HT
.
"
*
N
1.
11
-
.
1
-
..
.
IN
1
M
L'
U
w
i. Nih
'
.
.
2017
"
11
.
TL
ma
.
.
UNCLASSIFIED
ORNL
:
2
20
YR
479
ORNLP-499
"
DTIES
OCT 8 1989
CONF.641001- 27
MASTER
we
..
2-.
.
.
-
.
THE USE OF ISOTOPIC NEUTRON SOURCES FOR CHEMICAL ANALYSIS
J. E. Strain
W. S. Lyon
.
Analytical Chemistry Division
Oak Ridge National Laboratory
Oak Ridge, Tennessee
droniikin
in
LEGAL NOTICE –
The more wordgundamano wa wurth. More on
www, worden, maar omdat
A. Wahns my
w o rm e n, rood wit hemptiedo will respect to work

B. hocemos nag liabilities with respect to the nao et, er for damage resulting from the
As we won per months old a Comment * webedience ang •
warna ar auteur d e chambre, plays a roca tratar, the man who
i wa mare a real e states, who worector para
a mans, « martina www. we were more so we can
contoh
"Research sponsored by the V. S. Aconic Energy Commission under contract
4
LYFM21
4
M
.
2
1. Introduction
Though the nuclear reactor is considered the primary source of neutrons
for analytical application, there are many applications of the low intensity
isotopic-neutron sources. The simplicity, portability, and low cost very
of ten offset the disadvantage of their low-neutron yield. In this report, a
few examples of the use of isotopic neutron sources are presented. There
has been no effort to summarize all neutron source applications but only
those which have been proven economically practical at the Oak Ridge National
Laboratory.
of uliotola neutron sources are present out o
2. Iyes of Neutron Sources
There are presently three distinct types of 18otopic neutron sources.
They are alpha initiateå, photon initiated, and spontaneo'18 neutron emitting
sources.
The alpha-initiated source produces neutrons from the reaction of alpha
particies emitted in the decay of a heavy radionuclide with the nucleus of a
low mass element. Typical alpha-emitting radionuclides used and their hall-
lives are:
16,000 years
138 days
24,400 years
470 years
2
Ra 26
Po 220
Pu239
Am 242
The r.eutron energy and total yield is dependent upon the low mass tar-
get element used. Beryllium metal is most often used because of its high
neutron yield, ~ 77 a/10° alpha particles (J. The neutron yields and aver-
age neutron energy produced by typical target elements are:
Element
o/100 Alpha Particle
77
En (MeV)
4.5
Be
2.5
22
12
0.9
2.7
1.0
Recently, E. H. Acree of the Isotopes Division of Oak Ridge National
Laboratory has completed a study of the increased neutron yield from Am C4-
Be sources when irradiated in a thermal reactor (2) This irradiation produces
cm * in the source which greatly increases the internal aj.pha emission and
1
-l-
w
RASI
thereby increases the neutron yield. The Cra sus has a half-life of 162.5
days and is produced primarily by the beta decay of 16-bour Amcc.
There is ~ 5% Amat Pission which Increases the gamma background of
the source. The neutron yields and gamma backgrounds of a typical irradi- .
ated source are shown in Table I. Toe Amoz-Be source was fabricated by
blending the Amo, and Be metal in a 1:10 weight ratio and then compacted
into a 6-mm. diameter cylinder. This cylinder was welded into an aluminum
container. Following a neutron-yield lieasurement, it was placed in the Oak
Ridge Research Reactor and irradiated for 30 days at an average thermal neu-
tron flux of 2 x 1024 n/cm?/sec.
The neutron energy of the Am 4-cm 42-Be source was measured using 11°-
drifted silicon diodes, and except for a slight energy increase due to the
increased alpha energy of Cnce, it is essentially the same as that observed
for an Am "*--Be source (average E = 4.5 MeV) (Fig. 1).
Photon-Initiated neutron sources consist of a radionuclide which emits
> 1.6-lleV gamma radiation and a beryllium target. The sources use such
radioisotopes as Ra226, 56224, Mn54, Na24, etc., to produce neutrons of
< 100-keV energy. The primary disadvantage of this type source is the in-
tense gamma radiation associated with the source which requires rather mas-
sive shielding. The major advantages are the low cost (except Raccº sources)
and the ability to re-irradiate the source in a reactor to restore it to its
initial neutron emission level.
. The spontaneous Pission sources, such as cp , show promise as high-
intensity sources although they are costly and are handicapped by a high-
gamma background. The neutron energy distribution 18 essentially an un-
moderated fission spectrum with a specific neutron emission rate of w 109
a/8/mg. As gran quantities of this element become available (w 1970), its
use in neutron sources of high-neutron yield will accelerate the use of neu-
tron sources in industry.
226
- 3. Applications of Isotopic Neutron Sources
In the following examples of neutron source application, An --Be..
sources were used almost exclusively due to their long half-life, high yield
per cubic centimeter, and low ganna levels (37. Almost any type of neutron
source can be used in the examples presented except those activation analysis
care

