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'Hoy 2 9 1966
DOSIMETRY, SHIELDING AND SCATTERING II*
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J. H. Thorngate
Health Physics Division
Oak Ridge National Laboratory
Oak Ridge, Tennessee
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RMLEASED FOR ANNOUNCEMENT
IN NUCLEAR SCIENCE ABSTRACTS
Introduction
The preceding paper() gave a summary of dosimetry and some of the
units used. It also gave an insight into what must be considered the
central problem of dosimetry and radiation protection, that is, the
correlation of observed biological effects with measured quantities.
Surprising as it may seem, orly recently has this begun to receive its
proper share of study by health physicists. Because some of the current
approaches for the solution of this problem have already been outlined,
this discussion will cover other work, both past and present, being done
in dosimetry. Some is related closely to basic radiation protection, and
other is more typical of what research in dosimetry has generally been in
the past, that is, research and development of dosimeters and related
dosimetric systems. Typical research problems encountered are nuclear
accidents, radiobiological and radiochemical experiments, and experiments
to measure depth dose: a term used to describe dose as a function of
position in an animal or its phantom equivalent.
Electroscopes and ion chambers (4) were the first instruments used to
measure ionizing radiation quantitatively. These irstruments were sufficient
*Research sponsored by the U. S. Atomic Energy Commission under contract
with the Union Carbide Corporation.
LEGAL NOTICE
This report me propered as an account of Government sponsored work. Neither the United
A. Makes any warranty or representation, expressed or implied, with respect to the accu.
racy, completeness, or usehuness of the information contained in this report, or that the use
of any information, apparatus, method, or process disclosed in this report may not infringe
printoly owned righto; or
B. Asnimos any liabilities with respect to the use of, or for damages resulting from the
use of any information, apparatus, method, or process disclosed in this report.
As used in the abovo, "person acting on behalf of the Commission" Includes way om-
ployee or contractor of the Commission, or employee of such contractor, to the oxtent that
auch employee or contractor of the Commission, or employee of such contractor prepares,
disseminates, or provides accon to, any tuformation pursuant to hio employment or contract
with the Commission, or his employment with such contractor.
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when the sole radiation to be measured was x rays or radium gamma rays.
In 1942, these needs were expanded drastically hy the advent of nuclear
reactors which are abundant sources of neutrons. Eariy solutions to this
added complexity in the problem were simple and straightforward. Ion
chambers or electroscopes were lined with some material which would yield
a charged particle when irradiated with neutrons. Typical approaches were
to line the detectors with boron or a hydrocarbon to provide a supply of
charged particles. Natural boron contained enough 108 to provide reasonable
sensitivity for the detection of neutrons by the (nga) reaction. While
this approach still finds some practical applications, such as reactor
control, it suffers as a dosimetry system because it merely detects
neutrons. Only if the energy spectrum of the neutrons is known can dose
information be obtained. In addition, nothing had been done, or could
be done, to reduce the inherent sensitivity of these devices for gamma rays.
Thus, two units were always operated, one with and one without the neutron-
sensitive coating.
Because it was necessary to know the neutron energy spectrum to
determine neutron dose, it became clear that a spectrometer was desirable..
The means had been available since before the war in the form of the nuclear
emulsions developed for cosmic-ray studies. Recoil protons from the
hydrogenous film backing form tracks in the emulsions which can be correlated
with the incident neutron spectrum and used for neutron dos imetry: More-
over, the tracks could be separated from the general clouding of the
emulsion caused by, gamma rays. That this system is successful is illustrated
by the fact that it is still widely used for personnel monitoring, and
indeed a nuclear track film is a component in the film badges worn by
Oak Ridge employees. (3)
Regardless of how successful this approach was for personnel moni-
toring, it has limitations, and the quest for other systems continued.
One that measured spectrum, in a crude way, was developed. This was
the threshold detector syster in which foils of different elements were
used. . The most widely used system has six foils: gold, gold enclosed
in cadmium, plutonium-239 enclosed in boron-10, neptunium-237, uranium-238,
and sulfur, (4) The first figure shows the reason for this selection.
