and is in use today. This counter is in Room
16 of Building 2008, and is the only in vivo monitoring device outside the shielded counting room
(Watts et al. 1995). It has three germanium detectors facing downward and two Nal detectors on the



DISCLAIMER

This report was prepared as an account of work sponsored by an
agency of the United States Government. Neither the United States
Government nor any agency Thereof, nor any of their employees,
makes any warranty, express or implied, or assumes any legal
liability or responsibility for the accuracy, completeness, or
usefulness of any information, apparatus, product, or process
disclosed, or represents that its use would not infringe privately
owned rights. Reference herein to any specific commercial product,
process, or service by trade name, trademark, manufacturer, or
otherwise does not necessarily constitute or imply its endorsement,
recommendation, or favoring by the United States Government or any
agency thereof. The views and opinions of authors expressed herein
do not necessarily state or reflect those of the United States
Government or any agency thereof.



Special Exposure Cohort Petition U.S. Department of Health and Human Services

under the Energy Employees Occupational Centers for Disease Control and Prevention
iiness Compensation Act National Institute for Occupational Safety and Health
‘ OMB Number: 0920-0639 Expires: 09/20/2013

Special Exposure Cohort Petition — Form B.

Continuation Page — Photocopy and complete as necessary

ppendix — Continuation Page

in dose reconstructions.

itis time for NIOSH to finally realize that dose reconstructions cannot be accurately -

performed for workers at ORNL from January 41, 1943 to December 31, 1952, and
accept this class to the Special Exposure Cohort.

Attachments:

ORAUT-TKBS-0012-5 Oak Ridge National Laboratory Occupational Internal Dose

ORAUT-OTIB-0018 Internat Dose Overestimates for Facilities with Air Sampling
Programs

ORAUT-OTIB-0034 internal Dosimetry Coworker data for X-10
AECU-817 Lecture Notes; Health Physics Training Lectures, 1948-1949, Health
Physics Division, Oak Ridge National Laboratory, Oak Ridge, TN

Name or Social Security Number of First Petitioner:

Continuation pg. 5


o—
Special Exposure Cohort Petition U.S. Department of Health and Human Services

under the Energy Employees Occupational Centers for Disease Control and Prevention
lliness Compensation Act National Institute for Occupational Safety and Health

OMB Number: 0920-0639 Expires: 09/20/2013
Special Exposure Cohort Petition — Form B Page 6 of 7

F.3 [| We have attached a report from a health physicist or other individual with expertise in
radiation dose reconstruction documenting the limitations of existing DOE or AWE records on
radiation exposures at the facility, as relevant to the petition. The report specifies the basis for
believing these documented limitations might prevent the completion of dose reconstructions for
members of the class under 42 CFR Part 82 and related NIOSH technical implementation
guidelines.

(Attach report to the back of the petition form.)

F.4 L] We have attached a scientific or technical report, issued by a government agency of the
Executive Branch of Government or the General Accounting Office, the Nuclear Regulatory
Commission, or the Defense Nuclear Facilities Safety Board, or published in a peer-reviewed
journal, that identifies dosimetry and related information that are unavailable (due to either a lack
of monitoring or the destruction or loss of records) for estimating the radiation doses of
employees covered by the petition.

(Attach report to the back of the petition form.)

G Signature of Person(s) Submitting this Petition — Complete Section G

All Petitioners snouid sian and date the petition. A maximum of three persons may sign the petition.

— 7-13 -RO//
Signature Date
Signature Date
Signature Date
Notice: Any person who knowingly makes any false statement, misrepresentation, concealment of

fact or any other act of fraud to obtain compensation as provided under EEOICPA or who
knowingly accepts compensation to which that person is not entitled is subject to civil or
administrative remedies as well as felony criminal prosecution and may, under appropriate
criminal provisions, be punished by a fine or imprisonment or both. | affirm that the information
provided on this form is accurate and true.

Send this form to: SEC Petition
Division of Compensation Analysis and Support
NIOSH
4676 Columbia Parkway, MS-C-47

Cincinnati, OH 45226

Name or Social Security Number of First Petitioner:



[Document No. ORAUT-OTIB-0034 | Revision No. 00 __| Effective Date. 12/12/2005 | Page 7 of 29

3.0

40

DATA
3.1

3.2

Selected Bioassay Data

Data for each of the EDP codes considered were extracted from a series of Microsoft®
Access files that contain a version of the ORISE/CER Dosimetry Database. These
files were titled “tbK\ORNL_Urinalysis_rawData,” for the periods 1951-1978, 1979-1985,
and 1986-1988.

Analysis

A lognormal distribution for the annual data for each of the four radionuclides was
assumed, and the 50th and 84th percentiles were calculated, using the method
described in ORAU (2004a). Tables A-1 through A-4 in Attachment A show the
Statistical analysis results for strontium, uranium, plutonium, and transplutonium.

INTAKE MODELING

41

4.2

43

Assumptions

All results were assumed to be representative of a full day (24 hr) of urinary excretion.
Each result used in the intake calculation was assumed to be normally distributed, and
a uniform absolute error of 1 was applied to all results, thus weighting all results
equally. A chronic exposure pattern was assumed; while this is unlikely for workers at
ORNL, it will approximate a series of acute intakes with unknown intake dates. Intakes
were assumed to be via inhalation using a default breathing rate of 1.2 m°/hr and a 5-
um activity median aerodynamic diameter (AMAD) particle size distribution.

Bioassay Fitting

The IMBA Expert OCAS-Edition computer program was used to fit the bioassay results
to a series of inhalation intakes. Data for each radionuclide were fit as a series of
chronic intakes.

Because the Type S strontium, uranium, and plutonium isotopes and Type M plutonium
and americium isotopes present at ORNL have very long half-lives and because the
material is retained in the body for long periods, excretion results are not independent.
For example, an intake in the early 1950s could contribute to urinary excretion in the
1980s and later. To avoid potential underestimation of intakes for people who worked
at ORNL for relatively short periods, each intake was fit independently, using only the
bioassay results from the single intake period. This will result in an overestimate of
intakes, particularly for assumed Type S exposures extending through multiple
assumed intake periods.

Radionuclides And Material Types

For each radionuclide considered, the bioassay results were entered into IMBA with
assumed material types as discussed in more detail in the following sections. The
assumed 50th-percentile intakes that result are displayed in the figures in Attachment
A. In these figures, annual bioassay data used in the fits are shown as the dark dots
(e), and data that are not used in the fits are shown as lighter dots (e). For certain
radionuclides and material types, it is necessary to show figures for both the composite
intake over the entire data period and the individual plots of subsets of selected groups
of years during the data period that were used to develop the composite figure. The
composite plot shows the merged fit for all of the intake periods to the bioassay data.



AECU-817 87

The problem of shielding against neutrons is more complicated, Fast neutrons are very poorly
absorbed by most materials. It is therefore necessary to slow them down to speeds at which they
are absorbed efficiently. For this reason a neutron shield usually consists of a mixture of hydrog-
enous material with a neutron-absorbing substance which gives off nonpenetrating secondary
radiation, A typical example is paraffin containing borax. The hydrogen atoms in the paraffin slow
down the neutrons, and the boron absorbs them while giving off an alpha particle. The alpha particle,
of course, has a negligible range in the paraffin.

For small-scale laboratory work it is much more convenient to pile shielding around a source
until the measured radiation is below tolerance limits than to try to calculate accurately the amount
of shielding required, The measurement will account for all the leakage radiation and scattered
radiation much more simply than any calculation can,

In erecting shielding it is essential to remember that shields cast shadows and that if you stand
outside of the shadow you are receiving radiation exposure. It is necessary to remember that radia-
tion can effectively bounce around corners, although such scattering reduces both the intensity and
energy of the primary rays.

In agreement with the standard laboratory policy that health protection responsibility rests with
supervision, the Health Physics Division has no formal responsibility for the adequacies of shielding
unless it is specifically invited to make recommendations. However, any dangerous situation dis-
covered should be reported {mmediately to supervision, and, if it is not corrected, it should be re-
ported to Survey Headquarters.

Appendix A. Supplementary Study Material for Lectures 2.6 and 8.5—-Preliminary Notes on Radia-
tion and Shielding Problems

Dosage Rate from an Unshielded Point Source of Gamma Radiation. Within the range of energies
0.2 to 1.5 Mev, the dosage rate at 1 ft from an unshielded source is given approximately by

Ri =6CE

Where Rj is the dosage rate in roentgens per hour at 1 ft, C is the number of curies of the radio-
isotope emitting the gamma radiation, and E is the gamma energy per disintegration.

For a wider range of energies the following table, giving the dosage rate per curie in roentgens
per hour at 1 ft, may be used:

Table 8.1
5,* R}/curie, E,* R;/curie,
Mev r/hr at 1 ft Mev r/hr at 1 ft
0.1 0.5 1.5 8.2
0.2 Ll 2.0 10
0.3 1.8 2.5 11.6
0.5 3.1 3 13
0.8 5.0 4 15

1.0 6.0

*One quantum of this energy per disintegration
is assumed, If the number of quanta per disingegra~
tion is different from 1, or if more than 1 energy is
present, the value of Ry is adjusted accordingly.
Dosages, not necessarily quantum energies, are ad~
ditive. In the absence of more definite information,
it is sometimes assumed that the radiation is equiv-
alent to 1 quantum per disintegration of the most
energetic gamma known to be present. This assump-
tion may result in an estimate that is either too
great or too small.



UNIT 5. RADIATION-DETECTION TECHNIQUES

5.1 COUNTING TECHNIQUES
By P. E, Brown

The scope of this talk will necessarily be limited to counting done in the Health Physics counting
room and the application of this technique to field conditions.

Several types of samples are submitted to the counting room for counting:

A. Air samples
1, Alpha counting
a. Precipitron samples (Al cylinder)
b, Filtron samples (filter paper)
2, Beta counting
a. Precipitron sample (A} cylinder)
b. Filtron sample (filter paper)
B. Flat disk samples less than 2 in. in diameter, activity one side

1. Isotope-carrier smears

2, Floor smears

3. Miscellaneous

Cc. Mud, water, biological tissues

‘The equipment used in this counting is probably famillar to most of you, and some have used
all these types of counters. For the benefit of those who have not used them, the following descrip-
tion will be given:

For precipitron samples and filtron samples, both alpha and beta counting are done. The alpha-
counting equipment consists of a California sealer (scale of 8) unit connected to an ionization cham-
ber made so as to hold the cylindrical sample foil or filter paper upright, with the electrode running
the length of the cylinder and carrying a high voltage (approximately 1250 volts). This diagram shows
the sample in place between the two electrodes and single stage of amplification that carries the
weak pulses from the chamber through the relatively long shielded cable. The chamber is suspended
by rubber supports to prevent jarring. The geometry of this chamber is 50 per cent.

Auxiliary equipment consists of oscilloscope and automatic timer wired to count rate switch and
recorder.

‘The counting of air samples for beta-gamma activity is done with a Higginbotham scaler and a
long thin-walled tube (7 in. sensitive area), The tube is mounted ina horizontal lead pig to reduce
the background radiation. Auxiliary equipment includes recorder and timer. The geometry of this
type is 21 per cent.

Flat disk samples are counted for both alpha and beta activity if desired.

For alpha counting, a (scale of 8) scaler and paralle}-plate ionization chamber are used. Aux-
illary equipment includes recorder, oscilloscope, and timer.

46 AECU-817



(Document No. ORAUT-TKBS-0012-5 | Revision No. 01 __| Effective Date: 10/01/2007 | Page 18 of74 |

5.2.2.3 Zinc-Sulfide Scintillation

ORNL used zinc-sulfide scintillation counting to count alpha emitters such as the trivalent alpha
actinides (see Sections 5.2.3.1 and 5.2.3.2).

5.2.2.4 Gamma Spectrometry

ORNL uses gamma spectrometry to identify and quantify radionuclides that emit photons with
energies greater than 60 keV. A high-resolution, hyper-pure, intrinsic germanium (HPGe) detector
with a beryllium end-window is used. Urine samples are placed in a 1-L Marinelli beaker, which is
placed over the detector for counting. If the total sample volume is less than 1 L, distilled or deionized
water is added to bring the volume up to 1 L. Fecal samples were counted directly in the sample
counter for screening purposes, and they might have been ashed and placed into a 2-in. Petri dish for
quantitative results.

§.2.2.5 Gas Flow Proportional Counter (Beta Counting)

ORNL uses the gas flow proportional counter system for the analysis of strontium. Strontium is
chemically separated from the sample and filtered onto a glass fiber filter. The filter is placed on a
planchet for insertion into the counter. The counting system does not distinguish between beta
energies, so the reported result is total strontium (*Sr plus °°Sr).

5.2.3 Notes on Measurements of Alpha Emitters
5.2.3.1 Trivalent Alpha Actinides

Before 1989, ORNL did not perform radionuclide-specific analyses for americium, curium, and other
high atomic number elements beyond plutonium. [These radionuclides were typically recorded on the
HP Body Fluids Analysis Request cards as transplutonium (TPO or TPL).] Rather, the Laboratory
separated trivalent alpha actinides as a group and analyzed by zinc-sulfide scintillation counting.
Therefore, monitoring of transplutonium elements was unable to differentiate between such nuclides
as ‘Am and “Cm. The default radionuclide to use with measurements involving trivalent alpha
actinides would be “Am. The detection sensitivity of the transplutonium analysis technique is not
well-documented for samples processed before 1985. However, laboratory records suggest that the
transplutonium detection level was about 0.2 dpm through 1985.

For analyses performed after 1985, an estimate of the sample-specific MDA is generally reported.
Alpha spectroscopy analysis for transuranic elements in bioassay samples jan in the early- to mid-
4980s. Differentiation, by alpha energy separation, for isotopes such as TMU, *°U, ?°U, ?**Pu, 2*Pu,
241m, and “Cm became possible.

5.2.3.2 Plutonium

Before 1989, ORNL did not routinely perform isotope-specific analyses for plutonium. Rather, the
Laboratory separated plutonium as an element and analyzed it by zinc-sulfide scintillation counting.
Therefore, the historic plutonium analysis technique was unable to differentiate among the alpha
emitting isotopes Pu, “Pu, and “Pu. Site analytical personnel assert that in the early 1980s,
positive total Pu measurements were recounted on the limited number of alpha spectrometers that
were available at the site. Though many HP Body Fluids Analysis Request cards in the claims files
gave results for the isotopes Pu and ***Pu, this was not consistently observed. The default isotope
for positive, total Pu measurements should be 239y [5].



LECTURE NOTES; HEALTH PHYSICS TRAINING LECTURES, 1948-1949

PREFACE

The following notes were issued in conjunction with a series of lectures given during the fall of
1948 to persons working in the field of health physics at the Oak Ridge National Laboratory, either as
employees of the Health Physics Division or as trainees assigned from agencies outside the labora-
tory. Some of these notes were written by the lecturers, while others were taken by a stenographer and
edited by the lecturer. As nearly as possible, the notes were issued concurrently with the lectures,
and greater stress was placed on getting the material into the hands of the students than on the quality
of preparation,

For the purpose of reproduction by the Technical Information Division of the Atomic Energy Com-
mission, some deletions have been made by the AEC Declassification Branch, and minor corrections
or changes have been made by the Health Physics Division. However, except for deletions, the notes
are essentially in the form written or approved by the lecturer. Since the demand for the notes makes
it expedient to release them without thorough editing, it should be understood that some of the material
contained may not accurately reflect the policies or the-considered thinking of the Health Physics Di-
vision, Oak Ridge National Laboratory.

The lecture series to which these notes apply was planned and coordinated by T. H. J. Burnett of
the Health Physics Division, While most of the lectures were given by members of the Health Physics
Division, Lecture 4.4 was given by a member of the Instrument Department and Lecture 5.10 by a
member of the Physics Division. Unit 7 consisted of a series of lectures by representatives of other
divisions of the laboratory. Since these dealt primarily with the work of the laboratory, no notes were
issued.