..
applications which make use of high energy noutron reactions. The (7,0)
sources cannot be used for those applications.
3.1 Neutron Absorption Analysis
The basis of neutron absorption analysis is the measurement of neutron
flux depression within a sample. The depression of flux 18 proportional to
The theory of operation may be mathematically explained by considering
that a neutron source produces in a moderator a thermal neutron flux, g.
a (N2010.)
CR -
xx
e
&
-
-
-
-
X
X
.
-
4
A
4-
where
9 - thermal neutron flux,
- the product of the number of blº atoms in the
detector and the total neutron absorption cross
section of 320,
NO, = the product of the number of atoms and their
cross sections of the moderator, structural
materials, etc., which are assc ciated with the
system, and
N = the number of atoms of absorbing elements in the
sample times its total absorption cross section.
In a given arrangement, only the term, NC, 18 variable so that the
equation may be rewritten:
a
8
CR =
KK?
+ N
(K)
...(9200200
where
K
M201220B
X
X
x2 = (120,0103
and

.
If there is ac absorbing element within the sample, the couat rate will be:
CR - Kq
Dividing the count rate obtained with no <bsorbing element present by
the count rate obtained with an absorbing element present:
к
+
KON
O
Substituting the molar concentration (M) for . and an all inclusive
constant/(B) to replace K/K and to include Avagadro's number:
- 1 + B(o)(M)
This linear equation is the basis of absorption measurement. The value
of the slope constant, B, is controlled by the physical dimensions of the
sample. The larger the sample (the greater the sample thickness) surround-
ing the detector, the greater will be the value of B.
One arrangement used is shown in Fig. 2. The samp..e cell contains 100
m2. of sample. The curves obtained with several compounds in an aqueous sys-
tem are shown in Fig. 3.
The lower limits of detection of a few compounds and their measured
total absorption cross section are summarized in Table II. Using this same
principal, it has been possible to determine isotopic ratios in uranium,
lithium, and boron. If one measures the molar concentration of the element
by a conventional technique, 1.e., potentiometry, flame spectrometry, or
emission spectrography, and the total absorption cross section using neutron
absorption measurement, the results may be combined to calculate the 180-
topic ratio:

.
Isotopic Ratic
Experimentally Measured o
o of Absorbing Isotope
This method is particularly well adapted to uranium, lità:1 um, and boron,
since one isotope has a much higher absorption cross section than the other
(Table III).
To extend the usefulness of this technique for continuous measurement
of selected elements in a flowing process stream, the n tron source and
BF, were incorporated in a tank installed in the process line. The tank
volume was approximately 1 liter. A 30 cm. x 2.5 cm., one atmosphere B?UF,
detector, and a 10° n/s source were used. Both orcanic and aqueous solu-
tions of absorbing elements were 2n.nlyzed for boron, caclmium, chlorine,
lithium, and uranium continuously. In this situation, the hydrogeneous sol-
vent acted as the moderator, and because of the lurce sample size, the sensi-
tivity is greatly increased. Typical calibration curves and equipment are
shown in Fig. 4.
The latest application of neutron absorption techniques is in the analy-
sis ci reactor construction materials for total absorption impurities. Using
a large sample volume (7 liters) surrounding a 30 cm. x 2.5 cm. B?UF, counter,
it has been possible to detect total absorption cross section chances equiva-
lent to - 2 us of boron/cm". Usinä pure water as a calibration standard, the
practicability of analysis has been demonstrated in water, aluminum, and
beryllium. Using four isotopic sources (~ 4 x 10° n/s each) in a water nod-
erator around the 23-cm. diameter sample cell, sufficient counts were recor-
ded in 10 minutes to detect cross-section changes equivalent to + 2 ug B/cm".
The sample cell may be constructed in any shape desired to accommodate the
material whose cross section is to be measured. The only stipulation is
that the B+UT, detector must be surrounded by sample.
-
-
...-.
..
me

