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IN NUCLEAR SCIENCE ABSTRACTS
THE PRESENT STATUS OF NEUTRON MONITORING
FOR PERSONNEL PROTECTION*
J. A. Auxier
Health Physics Division, Oak Ridge National Laboratory
Oak Ridge, Tennessee
When science reached the threshold of the nuclear age, about 1942, the
record of injuries and deaths resulting from the misuse of radium and x rays
was a matter of concern to the scientists involved. Not only would nuclear
reactors produce millions of times more radioactivity than all of the radium
sources in the world's hospitals and laboratories, but they would yield co-
pious supplies of neutrons. Consequently, from the time of the first sus-
tained fission reaction, provision has been made at nuclear installations
for neutron dosimetry for personnel.
Twenty years ago, the chief instruments for neutron dosimetry were
nucisar emulsions and matched pairs of ionization chambers (or electro-
sacoes), one of the pair having walls and/or gaseous content such that its
ro.soonse to neutrons was greater than the other. Today, the most widely
15€ personnel dosimeter for neutrons is the nuclear emulsion, and some
casurements with pairs of ion chambers are still made, especially when
tissue-equivalent chambers are used. There are a number of papers to be
presented at this symposium which are concerned with these two devices.
*Research sponsored by the U. S. Atomic Energy Commission under contract
with the Union Carbide Corporation.
.
1
;
.
The principles of application are almost always the same as those used twenty
years ago; only some of the technical details have been improved to permit
faster, more convenient, or more precise measurements.
How, then, has the dosimetry for neutron exposures changed? Are better
systems now available? What changes are probable in the near future? Some
answers to all of these questions will be given in the course of this meeting.
However, to provide a framework for the more specific papers which are to
follow, let us look at the general development of the field during the last
twenty years.
The rapid development of scintillation systems, commencing about 1948,
promised an early and direct advancement of the field of neutron dosimetry.[1]
llowever, due to the greater sensitivity of the scintillators to gamma rays
and the lack of pulse-shape discrimination, the various systems were never
made practical and, in general, never performed satisfactorily for neutron
measurements in the laboratory. The first significant aplication of scin-
tillation systems in neutron dosimetry was in the analysis of induced or
fission-product gamma rays in activation and threshold detectors. [2,3] For
example, the use of two scintillation detectors with dimensions of 5 cm thick
by 10 cm diam. yielded sufficient sensitivity and resolution to permit much
wider and more definitive use of threshold detectors. Another significant
use of scintillation detectors near nigh-energy accelerators is that associa-
ted with the 12C(n, 2n) 11C reaction. The carbon containing crystal can be
irradiated and then transported to a counting room, placed on a photomulti-
plier, and analyzed for carbon-11. This technique is especially useful for
sharply pulsed fields.
In recent years, with the development of ultrafast electronics
(resolution of the order of 10 nsec) and pulse-shape discrimination, scin-
tillation detectors are important for neutron spectrometry (e.g., in 6Lil (Eu)
crystals). [4]
Threshold and activation detectors were used also from the beginning of
neutron dosimetry. The methods and techniques have improved, largely because
oi improved scintillation systems, and several papers on this subject will be
given here. Due to the finite slopes of the neutron cross sections of the
isotopes used in the activation and threshold detectors, they are more useful
for dosimetry than for neutron spectroscopy. A set of energy-fluence histo-
grams is depicted in Fig. 1. This is based on the ORNL threshold-detector.
system and shows only a crude approximation to the neutron spectra from the
Health Physics Research Reactor, the Y-12 accident, and the Yugoslavian
accident. If, due to variations in assumed effective cross sections or to
errors of other nature, an assignment of fluence to one energy group is in
error by as much as 25%, the error in dose will be less than 10% in most
cases. If 25% too many neutrons are assigned to the 1.5 MeV threshold group,
then too few neutrons will be assigned to the 0.75 to 1.5 MeV group, and this
will compensate partially for the error. For spectrometric purposes, the
errors are generally serious and the resolution poor. Because of the short
half-life of reaction products and/or the small cross sections of some
commoniy used detectors, the system is best suited to high intensity and
pulsed ficids. They are used more widely for nuclear accident dosimeters
than any other system. [5] Some investigators have used as many as thirty
or more detectors in a single unit to improve the resolution of the system, (6)
but little is gained by such extensive use of foils; there are few good ones,
and the use of additional poor ones can, in some analysis techniques, result
in less definitive results.