AECU-817 1



[Document No. ORAUT-OTIB-0034 | Revision No. 00 _| Effective Date: 12/13/2005 | Page 18 of 29 |

annual bioassay
value (50th percentile)
not used in fit
fit

Figure A-3. Assumed “Sr intake, 1954, 50th-percentile results, Type S.

* annual bioassay
value (50th percentile)
© not used in fit
fit

Figure A-4. Assumed “Sr intake, 1955 to 1964, 50th-percentile results, Type S.

» annual bioassay
value (50th percentile)
© not used in fit
fit

r intake, 1965 to 1983, 50th-percentile results, Type S.


(Document No. ORAUT-TKBS-0012-5 | Revision No. 01 __| Effective Date: 10/01/2007 [Page 50 of 74 ]

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 16 of 36

z

‘Activity (dpim/24‘h sanaple)

1 E#2 +

1 EH
1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Visual Comparison of Data to Estimated Gross Beta MDAs in Urine

Figure A-5. MDA data for gross beta in urine (part 1 of 3).

Annual Average MDAs for Gross Beta in Urine
Calculated from Recovered Data Between 1945 and 1990
1600

1400

1200

3

MDA (dpm/24 h)

g

8

8

Figure A-5. MDA data for gross beta in urine (part 2 of 3).



Special Exposure Cohort Petition U.S. Department of Health and Human Services

under the Energy Employees Occupational Centers for Disease Control and Prevention
lness Compensation Act National Institute for Occupational Safety and Health
; . OMB Number: 0920-0639 Expires: 09/20/2013
Special Exposure Cohort Petition — Form B Appendix — Continuation Page
Continuation Page — Photocopy and complete as necessary.
F-1 (cont.)

ORAUT-TKBS-0012-5 reflects that there was only one urine sample for Am- 241 in
1952, no Cm-244 urine samples were found until 1964, no gross alpha samples
found until 1953, two gross beta samples were found in 1952, no H-3 samples were
found until 1961. no Pm-147 samples were found until 1959. no rare earth samples
were found untit 1951 and only about 30 samples were found for each year from
1951-1952, no Sr-90 samples were found until 1950 with only one sample in 1950
and less than ten samples for each year 1951-1952. Appendix B, of ORAUT-TKBS-
0012-5, confirms that there were "...no bioassay results appearing in the DOE
worker files" from 1943-1947. Appendix B was attached to ORAUT-TKBS-0012-5
to justify utilizing the ORNL air sampling data to extrapolate internal dose. Although
the air sampling data has been utilized for dose reconstructions, as justified in
Attachment B of ORAUT-TKBS-0012-5, the air sampling was only for 1944-1947
and was only performed for alpha contamination. ORAUT-TKBS-0012-5,
Attachment B justifies utilizing the alpha air monitoring data for beta contamination.
Dose reconstructions cannot accurately be performed for X-10 workers utilizing
this very limited alpha air sampling dataset and extrapolating it to additional years
and to the beta emitting radionuclides. Attachment B, of ORAUT-TKBS-0012-5,
references ORAUT-OTIB-0018, Rev. 01, Internal Dose Overestimates for Facilities
With Air Sampling Programs for authority to apply air monitoring data to the X-10
internal dose. ORAUT-OTIB-0018 can only be utilized for "Claims for which it is
likely that the covered emplovee had no significant intakes of particulate radioactive
material." ORAUT-OTIB-0018 can only be utilized for “Intakes of particulate
radioactive material only." (page 8 of 35) Confounding the previous dose
reconstructions are the claimants who should have been assigned co-worker dose,
because they had recorded external dose which indicated they should be assigned
co-worker internal dose. Since there was no bioassay monitoring data available
until 1951, and then only a limited amount, NIOSH has been assigning ambient
dose for claimants and stating that the ambient dose is higher than the co-worker
dose. ORAUT-OTIB-0034, Internal Dosimetry Coworker Data for X-10 documents
that the co-worker dataset begins in 1951. There is simply no means to assign
co-worker data prior to 1951. | have previously noted the deficiency of the internal
dose from 1951-1953, which is being assigned with little or no bioassay data for
the vast majority of radionuclides found at X-10 during that time frame.

An accurate dose reconstruction cannot be performed for X-10 workers from
1943-1953. NIOSH has previously performed dose reconstructions for claimants
during that time period, using assumptions based on a very limited dataset.

Name or Social Security Number of First Petitioner:

Continuation pg. 1



[Document No. ORAUT-OTIB-0034_ | Revision No.00__| Effective Date. 12/13/2005 |

Page 16 of 29 _|

Table A-3. Statistical summary of annual ORNL plutonium 24-hour urinary excretion rates, 1951-

1988.
Effective GM (50th) | GM*GSD (84th) Effective GM (50th) | GM*GSD (84th)
bioassay date | dpm/day dpmiday bioassay date dpmiday dpmiday
7/1951 0.13 0.30 7/111970 0.08 0.19
TANN952 0.11 0.31 7/1971 0.06. 0.23
71953 0.04 0.14 THAIA972 0.06 0.18
7AI1954 0.07 0.36 THA973 0.03 0.11
TAI955 0.07 0.26 FAN 974 0.06 0.12
TAIN956 0.04 0.36 THAM975 0.04 0.12
TAIN957 0.14 0.73 THIN976 0.05 0.13
7THAN958 0.15 0.52 TAN977 0.07 0.15
TAN959 0.12 0.62 TAIN978 0.06 0.15
7/1/1960 0.13 0.28 TAHN979 0.08 0.14
7TAMN961 0.07 0.21 THIM980 0.05 0.11
TAIN 962 0.01 0.08 7HH1981 0.09 0.16
THAN 963 0.01 0.09 THIN982 0.10 0.17
7AN964 0.08 0.23 7T/1N983 0.07 0.13
TAN 965 0.13 0.26 7/1984 0.07 0.11
TAN966 0.12 0.21 7AH1985 0.01 0.05
TAN 967. 0.10 0.19 7H1H1986 0.01 0.04
TAA 968 0.12 0.24 THIN987 0.01 0.04
THA969 0.04 0.17 7AIN1988 0.00 0.03
Table A-4. Statistical summary of annual
ORNL transplutonium 24-hour urinary
excretion rates, 1968-1987.
Effective GM (50th) | GM*GSD (84th)
bioassay date dpmiday dpmiday
THI1968 0.12 0.31
TAIM969 0.10 0.42
TAN970 0.10 0.33
THA971 0.09 0.24
71972 0.12 0.27
THIM973 0.09 0.16
THI1974 0.08 0.15
TAI975 0.12 0.30
TAN976 0.15 0.52
TANG77 0.11 0.18
THAM978 0.10 0.18
TAN979 0.10 0.17
7Hi1980 0.10 0.14
7AN981 0.07 0.17
1/1/1983 0.10 Q.17
7HAIN984 0.10 0.15
TAN985 0.05 0.09
TAN 986 0.01 0.05
7TAN987 0.01 0.04



[Document No. ORAUT-OTIB-0018 _[ Revision No.01 __| Effective Date: 080972005 | Page 300f35 _|

LANL (Los Alamos National Laboratory), 1948, “Group CMR-12 Monthly Report January 20, 1948.”

LANL (Los Alamos National Laboratory), 1950, Group CMR-12 Monthly Reports for 1950.

LANL (Los Alamos National Laboratory), 1952, Group CMR-12 Monthly Reports for 1951-52.

LANL (Los Alamos Nationa! Laboratory), 1954, Group CMR-12 Monthly Reports for 1953-54.

LANL (Los Alamos National Laboratory), 1969, Weekly, Monthly, and Annual Reports for 1960-69,

ORAU (Oak Ridge Associated Universities), 2004a, Technical Basis Document for the Los Alamos
National Laboratory ~ Occupational Internal Dose, ORAUT-TKBS-0010-5, Oak Ridge,

Tennessee.

ORAU (Oak Ridge Associated Universities), 2004b, Technical Basis Document for the Rocky Flats
Plant — Occupational internal Dose, ORAUT-TKBS-0011-5, Oak Ridge, Tennessee.

Parker, H. M., 1944a, “Product Concentration in the Air,” Memo to |. Perlman, ORNL CF 44-7-78,
Clinton Laboratories, Oak Ridge, Tennessee, July 1.

Parker, H. M., 1944b, Review of Air Monitoring Procedures at Clinton Laboratories, CH-2562, Clinton
Laboratories, Oak Ridge, Tennessee.

Patterson, C. M., 1849, “General Operating Tolerances,” Letter to Lauriston S. Taylor (National
Bureau of Standards), HW-12710, Hanford Works, Richland, Washington, March 11.



6 AECU-817

A department breakdown was given as follows:

It is the primary function of the Survey-Monitoring Section to prevent radiation damage in any
of its various forms to living matter, The present Survey-Monitoring organization was developed
along these lines with a design which would make individual groups independent, with a minimum of
overlapping functions, ft has been agreed that four distinct groups are essential to achieve this ob-
jective, They are as follows:

Survey. It is the responsibility of the Survey group to assay or survey conditions throughout
the laboratory for the purpose of establishing tolerance levels, contamination control, etc. It can be
said that the successful contro! of the exposure of personnel to radioactive material and resultant
control over contamination lies with the Survey group. With instruments designed to aid in the quan-
titative and qualitative analysis of various types of radiation, the surveyor aids the laboratory op-
eration by keeping informed of potential damage which any given operation might produce.

Area Monitoring, In defining the duties of the Area Monitoring group, it can be said that this
group is interested in radioactive matter as associated with things. An example of this is a con-
taminated chair, wastebasket, or piece of laboratory apparatus. Other examples, in a broader sense,
are air, earth, and water as related to area contamination.

Currently this group consists of three major units. Field Analysis is interested in sampling of
soil surrounding the X-10 area and analyzing contents of the White Oak Creek drainage system and
on down the Clinch River valley. Incidents of high concentrations of contamination are investigated
by this group when found, and proper precautions are instituted. Meteorological Service is interested
in air contamination locally and regionally. For operational convenience the laboratory personnel
which provides analyses on samples from the field is under the same supervision as is the
meteorological group. Two other units of operation, currently under common supervision, are known
as Laundry Monitoring and the Counting Facility, employing a staff of technicians skilled in the use
of the Geiger counter, The Laundry Monitoring group checks all clothing issued by the various oper -
ating departments for protection against contamination. Before the garments are washed the gar-
ments are then divided into two groups: those which are hot and those which are free of contamina-
tion, Hot clothes are given a special wash designed to remove most types of contaminants, after
which the garments are rechecked, and, when thoroughly decontaminated, these are put back into
circulation. The Counting Facility could perhaps function in any of the Survey~Monitoring groups
or even exist as an independent group. However, since the majority of samples originate with the
other units within Area Monitoring, the Counting Facility has been placed there.

Personnel Monitoring. As the name implies, the Personnel Monitoring group is interested in
the quantitative doses of radiation which involve people. This group attempts to determine the extent
of exposure to radiations from external sources and, to a limited extent, to make some attempt at
deterosining internal exposure. External exposures are measured with the use of the pocket meter
(ion chamber) and film meter. Internal exposures are determined in a limited way by a study of body
fuids, .

Instruments Group. In a larger sense, the Instruments group is 2 service group catering to the
three above-mentioned groups. It is the responsibility of this group to provide for the procurement,
maintenance, and development of radiation-detection instruments.

Procurement requires an up-to-date study of available instruments and an accurate prediction
of the needs within the groups in order that adequate instruments are available to them, Routine
maintenance is the major function in that radiation-detection instruments require lengthy and fre-
quent calibrations, Development is activated in proportion to the birth of new ideas, The Instruments
group currently coasists of three units: a calibration unit utilizing four technicians performs precise
calibration techniques using beta, gamma, and alpha rays of varying intensities and energies; a re-
palr unit consisting of three instrument mechanics performs periodic examinations on the instru-
ments for the purpose of detecting defects; and a development group consisting of one engineer, two
mechanica, and one technician is in constant search for minor improvements on the instruments
currently in use,



Document No. ORAUT-OTIB-0018 __| Revision No. 01 [Effective Date: 08/09/2005 [ Page 12 of 35

Table 4-3. Inventory of radionuclides and absorption types.
Radionuclides/absorption types

associated with the alpha choice

Am-241 [M.S Pu-240 [M.S

CF252 M Pu-242 [MS

Cm-244 | M Ra-226 [M

Np-237_ | M Th-228_ | M,S

Po-210 [FM Th-230_ [| M, S

Pu-238 | M,S Th-232_[M,S

Pu-239_ | M,S U-234 EMS

Radionuclides/absorption types associated with the beta
choice

Ag-110 MS Fe-59 ELM $-35 FM
Ba-140 | F HF181 FM Sb-125_ [| F,M
Ce-141 M.S La-140 FM Sc-46 Ss
Ce-144_ | MS Mn-54 FM Sn-113_ [F/M
Co-58 M,_S Mo-99 FS Sr-89 F
Co-60 M,S Na-24 F Sr-90 F
Cr-51 F,M,S_| Nb-95 M.S Tb-160 M
Cs-134_ [| F Ni-63 EM To-99 ELM
Cs-137_ | F P-32 FM Y-91 M.S
Eu-152_ | M Pm-147_ | MS 2-65 Ss
Eu-154_ | M Ru-103 FLM,S [2-95 F.MS
Eu-155 | M Ru-106 F_M,S

Radionuclides/absorption types associated
with the uranium choice
Th-232 M.S RU EMS
a, RU=-234 plus the contaminants from Table 4-
2 in the ratios provided in Table 4-2.

default list for the alpha choice. it should be added to the list if "Pa or “”Ac was handled in a pure or
concentrated form at the site in question.

Th-232 is included in the uranium list because many uranium facilities also processed thorium at
some time. Th-232 can be removed from the list if it is known for sure that the facility did not handle
thorium. Removal of 2?Th will reduce dose from 2 to 700 times depending on organ and length of
exposure.

For strontium, type S is applicable to the titanate form only and is unlikely to be present at most sites.
Type S was not included in the default list for the beta choice. If it is established that a site had
strontium titanate, type S strontium will have to be added to the list. Check the Site Profile for
applicability. Adding type S strontium will significantly increase the dose for some organs.

5.0 INGESTION INTAKE

The possbility of ingestion intakes has to be considered separately from inhalation intakes when
intake is not based on bioassay measurements. Estimation of Ingestion intakes (OCAS 2004)
provides the method for estimating ingestion intakes based on air concentrations:

Ingestion intake = 0.021 (inhalation intake) (4)



[Document No. ORAUT-OTIB-0018 __| Revision No.01 | Effective Date: 08/09/2005 Page 320f35_|

2 “| @ ¥-12 median results:

> 40 emer Long-term OTIB-18 result
Zak

5 z —— 3-year OTIB-18 result

‘WU7I52 6/2357 F2i20/62 6/1168 W2HI3 Si25!79
Date
Figure B-2. Y-12 versus OTIB-18 uranium (type S).

The lowest recorded positive value varied through the years from 0.02 dpnvday (0.009 pCi/day) to
0.09 dpm/day (0.04 pCi/day) with most years at 0.02 dpmvday. Figure B-3 shows the assumed intake
in this TIB for both a type S and a type M inhalation. The figure also shows the 0.009-pCi/day and the
0.04-pCi/day values. The figure indicates that the TIB values would be likely to exceed the measured
data soon after the start of a chronic type M inhalation. The type S inhalation, however, might not
resultin detectable urine samples.

0.20
0.18
0.16
aaa |
ot |
0.10
0.08
0.06
0.04 beg
0.02
9.00

—e— Expected Type M
—®-— Expected Type S
a 009 pCiday
“04 pCitday

Urinary Excretion (pCi/day)

oO 2000 4000 6000 8000 10900 12000
Days after start of chronic intake

Figure B-3. Rocky Flats Pu expected urine versus actual results.

The Hanford site produced much of the plutonium used by the weapons program. It appears to be
reasonable, therefore, to review Hanford site plutonium urine results against the assumptions in this
TIB.