ing
RA
3.2 Neutron Transmission Measurements
Analysis of high cross-section materials by use of neutron transmission
is similar to neutron absorption except that the neutrons must be collimated
and thermalized before reaching the sample. Also, the neutron scattering,
as well as the neutron absorption cross section, is involved in the measure-
ment.
SH
1
www
Whi
11.11
12
IN
CM
.
.
en
4
The advantage of transmission measurement is that nondestructive analy-
sis of planer samples is possible without special preparation of soliå samples.
Analyses have been performed on boron-loaded plastic sheets (0.1 cm. to 2.5 cm
in thickness) and stainless steel. Figure 5 shows the apparatus used and the
type of calibration curve obtained with a 2.5-cm. thick boron-loaded plastic
sheet. The statistical limits are quite broad due to the low neutron source
intensity, short counting time (100 sec.), and the use of a low efficiency
Blur, tube. Transmission measurements should be considerably improved through
the use of a newly-developed highly-efficient neutron detector of small physi-
cal dimensions. The detector developed by H. H. Ross, Oak Ridge National
Laboratory, will be available on conclusion of patent arrangements 247.
Table IV shows the inclusion of the scattering cross section into the
total effective transmission cross section. In the geometry employed, only
about 30% of the scatter cross section is effective in reducing transmitted
neutrons. Therefore, it is necessary to experimentally determine the effec-
tive cross section of each element within the sample whose concentration is
expected to change.
3.3 Neutron Activation Analysis
lhe performance of thermal neutron activation analysis is seriously
limited by the low neutron flux available from isotopic neutron sources.
Using "alpha-n" sources, i.e., POS--Be, Amst-Be, etc., whose average neu-
tron energy is approximately 4.5 MeV, the available neutron flux is a iactor
of one hundred less than the 48 neutron yield. In order to surmount the
problem of non-uniform neutron exposure in a large sample, several small
sources have been arranged in a three-dimentional array as shown in Fib. 6.
This arrangement of sources produces a 183-cmo volume of fairly uniform ther-
mal neutron flux.
With this source array, 50-gram samples, and measurement on a 7.6 cm.
x 7.6 cm. crystal of the radioactivity produced. the following analyses have
proven practical from the standpoint of speed and accuracy.
.
.
-
ant
3.3.1 Hafnium in Zirconiun. Using a 50-gram sample of zirconium,
a neutron flux of 8 x 10* n/cm sec, it is possible to determine as little as
100 ppm of hafnium in zirconium. The sample is placed in a polyethylene
vessel, irradiated for one minute, and counted for one minute after a 20-
itu
....
! WY
Lh