This graph represents the probability of a neutron producing an inter-
action of interest as a function of neutron energy. Gold is activated to
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produce radioactive gold-198; the sulfur undergoes an (n,p) reaction and
produces phosphorous-32, a beta-ray emitter. The other elements undergo
fission, producing radioactive fission products. In each case, the
radiations can be measured with a scintillation detector, and, as the
curve shows, this activity is related to the number of neutrons above a
certain threshold level. (5) Thus, when enclosed in the proper thickness
of 10B, 239 pu interacts in a fairly uniform manner with all neutrons
that have an energy, greater than 1 kev. Neptunium-239 reacts with those
above 750 kev, 2380 with those above 1.5 Mev, and 32s with those above
2.5 Mev. By subtracting the number of neutrons above 750 kev from those
above 1 kev, those in the range from 1 kev to 750 kev are determined.
Cimilar calculations will produce the number of neutrons between 750 kev
and 1.5 Mev, 1.5 Mev and 2.5 Mev, and above 2.5 Mev. Multiplying these
fluxes by the average dose per neutron for the related energy increments
and summing provides the required fast neutron dose measurement. Gold
is used to determine the number of neutrons with thermal velocities. This
system is widely used as a nuclear accident dosimeter, but it has short-
comings. Because the thresholds are not sharp, the sensitivity of the
foils in these regions depends upon neutron spectrue. The average cross
sections and the average dose per neutron are also spectrum dependent.
Thus, one has a device, the calibration of which depends on spectíum,
being used for spectrum measurements. As a spectrometer, this is a
severe weakness; as a dosimeter, some of the effects tend to cancel su
that specific knowledge of the shape of the spectrum, before measuring
it, is not required. As a result of the energy increments used, this
instrument was designed to be most useful for measuring doses from normal
fission spectra. Spectra with a predominance of neutrons below 0.5 Mev
are difficult to evaluate with this system as are spectra with large
numbers of neutrons greater than 2.5 Mev. The second problem may be
alleviated somewhat by using additional foils with higher energy thresholds.
This system is insensitive. Evaluation of long exposures is complicated
by the complex decay of the fission products occurring concurrently with
irradiation. A related problem arises from the need to count the fissile
foils within a short time after exposure due to their rapid decay. Prompt
recovery is not always feasible, such as after a nuclear accident or during
a nuclear weapons test. Finally, some rather expensive foil counting
equipment is required. The fission foil system described is shown in the
second figure. A variation using thin targets of the fissionable material
and a silicon surface barrier diode to detect the fission fragments is
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shown in the third figure."
This arrangement has some advantages over
the basic foil system in that it may be used to measure lower levels and
lengthy runs. In fact, this equipment was used to provide an accurato
calibration of the standard foil counters.
This system requires either
a considerable amount of electronic equipment, if all three fissile foils
are to be operated simultaneously, or a great deal of exposure time. A
more recent variation is given in Fig. 4.M Thin fissile foils are used,
and the fission fragments are detected by the tracks they produce in a
piece of glass or plastic. This detection system was developed by the
people at Jülich. The advantages make it a quiet revolution in nuclear
accident dosimetry. Orly small amounts of fissile material are needed,
but the good geometry for detection increases the over-all sensitivity.
No expensive apparatus is required to read the glass, for when it is
etched, the tracks become readily visible with an ordinary microscope.
What a glass looks like after irradiation and etching with hydrofluoric
acid is shown in Fig. 5. Early recovery is no longer important. Klaus
Becker of Jülich estimates the half-life of the unetched tracks to be
similar to the age of the earth. When used for nuclear accident dosimetry,
it can be recovered at any time after an accident and sent elsewhere for
interpretation. In fact, this system is so appealing that we have begun
studies to determine if a similar approach could be used to replace the
nuclear track film in the personnel monitoring system. In certain plastics,
it is possible to detect tracks left by a particles of particular energy.
This increases the possibility that a suitable personnel monitoring system
may be designed around solid-state detectors.