The TIB analysis evaluated plutonium urine samples from Hanford from 1953 through 1969. The
median sample for each year varied. The median result was never more than 0.023 pCi/day and for
most years was less than that. Figure B-4 shows the assumed intake in this TIB for both a type S and
a type M inhalation. The figure also shows the 0.023-pCi/day value. As with the Rocky Flats data,
the graph indicates that the TIB values would be likely to exceed the measured data soon after the



AECU-817 ci

age can often cause great health and economic damage, e.g., polluted water supplies (epidemics),
streams ruined, property depreciated, odor nuisances, etc,

There is a great variety of wastes, from sewage of 2 city to shavings and sawdust in a carpenter
shop. Typical and usual properties are listed below:

1, Domestic sewage: liquid, organic, disease germs, odors

2, Industrial wastes: liquid, organic or inorganic, no disease germs, odors, bad appearance, etc.

3. Refuse: solid or semisolid, organic and inorganic, few disease germs, odors, fires, insect
breeding, etc.

4, Atmospheric pollutants: gases and dusts, organic or inorganic, no disease germs, odors, and
sometimes toxic substances, etc.

This classification indicates the character of wastes to some extent, but it is based mainly upon
the source and responsibility for disposal.

Wastes may be classified in other ways if more useful in treatment or disposal, e.g., organic
versus inorganic; radioactive versus nonradioactive; infectious versus noninfectious, etc.

‘Who is concerned with problems of waste disposal ?

1, Individuals who may be damaged or annoyed

2. Owners, i.e., those who produce wastes and are responsible for damage, if any

3. Public health authorities, state and municipal health officials, health divisions of industry,
and other regulatory agencies

Waste disposal does not mean just getting rid of and passing on to others; it implies satisfactory
(especially safe) disposition of wastes, Waste disposal means the handling, processing, and ultimate
disposa) in such a manner as to prevent danger to health, objectionable nuisances, interference with
the rights of others, or economic damage.

Listed below are a few examples of improper disposal:

1, Sewage containing disease germs (and all sewage may at times) discharged above a water
supply or swimming place results in danger of disease transmission.

2. Excess organic matter (as from sewage, tannery wastes, milk wastes, etc.) into streams or
lakes results in decay which causes odors, killing of fish, depreciation of property, interference
with use of water, etc.

3. Excess of smoke, fumes, gases, chemical irritants, or dusts into atmosphere result in nui-
sance and damage and may cause danger to health and safety.

4, Burial of infectious, or offensive, material where it may pollute underground water and reach
springs and wells, or where it may be uncovered as by washing out or tampering by people.

Available methods of waste disposal are few in number although infinite variety of conditions
are involved, Some of these are listed below:

1, Mix with surface water, ¢.g., add industrial wastes to streams or lakes to take advantage of
dilution, time for decay, die-away of germs, etc.

2. Deposit on surface of the ground within certain limits, e.g.,
land, etc., to take advantage of sunlight, drying, decay, etc.

$. Burial or absorption underground limits access of people, but may affect underground water,
This takes advantage of segregation and time.

4. Discharge to the atmosphere, e.g., incineration or evaporation of solids or liquids and stacks
for gas. This takes advantage of diJution: there {s the disadvantage of wide dissemination and some-
times of concentration (meteorological conditions),

5. Recovery at the source for by-products, process use, or storage indefinitely. Certain con-
stituents or the water may be valuable for use, Otherwise storage becomes more am more trouble-~
some, expensive, and uncertain of results, Storage may be necessary for some strongly toxic chem-
icals and radioactive materials.

All these methods may be satisfactory, but only if the quantity of wastes is within permissible
limits in order to prevent danger of damage or nuisance, for example, tolerable versus excessive
amounts of sewage in water, too much smoke or gases in air, burial of more material than soil con-
ditions can assimilate,

lagoons, dumps, irrigation of



[Document No. ORAUT-OTIB-0034_ [Revision No. 00 __| Effective Date: 12/13/2005 | Page 150f29 |

ATTACHMENT A

Table A-1. Statistical summary of annual ORNL strontium 24-hour urinary excretion rates, 1951-
1988.

Effective GM (50th) | GM*GSD (84th) Effective GM (50th) | GM*GSD (84th)
bioassay date | dpm/day dpmi/day bioassay date | dpmiday dpmiday
T1952 109.15 1,008.55 71971 6.91 15.08
TINN1954 24.68 241.98 7972 7.54 22.75
T/N111955 13.10 63.17 TH11973 5.18 13.45
T/11956 13.54 82.88 TIMN974 7.66 24.76
T/AN957 15.71 68.85 TAN975 5.27 10.68
TAN958 17.01 75.45 7/1976 5.14 11.59
TAIN959 12.20 59.03 T/AN1977 §.19 11.58
7/AMN960 15.67 65.07 7H11978 5.04 9.61
7/1/1961 9.48 28.06 7/1979 2.44 6.68
TAN 962 12.04 29.49 7AN980 5.69 23.05
7/AN963 10.08 26.23 7/1/1981 3.04 7.22
N/A* N/A* N/A* 7/1/1982 5.23 12.06
THNN965 4.86 15.75 7/1/1983 6.71 14.48
7/966 3.62 17.78 7/111984 4.38 14.14
TIAN967 3.35 12.86 TAIN985 2.11 5.93
71968 1.02 777 7111986 0.67 2.88
TAN969 3.89 8.96 THAN987 3.37 15.98
7111970 6.07 13.96 7/1/1988 0.56 1.96

* Insufficient data for analysis.

Table A-2. Statistical summary of annual ORNL uranium 24-hour urinary excretion rates, 1951-1987.

Effective GM (60th) | GM*GSD (84th) Effective GM (50th) | GM*GSD (84th)
bioassay date | _dpm/day dpmiday bioassay date | _dpm/day dpmiday
THM951 31 7.84 THM970 0.16 065
TAM952 2.88 8.01 TAM971 0.09 0.26
7A1953 3.64 7.54 TANG72 0.13 0.39
7AM954 2.27 5.46 TANTS 0.07 0.30
TAA95 3.07 9.34 TAM974 0.06 0.22
TAM986 1.20 8.16 TAA975 0.08 0.32
TAM957 0.50 3.65 TAN976 0.10 0.46
THM958 0.64 4.74 TAM9T7 0.20 0.48
TAH959 1.43 6.93 TAM978 0.14 0.31
7AM 960 1.36 3.97 TAN979 0.14 0.52
TAM961 1.52 6.73 7/980 0.11 0.39
TAM962 0.52 3.57 7HAM984 0.13 0.22
TAA963 0.58 3.24 TAM982 0.17 0.31
TAM964 0.25 1.29 TAN983 0.12 0.28
TAN965 0.17 0.62 7/984 0.10 0.19
TAM966 0.18 0.52 7/AM985 0.07 0.17
TAM967 0.09 0.21 TAN986 0.04 0.08
TAA968 0.13 0.34 TAM987 0.00 0.03

TAN969 0.10 0.33



68 AECU-817

Disadvantages

(These are less serious now since better emulsions have been obtained recently which are highly
concentrated in AgBr.)

1, ‘‘Fog’’ grains, dense background in some experiments. The variety of agents to which emul-
sions are sensitive include (a) mechanical shock, (b) thermal change, and (c) cosmic rays. Often one
gets other effects along with the desired ones.

2, Tedium involved in inspection (microscopy). Particles cannot be bent enough with magnetic
fields onte they are in emulsion (however, these can be bent outside plate before impingement), Get
only ranges, directions, and number, .

Applications in Health Physics

Monitor for neutrons: both slow and fast neutrons can produce secondaries to make tracks.
Slow Neutrons.

n+qni4opsci4sg

(Q = 1/2 Mev) 6- to 7-4 length protons, 4 to 6 grains.

Number of tracks gives estimate of exposure.

Fast Neutrons. These give recoil proton, seen in the same way. Even 50 per cent accuracy is
hard to get for one-third or one-fourth of tolerance,

Alternative Method for Slow Neutrons, (This method was proposed and demonstrated by M.
Shapiro and J. Barnes.) The plates were impregnated with

Li6 +n + He? + H® 4 5 Mev (alpha and triton, principal path ~54 4, some denser)
B10 ,n ~ Lil + He? 4 2.5 Mev (alpha, principal path 7 to 8 1, much denser)

With Li, for a given neutron exposure, it is possible to get 150 times as many tracks as from N
interaction. With B it is possible to get 1500 tracks compared with N interaction.



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 | Effective Date: 10/01/2007 | Page 30 of 74 )

for gross alpha (the OTIB-0018 intake rate is the same for both Type M and Type S). This supports
extending the use of OTIB-0018 from 1947 through 1950, assuming that the exposure potential did
not increase between 1947 and 1952. A-review of ORAUT-TKBS-0012-2, “Oak Ridge National
Laboratory — Site Description” indicates that, while the site was by no means static during this period,
the assumption of relative constant exposure potentiat is reasonable. Thus, the OTIB-0018 protecol
should be applied to calculate intake rates for 1/1/1948 through 12/31/1950 with the same
Modifications used for 1/1/1944 through 12/31/1947 [7].

5.5 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information,
conclusions, and recommendations provided to assist in the process of worker dose reconstruction.
These callouts are listed here in the Attributions and Annotations section, with information to identify
the source and justification for each associated item. Conventional References, which are provided in
the next section of this document, link data, quotations, and other information to documents available
for review on the Project's Site Research Database.

Kenny Fleming served as one of the initial Subject Experts for this document. Mr. Fleming was
previously employed at ORNL and his work involved management, direction or implementation of
tadiation protection and/or health physics program policies, procedures or practices related to atomic
weapons activities at the site. This revision and earlier revisions have been overseen by a Document
Owner who is fully responsible for the content, including ail findings and conclusions. In all cases
where such information or prior studies or writings are included or relied upon by Mr. Fleming, those
Materials are fully attributed to the source.

Michael Bollenbacher served as one of the initial Subject Experts for this document. Mr. Bollenbacher
was previously employed at ORNL and his work involved management, direction or implementation of
radiation protection and/or health physics program policies, procedures or-practices related to atomic
weapons activities at the site. This revision and earlier revisions have been overseen by a Document
Owner who is fully responsible for the content, including all findings and conclusions. In ail cases
where such information or prior studies or writings are included or relied upon by Mr. Bollenbacher,
those materials are fully attributed to the source.

Elizabeth Brackett serves as a Site Expert for this document. As such, she is responsible for advising
on site-specific issues and incidents as necessary and ensures the completeness and accuracy of the
document. Because of her prior work experience at the site, she possesses, or is aware of
information that is relevant for reconstructing radiation doses experienced by claimants who worked at
the site. In all cases where such information or prior studies or writings are included or relied upon by
the document owner, those materials are fuily attributed. Ms. Brackett’s Disclosure Statement is
available at www.oraucoc.org.

[1} Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007.
Section 2 of this Site Profile (ORAUT 2006) provides a more detailed discussion of the key
radionuclides workers encountered.

[2] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007.

This list was compiled from queries of the database of in vitro bioassay results provided by
ORNL.

{3] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007.
The system of absorption types promulgated in ICRP Publication 66 (ICRP 1994) is used for



[Document No. ORAUT-TKBS-0012-5 | RevisionNo. 01 __| Effective Date: 10/01/2007 |__ Page 4 of 74_]

A.1.1 Data Description...
A.1.2 Data Classificatio
A.1.3. Synopsis of Recovered Data

36
37
37

A.2_ Calculation Method for Sample MDAs....

A.2.1_ Standard Deviation of Backgrount

A.2.2 Background Count Time

A.2.3 Instrument Efficiency .

A.2.4 Yield..

A.2.5 Volume iquot Analyzed.

A.2.6 Volume of 24-Hour Sample .
A3 Calculation Method for MDAs for Various Periods Between 1945 and 1988 .........-:-csee 44
A.4 Additional Considerations.........c:.scssssscsssssssscessseeecesssesessnneessssesssseesssseesssnesseennnecteanneecesssee AY
ATTACHMENT B, ASSESSMENT OF INTERNAL DOSE AT ORNL: 1944 to 1947.00... esssesesessees 71

LIST OF TABLES

TABLE TITLE
5-1 Radioanaytical results between 1947 and 1988...
5-2 Solubility classifications used by ORNL for reevaluation of historical bioassay ‘results. 1
5-3 Routine bioassay monitoring methods and frequencies from 1973 to 1982............. 12
5-4 Historical division codes 13
5-5 New division codes .... 14
5-6 DP urinalysis abbreviations and codes 14
5-7 Fecal analysis codes 15
5-8 Result codes for whole-body counting 15
5-9 Recommended in vitro MDAs (dpm/2:

1989... ae 16
5-10 MDAs for i in n vitro samples after 1989. 17
5-11 Construction history of the counting room 20
5-12 1965 selection criteria for whole-body counting... 21
5-13 Qualitative information concerning radioactive mat

facility from 1966 to 1983 . seteeesenenee 22
5-14 Qualitative information concerning radioactive material detected utilizing the WBC

facility through 1978....... 23
5-15 Maximum measured in vivo ‘activity from 1961 to 1966 24
5-16 Mean body burdens of '’Cs from fallout in United Stat 27
5-17 MDA values for six-detector HPGe array............... 27
5-18 Current MDA values for ORNL scanning bed counte! 27
5-19 Current MDA values for ORNL germanium thyroid counter. 28
5-20 Current MDA values for ORNL fixed bed counter .28
5-21 Frequency of in vivo monitoring
A-1 Data used to determine historical MDAs for radionuclides in urine and feces.
A-2  MDAs calculated from recovered data for radionuclides... seeseeeeene



Special. Exposure-Cohort Petition U.S. Department of Health and Human Services

Under the Energy Employees Occupational Centers for Disease Control and Prevention
Iliness Compensation Act National Institute for Occupational Safety and Health
; OMB Number: 0920-0639 Expires: 09/20/2013

Special Exposure Cohort Petition — Form B ppendix — Continuation Page

Continuation Page Photocopy and complete as iiecessary

| am attaching a copy of AECU-817, LECTURE NOTES; HEALTH PHYSICS
TRAINING LECTURES, 1948-1949, September 29, 1950, Health Physics Division,

Oak Ridge National Laboratory, Oak Ridge, Tennessee. This appears to be a
series of lectures, for student Health Physics Technicians at ORNL. This document
provides great insight into the Health Physics practices at ORNL, during the mid
1940's. Page 6 of this document, under Personnel Monitoring, states "External
exposures are measured with the use of the pocket meter (ion chamber) and film
meter. Internal exposures are determined in a limited way by a study of body
fluids.” All throughout this document it is evident that ORNL Health Physicists did
not know the extent of radiation damage to tissues, simply because no bioassay
procedures were in place. Pages 29-34 provides the rational for converting air
sampling to tissue dose, under the section entitied “Relationship between
Concentration of a Radioisotope in Tissue and Dosage Rate." This lecture provides
several calculations to derive tissue dose from air samples and includes the
assumption that a very certain amount of the quantity of the radioisotope is passed
from the body by physiological elimination. Key to these calculations are the
ability to determine C (= number of curies of the radioisotope in the tissue) and E
(= energy per disintegration, measured in Mev). Page 32 provides insight into

the difficulties of determining C & E for these equations. Page 32 states: "The
principle difficulty involved in making estimates of dosage rates due to
radioisotopes within the tissue is that of obtaining reliable values for C and E. It
has already been indicated that the above equation is valid only if the material is
uniformly distributed through the tissue. In many cases the extent to which this is
true is difficult to estimate. Further difficulty arises from inadequate knowledge of
tapidity and percentage of absorption into the tissues and the rates of elimination
from the tissues." "For many radioisotopes, values of E are only roughly known."
“Sometimes, however, the disintegration scheme is quite complex, and, more
frequently than not, it is not definitely known.” On pages 34-36 we find a lecture
concerning "CONTAMINATION AND DUST PROPOGATION." This lecture
provides some insight into the importance of particle size and surface area. This is
stated on the bottom of page 34 as "In other words, relatively small amounts of
material will give large amounts of particulates." Page 36 continues to stress the
fact that particle size is important in regard to whether the particle is retained in the
lung. Another telling statement, on page 36_is “The lack of pertinent biological
data has been such that it is impossible to properly evaluate the actual
physiological hazard caused by the inhalation of insoluble radioactive particles.”
Page 61, under the section entitled "CALIBRATION TECHNIQUE" in regard to

AS

Name or Social Security Number of First Petitioner:

Continuation pg. 2


"4 AECU-817

X-10 area laboratories desiring to obtain some of these rings for their use may order them
through J, C. Ledbetter, Personnel Monitoring Section, Health Physics Division, Oak Ridge National
Laboratory.