seconå delay to allow for sample transier. Using a multichannel pulse-height
analyzer and a 7.6 cm. x 7.6 cm. MAI(TI) detector, the analysis can be com-
pleted in 6 minutes with a precision of 34 and an overall accuracy of 7% at
the 95% confidence level.
re,
3.3.2 Soil Analysis i'or Silicon, Aluminum, and Manganese. Weiched
3011 samples of approximately 30 grams are irradiated for 5 minutes in the
irradiation unit fitted with a O-oll. cadmium liner. Following irradiation
and a 20-second transfer delay, the sample is countei on a 3" x 3" NaI detec-
tor for 5 minutes. The number of counts under the 1.78-MeV AIC) photopeak is
proportional to the amount of silicon present.
The 2.3-min. Al activity is produced by the fast neutron reaction on
Si, i.e., (n,p) AIC while the cadmium liner prevents the thermal neutron
(n,x) reaction on any Al' which may be present in the sample.
Using the present system, it is possible to determine as little as 0.58
oi silicon with an accuracy or 20 percent. Il a 50-gram sample is employed,
one may then determine Si content down to 1% without chemical processing
within 15 minutes.
Aluminum may be determined in the same sample by removing the cadmium
shield and repeating the irradiation and counting. The aluminum is deter-
mined from the count data after subtracting the ALC activity due to the
(n,p) reaction on silica from the cadmium-shielded irradiation. Using the
above procedure, it is possible to determine as little as 0.1 grams of alu-
minum with an accuracy of 20%. And if a 50-gram sample is used, then alu-
minum may be determined at concentrations of 0.2%.
Manganese is determined in a sirilar manner except that an irradiation
12
NH
to decay prior to counting. A counting time of 5 minutes is used and the
concentration of manganese is proportional to the counts beneath the 0.85-
MeV photopeak of 2.6-hr. Mnº. Using this irradiation time, delay, and
counting, as little as 0.006 grams of manganese may be determined with an
accuracy of 20%. If one uses a 50-gram sample, then the manganese content
may be determined in concentrations as low as 0.01%.
In addition to the above analyses, which have applied to some 50 soil
samples, a brief effort was made to establish a table of sensitivities for
Ch
NL
.
several of the most promising of the widely distributed elements. Table V
sumarizes the observed sensitivities.
As can be seen in Table V, it would be possible to analyze a 50. Eram
sample for indium in the range of 0.001%, vanadium in the range of 0.0026,
etc. This, of course, assumes no interference radioactivity being produced
in the sample.
3.3.3 Identification of letals. Using the eight-source irrudia-
tion unit, a 15-minute irradiation, a 20-second delay, and a 5-minute count,
on a multichannel pulse height analyzer, it was possible to distinguish
Hastelloy alloys. The cpectra (2 Hastelloy , B, C, and D are surficiently
different to allo: positive Identification oñ an unkno:n Hastellcy sample.
The same system of irradiation time and counting was performed on nine
varieties of stainless steel (347, 330, 316, 302, 304, 405, 420, 420, and
446) which show sufficient difference to allow identification of an alloy in
question.
-
A
VS .
4. Conclusion
Isotopic neutron sources are a rugged, safe, and dependable source of
neutrons without the electronic complication of accelerator sources or the
high cost of nuclear reactors. They are easily portable and uniquely adapted
for the field or industrial application. While the present sources are more
than adequate for neutron absorption or transmission applications, the advent
of cheaper sources of higher-neutron yield will increase the userulness of
isotopic sources in activation analysis applications. When isotopic neutron
sources are recognized for the valuable industrial tools they are rather than
a neutron physics teaching device, their use will greatly increase.
5. Bibliographical References
[1] "preparation, Properties, and Availability of Polonium Neutron Sources,"
TID-5087, July 1952.
[27 ACREE, E. H., Oak Ridge National Laboratory. Unpublished data, 1964.
[ 37 STRAIT, J. E. and LEDDICOTTE, G. W., "Preparation, Properties, and Use
of AmC4 - Alpha, Gama, and Neutron Sources," ORNL-3335 (1960):
[47 ROSS, #. #., Oak Ridge National Laboratory: Unpublished data, 1964.
.
,
-8-
E
ELA
TASLI I
241
DIFECT OF YOUTRO:! IRRADIATIO: OX An-4--Be IILUTRON SOURCE YELD
Irradiation fine,
23 2024 cm2sec.
Neutzon. Yield
hiter
Irradiation
y Radiation
Present After
Irradiation
Source
15.3 mg. Aino, Be
15.3 mg. Amoz-Be
1.1 x 10n/s
3.9 x 200 n/s
3 mr./br.
5.5 R/nr. 2 6**
30 days
"Radiation half thickness is 5 min. of Pb; 22 mm. of Po reduces the rzdiation
by a factor of 10 after a 30-czy decay.
LUL
-
11
.
IN
1.
"
TL
.
mu
..
mo
i
vraison restaurants
..
L
WHY
.
dos
"
..
.
Mny
C
nh
HY
UK
I
.
YOU
TRN.
''
ile
TABLE II
NOUTRO: ABSORPTIO:: S313IZIVITIES FOR SELECTED ELET.COMTS
Apparent tolar
Cross Section, barns
Lower Limit op Detection,
molar concentration
Ccาววนาน
IC
0.02
HNO3
нсі
0.74
0.067
0.035
LIOH
0.0065
HG(NO3)2
H%B03
0.0033
vo (C2H302)2:24,0
0.052
-10-