A widely used detector system is that employing a cylindrical or spheri-
cal moderator, usually paraffin or polyethylene, with a thermal neutron
detector located on the axis or at the center. [7,8] These devices are used
as both active (e.g., with lil scintillation detectors) or as passive (e.g.,
with gold foils) systems. Because the neutron diffusion length is not a
sensitive function o. energy, such systems can be made to have a nearly flat
response vs neutron energy. It is not feasible to construct a spectrometer
based on a variety of sizes of moderators, but a sufficiently good approxi-
mation can be made to estimate "dose." In general, the single sphere "REM"
detector yields a greatly enhanced response for neutrons in the energy range
from thermal to a few hundred kev. However, because of the simplicity of
the system, several laboratories are participating in developmental work
with it. One of the most widely used systems of this type was developed by
Hankins, [9] and the response of the instrument is shown in Fig. 2. In the
approximately five. Of course, many systems have little or no response in
this region (e.g., some proportional counters and all nuclear e:nulsions).
It is apparent that the neutron energy region which poses the most problems
in dosimetry is, at present, that between 1 ev and 100 keV and that several
types of detectors are required for definitive measurements.
One of the major advances in neutron dosimetry was the development of
a good proportional counter for neutron dosimetry. (10,11] This instrument,
depicted in Fig. 3, has been the basic standard at several laboratories for
nver a decade. Essentially, it is a good Bragg-Gray cavity chamber with a
built-in al ha source for standardization of gas and electronic gain, and
field tubes for accurate delineation of the electric field. Typical pulse-
height distributions are shown in Fig. 4. The upper portion shows the
distribution due to the Pu alpha particles for four different applied
potentials, and the lower portion shows a typical proton pulse spectrum for
a PoBe source. Energy resolution with this counter is typically 4% or
better. Another important application of the proportional counter is in
LET spectrometry. [12] The basic phenomena of proportional counters have
been researched extensively and the radiation physics of gases continues to
be an important field of health physics research. However, except for
improved electronics, no significant advances in proportional-counter
dosimetry have been reported recently.
Various "solid-state" detectors for neutrons have been proposed and
investigated during the last twenty years. [13] Changes in impedance and
carrier lifetimes of diodes can be used as a measure of a radiation field,
the accuracy depending on many paranieters. Some of the best detectors will
be described during this symposium. Without proton radiators, these systems
generally yield a better approximation to fluence or energy fluence than to
absorbed dose. Also, they have a low response to neutrons with energies
less than about 400 keV.
However, Davy (14) has used a combination of silicon-diode detectors,
proton radiators, and operational analyzers to develop an absolute dos imeter
based on Ilurst's generalized principle of dosimetry. (15) Figure 5 shows a
sketch of one of his detectors and a block diagram of the apparatus used
with it. This system can be calibrated for monodirectionad neutron beams
only, but the logic of the system is more general. One of the s:accessful
configurations used by Davy was that of a proton radiator (108 mg/cm2
polyethylene) and a "thin" silicon detector (63 um). The expected pioton
spectra from the radiator for various neutron energies are shown in Fig. 6.
Good agreement was obtained between the calculated and measured values. For
tliis configuration, the choice of operational parameters required to compute
dose was straight forward and excellent agreement with other accurate systems
was obtained. The importance of these experiments was the demonstration of
the principles rather than the development of a practical system. However,
this system of detectior. has been extended to depth-dose distributions in-
cluding LET. One study utilized both planar (200 um thick) and thick cubical
(1.5 x 1.5 x 1.5 mm) silicon detectors. Figure 7 shows the detector assem-
blies used in some of the most recent studies. In order to minimize the
perturbation of the proton distribution in tissue-equivalent phantoms, the
support rod for the detectors is of tissue-equivalent plastic and the protec-
tive film over the detectors was as nearly tissue equivalent as possible.