(5. C. Ledbetter)

6.3 MEASUREMENTS ON INTERNAL EXPOSURE
By L. B. Farabee

The chief purpose of the Health Physics Division is to protect personnel from bodily injury due
to radioactive hazards. The many branches of Health Physics coordinate into a unit, with each branch
taking care of its specific duties. Instrument research and development, survey monitoring of con-
taminated working areas and materials, and personne] monitoring are integral parts which contribute
to the above objectives. Body-fluid analyses rightly becomes a part of personnel monitoring.

Health physics surveyors in the field are constantly alert to survey areas and give advice to
chemists, technicians, and laborers as to what conditions of safety from radiation exist, It is their
purpose to help, instruct, and give advice to workers who are less familiar with radiation dangers,
The department which carries out body-fluid analysis is the final link in the personnel-protection
program. The purpose of this group is to measure internal exposure by means of analyses of bodily
excretions.

The analysis of the human body shows that about every element found in nature goes to make up
the body. Some are harmless in small amounts, but hazardous in excessive amounts. These elements
can be detected in excretions from the body provided that enough is present so that detection can be
made possible. The fact that these elements can be detected provides the basis for organization of
a group in Health Physics for making such measurements.

‘The body~-fiuid analysis group in Health Physics is primarily interested in determination of
radioactive materials. Therefore this group must be closely linked with the medical department
and biological research, Chemical isolation of the radioactive material must first be done; then ex~
trapolation to internal exposure is made on the basis of biological research done on experimental
animals, :

Since the organization of a group for body -fluid analysis at Clinton Laboratories in July 1945,
the main concern has been the development of procedures of chemical] isolation of radioactive ele-
ments, followed by the perfecting of these procedures to a point where they were usable. The de-
velopment of instruments which are rellable has been one of the major problems, The quantity of
any radioactive material to be found in either urine or feces is exceedingly small. Therefore the
connting instruments must be dependable over a long period of time.

6.4 WASTE-DISPOSAL THEORY AND PRACTICE
By R. J. Morton

‘There are two ways in which I might discuss this problem: (1) a detailed discussion of radio-
active waste problems here or (2) a more general discussion of the disposal of all types of waste
here and elsewhere. My purpose is to discuss this problem in a general way so as to present a
broad picture of waste-disposal problems as a whole and then speak more particularly of radio-
active wastes.

When we have to deal with new problems, the tendency is to think of them as separate and en-
tirely different from previous ones. After a while we usually find that there are many more points
of similarity than we had realized —similar in principle, the same people are concerned, the same
basic interests prevail, and previous experience gives valuable guidance in studying and solving.

Waste disposal 1s 2 universal problem, for example, in communities, industries, etc. It is axi-
omatic that communities and industries must find methods of satisfactory disposal for health, growth,
and existence. Cities spend a large part of their income for sewerage, refuse disposal, and control
of atmospheric pollution. Some industries have been forced to close or move, Lack of proper sewer-



[Document No. ORAUT-O1IB-0018 __| Revision No. 07 [Effective Date: 08/09/2005 [Page 33 of 35 |

start of a chronic type M inhalation but a type S inhalation might not result ina detectable urine
sample.

Urinary Exeretion (pCi/day)

0 2000 4000 6000 8000 10000 12000
Days after start of chronic intake

Figure B-4. Hanford Pu expected urine versus actual results.

B.3 FISSION PRODUCTS

The fission product that was primarily sampled in urine at Hanford was Sr. In the 1960s, the

recording limit at Hanford for °°Sr urine samples appears to be 16.7 pCi/day. More than 95% of the
samples in 1966, 1967, and 1968 were recorded as less than this limit. After that period, the

recording limit decreased and enabied the evaluation of the median sample result.

Figure B-5 shows the median and 95"-percentile urine result predicted from the values used in this
TIB for a chronic inhalation of type F °Sr. The 95th-percentile results were determined using the
assumption in this TIB that the intakes are lognormally distributed with a geometric standard deviation
of three. The figure also shows the recording limit of 16.7 pCi/day as well as the median sample
results for 1970 through 1974.

Figure B-5 indicates that the results predicted from the 95th percentile of the intake in this TIB exceed
the reporting limit from the 1960s. More than 95% of the values were reported as less than this value.
In addition, it shows that when the reporting limit was reduced, the median result was less than the
mean result predicted from the intake in this TIB.

B4 CONCLUSION

The analysis discussed above indicates that the assumption made in this TIB is valid if the material to
which an individual is exposed has type F or type M absorption characteristics. Because these are
the limiting absorption types for systemic organs, it appears the assumption is valid for systemic
organs.

Type S, however, is the limiting absorption type (when comparing the same intake values for different
absorption types) for the respiratory tract. Due to the detection limits of urinalysis, it cannot be shown



AECU-817 a

b. Prompt action
¢. Patient explanation to extent needed
d, Intelligence
4, Cooperation
a. No backbiting
b. Mesh with efforts of others on your side
¢. Be open minded enough to use the good ideas of others
B. Dominance
1, Maintain leadership in knowledge of field
2. Report to top authority (in T.O.)
3. Be practical as well as ideal

TV. Measuring success of Health Physics

A. Any job can be done safely under any tolerance level that may be established if the will to do so,
time, and funds are available

B, Failures may not show up for generations or may not be recognizable, Also false evidence may
be hard to disprove.



i
'

[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 __| Effective Date. 10/01/2007 [ Page 340f74 |

Morgan, K. Z., W. S. Snyder, and E. G. Struxness, 1965, Health Physics Division Annual Progress
Report for Period Ending July 31, 1965, ORNL-3849, Oak Ridge National Laboratory, Oak
Ridge, Tennessee.

Morgan, K. Z., W. S. Snyder, and E. G. Struxness, 1966, Health Physics Division Annual Progress
Report for Period Ending July 31, 1966, ORNL-4007, Oak Ridge National Laboratory, Oak
Ridge, Tennessee.

Morgan, K. Z., W. S, Snyder, and D. M. Davis, 1967, Health Physics and Safety Annual Report for
1966, ORNL-4146, Oak Ridge Nationai Laboratory, Oak Ridge, Tennessee.

NCRP (Nationa! Committee on Radiological Protection and Measurements), 1988, Exposure of the
Population in the United States and Canada from Natural Background Radiation, Report 94,
Bethesda, Maryland.

ORAUT (Oak Ridge Associated Universities Team), 2005a, Intemal Dose Overestimates for Facilities
With Air Sampling Programs, ORAUT-OTIB-0018, Rev. 01, Oak Ridge, Tennessee, August 9.

ORAUT (Oak Ridge Associated Universities Team), 2005b, Intemal Dosimetry Coworker Data for X-
10, ORAUT-OTIB-0034, Rev. 00, Oak Ridge, Tennessee, December 13.

ORAUT (Oak Ridge Associated Universities Team), 2006, Oak Ridge National Laboratory — Site
Description, ORAUT-TKBS-0012-2, Rev. 01, Oak Ridge, Tennessee, August 30.

ORNL (Oak Ridge National Laboratory), 1961, Procedures and Practices for Radiation Protection,
Health Physics Manual, Oak Ridge, Tennessee.

ORNL (Oak Ridge National Laboratory), 1982, Guides for Intemal Exposure Monitoring, Procedure
A-12, in Procedures and Practices for Radiation Protection, Health Physics Manual, revised
May 15, Oak Ridge, Tennessee.

ORNL (Oak Ridge National Laboratory), 2002, Oak Ridge National Laboratory Intemal Dosimetry
Program Technical Basis Document, Revision 4, Oak Ridge, Tennessee.

Thompson, W. E., 1963, History of the Oak Ridge National Laboratory 1943-1963, First Rough Draft,
Oak Ridge, Tennessee.

Watts, J. R., J. M. Barber, J. A. Biggerstaff, and G. R. Rao, 1995, Oak Ridge National Laboratory
intemal Dosimetry Group In Vivo Monitoring Technical Basis Document,
TB/IDG/WBCL-95-RO, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Wirth, J. E.,1945a, “Present Status of Relationship of Analyses of Plutonium in Urine to the Plutonium
Hazard,” memorandum to Medicat Files, Oak Ridge, Tennessee, May 28.

Wirth, J. E., 1945b, “Product in Urine,” memorandum to Medical Files, May 11, Oak Ridge,
Tennessee.



(Document No. ORAUT-TKBS-0012-5 |

Revision No. 04

[ Effective Date: 10/01/2007 | Page 12 of 74_}

Table 5-3. Routine bioassay monitoring methods and frequencies from 1973 to 1982° (ORNL

1982).
Routine sampling categories

Radioactive material \ u i
Pu-241 and alpha emitters | Urine 3-4 wk Urine 6-13 wk Urine 6-12 mo

other than uranium Whole-body 3-4 mo | Whole-body 6-13 mo_| Whole-body 6-12 mo
Sr-90 and uranium Urine 4-6 wk Urine 4-13 wk Urine 6-12 wk
H-3 Urine each wk Urine each mo Urine each gtr
1-134 Whole-body each wk _| Whole-body each qtr_{ Urine each qtr
Co-60 and Cs-137 Whole-body 3 mo. Whole-body 6 mo Whole-body each 6 to 12 mo
All others Consult wiinternal Dose Group

a. Frequency of sampling should be in accord with employee's potential for exposure as determined by health
physics surveyors. The following is a guide:

Category |

Persons actively involved in operations or processes containing quantities of radioactive material,

and when there is some evidence of contamination (i.e., positive results from smear and air

samples).
Category It

Category II!

5.1.6.2

ORNL has collected whole-body, lung, and wound counting data for employees since 1959. For the

In Vivo Monitoring Program

Employees working with relatively small quantities of materials that are confined or when there is
no evidence of contamination or activity.

Employees working with radioactive material or in the vicinity of material when there is no known
exposure but some potential for exposure.

most part, in vivo counting was used until the late 1980s to confirm potential intakes from known
incidents or identified by the in vitro monitoring program. Though Morgan, Snyder, and Struxness
(1965) indicated that routine in vivo monitoring for all site radiological workers began in 1965, it

appears that a forma! program did not begin until the mid- to late 1980s. Section 5.3 discusses the in
vivo monitoring program and the data it produced.

5.1.7

ORNL used various formats on paper records to record bioassay results through most of the period of

Recordkeeping

interest. in addition, ORNL has entered much historical in vitro monitoring data into a database.
These data were used to estimate MDAs as described in this document (see Attachment A).

Discussions with personnel responsible for radiological records (Dixon 2004) indicated that hard-copy
data are maintained at ORNL and that upon formal request these records are made available to the

requester. These records will be used by dose reconstructors in estimating intakes.

It is not practical for this document to provide examples of all record formats used in the past six

decades; MMES (1995) contains a guide to historical record formats. Most bioassay records share
the following information: name, badge or other ID number, division code, health physics area, date,
analysis code, results (in disintegrations per minute per 24-hr sample), and a reason for the analysis.
A review of ORNL claim files indicated that many bioassay forms had slight revisions throughout their
use, but that the information they maintain is similar from one revision to the next and should be easily

interpreted.

6.1.7.4

The division code sometimes provides information on individual locations and job assignments. Table
5-4 lists historical division codes, and Table 5-5 lists more recent codes.

Division Codes



[Document No. ORAUT-TKBS-0012-5 |_Revision No. 01 _| Effective Date: 10/01/2007 | Page 3of74

TABLE OF CONTENTS
SECTION TITLE PAGE

Acronyms and Abbreviations

5.1 Introduction
5.1.1 Purpose
Scope...
Radionuclides of Concern.
Solubility Classes ........
Route of Intake and Particle Sizes
Bioassay Programs ..................
5.1.6.1 /n Vitro Monitoring Program .
5.1.6.2 In Vivo Monitoring Program..
5.1.7 Recordkeeping ........
5.1.7.1 Division Codes.
5.1.7.2 Radioanalytical Abbreviations and Codes .
5.1.7.3 Whole-Body Counting Results Codes ...

AAA ANS

4.
1
1.
1
1

Onkwbr-

5.2 In Vitro Minimum Detectable Activities, Counting Methods, and Reporting Practices
5.2.1 In Vitro Minimum Detectable Activities...
5.2.1.1 Minimum Detectable Activities between 1947 and 1989
5.2.1.2. Minimum Detectable Activities after 198:

5.2.2 Counting Methods for /n Vitro Sample:
5.2.2.1 Alpha Spectrometry.
5.2.2.2 Liquid Scintillation ..
5.2.2.3 Zinc-Sulfide Scintillation
5.2.2.4 Gamma Spectrometry...
5.2.2.5 Gas Flow Proportional Counter (Beta Counting) .

5.2.3 Notes on Measurements of Alpha Emitters
5.2.3.1 Trivalent Alpha Actinides
5.2.3.2 Plutonium............
5.2.3.3 Environmental Uranium.

5.3. In Vivo Minimum Detectable Activities, Counting Methods, and Reporting Practices..............19
5.3.1 Shielded Counting Room
5.3.2 Detectors, Geometries, and Techniques .21
5.3.3 Cesium Counting Artifact
5.3.4 Minimum Detectable Acti

y.
5.4 Using Air Monitoring Data To Assign Best Estimates of Dose

5.5 — Attributions and Annotations ...........ccccscceseseseesseeseseeseeteeserenesseeeseeseatensenertantiesesseeteessessese: OO.
RefEPENCES oo... es cseescsesescsesesesesenenssesesaseveossescseesaneveasssaeecscstssesssecesaeseseecassestecsuseaenassseessesrecsegesseesans OD
ATTACHMENT A, CALCULATION OF MDAS FROM HISTORICAL DATA.........:ssessessseseessesesseesieese 35

A.1 — Recovery of Historical Data 0.0.2... ceseessessssseesesseseeseesesssnseneseeeeseensnesecsertenssesssseescseeesseereansante 36



AECU-817

5. Handling telephones, reports, etc., while wearing protective gloves
6. Building of inadequate shields, i.e., a lead shield with one or more sides missing
1. Failure to wear coveralls, gloves, rubbers, respirators, and such protective chothing when
job conditions warrant contamination protection
8. Continued wearing of coveralls, etc,, that are highly contaminated
9. Failure to wear personnel-monitoring meters, especially for ‘‘hot’’ operations
10. Tampering with personnel monitoring meters to confuse the health physics records
11. Pipeting by mouth of radioactive solutions
12. Handling radioactive sources without proper protection by distance or shielding
13, Storage of radioactive material without use of proper danger signs
14, Leaving radioactive samples on benches when not in use
15, Storage of radioactive material in rooms assigned to other men without proper notification
of person responsible for that room
16, Improper disposal of contaminated apparatus, e.g., ‘hot’? material in an ordinary trash
basket, or mixing of contaminated tools or equipment with clean items
17. Discharge of ‘‘hot”’ solutions via a ‘‘cold’’ sink
18. Failure to check for hand contamination on a routine basis
19, Assuming because an operation has been done before without observed ill effects that it is
therefore safe
20, Continued acceptance of a region approved by a survey as safe without attention to changes
caused by activity growth or transfer of radioactive materials from safe enclosures
21, Depending upon a statement of safety regarding a particular material without concern for
the interpretation of safety relative to the working conditions

Contamination Must be Controlled

Restricting radioactivity to as small a zone as possible makes it easier to protect people from
health hazards and simplifies material accounting. Immaculate housekeeping is almost synonymous
with close control of contamination.