TIL
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Another solution to the problems of neutron dosimetry, developed in
the early fifties, was the polyethylene-lined, ethylene-filled proportional
counter. As shown in Fig. 6, Hurst found that the ratio of energy deposited
per gram by fast neutrons in polyethylene to that deposited per gram in
tissue remained essentially constant over a wide range of neutron energies. (*)
A cross-sectional drawing of the counter is given in Fig. 7. Thus, if the
energy deposited in polyethylene could be measured, a simple factor would
convert it to the amount of energy wnich would be deposited in tissue
irraciated ulder similar circumstances. Bragg and Gray had developed a
theory that the energy deposited in a small void within an equilibrium
thickness of material would be related to that deposited in the material. (0)
he material (10)
In addition, if the void were filled with a gas with the same composition
as the material, the void could be any size. Ethylene and polyethylene
(C,H,) fulfill these requirements. Both are hydrogenous materials so that,
in a neutron fluence, recoil protons will be produced equally in each for
a unit mass. The protons lose energy by ionizing atoms along a fairly
straight path. For that part of the path in the gas, an applied electric
field separates the electrons and ions and accelerates the electrons to
the anode. Near the anode, the electric field is intense and the electrons
obtain enough energy from it to produce additional ionizations. The total
number of ions formed is proportional to the initial number of electrons
produced along the proton track which is, in turn, proportional to the
amount of energy lost by the proton. Thus, the work done by the electric
field in moving the positive ions out of the region near the anode is
proportional to the energy deposited by the proton in the gas. A voltage
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pulse developed across the capacitance to the anode is related to the work
done by the electric field. By integrating the number and size of these
pulses, the energy deposited by the recoil protons, and thus the incident
neutrons, ca
be measured. An electron will lose much less energy in
traversing the counter than a proton, so pulses produced by gamma rays
will be smaller than those produced by neutrons. Rejecting small puises
limits the number of gamma-ray interactions recorded. Unfortunately, this
also affects the minimum neutron energy which can be measured, a fact of
varying importance depending upon the neutron spectrum. (14) When the actual
amount of gas present in the sensitive volume of the counter is known,
absolute measurements of energy deposited per gram may be made. Energy
calibrations are simplified by an internal a source of known energy since
the a particle will be stopped before it has traversed the sensitive volume.
This detector has found wide application where careful measurements of low
dose rates are required, although it is not suited to measurements of high
dose rates or low energy neutrons. Because absolute measurements may be
made with it, it has been used to determine the average dose per neutron
from a number of commonly used neutron sources. (12) A smaller counter
operating on the same principle was built for measurements in a tissue-
equivalent phantom. These measurements are of considerable importance to
determine how neutron dose builds up and is attenuated in animals and
humans. A computational program has been conducted as well to determine
the neutron dose as a function of position in a man-sized tissue-equivalent
phantom. (13)
Hurst and Ritchie have developed a generalized concept of radiation
dosimetry that describes a system of operators that makes it possible to
use any detector system which has an energy-dependent response to measure
any type of dose. (14An application of this principle was used to make
neutron dose measurements with the simple apparatus shown in Fig. 8. (15)
Measurements of average dose per neutron for several sources were made
which agreed well with those made with the proportional counter. This
particular system is not convenient for routine measurements but served
to illustrate the generalized concept of dosimetry.
Finally, a system of neutron dosimetry which deserves mention, primarily
due to its popularity, involves the use of a thermal neutron detector within
a sphere of moderating material. Measurements with it are expressed in a
dose unit theoretically related to the biological effectiveness of the
radiation. This system will measure over the entire energy region of
thermal, intermediate, and fast neutrons but generally overestimates the
dose in the intermediate region. A variation on this technique has been
used by Japanese investigators to determine the neutron dose versus distance
from the nuclear detonations at Hiroshima and Nagasaki. They measured 60 Co
activity in iron, such as reinforcing rods, buried in concrete.
Until good sys:: ms of neutron dosimetry were developed, it was not
evident that gamma-ray dosimetry was perhaps the most difficult of the two,
particularly for coexistent neutron and gamma-radiation fields. The
problem can be stated simply. Almost all elements produce charged particles
or gamma rays when irradiated with fast neutrons. Thus, gamma-ray measure-
ments in a mixed field are complicated by the contribution caused by neutrons.
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Ion chambers and electroscopes built with hydrogenous materials were
recognized early as being poorly suited for gamma measurements in mixed
fields. Carbon ion chambers filled with co, gas were tried and considered
adequate until carbon recoils were recognized as a severe problem. Thus,
a solution evolved based upon the premise that if neutron interactions
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cannot be eliminated, perhaps a system can be devised which minimizes
their contribution to the measured dose. Such a system is the Singie Ion
Detector shown in Fig. 9.'10) Basically, this is a small proportional
counter with a low pressure filling, which is made of materials known to
have a low probability of undergoing neutron interactions. Its dimensions
are such that when a gamma ray produces an electron in the sensitive
volume, the ion path produced by slowing down of the electron is only
one ion long. This produces a pulse of a particular small size from the
detector. A large charged particle produced by a neutron interaction will
leave more energy in the chander but will also pr 'uce only a single pulse.