TABLE III
ISCTOPIC ABSORPTION CROSS SECTIONS
FOR BORON, LITHIUM, ANU VA NIUM
Isotope
820
ܐܐܩ
Total Absorption Cross
Section, barns*
4 x 103
<0.05
945
0.03
2600
9.5
2235
0238
Neutron cross sections, Hughes and Harvey,
BNL-325.
-11-
TV
..
.
:
.
.
.
..
h
MR
.
0.1
1.
SILV
21W
L"
12
TABLE IV
CROSS 52CTIO.:3 703 !MURO: TUSIS ICI!
Cross Sectiori
Absorption, Scattering,
barns
barns
Efective Transmission
Cross Section, barns
Element
190
2.2
192 ***
62
b
65
2550
3400
800
755
0.33
10.3
< 0.0002
1.2
"W. 11. Sullivan, Trilindar Chart of juclides, Oak Ridge National
Laboratory.
**
This cross section used as standard and all values determined
relative to this value.
.12.

1
TABLE V
OBSERVED ACTIVATIC!I NIALYSIS SEUISITIVITICS
AT A FLUX 02 8 x 104 n/cm/sec
Irradiution
Time, min.
Element
10
1000
Integrated Counts/
Min./Grams Under
the Mejor Photoveak
2.9 ä 100
1.8 x 135
4.5 x 20"
1.6 x 103
5.3 :: 103
i
2.5 x 203
7.3 x 103
-13-
.
ORNIL-DIG. 63-2524
Fig. 1. Neutron energy distribution from an Amc*--Be neutron source.

1
.
I
',
NHW
. VI
:)
UNCLASSIFIED
ORNL-DWG. 63-2524

1600
1400
Low Energy
Neutrons
Produced by
Moderation
| ! In Be and
Scatter
Am741 - Be Neutron Spectrum
LibF - Si Diode Sandwich Detector
24-hour Count
Six-Source (2 Curies Each) Cluster
Total Am 241, 3.6 grams
Total Be , 36 grams
J.E. STRAIN
E.H. ACREE
NEUTRON FLUX, relative
Lect--I--I-
2006
ó
í
3
2
.
é
1
4 ó Ź
NEUTRON ENERGY, Mev
0
11
12
S
SS
3
.
.
SL
i
E
7

...
- -
-
-
-
-
-
-
-
-
-
-
-
- -
-
-
-
...
_
..
-
-
.
F
Fig. 2. Apparatus for neutron absorption measurements.
ORNL- LR-DWG. 61054Å
-
-
.
.-
UNCLASSIFIED
ORNL-LR-DWG. 61054A
han
THERMAL
WELLS
-NEUTRON
ABSORPTION CELL

REMOVABLE
PARAFFIN
PLUGS
BF TUBE
TEFLON SAMPLE
CELL

PARAFFIN
SAMPLE LEVEL
Am-Be NEUTRON
SOURCE
EBF COUNTER
THERMAL
WELLSN

-NEUTRON
ABSORPTION
CELL
PARAFFIN
MODERATOR
SOURCE
AMPLIFIER
H.V.
POWER
SUPPLY
TO PREAMPLIFIER
SCALER
SAMPLE CELL
AND COUNTER DETAIL
ORNL-LR-DAG. 1741B
Fig. 3. Neutron absorption ratios as a function of molar concentration.
.
:.
::
-
-
-
.
L
M
T
M
.
..
1
W
.
WW
W
.04.
With
....
.
.
.
1.25 am
UNCI ASSIFIED) :
ORNI-IR-OWG 17916

hN
N
077----------
Cd (NO3)2
H,803
Hg(NOZ 2
In(NO,),
1.20H
LiOH
AgNO3
1.15
ole
! 1.1041
1.08+//
COINO2
14.0611!
NICI
MnSOR
1.041
4.0214
)
o
NHAI
NHANO:3
0.5
HNO3
1.5
1.5 2.0
2.0
MOLAR CONCENTRATION
2.5
- 3.0
2.5
3.0
3.5

.
.
LA
1
TW
.
.."
1
"
TAN
-
H
ow
--.---...-..
-.-.-.-
.................-..-----.-.--............
Fig. 4. Neutron absorption measurements in flowing streams.
ORNL-LR-D1 . 61053A

UNCLASSIFIED
ORNL-LR-DWG. 61053A
OUNTER
AIA PREAMP.