An ideal is approached when all radioactive materials are kept in a space sealed away from
personne!. Hoods with an adequate air flow (in excess of 100 ft/min) cannot be expected to prevent
tactile transfer of contamination. A cell to provide a contamination seal may take the form of a box
equipped with windows in which the work with radioactive materials is performed manually through
rubber gauntlets built into the walls of the enclosure. The colloquial term for this device, which
permits work under a controlled atmosphere, is ‘‘dry box.’’

A more elegant dry box may be the size of a hood and be provided with built-in utilities. A
service corridor should provide access to the rear, where removable panels permit changes of
equipment without spreading contamination to the operating room. The service corridor can be ar-
ranged to accommodate any number of units. It terminates in a decontamination room and a decon-
tamination change room. No occupancy in the service corridor is permitted without complete pro-
tective clothing and alr-supplied hood, All items removed must pass through the decontamination
room and be effectively cleaned or acceptably packaged.

The infiltration of contamination into walls and furred spaces must be controlled to minimize
hazards to fire fighters and to facilitate razing and remodeling when they occur. Air-borne con-
tamination has been known to travel through wall spaces along plumbing and to emerge in zones
where no provision for contamination had been prepared. This phenomenon can be overcome by
maintaining the atmosphere in spaces occupied by radioactive materials at pressures below ambient.

Smooth impervious simple surfaces facilitate the recovery of contamination that may escape
experimental apparatus by accident. Stainless steel, glass, and lead can be subject to cleaning with
strong acids. Hf linoleum surfaces fail to respond to decontamination, they can be economically re-
placed. Plastered surfaces should be sealed with a glossy paint. Seriously contaminated concrete
requires physical removal. Obscuring of contamination, such as the covering of alpha emitters with
paint, is unacceptable.

95



42 AECU-81T

It is shown in Fig. 4.2 that the plateau (that is, the region in which the counting rate is independ-
ent of the voltage) is a linear function of the value of this quenching resistor. In practice it is very
desirable to have a large plateau, but this cannot be done with a resistor since the presence of the
high resistor presents a serious limitation on the counter from the viewpoint that time constants
would become too long for effecting a system that is capable of recording high counting rates.

Neher-Harper Quenching

Various electronic circuits have been designed which will momentarily remove the potential
across the counter when current begins to flow. These circuits are very simple and are much more
satisfactory than resistance quenches. Famous among these is the Neher-Harper circuit.

Self-Quenching Counters

Today the self-quenching counter is almost universally used because of its simplicity. This
counter is constructed in a manner similar to the non-self-quenching counter. The only difference
is that a polyatomic molecule is added to the inert gas, A common example is alcohol added to argon
{usually 10 parts of argon to 1 part of alcohol, filled to a total pressure of 8 to 10 cm Hg). The func-
tion of the polyatomic vapor is to absorb the photons that are emitted during the discharge. This
absorption eliminates the need for external quenching.

However, the very generous participation on the part of the polyatomic molecule presents a
useful lifetime limitation to the counter. During each discharge a large number of these organic
molecules are decomposed. This means that after a certain number of counts the counter will n0
longer be useful. The counter characteristics change gradually owing to an increased pressure
brought about by the broken-down organic molecule.

Indicating Devices

The pulse produced by G-M tubes may be recorded in various ways. The most commonly used
methods are the following:

1. For indication only, a G-M tube may be used to fire a thyratron, which in turn operates a set
of earphones.

2, For a fairly quantitative measure, a rate meter may be substituted for the earphones.

3. Very accurate records may be obtained by using a scaler in conjunction with a mechanical
recorder.

44 THE PROPORTIONAL COUNTER
By R. A, Dandl

In any discussion of a gas counter it is desirable to introduce the gas characteristic, and then
stress the region one is interested in.

LOAD
RESISTOR

aA -———_| ay AMPLIFIER |

SUPPLY shy
(Cf fsaiwea mi

Fig. 4.3

Figure 4.3 is a block diagram of a typical counter circuit. Qy is the charge produced in the gas by
one particle and is independent of the high voltage, Q2 is the total charge on the chamber capacity

as a result of a particle, ¢ is the sum of the chamber capacity and the input capacity of the amplifier,
vy is the voltage change across the chamber due to collection of Qg, a is the gain of the amplifier, and
Ep is the pulse voltage out of the amplifier.



[Document No, ORAUT-TKBS-0012-5 | Revision No. 01__| Effective Date: 10/01/2007 [ Page 53 0f74_|

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 19 of 36

TEs =
+ Sample MDA
—H— Est, MDA of 396 (95%CL on Mean)
+ > 50th Percentile of 40
1B | —________—_- - +
z —————___i____s
—
i i
§ 1B -
=
2 One rece eee eee “0
3 :
1 Ea) |-—-—_— + —
1 E+ —— +
1945 1950 1985 1960 1965 1970 1975, 1980 1985 1990

Year

Visual Comparison of Data to Estimated Pm-147 MDAs in Urine

Figure A-7. MDA data for Pm-147 in urine (part 1 of 3).

Annual Average MDAs for Pm-147 in Urine
Calculated from Recovered Data Between 1945 and 1990

300

MDA (dpmv24 bh)

Figure A-7. MDA data for Pm-147 in urine (part 2 of 3).



48 AECU-~-817

9.55
200 7 0.0477 or 5%

0.05 x 200 = 10 ct/min

Therefore, counting rate = 200 + 10 ct/min.

Tf all counts are made for the same length of time when plotting a decay curve or absorption
curve, a correction would need to be applied for each count, and the error would be increasing as
the curve slopes downward (see Fig. 5.2).

Fig. 5.2

If the per cent of error is specified when the counting requisition is made, all counts will have the
same accuracy. This is done by the counting technician by referring to tables which have the ratio
on horizontal axis and total number of counts needed to get the required accuracy on the vertical
axis, Note on the charts that to get low probable error, long counts are necessary unless the ratio
of counts to background is great. This shows why a low background is necessary for accurate count-
ing (Fig. 5.3 1s a rough copy of actual charts used),

NN 2% enmon

3% ERROR

3
8

2000

1000

pr

TOTAL COUNTS NEEDED

°
o12345678

RATIO: BRURGRGUND COUNF RATE
Fig. 5.3

Coincidence Loss. Where probable error is a loss of accuracy due to a very low count, coinci-
dence Loss is due to a high count or directly proportiona! to the counting rate. This is due to the fact
that the resolving time of the counter is such that one count might be counted with another, as one,
The amount of correction used is 1%/1000 ct/min for all counting in this counting room.

Special Counting Services

Long-lived Alpha and Beta Activity. In the determination of long-lived activity in radioactive
dusts and mists, a correction must be made for the presence of radioactive contamination from
natural radon and thoron sources. Four hours after collection of an air sample, the radon decay
products will have decayed to a negligible quantity (27-min half-life, Ra B, is the controlling half-
life). The 10.6-hr thorium B controls the thoron chain decay so a correction for this activity must
be made to get the long-lived activity. This correction is applied in the following equation for long-
lived activity

C2 -€1(e-ASt)

Cus
oy — (ety



[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 _| Effective Date. 10/01/2007 | Page 11of74 |

Table 5-2, Solubility classifications used by ORNL for reevaluation of historical bioassay results
ORNL 2002).

Material Intake
class mode Radionuclides
V (Very soluble) Ingestion [| H-3.
L (Labeled organic) | Inhalation | C-14
D (Days) Inhalation { Na-22, P-32, P-33, S-35, Rb-86, Sr-85, Sr-89, Sr-90, 1-125, 1-129, 1-131,
Ba-133, Cs-134, Cs-137, Eu-152, Eu-154, W-188
W (Weeks) Inhalation | Ca-45, Cr-51, Mn-54, Fe-55, Fe-59, Ni-63, Ge-68, Sc-75, Tc-99, Gd-153,

Hg-203, Bi-207, Po-210, Ra-226, Np-237, Am-241, Am-243, Cm-242,
Cm-244, Cm-248

Y (Years) Inhalation | Sc-46, Co-57, Co-58, Co-60, Cu-64, Zn-65, Cu-67, Y-88, Y-90, Pd-103,
Ru-106, Cd-109, Ag-110m, Pm-147, {r-192, Os-191, Th-228, Th-229,
Th-230, Th-232, Pa-231, U-232, U-233, U-234, U-235, U-236, U-238,
Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Cf-249, Cf-252

5.1.5 Route of Intake and Particle Sizes

Unless additional information is provided, it should be assumed in all cases that the route of intake for

internally deposited radionuclides was via inhalation of 5 ym Activity Median Aerodynamic Diameter
(AMAD)-sized partictes.

5.1.6 Bioassay Programs

§.1.6.1 in Vitro Monitoring Program

ORNL has collected urine and fecal samples from individuals suspected of potential intakes from
1947 to the present. Urine samples have been and still are the preferred method. While fecal
samples can provide good supplementary information to determine when an intake occurred, the
chemical solubility of the material, and particle size, there typically is more variation associated with
these samples than with urine samples. When fecal samples are obtained with urine samples after a

known intake, the results can be used to better understand the intake parameters and provide a more
accurate estimate of intake.

Urine samples were collected in the early years of the bioassay program based on the area health
physicist’s knowledge of field conditions (e.g., known spills/incidents, air and contamination sample
results, etc.). This practice of scheduling did not utilize a specified sampling frequency (Auxier 2003;
Henley 2003). A 1961 procedure manual (ORNL 1961) references procedures and practices
governing the health physics program at that time, including internal and external exposure
monitoring. Although referenced, the procedure detailing internal dosimetry was not among the
documents available for review.

A similar manual of health physics procedures and practices from 1982 lists a detailed set of
requirements for graduated routine sampling, depending on the frequency and extent of a potential
exposure (ORNL 1982). Table 5-3 lists ORNL-published routine bioassay monitoring methods and
frequencies from 1973 through 1982. NOTE: Historical information of the in vivo monitoring program
(see Section 5.3) indicates that the whole-body counting frequencies listed in Table 5-3 were not
consistently followed. Discussions with previous site personnel indicate that no formal counting
frequency was used at ORNL until the later 1980s (Berger 2004).

Section 5.2 discusses the in vitro monitoring program and the data it produced.



*

Special Exposure Cohort Petition U.S. Department of Health and Human Services

under the Energy Employees Occupational Centers for Disease Control and Prevention
illness Compensation Act National Institute for Occupational Safety and Health

iti OMB Number: 0920-0639 Expires: 09/20/2013
Petitioner Authorization Form Page 2 of 2

Public Burden Statement

Public reporting burden for this collection of information is estimated to average 3 minutes per response,
including time for reviewing instructions, gathering the information needed, and completing the form. if you
have any comments regarding the burden estimate or any other aspect of this collection of information,
including suggestions for reducing this burden, send them to CDC Reports Clearance Officer, 1600 Clifton
Road, MS-E-11, Atlanta GA, 30333; ATTN:PRA 0920-0639. Do not send the completed petition form to this
address, Completed petitions are to be submitted to NIOSH at the address provided in these instructions.
Persons are not required to respond to the information collected on this form unless it displays a currently
valid OMB number.

Privacy Act Advisement

In accordance with the Privacy Act of 1974, as amended (5 U.S.C. § 552a), you are hereby notified of the
following:

The Energy Employees Occupational Illness Compensation Program Act (42 U.S.C. §§ 7384-7385)
(EEOICPA) authorizes the President to designate additional classes of employees to be included in the
Special Exposure Cohort (SEC). EEOICPA authorizes HHS to implement its responsibilities with the
assistance of the National Institute for Occupational Safety (NIOSH), an Institute of the Centers for Disease
Control and Prevention. Information obtained by NIOSH in connection with petitions for including additional
classes of employees in the SEC will be used to evaluate the petition and report findings to the Advisory
Board on Radiation and Worker Health and HHS.

Records containing identifiable information become part of an existing NIOSH system of records under the
Privacy Act, 09-20-147 “Occupational Health Epidemiological Studies and EEOICPA Program Records.
HHS/CDC/NIOSH.” These records are treated in a confidential manner, unless otherwise compelled by law.
Disclosures that NIOSH may need to make for the processing of your petition or other purposes are listed
below.

NIOSH may need to disclose personal identifying information to: (a) the Department of Energy, other federal
agencies, other government or private entities and to private sector employers to permit these entities to
retrieve records required by NIOSH; (b) identified witnesses as designated by NIOSH so that these
individuals can provide information to assist with the evaluation of SEC petitions; (c) contractors assisting
NIOSH; (d) collaborating researchers, under certain limited circumstances to conduct further investigations;
(e) Federal, state and local agencies for law enforcement purposes; and (f} a Member of Congress or a
Congressional staff member in response to a verified inquiry.

This notice applies to alt forms and informational requests that you may receive from NIOSH in connection
with the evaluation of an SEC petition.

Use of this form is voluntary. Failure to use this form will not result in the denial of any right, benefit, or
privilege to which you may be entitled.

Name or Social Security Number of First Petitioner:



[Document No: ORAUT-OTIB-0018 | Revision No. 01 [Effective Date, 0/0072005 | Page 60f35__|

1.0 PURPOSE

The purpose of this technical information bulletin (TIB) is to provide an alternative method to that
discussed in Technical Information Bulletin 002, Maximum Internal Dose Estimates for Certain DOE
Complex Claims (ORAU 2004), for estimating intakes and internal doses when overestimated doses
are acceptable. The internal doses estimated using this TIB will be smaller than those obtained using
ORAU (2004) under some situations but are still overestimates. See Section 7.0 for additional
discussion about when to use this TIB rather than ORAU (2004).

Technical Information Bulletins (TIBs) are general working documents that provide guidance
concerning the preparation of dose reconstructions at particular sites or categories of sites. They will
be revised in the event additional relevant information is obtained. TIBs may be used to assist the
National Institute for Occupational Safety and Health in the completion of individual dose
reconstructions.

In this document the word “facility” is used as a general term for an area, building, or group of
buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons
employer facility’ or a “Department of Energy facility’ as defined in the Energy Employees
Occupational llness Compensation Program Act of 2000 (42 U.S.C. § 73841 (5) and (12)).

2.0 BACKGROUND

ORAU (2004) provides default overestimated intakes based on the assumption ‘that an intake
resulting in 10% of a MPBB [maximum permissible body burden] would not have likely occurred to an
unmonitored worker or would have likely resulted in a readily noticeable bioassay result in a
monitored worker, readily noticeable air sample, or other indicators of personnel contamination.”
ORAU (2004, Table 3.1.1-2) applies this assumption to each of a list of radionuclides to which
workers at most U.S. Department of Energy (DOE)' sites might have been exposed. The list includes
three beta/gamma-emitting radionuclides and nine alpha-emitting radionuclides for nonreactor sites
and an additional 16 beta/gamma-emitting radionuclides for sites with reactors. However, the ORAU
(2004) method resulted in unrealistically high intakes of long-lived, long-retained radionuclides when
the exposure time was only a few years; that is, the air concentrations would have had to have been
unrealistically high for a worker to have incurred unnoticed intakes resulting in 10% of the MPBB.

For efficiency in processing claims, another method was needed to estimate intakes that are

« Overestimates
© Derived from more realistic air concentrations

Most DOE sites had air sampling programs for particulate radioactive contamination and, with the
exception of short-term, unplanned situations, controlled air concentrations to limiting values
established by the National Council on Radiation Protection and Measurements (NCRP) or DOE. The
first official national guidelines on air concentrations were produced by the NCRP and published in
4953 in National Bureau of Standards (NBS) Handbook 52 (NBS 1953). The NCRP revised its
methodology and expanded the list of covered radionuclides in 1959, published as NBS Handbook 69
(NBS 1959). Both of these documents were based on and consistent with similar guidance from the
International Commission on Radiological Protection (ICRP 1959). Subsequent to the publication of
the NBS handbooks, DOE issued orders or regulations establishing air concentration limits applicable
to its sites [for instance, in U.S. Atomic Energy Commission (AEC) Manual Chapter 0524 (AEC 1968)

‘In this document, reference to DOE includes its predecessor agencies.