Thus, while more dose was deposited, its contribution had been minimized.
This is a difficult device to operate and use, but it led directly to what
is probably the simplest device possible. As shown in Fig. 10, this is
a small geiger counter. Its small size reduces the probability of unwanted
neutron interactions, and, like the SID, it produces only a single output
pulse, of uniform size, for every charged particle reaching its sensitive
volume. What's more, the halogen-quenched tube used is stable and produces
large output pulses. A graded shield encloses the detector to decrease the
response at low energy. Any gamma-ray detector composed of high z materials
will have an increased response at low energy due to the increase in the
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probability of a photoelectric interaction. The shield makes the detector
response uniform from 100 key to over 8 Mev. (10. Because it is more
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sensitive to thermal neutrons than desirable, a bli shield is used when
measurements in a mixed radiation field are made. The small size of this
detector make it, like the small neutron proportional counter, admirably
suited for depth-dose measurements.
Other properties of materials have been exploited for use in gamma-ray
dosimetry, One which has found wide use is radicphotoluminescence. (18) In
this case, a material is put into a metastable state by the gamma rays and
then returned to the ground state by irradiation with ultraviolet light,
with the simultaneous emission of energy. Silver-activated metaphosphate
glass is the medium most often used for this type of measurement. These
detectors can be quite small, such as the l-mm-diameter-by-6-mm-length
rods used as accident dosimeters in the film badge unit. This type of
dosimeter, in other physical configurations, can be reasonably sensitive
and also capable of high dose readings. However, as gamma-ray sensitivity
increases, so does neutron sensitivity, mostly as response to thermal neutrons.
This is generally minimized by shielding with 6Li. The increased response
at low energies is usually compensated by a graded tin and lead shield
similar to that used with the geiger tube.
Another approach, to be mentioned only briefly, is thermoluminescence.
This system is not basically different from glass except the luminescence is
obtained by heating the sample. This appears to yield greater sensitivities
than glass but has an uncertain response to neutrons. For some applications,
the fact that readout is destructive can be a severe handicap. Until lately,
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thermoluminescent materials were also generally more difficult to handle
than glass, but considerable improvements in packaging have occurred. Lif
and Teflon can be compressed to give rigid, easy-to-handle detectors. It
is interesting to note that we use 6Lif-Teflon containers to shield glass
from excessive thermal neutrons.
An interesting application of thermoluminescence has been made by
Japanese investigators. 120) Using the nonferrous portions of ground-up
roof tiles, they have been able to determine the gamma doses as a function
of distance from the nuclear detonations at Hiroshima and Nagasaki. Their
results are in good agreement with similar data obtained by other methods.
One final note before we leave the subject of gamma dosimetry; the
generalized concept of dosimetry is just as applicable to gamma dosimetry
as to neutron dosimetry. One of the first tests of the principle suggested
was to use a scintillation datector for dose measurements.
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spectrometry. Measurements of the energy spectrum of the incident radiation
can be used to calculate almost any dose desired, e.g., depth dose, kerma,
etc. for several years, the fluid state of dose units and terminology made
it appear that perhaps spectrometry would be the only satisfactory solution
to the dosimetry problem. Fortunately, this situation has eased somewhat,
fortunate not only because there is considerably less confusion, but also
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because spectrometry is a difficult problem, and to supply a practical,
easily operated device for field use seems beyond the limits of present
technology
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Even limited in its use, spectrometry has been particularly useful in
solving dosimetry problems associated with specific experiments. In gen-
eral, neutron spectrum measurements have been important due to the varicus
energy-dependent effects of the neutron dosimeters available. Some of the
spectrometers used were developed to solve specific dosimetry problems,
others were borrowed from the nuclear spectroscopists. An example of the
first type is shown in Fig. 11. This device was designed more for
sensitivity than resolution. (21) It is a thick radiator, recoil proton
telescope that was built originally for use during a field experiment at
the Nevada Test Site. It will be used again this fall for another
experiment at the NTS. Neutrons are collimated so that they impinge
normally upon the first organic phosphor. Recoil protons are produced
with those scattered in the near forward direction traversing the evacuated
space and being detected in the second phosphor. When a coincidence is
obtained and identified as a proton by a special pulse shape selection
circuit, the signals from the detectors are summed and analyzed. Thus,
information about the e? rgy deposited by the proton in leaving the radiator
is not lost. As the range of the protons increases with energy, so does
the useful radiator thickness and thus efficiency of this spectrometer, a
rather unusual situation. Unfortunately, the recorded pulse-height spectrum
has little in common with the neutron spectrum due to the shape of the
efficiency curve and the nonlinear light output versus proton energy of
the phosphors.