A proton energy distribution measured in a 30-cm-diam. by 60-cm-long cylinder
of tissue-equivalent solution is shown in Fig. 8. These data are for the
1.5-mm cubical detector and a 14.6-MeV source of neutrons incident laterally.
Thermoluminescent alid photoluminescent systiims have been and are used
for neutron measurements. Either system with lithium-6 incorporated in it
is, of course, a sensitive thermal detector and can be used with a moderator
as described above. The Lif thermoluminescent detector has been used in a
solution of alcohol as a neutron detector, the alcohol serving as a source of
protons. This type of detector involves subtraction of the response of the
"neut ron insensitive" gamma detector, and the accuracy is thus a sensitive
function of the ratio of neutron to gamma-ray dose. By using teflon plastic
as a carrier matrix for Lif crystals, a wide variety of shapes and sizes are
available.
Recent work by Blanc, et al. (16,17) has shown that liquid ionization
chambers may be used for coexistent neutron and gamma-ray fields. There is
promise that these chambers can be used to measure the effective quality
factor for such fields as well. Excellent papers on this subject will be
presented during this symposiwn.
Perhaps the most exciting new development ir neutron detection is the
discovery and use of the heavy ion tracks in insulating solids. [18] A
heavily ionizing particle (e.g., a fission fragment), in traversing a layer
of the solid (e.g., glass), produces a track which can be rendered visible
under the optical microscope hy etching in a suitab’e reagent (e.g., hydro-
fluoric acid). Recently, the technique has been extended at ORNL to include
particles as light as deuterons by use of suitable plastic solids. Figure 9
shows a photomicrograph of alpha particle tracks etched in plastic with NaOH.
One of the first applications of the technique was with the fission foils
in the ORNL threshold-detector system. (19} The modified fission-foil
assembly is shown in Fig. 10. Fissile materials are plated on metallic back-
ings, and assembled in the sintcrcd boron ball such that fragments from fission
hombard the surface of the plastic or glass disks. Standard microscope slide
covers make economical and excellent detectors for heavy ions. The track
technique makes possible the reduction of the fissile material from gram to
microgram quantities, increases the sensitivity, negates the need for imme-
diate analysis, and decrcuscs barkedly the amount and complexity of the
clectronic equipment required. The application of similar techniques for
personnel monitoring is being explored.
The major developments of the recent past and the pattern of present
rescarch lead me to expect that the most significant change in the field of
radiological monitoring will be the substitution of solid-state devices for
the film and nuclcar cmulsions in personal dos imcters. The scientific prob-
lems associated with the gamma detectors have been solved, and only engineer-
roider system which is largely automatic has been reported by Buttler. [20]
for tlic noutron detectors, there are further technical problems to be solved,
but the progress of the immediatc past is encouraging. Within five years, a
worker should be able to insert his "badge" in a slot nad have displayed
hefore him both neutron and gamma-ray "dose" since his last reading, plus an
indication of his integrated "dose" since he started wearing the badge (i.e.,
the "lifetime" dosimeter should be available). In the broader field of
dosimetry, the use of solid-state detectors and operational analyzers will
be used for measurements of absorbed dose, LET, and other parameters.
REFERI:NCES
.
11] HORNYAK, W. F., Rev. Sci. Instr. 23 (1952) 264.
HURST, G. S. et al., Rev. Sci. Instr. 27 3 (1956) 153.
(3) REINHARDT, P. W., DAVIS, F. J., llcalth Phys. 1 (1958) 169.
141 MURRAY, W. F., Nucl. Instr. 2 (1998) 237.
(5) HURST, 6. S., RITCHIE, R. II. ; Radiation accidents: dosimetric aspects
of neutron and gamma-ray cxposurcs, USAFC Rep. ORNL-2748(Pt.A) (1959).
16) BARRALL, R. C., McELROY, W. N., "Nuclear reactions for determining
neutron spectrum and dose; a world iterature search," pp. 251-277,
Personnel nosimetry for Radiation Accidents, IAEA, Vienna (1965).