Revision No. 01 _| Effective Date: 10/01/2007 | _Page 43 of 74_ |

[Document No. ORAUT-TKBS-0012-5]

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6 AECU-817

The guiding principle is that the quantity of the objectional constituents must not be too great
for the capacity of the method of disposal employed under the most adverse conditions. The capacity
of the method depends mainly upon harmful exposure of people, i.e., potential health hazard.

The purpose of waste treatment is to remove or neutralize the objectionable constituents to ~
bring them within the capacity of the disposal method, for example, sewage bacteria in a stream
near water suppiy or swimming place must be removed by settling, etc.; then, if necessary, kill
remaining bacteria by chlorination of the sewage. -

The same avatlable methods and guiding principles apply to industrial waste disposal, although
the constituents and their effects vary widely, depending upon the type of industry and its particular
processes.

The criteria of proper disposal are;

1, Prevent danger to health

2. Prevent nuisance

3. Prevent interference with rights of others

4, Prevent economic damage

‘The present policy and practice tends toward disposal of industrial wastes together with munic-
ipal domestic sewage wherever feasible, with pretreatment, if necessary, to prevent damage to
sewers or upsetting of sewage-treatment processes.

Certain radioactive wastes contain toxic chemicals which must be considered just as other toxic
industrial wastes, The peculiar property is radioactivity, for example, radioactive toxicity.

Radioactive materials must not be disseminated promiscuously in considerable quantities into
surface waters, underground waters, upon the soil or in the soil, or to the atmosphere unless reli-
able knowledge shows no present or future danger to man or other living things.

This policy sharply limits use of the natural methods of disposal so widely used for other wastes,
namely, water, soll, and air, because knowledge is still inadequate, tolerance values are not well es~
tablished (effects not certainly known), and factors of safety have to be ample.

Five methods of disposal of radioactive wastes mentioned by Forrest Western in a paper pub-
lished in Nucleonics, August 1948, are as follows:

1. Dilution in water reduces amount of ingestion or exposure,

2. Dilution with stable isotopes limits reconcentration.

3. Retention for partial decay to reduce level of activity, as in tanks,

4, Permanent retention by various methods, including burial in earth (presumably will be re-
tained) and concentration and retention in tanks.

5. Making into concrete (if highly concentrated activity) and disposal in oceans (really slow di-
Jution).

All these include uncertainties and present more or less practical difficulties,

Health physicists agree that it is essential to adopt very conservative maximum permissible
limits, to observe them strictly, with constant monitoring and supervision, and to carry on much
more research and investigative survey work to improve our knowledge of radioactive waste-dis-
posal practice.

General purpose and approach in waste monitoring and waste-disposal studies programs of
Health Physics are outlined below:

A. Area monitoring group
1, Purpose or general function: observations, records, and reports to see that
a. Discharges of activity are not greater than prescribed limits,
b, Accumulations or reconcentration of activity do not result in conditions that are actually
or potentially hazardous
2, Approach and activities
a, Liquid wastes
(2) Settling Basin effluent: daily sampling and continuous sampling for daily average. Limit
for beta-emitting materials set at average of § curies/day. Recent study by Setter (Jan-
uary to November, 1948) shows that for 11 months the average was less than $ curies/day



AECU-817

CONTENTS

Page
PREFACE 1

UNIT 1. INTRODUCTION 2

1.1 A Blueprint for the Development of a Health Physicist
(Place of Health Physics in Industry)
By K. Z. Morgan

1.2 Heaith Physics Organization at Oak Ridge National Laboratory
By J. C. Hart

UNIT 2. FUNDAMENTAL PHYSICS 7

2.1 Atomic Structure
By R. S. Thackeray
2,2 Nuclear Structure
By R. 8S. Thackeray
2.3 Transmutation and Nuclear Reactions
By R. 8. Thackeray
2.4 Chain Reaction and the Pile
By W. H. Ray
2.5 Radioactivity Defined
By W. H. Ray
2.6 How to Compute Half-life and Half-thickness
By T. H. J. Burnett
2.7 Fundamental Particles and Units of Measurement
(Activity Evaluation)
By R. 8S. Thackeray
2.8 Interaction of Radiation with Matter
By W. H. Ray

UNIT 3. APPLICATION OF PHYSICS TO HEALTH PHYSICS 22

3.1 The Roentgen and its Derivatives
By R. 8S. Thackeray

5.2 Permissible Radiation Exposure
By L. Leibowitz

3.3. Permissible Contamination Concentration
By W. H. Ray

8.4 Permissible Inhalation Concentration
By F. Western

3.5 Contamination and Dust Propagation
By H. J. McAlduff

3.6 Permissible Waste-disposal Concentrations
By F. Western



AECU-817 27

When two or more different types of radiation contribute to the dose, the total dose is determined
by the combined effect. For example, if a man is in an area in which he is exposed to neutron flux
and gamma radiation at the same time, he will be permitted to work for 1 hr if the neutron dose is
0.05 rem/hr and the gamma dose is also 0.05 rem/hr. If the neutron dose was 0.05 rem/hr and the
gamma dose was 0,02 rem/hr, he wouid be able to work approximately 1 hr and 20 min in the area.

Some people have radioactive substances fixed in their bodies. Such people are limited to a dose
which is the difference between the tolerance dose and the dose from the amount fixed in the body, A
man with a concentration of plutonium in his bones amounting to 0.06 rem/day will be limited to a
tolerance dose of 0.04 rem/day from external exposure.

‘The above discussion applies to total-body exposure. The only tolerance defined for an individual
part of the body in industrial workers is a dose of 1.0 rep/week for the hands.

3.3. PERMISSIBLE CONTAMINATION CONCENTRATIONS
By W. H. Ray

Contamination concentration ‘is usually thought of as radioactivity per unit of surface area and
should have curies cm~2 as a fundamenta) unit of measure. When specifying scrap material for sal-
vage, curies ge”! or curies cm73 may be more significant. Volumetric determinations involve labo-
ratory analysis of samples.

‘The earth’s crust has been estimated to contain 1,3 x 10-12 g of Ra and 1.6 x 10°5 g of Th on the
average per gram of matter. This would indicate 3.6 x 10-1 curie/g (or 79 dis/min/g) if all the
members of only these two naturally radioactrve decay chains were considered, i.e., 46 dis/min al-
pha plus 38 dis/min beta, It is evident that the concentration varies widely from the above average.
Some ores of pitchblende may comprise 84 per cent uranium, Therefore we are ever beset with a
background of radioact:ve contamination above which we must work.

Reasons for limiting radioactive contamination are based on both health and technical objectives.
The basis for health considerations is 300 mrem/week. Contamination should be limited to prevent
overirradiation to personnel, Control of contamination is also imperative to keep radioactive sources
out of the anatomy. The mechanism of tactile transfer of contamination is as impressive as King
Midas’ touch. To keep things out of the mouth is simple, but to keep an environment comfortably
occuptiable, immaculate techniques must be instituted. It is reasonable to expect that surface con-
tamination may become air-borne as a result of erosion. Practically no correlation between surface
activity and concentration of air-borne contamination has been observed thus far, The possibility of
inhaling transient ‘‘slugs’’ of radieactivity lead one to measures of absolute prevention.

Consequently the permissibility of any radioactive contamination is a matter involving profes-
sional appraisal of the specific situation considering the above factors. Sometimes quantities have
been expressed for specific situations, and these values have been unadvisedly published, Such values
are derived from the basic limitation of 300 mrem/week by an appraiser’s estimate of the conditions
pertaining to a special example and should be considered in this light. As a suggestion they are less
reliable to apply to another situation than reconsideration from the basic fundamentals to arrive at
an appropriate recommendation.

For the various reasons stated above, contamination concentrations recommended are usually
very low, sometimes expressed as none. The determination of small values of randomly occurring
events above an appreciable background of also randomly occurring events has practical limitations.
An indication of twice background takes no more time to determine than background and is almost
certainly positive, As a consequence twice background has been suggested as a permissible con-
tamination limit where the intent was zero. Because background is not a universal value, some spec-
ification as to the normal value intended should accompany such a requirement,

Before less than twice the background was suggested, it was evident that contamination limits
were usually suggested at levels dependen upon the sensitivity of available instruments. As sensi-
tivities have increased, the trend has been downward; therefore we may expect the twice background
level to be the limit being approached.



12 AECU-817

Dark Room Equipment
A. Developing Tank

Running-water

Fixer coolant and rinse Developer
68°F 68°F 68°F
5 gal 10 gal 5 gal

10 min 1 min 3min

(Room temperature, 68°F)

B. Water bath = -+--—-—~~-——~~-----~~- woeeee
Running water
54-gal capacity
68°F
(Controlled by hot and cold water mixer)

C. Film dryer
1, Commercial type
2. Provides moderate air flow at 85°F

3. Dries film in approximately 20 to 60 min, depending upon the humidity of the room in which
it is located

Densitometry

A. Derivation of density

I
Tes
Ig
Ig
O=7
Io
D =log10 7

where T = transmission
O = opacity
I = intensity after filtering
Ig = intensity at source
D = density

B, Definition of density

Den. 1 = 5 transmission
1 + cet,
Den, 2 = Too ttansmission

1
Den. 3 = Tooo transmission

1
Den, 4 = 70,000 transmission
Ete.



[Document No. ORAUT-TKBS-0012-5 | Revision No.01__| Effective Date: 10/01/2007 | Page 49o0f74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 15 of 36

Annual Average MDAs for Gross Aipha in Urine
Calculated from Recovered Data Between 1945 and 1990

as

MDA (dpmv24 hy

Figure A-4. MDA data for gross alpha in urine (part 2 of 3).

Number of Sample Resuits Recovered per Year
Between 1945 and 1990 for Gross Alpha in Urine
(2500

2000

1500

1000

Number of Samples (N)

Figure A-4. MDA data for gross alpha in urine (part 3 of 3).



1 AECU-817

adding neutrons, This gives rise to the concept of isotopes in which the atoms of a chemical element
can have a variety of atomic weights without changing the nature of the element itself.

2.2 NUCLEAR STRUCTURE
By Ross 8. Thackeray*

We shall have to study the nuclear structure without having a convenient picture of the depend-
ence of energy on position such as is available in atomic structure, There is no experimental in-
formation about what goes on inside the nucleus. It is possible, however, on the basis of analogy
with atomic structure, to construct a theory which helps to explain what is known as the “‘nucleus.””

In the last lecture we studied the idea of a nucleus constructed of positive and neutral particles
with negatively charged particles moving in planetary orbits around the heavy core. If we restrict
ourselves to the consideration of the Bohr model of the atom, we have an expression for the total
energy of an electron moving in an orbit of radius r. This expression is

~e2

Ero

This shows that as you go closer and closer to the nucleus, the negative energy of the electron be-
comes greater.
Let us now discuss the situation when there are two possible orbits which will be numbered 1
and 2 (see Fig. 2.7). We have then
2
-e
21-3

x2
arg

where e is the electronic charge and r2 is greater than ry. In this situation E2 is greater than E;
since Eo corresponds to a smaller negative energy. In order to transfer an electron from orbit 1 to
orbit 2, positive energy must be added. This can be done by bombardment with radiation, excitation
in an arc, and other means, An electron which has been lifted from its ‘‘home’’ orbit will be at-
tracted to return by the unlike charge on the nucleus. If it falls back to the state of higher negative
energy, the difference in the energy levels Ey and Eg will be available for radiation as a photon of
electromagnetic energy. This difference in energy is radiated as a photon of which the wavelength
or “color’’ is related to the energy by the equation

Eg=

c
SE =hy

in which h is Planck’s constant, c is the velocity of light, and A is the wavelength,

‘The identity and characteristic of the fundamental particles used in building up heavy nuclei
were determined as the combined result of many individual experiments. We can outline the way in
which one such characteristic was determined as an illustration of the method of procedure. As a
stream of electrons bombards hydrogen gas it will strip electrons from some of the hydrogen nuclei.
‘The charged hydrogen ions resulting can be accelerated electrostatically and run as a beam into a
magnetic field. If the charge on the particle, the velocity on the particle, and the strength of the
magnetic field are known, the mass of the particle may be calculated from the curvature in its path
due to the magnetic field. This process is the basis of the mass spectrometer (Fig. 2.6).

No particle mass has been observed which is a submultiple of the mass of the hydrogen nucleus,
nor has any submultiple of its charge been observed. This Jeads to the conclusion that the hydrogen
nucleus is a single fundamental particle, and as such it is known as a “‘proton.’’ The other fundamen-

*Notes taken by C. Scott; edited by R. S. Thackeray



Document No. ORAUT-OTIB-0078 _[ Revision No. 01 [Effective Date: 08/09/2005 | Page 14 of 36 |

7.0 WHEN TO USE THIS TIB

Dose reconstructors should use this TIB for workers who were not on a bioassay program or were
monitored but had no indication of intakes in the bioassay data, except as discussed below and
subject to the limitations in Section 3.0.

Some bioassay methods were capable of detecting small intakes, well below the intakes determined
using this TIB. Whole-body counting for gamma emitters with high gamma abundance and energies
greater than about 200 keV is a good example. Use of this TIB is still an overestimate even if there
are bioassay indications of intakes (i.e., bioassay results exceeding the decision level, minimum
detection level, or recording level, as appropriate), provided the results meet the conditions listed in
Table 7-1 (whole body counts) or Table 7-2 (urinalyses). The values in Tables 7-1 and 7-2 are
actually a factor of 10 less than that which would equal the dose produced by the assumed intake to
account for uncertainties in counting, the possbility of several radionuclides in a single count,
undetected radionuclides, and the different impacts between inhaled or ingested material.

Urinalysis results from the intakes listed in this TIB can be predicted using the current ICRP models.
If an individual has urinalysis results that are less than 0.1 times that predicted by the intakes of this
TIB using the most favorable absorption type, the TIB intake can be considered an overestimate.



{_ Document No. ORAUT-OTIB-0034 | Revision No.00__| Effective Date: 12/13/2005 | Page8of29 |

The IMBA output figures in Attachment A show the fit to the bioassay data for each
annual intake period in terms of days after the date of initial data reporting. For
example, for °’Sr, day 1 equals June 1, 1951, and day 13,515 equals June 1, 1988.

4.3.1

4.3.2

Strontium-90

Strontium results were assumed to be Sr-90. Because of the presence of Sr
both as an effluent from reactor operations and in a highfy insoluble titanite
form, bioassay results for °°Sr were fit using Type F and S material. Figure A-1
in Attachment A shows the individual fits to the 50th-percentile intake values for
Type F *°Sr. Table A-5 summarizes the intake periods and corresponding
intake rates for the 50th- and 84th-percentile values, and the geometric
standard deviations (GSDs) for Type F °°Sr . The GSDs were determined by
dividing the 84th-percentile intake rates by the 50th-percentile intake rates.
Figures A-2 through A-6 show the fits corresponding to each of the subsets of
years across the intake period used to develop the composite figure for Type S
°°Sr. Figure A-7 shows the composite of the subsets of selected groups of
years used to fit the 50th-percentile intake values for Type S °°Sr. Table A-6
summarizes the intake periods and corresponding intake rates for the 50th- and
84th-percentile values, and the calculated GSDs for Type S Sr. The same
intake periods were applied to the 84th-percentile values for both °°Sr Type F
and S$ material because the values followed a similar pattern; results of the
individual fits are not shown here since they were largely in agreement. It is
noted that the urinalysis data for 1964 were insufficient to support a complete
analysis and was therefore was omitted from the sample set.