For gamma rays, spectrometry in mixed radiation fields has proven
especially difficult. A spectrometer was built for use during a field
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experiment in 1962 which operated by magnetically analyzing the Compton
recoil electrons produced in a beryllium foil by the incident gamma
spectrum. (22) It is shown in Fig. 12. While this yielded low neutron
sensitivity, it also gave low gamma efficiency so little useful data
were obtained during this exper iment. This approach has warranted further
investigation and development due to its wide energy range and freedom
from neutron response. It shows considerable promise for high resolution
measurements.
For a pure gamma-ray field, standard scintillation detectors have
proven useful although they produce a complex response to a given gamma-
ray energy. During the field experiment previously mentioned, the gamma
spectra as a function of time after a burst of neutrons and as a function
of angle from a large 60 Сo source were measured. The next figure shows
one of the curves from the 60Co measurements and illustrates the complexity
of data from a scintillation detector.
More sophisticated scintillation spectrometers utilizing the pair
production interaction of high-energy, gamma rays or Compton recoil reactions
for wide energy ranges are also employed. As an example, the Compton
spectrometer configuration is shown diagramatically in Fig. 14.(25)
Because both systems operate with some sort of coincidence analysis,
neutron response is low. In fact, a PuBe neutron source, which is an abundant
source of 4.43-Mev. gamma rays, is commonly used to set up the pair
spectrometer.
Spectrum measurements are not limited to the incident spectrum of
neutrons and gamma rays. A good example of other measurements made in
dosimetry experimentation is shown in Fig. 15. This is the spectrum of
recoil particles produced by 15 Mev neutrons as a function of position
in a tissue-equivalent phantom. (24) The spectrometer in this case was a
lithium-drifted silicon diode in the form of a cube 1.5 mm on a side.
While most of the spectrum is due to protons, the increase in pulses at
lower energies is due to heavy ion recoils. These data are quite recent,
having been taken by a graduate student at our facility during the spring
and early summer (1966).
Nuclear accident dosimetry has been an important aspect of health
physics since the first probability of an accident developed. Dosimetry
is important as a guide to the proper medical treatment for survivors.
However, we tend to sift, vulture like, over all of the data we can obtain
from the accidents which do occur in an effort to increase the store of
human data dealing with radiation damage. As understanding of the effects
of radiation on the human system increases, more realistic, safer limits
on occupational exposures may be set. The most useful devices we have
for nuclear accident dosimetry have proven to be the threshold detector
systems for neutrons and glass for gamma rays. Considerable effort has
been put into a biological dosimeter which everyone always has with him--
his blood--through efforts to relate dose received to activation of blood
sodium. If something is known or can be assumed about the incident
neutron spectrum and the orientation of the victim, sodium activation is
a powerful dosimetric tool. Considerable laboratory and computational
work has been expended to increase our knowledge of this tool and more
is continuing. (25)
15
In 1964 we performed an experiment at Los Alamos to measure the radiation
leakage of a critical assembly designed to have leakage characteristics similar
to the Hiroshima weapon. (20) 'This input information was another basic piece
of information in evaluating the radiation doses received by the survivors
of the nuclear bombings of Hiroshima and Nagasaki. Careful medical records
have been kept concerning these people since shortly after the war which,
when coupled with good dose estimates, form the largest available pool of
human data on the effects of exposure to nuclear radiation. Work towards
determining these doses has continued since 1956 in many ways. Many experi-
ments have been performed in the laboratory and during weapons tests and
other experiments at the Nevada Test Site that have been concerned with
transport of weapons radiation through air and shielding by Japanese-style
houses. Backing up these extensive experiments have been many calculations.
As a result, an equation has heen developed which relates a survivor's
location, with all its geometric and shielding considerations, to the dose
he received.