11 WACHTIGALL, D., ROMILOFF, F., Nukleonik 6 7 (1964) 330.
isi BENEZE:CII, 6. et al., "Détermination des flux et des doses de neutrons
lors d'un accident de criticité a l'aide de la technique multisphère,"
pp. 349-368, Personnel Dosimetry for Radiation Accidents, IAEA, Vienna
(1965).
(9) HANKINS, D. E., "New methods of neutron-dose-rate evaluation," pp. 123-
139, Neutron Dosimetry, vol. II, IAEA, Vienna (1963).
(10) HURST, G. S., Brit. J. Radiol. 27 (1954) 353.
111) WAGNER, E. B., HURST, G. S., Rev. Sci. Instr. 29 (1958) 153.
(12) ROSSI, Il. H. , ROSENZWE!G, W., Radiation Res. 2 (1955) 417.
CASSEN, B. et al., "Development of a germanium crystal dosimeter,''
School of Aviation Medicine, USAF Rep. 57-90 (1957).
(14) DALY, D. R., PESHORI, li. H., An absolute neutron dosimeter based on a
generalized concept for radiation dosimetry, to be published in Health
Physics.
(15) HURST, G. S., RITCHIE, R. H., Health Phys. 8 (1962) 117.
(16) BLANC, D. et al., Health Phys. 11 (1965) 63.
(17) BLANC, D. et ai, Application a la dosimetrie pratique des rayons gamma
et des neutrons apides de chambres d'ionisation remplies d'un liquide
dielectrique, to be published in Health Physics.
FLEISCHER, R. L., Science 149 (1965) 353.
1191 KERR, G. D., STRÍCKLER, T. D., Health Phys. 12 8 (1966) 1141.
(20) BUTTLER, W., Possibilities in radiophotoluminescent dosimetry and a recent
development with particular attention to dosi.neter- and reader-techniques,
Proceedings of International Conference on Luminescence Dosimetry, Stanford
l'niversity, June 21-23, 1965, to be published by USAEC.
(13)
CAPTIONS
Fig. 1.
A typical neutron flucncc vs energy histogram based on the ORNL
threshold-detector system.
Fig. 2. Calculated response of a 10-in. sphere plotted against energy
taken from llankins, New methods of neutron-dose-rate evaluation,
Neut ron Dosimetry, vol. II, IAEA, Vienna (1963).
Fig. 3. The absolute fast neutron dosimeter: a proportional counter with
an internal alpha source.
Fig. 4.
Typical pulse-height distributions for the absolute proportional
counter.
Fig. 5.
The silicon diode detector and apparatus used as the basis for an
absolute dosimeter for neutrons.
Fig. 6. Calculated energy-loss functions for protons in silicon for neutrons
of energy up to 10 MeV. E is the neutron energy and € is the proton
energy.
Fig. 7. Silicon diode detector used in measuring energy distributions of
protons in tissue-equivalent phantoms.
Fig. 8. Measurod procon distributions in tissue-equivalent phantoms for
i4 MeV neutrons.
Fig. 9. Alpha particle tracks etched in the surface of a plastic disk.
Fig. 19. Components of the modified ORNL threshold-detector unit with glass
disks for fission-fragjent detectors.
"1
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Fig. 1
A typical neutron fluence vs energy histogram based on the
ORNL threshold-detector system
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Calculated response of a 10-in. sphere plotted against energy taken from
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Vol. II, IAEA, Vienna (1963)
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ORNL-DWG. 66-7603
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CALCULATED RESPONSE
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NEUTRON ENERGY
CALCULATED RESPONSE OF A 10 INCH SPHERE PLOTTED AGAINST ENERGY
TAKEN FROM HANKINS "NEW METHODS OF NEUTRON -DOSE-RATE EVALUATION"
NEUTRON DOSIMETRY Vol. II PROCEEDINGS OF A SYMPOSIUM. HARWELL, DEC.
1962, IAEA VIENNA, 1963. (RPG REPRESENTS "RADIATION PROTECTION GUIDE)
..
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Fig. 3
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39
ORNL-DWG. 66 - 7602

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for an absolute dosimeter for neutrons
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