Uranium-234

Because a variety of uranium enrichments and exposure conditions are
possible at the ORNL site, *“U was assumed for the IMBA intake modeling
even though the source data are for gross uranium in varying amounts,
depending on the enrichment. This does not affect the fitting of the data for
intake determination (i.e., the same total intakes would be obtained for any
enrichment that was assumed) because all uranium isotopes behave the same
biokinetically and the isotopes considered in this analysis have long half-lives
relative to the assumed intake period. The ICRP 68 (1995) dose coefficients
(also referred to as dose conversion factors) for 7“U are 7% to 31% larger than
those for “*U, 2U, and “8U. Because of the isotopic compositions of the
source terms, the *“U dose conversion factor will overestimate doses for any
combination of the uranium radioisotopes, but the assumption of intake of 100%
?%4U) is made to ensure claimant-favorability of the doses.

Because uranium is found in many forms at ORNL, the bioassay data were fit
using Type F, M, and S material. Figures A-8 and A-9 show the individual fits
for the 50th-percentile values for Type F and M @“U, respectively. Table A-7
summarizes the intake periods for the 50th- and 84th-percentile values, and the
calculated GSDs for Type F 7“U, while Table A-8 shows similar information for
Type M“U. Again, the GSDs were determined by dividing the 84th-percentile
intake rates by the 50th-percentile intake rates. For Type S TMU, Figures A-10
through A-16 show the fits corresponding to each of the subsets of years
across the intake period used to develop the composite figure. Figure A-17
provides the composite of the subsets of selected groups of years used to fit
the 50th-percentile intake values. Table A-9 summarizes the intake periods
and corresponding intake rates for the 50th- and 84th-percentile values, and the


;——— UNITED STATES ATOMIC ENERGY COMMISSION
AECU-817

LECTURE NOTES; HEALTH PHYSICS TRAINING
LECTURES, 1948-1949

September 29, 1950
Health Physics Division

Oak Ridge National Laboratory
Oak Ridge, Tennessee

®&

L

Technical Information Service, Oak Ridge, Tennessee



UNIT 4. RADIATION-DETECTION METHODS

4.1 IONIZATION MECHANISM
By R.S. Thackeray

Jonization, or the separating of neutral atoms into two charged portions, accompanies the bom~
bardment of any material by the radiations in which we are interested in this laboratory. The loni-
gation associated with uncharged radiations, neutrons or photons, is caused by the secondary elec-
trons which are set into rapid motion by the primary radiation. The radiations which consist directly
of charged particles, such as alpha, beta, and proton, can cause ionization as a result of the inter-
action of primary radiation with the absorbing material and can also give rise to secondary electrons
which will produce additional ionization,

‘The Wilson cloud chamber is a device for making visible the ionized tracks of radiation passing
through air. Such a device gives a good picture of the effectiveness of the various radiations in pro-
ducing ionization. Cloud-chamber photographs show that alpha rays give broad dense tracks, which
are filled with many ions per centimeter of path length. Beta particles give tortuous, sparsely popu-
lated tracks, which are sometimes so ill defined that they are hard to follow. The tracks of protons
have characteristics intermediate between those of the alpha and beta particles.

A concept which is useful in describing the radiations of the ability to ionize is to speak of their
specific ionization in a certain medium. Specific ionization is defined as the number of ion pairs
formed in a centimeter of track length. In this unit the alpha particle gives a value 1000 times great-
er than that given by the beta particle, which follows from the fact that an alpba tract is broad and
densely populated. There is an empirical relationship in using this concept which has proved to be
of some use as a rule of thumb. This rule states that the damage to tissue, which can be done by the
various radiations, is proportional to the cubic root of the specific ionization for the appropriate
radiation and medium, From this, it follows that an alpha particle with a specific ionization 1000
times greater than that of a beta particle will do damage 10 times greater than that done by a beta
bombardment which yields the same energy to the tissue.

Another empirical value, which has proved of great importance, is the value describing the av-
erage energy lost by an ionizing particle per ion pair formed. We know that it requires between 10
and 15 ev to ionize a single atom. However, the moving particles lose energy by means other than
ionization, and the value for electron volts lost per on pair formed is obtained by dividing the initial
energy of the particle by the total number of ions which it creates. For fast electrons or beta parti-~
cies, the average energy loss per ion pair formed is 32 1/2 ev. By using this constant we can obtain
the total charge to be collected in an ionization chamber simply by multiplying the total energy ab-
sorbed in the ion chamber from the incident radiation by

4.8 x 10-10 esu
32.5 ev

The actual value of charge collected will, however, be somewhat lower because some of the ions
will recombine to give neutral atoms before they can be collected by the lon chamber.

38 AECU-817



[Document No. ORAUT-TKBS-0012-5 |

Revision No. 01

Table 5-19. Current MDA
values for ORNL germanium 7000
thyroid counter. Pusss
Isotope MDA (nCi) 100
1-425 0.06
131 0.12 10 uess
Table 5-20. Current MDA By
values for ORNL fixed bed Am 244
counter. ot
Isotope MDA (nCi) U238
Co-60 0.34 0.01 + + + eames |
Ce-144 6.28 1.60 245 257 Sat 4.04
chert woll thickness

[Effective Date: 10/01/2007 [ Page 28 of 74_|

Figure 5-1. Current MDA values (nCi)} for ORNL germanium lung
counter versus chest-wall thickness (cm) (Watts et al. 1995).

net

a 3

ad

oa + —t———t +——+ “1 1h +

1 16 2 22 3 85 4 48 5 wb 2 238 0 8 884 AB OS
‘chest well chest wall
“Am Pu

Figure 5-2. Current MDA values (nCi) for ORNL phoswich {ung counter versus chest-wall thickness

(cm) for "Am and 2°*Pu (Watts et al. 1995).

Table 5-21. Frequency of in vivo monitoring.

Monitoring type

Applicable period

Frequency

1960-1961 INo set frequency. (Period of facility development —-limited
4- x 4-in. Nal (Tl) routine human counting, typically of incidents)
1961-1962 INo set frequency. (Development activities to change over
8- x 4-in, Nal (TI). to larger detector — typically incidents)
1962-1967-71 No set frequency. (Period spent obtaining baseline counts
8- x 4-in. Nal (TI) onsite employees and incident counts)

Lung counting 1965-7? Frequency stated in Annual Report (see Table 5-12), but
8- x 4-in. Nal (TI) discussions with site personnel indicate that it was not

consistently implemented.

ca 1967-71 to 1976

8- x 4-in. Nal (Tl) and 9- x 9-in.

Nal (TI)

Large detector was added below counting table and used
Iwhen conducting stationary lung counts only.

1962-1976

5-in. dia. x 1/16-in. Nal (Tl)

INo set frequency. (Thin Nat crystals used to quantify low
lenergy photons.)



L_Document No. ORAUT-OTIB-0034 | Revision No. 00 __| Effective Date: 12/13/2005 [Page 5of29 |

LIST OF FIGURES

FIGURE TITLE PAGE

A-1 Assumed °°Sr intake, 1951 to 1988, 50th-percentile results, Type F .
A-2 Assumed “Sr intake, 1951 to 1953, 50th-percentile results, Type S.
A-3 Assumed “Sr intake, 1954, 50th-percentile results, Type S...........
A-4 Assumed °°Sr intake, 1955 to 1964, 50th-percentile results, Type S.
A-5 Assumed °°Sr intake, 1965 to 1983, 50th-percentile results, Type S.
A-6 Assumed °°Sr intake, 1983 to 1988, 50th-percentile results, Type S.
A-7 Predicted strontium excretion rate from independently fit intakes, 1951 to 1988,

50th-percentile, Type S
A-8 Assumed “U intake, 1951 to 1988, 50th-percentile results, Type F.
A-9 Assumed 7“U intake, 1951 to 1988, 50th-percentile results, Type M
A-10 Assumed 7U intake, 1951 to 1952, 50th-percentile results, Type S.
A-11 Assumed 2“U intake, 1953 to 1956, 50th-percentile results, Type S.
A-12 Assumed 4U intake, 1957 to 1958, 50th-percentile results, Type S.
A-13 Assumed “U intake, 1959 to 1961, 50th-percentile results, Type S.
A-14 Assumed U intake, 1962 to 1963, 50th-percentile results, Type S.
A-15 Assumed “U intake, 1964 to 1982, 50th-percentile results, Type S.
A-16 Assumed ”U intake, 1983 to 1987, 50th-percentile results, Type S.
A-17 Predicted uranium excretion rate from independently fit intakes, 1951 to 1988,

50th-percentile, Type S..
A-18 Assumed 2°Pu intake, 1951 to 1952, 50t percentile results, Type
A-19 Assumed *°Pu intake, 1953 to 1968, 50th-percentile results, Type M
A-20 Assumed *°Pu intake, 1969 to 1984, 50th-percentile results, Type M
A-21 Assumed **Pu intake, 1985 to 1988, 50th-percentile results, Type M
A-22 Predicted *°Pu intake, 1951 to 1988, 50th-percentile composite results, Type
A-23 Assumed *°Pu intake, 1951 to 1988, 50th-percentite composite results, Type S .
A-24 Assumed *‘Am intake, 1968 to 1984, 50th-percentile results, Type M.
A-25 Assumed TM"Am intake, 1985 to 1988, 50th-percentile results, Type M.
A-26 Predicted **'Am intake, 1968 to 1988, 50th-percentile composite results, Type



AECU-817 91

B. Acids and salts
1, Increase solubility of the contaminants
2, Complex the contaminants
3. Remove contaminated surface by corrosive action
4. HNOs, solutions varying from a 30-min boil in a 70 per cent solution to dipping them in 10
per cent solution at room temperature
5. HgSO4, HCl, and HgPO, are usually poor agents because of corrosive actions
C. Organic acids
1. Citric, oxalic, and tartaric acids are considered good reagents on corrodible materials
D. Bases
1. NaOH and NagCO3 used on corrodible surfaces
2. NaOH and NajCO3 used as follow up on HNOg
E. Soaps and detergents
1, Zepco (commercial name), good foam action when used with citric acid
2. Sterox (nonfoaming agent), used where air-borne contamination might be present, i.e.,
exhaust fans
F. Organic solvents and oils
1. Electronic instruments, acetone
2, Waxed and painted surfaces, acetone, CCl4
3. Active dusts have been removed by using Spindle Oil, or by a soap solution of glycerine,
followed with a damp cloth
G. Common-ion solution
1. A specific radioactive contaminant may be removed by using a solution of its nonradioactive
isotope
a. KI removes 1131
b. NagPOq removes P32
c. Barium salts may remove Ba contamination
H, Short-lived products may be left to decay
I. List of materials and suggested reagents to be used in the decontamination of a particular ma~
terial
IV. Use, storage, and disposal of materials that are unsuccessfully decontaminated
A. Burial ground
1, Stored above ground
a. Process tanks
b, Machinery
ec, Lead
d, Concrete blocks
e, Red cans
2. Buried material
a. Highly contaminated equipment
b. Contaminated obsolete equipment
B. Salvage yard
1, Valuable contaminated equipment.
2, Slightly contaminated scrap metal, less than 5 mr/hr
3. All materials going to salvage yard should have green tag on them
C. A large percentage of equipment stored in the burial ground is reused; after some decay and
subsequent decontamination materials are safely used
D. Clarification of contamination procedures
E. Equipment to be excessed
1, Other AEC contractors
2. Other government agencies
3. Roane-Anderson for disposal through commercial channels; limits —beta-gamma, twice
background on V-263 at their scrap yard; alpha, none detectable with Poppy or Zeuto



[Document No. ORAUT-TKBS-0012-5 [| __Revision No. 01__| Effective Date: 10/01/2007 [ Page 71of74 |

ATTACHMENT B
ASSESSMENT OF INTERNAL DOSE AT ORNL: 1944 TO 1947
Page 1 of 4
LIST OF FIGURES

FIGURE TITLE

B-1 Plutonium in urine samples ffoM 1945 20.00... cccseessecssssssscnsecssssesssseessesseeseseessevsessaceeaesseeeeeaee
B-2 Plutonium urine sample data through April 18, 1945 ....
B-3 Plutonium urine sample data for April 25-May 30, 1945
B-4 Alpha air-monitoring results for 1944-1947

Information about internal dose at ORNL from 1943 to 1947 is somewhat limited, with no bioassay
results appearing in the DOE worker files. ORAUT (2005a) currently extends back to 1944 for use as
an overestimate at the site because tolerance levels are well documented and there is strong
evidence (memoranda and monthly reports) that exposure was controlled based on the results of the
samples. These memoranda include weekly reports on the airborne activity levels in various areas,
as well as precautions taken or to be taken in the future. Examples of these reports include Bloss
(1944a) to Bloss (1944g).

ORAUT (2005a) generally provides an overestimate for a number of reasons. Exposure at the
maximum permissible air concentration is assumed for 40 hr per week, 50 weeks per year. While
exposure to concentrations greater than this is likely over short periods for some occupations, the
overriding control factor was to limit the average exposure over longer timeframes. In addition, the
nuclide that results in the largest dose is assumed to comprise 100% of the intake; the list of possible
nuclides includes some that are very unlikely to be present in significant quantities. During the early
years of ORNL, several of the nuclides on the list were just being discovered or were yet to be
discovered. The assumed nuclide might change from year to year to ensure that the largest possible
dose is assigned in each year until the date of diagnosis.

Although not included in the individual files, there are limited urine sample results from ORNL in this
period. About 150 results for urine plutonium content were found; all were submitted in 1945. It is
postulated that these are not in the database because they precede the routine sampling program at
the site; they were collected as part of a study to determine “the amount of plutonium, if any, that may
be fixed in the bodies of any chemists at Clinton Laboratories” (Wirth 1945a). Urine samples were
collected weekly for about 6 months at both Clinton (the original name for ORNL) and the Met Lab in
Chicago. The entire data set was fitted to a lognormal distribution (Figure B-1).

The geometric mean of this distribution is somewhat large (1.7 pCi/1400 mL), and the fit to the results
indicates that there are possibly two distributions. If this value is assumed to be due to a constant,
chronic intake of Type M *°Pu beginning in early 1944, an airborne concentration of 8.1 x 10°" uCi
/mL is indicated. This is almost three times larger than the tolerance level of 3 x 10°" Ci /mL in
effect at the time. If Type S is assumed, the airborne concentration is expected to be 2 orders of
magnitude larger than the tolerance level.

Jn amemorandum dated May 11, 1945, it was reported that the findings appeared to be above control
levels, which resulted in a more rigid technique of collection and system of collecting samples being
implemented on April 25 (Wirth 1945b) . Results in the first 2 weeks of this new collection system
were not substantially different from previous results, but samples in the subsequent, final 4 weeks of



AECU-817 93

Material Treatment

Plastics (lucite, Some of the equipment is discarded, smooth surfaces wiped with acetone
tygon, neoprene, and emery cloth
and rubber)

Concrete Stiff brush using Bon-Ami paste, soap and water NagCOg, highly con-
taminated areas have to be chipped and removed, citric acid, trisodium
phosphate

Wood Sanding or planing, discarding wood materials

Tools and machine Boiled in 10% citric acid and 5% Zepco powder, scrubbed and rinsed to

parts > 50 mr/hr prevent corrosion and remove citrate film

Tools and machine Soaked in 10% citric acid and 5% Zepco, kept below the boiling point,

parts < 50 mr/hr scrubbed, and rinsed

Large machines and Smeared with a paste of 1 gal flake NaOH, 1 qt starch, 1 pt Zepco soap,

vehicles > 50 mr/hr allowed to soak, and removed with hot water under pressure

Large machines and Scrubbed with 2 parts citric acid, 1 part Zepco, made into a lather

vehicles < 50 mr/hr

Coated surfaces Treatment

Ucilon paint Decontaminated with water, trisodium phosphate, 5% ammonium citrate,
2% ammonium bifluoride, or remove Ucilon paint with organic solvent

Liquid envelope Stripable coating sprayed on Ucion paint; 10% nitric, 5% ammonium
citrate. If necessary, the coating can be stripped from material

Prufcoat paint 2% ammonium citrate

Waxed linoleum Apply organic solvent gently with cloth, avoiding the spread of contami-
nation

Asphalt tile Apply organic solvent; replace highly contaminated section of the tile

Appendix B: Hand Decontamination Procedure For Health Physics Surveyors
Procedure.