During our 1962 field experiment, measurements were made of the effect
of the air-ground interface on the attenuation of radiation as a function of
distance from a nuclear reactor. (26) A typical set of data is given in
Fig. 16 which compares the effect of the air-ground interface with the
transmission of the radiation through an infinite media. The data presented
in Fig. 17 were also taken during this field experiment and show the gamma
dose as a function of angle at a distance of about 800 yards from the
source. (28)
A number of experiments in radiation scattering have been conducted;
for example, the neutron albedo of various materials, such as concrete,
soil, and water, were measured. (29) Theoretical work was done to relate
the measured results to known physical phenomena. A more recent experiment
provided the same sort of data for gamma rays scattered by load, water, or
concrete. Typical data for concrete and 60Co gamma rays are shown in
Fig. 18. It is worth noting that a great deal of this latter work was
done by a graduate student and a visiting professor. (30)
Obviously all of the material I have presented was not for your
edification as dosimetrists but was directed toward increasing your under-
standing of the tools and research which has been conducted in the past,
directed at solving specific problems in dosimetry. I hope that the omissions
and simplifications were not overdone. Many widely used dosimetry systems
were not mentioned because we have done little with them. For the sake of
time, most details of our extensive field experiments were omitted. While
a great deal of that discussed seems applied, such an impression is incomplete
and may have resulted from the fact that my own work has been highly applied
in the last few months as we prepare for another field operation. The
operation itself is a large-scale, basic research problem in the transport
of neutrons in air and a study of the high-energy, gamma rays produced by
neutron interactions in air. (30) Moreover, as more is learned about
dosimetry, new approaches become more basic. In short, those of us in
dosimetry find it constantly challenging and are eager to encourage others
to join us.
REFERENCES
1. W. C. Roesch, Dosimetry, Shielding and Scattering I, ORAU Conference
on Principles of Radiation Protection, August 24-26, 1966, Oak Ridge,
Tennessee.
2. There are many sources of information on these counters. For example:
B. B. Rossi and H. H. Staub, Ionisation Chambers and Counters, McGraw-
Hill Book Co., Inc., New York, 1949, and J. W. Boag in Radiation
Dosimetry, edited by G. J. Hine and G. L. Brownell, Chapt. 4, p. 153,
Academic Press, Inc., New York, 1956.
3. Film dosimetry is reviewed by R. A. Dudley in Radiation Dosimetry,
edited by G. J. Hine and G. L. Brownell, Chapt. 7, p. 299, Academic
Press, Inc., New York, 1956.
4. G. S. Hurst et al., Techniques of Measuring Neutron Spectra with Thresh-
old Detectors - Tissue Dose Determination, Rev. Sci. Instr. 27, 153-
156 (1956).
5. P. W. Reinhardt and F. J. Davis, Improvements in the Threshold Detector
Method of Fast Neutron Dosimetry, Health Phys. 1, 169-175 (1958).
6. D. R. Johnson et al., An Experimental Calibration of Fission Foil
Threshold Detectors, Health Phys. 11, 759-762 (1965).
7. G. D. Kerr and T. D. Strickler, The Application of Solid State Nuclear
Track Detectors to the Hurst Threshold Detector System, Health Phys. 12,
1141-1142 (1966).
8. K. Becker, Nuciear Track Registration in Dosimeter Glasses for Neutron
Dosimetry in Mixed Radiation Fields, Health Phys. 12, 769-785 (1966).
9. G. S. Hurst, An Absolute Tissue Dosimeter for Fast Neutrons, Brit. J.
Radiol. XXVII, 353-357 (1954)..
10. This theory is discussed by P. W. Spier in Radiation Dosimetry, edited
by G. J. Hine and G. L. Brownell, Chapt. 1, p. 23, Academic Press, Inc.,
New York, 1956.
11. R. H. Ritchie, Calculations of Energy Loss under the Bias in Fast Neutron
Dosimetry, Health Phys. 2, 73-76 (1958).
12. T. D. Jones, D. R. Johnson, and J. H. Thorngate, Neutron Dose Conversion
Factors for Pube and PoBe Sources, Health Phys. 11, 519-522 (1965).
13. J. A. Auxier et al., Health Physics Division Annual Progress Report
for Period Ending July 31, 1965, ORNL-3849, pp. 166-171.
14. G. S. Hurst and R. H. Ritchie, A Generalized Concept for Radiation
Dosimetry, Health Phys. 8, 117 (1962).
15. D. R. Davy and L. H. Peshori (To be published in Health Physics).
16.' J. A. Auxier, G. S. Hurst, and R. E. Zedler, A Single Ion Detector for
Measurement of Gamma Ray Ionization in Cavities, Health Phys. 1,
21-26 (1958).