LINEAR
RATE METERS
AIC Amp.
BROWN
RECORDER
YHV SUPPLY
-NEUTRON SOURCE
SS SAMPLE CONTAINER
CENTRIFUGAL PUMP
SOLUTION
RESERVOIR
(a) Block diagram of equipment used for evaluation and
calibration.

TT
22 -
& neutron count/min
with pure water in stream
CR - neutron count / min
with U, B, or Cd
in stream
Natural Boron
Natural Cd
Natural U
0.5
3.5
1.0 1.5 2.0 2.5 3.0
U, Cd, OR BORON CONCENTRATION (9/1)
-
(b) Calibration curves obtained from pure solutions.
IT,
12
12%
2
ORNL-LR-DWG. 62065A
Fig. 5. Neutron transmission reasurements.
.
.
.
.
-
-
-
-
--------------.............................
i

satt
SIN
H
t
UNCLASSIFIED
ORNI-LR-DWG. 620 6 5 A
.
.
9
>
Am241 - Be NEUTRON SOURCE
-CD SHIELDING
-20 cm x 20 cm x 20 cm
PARAFFIN MODERATOR
2.5 cm SLAB SAMPLE
PRE AMP

2.5 cm x 6 cm
B10 F, DETECTOR
LEAD SAMPLE
SUPPORT
LINEAR
AMPLIFIER
H.V. SUPPLY
SCALER
(a) Thick slab neutron transmission
analyzer.

Сттттттт
-
-
-
-
$
106
50
100
150
200
250
1
.
-
-
-
. mg B /cm2
(b) Experimental result of neutron transmission
measurement in synthetic boron loaded samples.
A
*
.
.
.
"
.
2
.
.
V
2
w
22
:
-
---
-
-
--
--.
www.DE
.
.
.
.
.
-
-
-
...
*..
.
.-..
Fig. 6. Arrangement of Am-Be neutron sources in paraffin moderator.
ORNL-LR-LWG. 66052

1
A
2
UNCLASSIFIED
ORNL-LR-OWG. 66052
100 ml POLYETHYLENE
IR RADIATION CONTAINER
ob - 5cm IRRADIATION
- 19"
L WELL

5 cm
DIAMETER
-
-
MODERATOR
ASSEMBLED
23 cm
UPPER
SOURCE
PLANE

7.7 cm
ä
LOWER
SOURCE
PLANE
--
5cm
RADIUS
Arrangement of Am-Be Neutron Sources in Paraffin Moderator

WW
,
.
X
$
DATE FILMED
12/0 765
.
RY
TI
L
LEGAL NOTICE --
V4
AT
How
This report was propared as an account of Govornmont sponsorod work. Moithor the Unitod
Suatos, nor the Commission, nor any person acting on behall of the Commission:
A. Makos any warranty or roprosontation, exprosed or implied, with respoct to the accu-
racy, completonosi, or usohulnous of the laformation contained in this roport, or that the wo
of any information, apparatus, mothee or procosdisclosed in this report may not infringo
privately owned righto; or
B. Assumos any liabilities with rospoot to the use of, or for damages rosulung from the
Use of any information, apparatus, method, or procons disclosod in this roport.
As used in the abovo, "person noting on behall c! the Commission” Inoludos any om-
ployo. or contractor of the Commission, or omployee of such contractor, to the oxtont that
such omploy.. or contractor of the Commission, or employee of such contractor proparos,
dienominatos, or provides accous to, any Information pursuant to his omployment or contract
wild the Commission, or bio omployment with such contractor.
M
ST
TRIT
$
.
EW
W
TA
..
..
:
.
.
END