1, Obtain hand count on Hand and Foot Counter. If possible, locate specific hot spots with probe,

2, Wash thoroughly with soap and water, using a surgical scrub, Repeat Step 1. If tolerance is
exceeded, proceed with Step 3.

3. Pour a liberal quantity of cleanser on the hands, brush, and follow with soap and water rinse.
Repeat Step 1. Proceed with Step 4 if results remain unchanged.

4. Moisten hands with water, apply citric acid crystals and rub for several minutes, then rinse
and follow with a soap and water wash. Check results, then proceed with Step 5, if necessary.

5, Apply KMnO, crystals on moistened hands. Rub into skin thoroughly, Rinse the hands, Shake
citric acid crystals into the palm and mix well with the KMn0Oq. If hands become warm, immediately
rinse in tepid water followed by a soap scrub and water rinse. The KMnQgq stain can be removed by
NaHSOg.

6, 5 all the above methods fail after a fair trial of each, send the employee to the Health Di-
vision,

7, Lanolin is to be spread upon the hands after decontamination to soften the hands and prevent
chapping.

Suggestions.

1, Avoid prolonged use of any one method, Irritation of the skin might impede the success of
more suitable procedures. No one chemical treatment is known to be specific for all the elements
with which one may become contaminated.

2, Do not use organic solvents. These may make the skin more permeable to contamination.

8. Oxalic acid is poisonous, It is not to be used under any circumstances.



RE Blac ae

[Document No. ORAUT-TKBS-0012-5 | Revision No. 01 __| Effective Date: 10/01/2007 | Page 590f74 |

ATTACHMENT A
CALCULATION OF MDAS FROM HISTORICAL DATA
Page 25 of 36
EH ———
Ec ane
= @* 50th Percentile '
1E+3 i
Ele
H
3 LEHI
=
3
1 E+
1E-1 |

1945, 1950 1955 1960 1965 1970 1975 1980
Year

Visual Comparison of Data to Estimated Rare Earth MDAs in Urine

1990

Figure A-11. MDA data for rare earths in urine (part 1 of 3).

Annual Average MDAs for Rare Earths in Urine
Calculated from Recovered Data Between 1945 and 1990

350

8
|
|
|
i
i

8

|

MDA (dpm/24h)

1955

Figure A-11. MDA data for rare earths in urine (part 2 of 3).



——

Special Exposure Cohort Petition U.S. Department of Health and Human Services
under the Energy Employees Occupational Centers for Disease Control and Prevention
Uliness Compensation Act National Institute for Occupational Safety and Health

OMB Number: 0920-0639 Expires: 09/20/2013
Special Exposure Cohort Petition — Form B Page 7 of 7

Public Burden Statement

Public reporting burden for this collection of information is estimated to average 5 hours per response,
including time for reviewing instructions, gathering the information needed, and completing the form. If you
have any comments regarding the burden estimate or any other aspect of this collection of information,
including suggestions for reducing this burden, send them to CDC Reports Clearance Officer, 1600 Clifton
Road, MS-E-11, Atlanta GA, 30333; ATTN:PRA 0920-0639. Do not send the completed petition form to this
address. Completed petitions are to be submitted to NIOSH at the address provided in these instructions.
Persons are not required to respond to the information collected on this form unless it displays a currently
valid OMB number.

Privacy Act Advisement

In accordance with the Privacy Act of 1974, as amended (5 U.S.C. § 552a), you are hereby notified of the
following:

The Energy Employees Occupational Illness Compensation Program Act (42 U.S.C. §§ 7384-7385)
{EEOICPA) authorizes the President to designate additional classes of employees to be included in the
Special Exposure Cohort (SEC). EEOICPA authorizes HHS to implement its responsibilities with the
assistance of the National Institute for Occupational Safety (NIOSH), an Institute of the Centers for Disease
Control and Prevention. Information obtained by NIOSH in connection with petitions for including additional
classes of employees in the SEC will be used to evaluate the petition and report findings to the Advisory
Board on Radiation and Worker Health and HHS.

Records containing identifiable information become part of an existing NIOSH system of records under the
Privacy Act, 09-20-147 “Occupational Health Epidemiological Studies and EEOICPA Program Records.
HHS/CDC/NIOSH.” These records are treated in a confidential manner, unless otherwise compelled by law.
Disclosures that NIOSH may need to make for the processing of your petition or other purposes are listed
betow. .

NIOSH may need to disclose personal identifying information to: (a) the Department of Energy, other federal
agencies, other government or private entities and to private sector employers to permit these entities to
tetrieve records required by NIOSH; (b) identified witnesses as designated by NIOSH so that these
individuals can provide information to assist with the evaluation of SEC petitions; (c) contractors assisting
NIOSH; (d) collaborating researchers, under certain limited circumstances to conduct further investigations;
(e) Federal, state and local agencies for law enforcement purposes; and (f) a Member of Congress or a
Congressional staff member in response to a verified inquiry.

This notice applies to all forms and informational requests that you may receive from NIOSH in connection
with the evaluation of an SEC petition.

Use of the NIOSH petition forms (A and B) is voluntary but your provision of information required by these
forms is mandatory for the consideration of a petition, as specified under 42 CFR Part 83. Petitions that fail to
provide required information may not be considered by HHS.

Name or Social Security Number of First Petitioner:



70 AECU-817

It ig assumed that the lowest reading in a pair of pocket meters worn together is the most nearly
correct reading (an understanding of the ionization chamber —minometer proves this to be true). A
significant reading has been defined as the lowest reading in a pair, provided that it is not off scale,
It follows then that an insignificant reading has no meaning. There are four types of insignificant
readings, To simplify record keeping they have been symbolized as follows:

Number Symbol Meaning
1 oO Off-scale meter
pig DM Damaged meter
wm LM Lost meter
Vv cM Cap missing

When an insignificant reading is noted the meter involved is immediately checked for contami-
nation leakage or damage. A reading is considered insignificant where both entries in a pair are
insignificant. Similarly, we have significant and insignificant readings on film meters, Symbols
have not been developed to describe film-meter insignificants, and when they occur the condition
is noted by a statement,

6,2 EXPOSURE REPORTING AND PERSONNEL MONITORING RECORDS
(Demonstration of Personnel Monitoring Techniques)
By J. C. Hart*

Minometry

Reading. Two hands working in rhythm manipulating the controls on the face of the minometer
have been known to cause incomplete charging of pocket meters. For this reason the technician is
trained to use one hand at the control panel. The charging button being strategically located at an
appropriate distance from the minometer well makes it impossible to effect the above undesirable
conditions if one hand is used. This is usually the right hand. The left hand serves the purpose of
holding the pocket meter cap, a desirable feature in that there is no lost motion such as would be
encountered in searching for a cap deposited on the table, In addition, there is less likelihood of the
cap picking up dust contamination which in turn would cause leakage across the electrodes of the
pocket meter.

One recognizes incomplete charging of meters when, in a series of readings, there is a high
frequency in the off-scale region between 110 and 130 and between 40 to 80, Readings in the off-
scale region can be traced to (1) mechanical discharge, (2) loose connections in the minometer well,
and/or (3) leakage. Readings in the other two regions when due to incomplete charging are due to
(1) a loose minometer well and/or (2) the pocket meter being inserted in the minometer with removal
from the well before the charging button is depressed, The latter is the most commonly encountered
event when both hands are used at the control panel.

Testing. Although the vast majority of readings are legitimate (1 bogus reading in 500), all
pocket meters which read 20 or more are treated as ‘‘suspects.’’ Suspects fall into three classes:
(1) the legitimate reading, (2) a ‘leaker,’ and (3) loose connections, All three classes are cata-
logued, and the good meters are put back into service after the following tests are performed:

1, Vibration Test. When a meter gets loose parts or the graphite conducting liner begins to
slough off, leakage across the elgctrodes results, An ordinary commercial Boerner shaker has
been remodeled for the purpose of giving the meters a vibration test. When they become ‘‘suspects””
they are charged, put on a Boerner shaker for 1 min, and then read. Meters reading 20 mr or less
are set aside for leakage tests, Meters reading 25 mr or more are sent to repairs,

*Notes taken by C. Scott; edited by J. C. Hart



AECU-817

INCIDENT
PHOTONS

ELEGTRON LOST (IONIZATION)

ATOMIC AND NUCLEAR

MAGNITUDES
_~ EMITTED PHOTON, RETURN OF ELECTRON
LT FROM EXGITED TO GROUND STATE
evectron ~~ YY
TO ORBIT OF
EXCITED STATE Fig. 2.10

z
é
-
S
a
€
£
gl-- a
4 APPROXIMATE SHAPE
2 PACKING FRACTION or
a Z \ CURVE
ATOMIC MASS
Fig. 2.11 Fig. 2.12
o(4)
Fig. 2.13
Ino1uM
FISSION
|
SECTION RESONANCE
o \ | \CPEAKS FOR CERTAIN ENERGIES
Iv 238
<4 CAPTURE
> i
I
1 | ws ey OTHERS
se Q
a LT ~ ev NENT

ENERGY OF NEUTRONS

Fig. 2.14

15



[Document No. ORAUT-OTIB-0018 | Revision No.01 | Effective Date: 08/09/2005 | Page4of35_|

Table

24
1
42
43
TA

7-2

LIST OF TABLES

Limiting air concentrations for selected particulates from 1953 to present
Default limiting air concentrations............
Intakes of contaminants in recycled uranium
Inventory of radionuclides and absorption type:
Whole-body count (WBC) results below which this TIB method is still
considered an overestimate
Urinary excretion results below which this TIB method is still considered
an overestimate..

LIST OF FIGURES

Y-12 versus OTIB-18 uranium (type M)
Y-12 versus OTIB-18 uranium (type S).
Rocky Flats Pu expected urine versus actual results
Hanford Pu expected urine versus actual results..
Hanford Sr-90 expected urine versus actual results.



-U
1 AECU-817

experimental research on radiation damage to various tissues, organs, and functions of the body as
observed with laboratory animals, etc.

6. Correiation with industrial health and hygiene information, sanitary engineering as it ties in
with radioactive waste disposal, more sophisticated and detailed shielding studies and planning, and
many others are further topics for advanced study in the development of adequate qualifications as a
health physicist.

While much more might have been included even in the scope of this introductory course, such
as details of the characteristics of various specific isotopes, etc., it was felt that the program, as
it has been given, represented the optimum coverage for the time available.

The full measure of your achievements in this course can only be gauged in the progress you
make in the months to come and in the improvements you show as you fulfill our imposed obligation
to reduce radiation-caused injury to the individual and to the race to a mimimum.

END OF DOCUMENT



AECU-817

(2) White Oak Dam; daily sampling, e.g., floods; and more frequent series when needed,
No exact limit in curies is prescribed. Permissible discharge complex and related to
flows in Clinch River

(3) Special and periodic sampling: down Clinch River monthly for water, muds, algae, etc;
White Oak Creek and Lake muds monthly with complete survey annually. Examinations
of algae, fish, etc., from creek or lake as indicated (for reconcentration of activity)

b. Atmospheric, routine monitoring: continuous recording monitors (3 outside and some 10
to 12 inside various buildings). Number of outdoor stations to be increased (7 + new). Rou-
tine observation of weather instruments. Reports include major factors, e.g., wind, inver-
sions, rainfall, etc.

B. Waste-disposal studies program (Waste-disposal Section)
1, Program of research and fundamental studies

a. Survey studies concerning conditions that exist and behavior of radioactive materials in
water, soil, air, plants, and animais (sampling and analysis activities)

b, Laboratory studies of treatment processes (sanitary engineering approach, i.e., health
protection) include water-treatment processes, sewage and industrial-waste treatment
processes, and other related methods of treatment or disposal.

6.5 DECONTAMINATION LAUNDRY
By W. H. Ray*

Laundering of contaminated clothing started during the war. Originally protective clothing was
provided for maintenance men working on contaminated equipment. The clothing was then to be dis-
carded. As a result of shortages during the war, clothing was difficult to replace; therefore launder -
ing was attempted,

Paul Fisenacher, a trainee, began experimenting tn a beaker with decontamination solutions. He
found in the Physics-Chemistry Handbook that most of the acetates of the rare earths were soluble.
‘Thus, we started using acetic acid for decontaminating clothing. The clothes were washed several
times and then rinsed several times before being finally released.

The next problem was to set up a regular routine for inspection of garments with Geiger count-
ers. We were concerned only with reducing the activity of garments to such an extent that they would
not be harmful, It was later that we worried about the intake of alpha into the body.

Acetic acid proved to be irritating, and the odor was objectionable, It subsequently was found
that citric acid was as good or better and would also work with Pu, In addition, the garments lasted
longer. We are still using citric acid solutions.

The concentration of solution used and the number of washes are determined by the degree of
contamination on the clothing.

If contaminated garments are just dumped together with relatively clean ones, there results a
transfer of activity from one to another. Some materials have dyestuffs that have an affinity for
contaminants, A very contaminated white laboratory coat will come out of a wash much cleaner than
a virgin pair of coveralls. Cross contamination is also possible among garments of the same color.
To minimize this transfer, it is desirable to segregate the clothes inte categories —cool, warm, and
hot.

It is not reasonable to launder garments that are not contaminated in the same laundry where
contaminated clothes are washed, It is better to have separate laundries, Contaminated garments
should not be sent to public laundries.

The life of a garment should also be taken into consideration in laundering. One washing does
not necessarily decontaminate a garment, Strong solutions shrink the material and shorten the life
of garments.

Is it worth while to launder garments? It is a question of economies, and it is quite possible
that with careful study given to the way garments are used so that they would be only contaminated

*Notes taken by C. Scott; edited by W. H. Ray.

7


18 AECU-817

of this missing chain have been demonstrated, In addition, the chains have been extended by trans-
mutation to elements of higher atomic number than the 92 maximum found in nature. In Table 2.2
the three chains found in nature plus a new one called the “‘Neptunium Chain,” the missing (4n + 1)
series,

Table 2.2
Chain Name Prime ancestor Terminal isotope
4n+0 Thorium goTh252 g2Pb208
4net Missing g3Bi209
4n+2 Uranium 920238 ggPb206
4n +3 Actinium 92235 g2Pb207

2.6 HOW TO COMPUTE HALF-LIFE AND HALF-THICKNESS
By T. H. J. Burnett

«cHalf-life"’ is the length of time (in seconds, hours, days, or years) required for one-half the
amount of radioactive material being considered to disintegrate (or ‘“‘decay’’}. This is a constant for
each specific isotope, and this decay rate is unaffected by the amount of isotope present or by any
other factor.

«Half-thickness’’ is the thickness of some materia] which will (by interaction with the radiation)
reduce the intensity of incident radiation to one-half.

Mathematical relations which enable us to compute these values are formulas:

I = IgeTM* for half-life and decay calculations
2» =decay constant (given in units of reciprocal time, as ‘‘per hour,” etc.)
t = time elapsed from initial time at which
Ig = original intensity

1 = Ige"* for half-thickness and absorption calculations
4 =absorption coefficient (given in units of reciprocal distance, as ‘‘per centimeter,’’ etc.)
x = thickness of material which will cut down
Ip = incident intensity
I (in both cases) = resultant activity or intensity

This type of mathematical relation is catled a ‘‘function’”’ where for specific A or the intensi-
ties are a function of time t or thickness x. A function may be defined by a formula, by a table of
values, by a graph, etc. .

A graphical method of solving these types of problems was presented in principle and by exam~
ples,

Brief derivations of these relationships might be as follows:

Consider \ as the probability of disintegration in unit time of the atoms present where N is the
number of such atoms,

Then AN is the fraction decaying per unit time, which may also be considered as the rate of de-
crease of N, or dN/dt, per unit time.

Therefore

a differential equation of which a solution is