17. E. B. Wagner and G. S. Hurst, A Geiger-Müeller y-Ray Dosimeter with
Low Neutron Sensitivity, Health Phys. 5, 20-26 (1961).
18. J. H. Thorngate and D. R. Johnson, The Response of a Neutron Insensitive
Gamma-Ray Dosimeter as a function of Photon Energy, Health Phys. 11,
133-136 (1965).
19. C. H. Bernard, W. T. Thornton, and J. A. Auxier, Silver Metaphosphate
Glass for X-Ray Measurements in Coexistent Neutron and y-Radiation
Fields, Health Phys. 4, 236-243 (1961).
20. Y. Ichikawa, T. Higashimura, and T. Sidei, Thermoluminescence Dosimetry
of Gamma Rays from Atomic Bombs in Hiroshima and Nagasaki, Health Phys.
12, 395-405 (1966).
19
21. J. H. Thorngate et al., Health Physics Division Annual Progress Report
for Period Ending July 31, 1965, ORNL-3849, p. 159.
22. J. H. Thorngate et al., Energy and Angular Distribution of Neutrons and
Gamma Rays - Operation BREN, USAEC Report CEX-62.12 (To be published).
23. J. H. Thorgate et al., Health Physics Division Annual Progress Report
for Period Ending July 31, 1965, ORNL-3849, p. 159.
24. G. P. Stone, Experimentally Determined Proton-Recoil. Spectra in Tissue-
Equivalent Material from 3 and 14 Mev Neutrons, Submitted as a Master's
Thesis to the University of Tennessee, August 1966.
25. F. W. Sanders and J. A. Auxier, Neutron Activation of Sodium in
Anthropomorphous Phantoms , Health Phys. 8, 371-379 (1962).
26. J. H. Thorngate, D. R. Johnson, and P. T. Perdue, Neutron and Gamma-Ray
Leakage from the Ichiban Critical Assembly, USAEC Report CEX-64.7
(June 1966).
27. F. F. Haywood, Spatial Dose Distribution in Air-over-Ground Geometry,
Health Phys. 11, 185-192 (1965).
28. J. H. Thorngate et al., Energy and Angular Distribution of Neutrons
and Gamma Rays - Operation BREN, USAEC Report CEX-62.12 (To be published).
29. J. W. Cure and G. S. Hurst, Fast-Neutron Scattering: A Correction for
Dosimetry, Nucleonics 12, 36-38 (1954).
30. F. F. Haywood and J. A. Auxier, Technical Concept - Operation HENRE,
USAEC Report CEX-65.02 (March 1965).
B
30.
..
.
'
A
:
20
.
.
PIGURE CAPTIONS
Figure 1. Threshold Detector System Cross Sections.
Figure 2. TDU Pission Foil Configuration.
Figure 3. A Fissile Foil with a Semiconductor Diode Fission Pragment Detector.
Figure 4. Pissile Foils with Plastic Fission Fragment Detectors.
· Figure 5. Microphotograph of Etched Glass after Pitting by Pission Fragments.
Figure 6. Ethylene and Tissue Neutron First Collision Dose vs En and the
Ratio of These Two Doses.
Figure 7. Polyethylene-lined, Cyclopropane-filled Proportional Counter for
Fast Neutron Dosimetry.
:: Figure 8. Detector and Electronic Equipment Used to Test the Utility of
the Hurst-Ritchie Generalized Concept of Dosimetry.
Figure 9. Single Ion Detector.
Figure 10. "Phil" Gamma Dosimeter and Thermal Neutron Shield.
Figure 11. Thick Radiator Telescope Neutron Spectrometer.
· Figure 12. BREN Compton Y-Ray Spectrometer.
3-In.-by-3-In. NaI(T1) Scintillation Detector.
Figure 14. NaI(TI) Compton Spectrometer.
Figure 15. Recoil Proton Spectrum Produced by 15 Mev Neutrons as a Function
of Position in a Tissue-Equivalent Phantom.
Figure 16. BREN Air-Ground Interface Data.
Figure 17. BREN y Dose vs Angle at 750 Yards from the base of the Tower.
Figure 18. Gamma-Ray Albedo of 60Co on Concrete.
ORNL-LR-DWG. 9441-AR
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END
DATE FILMED
- 12/ 21 / 66
.
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