SMALL SAMPLE ASSAY SYSTEM
T. Gozani
Senior Scientific Advisor
C. N. Ingraham
Senior Engineer
Intelcom Rad Tech
San Diego, California
ABSTRACT
A new high-precision assay system for fissile nuclear material has
been built. The new system, called Small Sample Assay System (SSAS), is
based on the principle of thermal (or epithermal) neutron activation analysis,
where the observed response is the delayed fission gamma rays emitted from
the fission product a few seconds to a few minutes after the fission. The
system takes advantage of the high initial yields of the delayed gamma rays
by starting the counting very shortly after the end of the irradiation. The
SSAS operates with a basic timing cycle which begins with a background measurement, continues by irradiating the sample in the neutron flux of a properly
moderated 252(jf source, and concludes with a measurement of the signal plus
background. The basic physics of the system, which makes efficient use of
the delayed gamma spectrum and time dependence, yield very high count rate
per gram of fissile material, even when moderately small Cf sources (10 to
50 pg) are employed.
The SSAS can be operated with either or both of two types of electronic processing systems. The basic system employs a dead-time-less technique which allows handling of high count rates and intrinsically gives
increased weight to the higher energy portions of the gamma spectrum, thus
reducing the gamma self shielding. The other system involves a fast linear
system which retains pulse-height information, found to be a very useful
feature, especially for mixed oxide pellets. The prototype SSAS was subjected
to an intensive series of tests aimed at determining the precision of the
device and its reliability.
Extensive tests to determine the precision of the SSAS were performed with sintered LWR pellets and UC>2 powder spanning the enrichment range
from natural to 3.3%. The results show that the typically obtained precision
is about 0.5 and 1% at the 1-a confidence level for pellets and oxide powder,
respectively.
1.
INTRODUCTION
There is a definite need within the nuclear industry, especially in
fuel fabrication, for a rapid and accurate measurement of the total fissile
content of relatively small samples, such as sintered fuel pellets, green
pellets, vials of uranium oxide, HTGR green and fired rods, etc. Presently,
the total fissile assay is based, in most cases, on the total weight and the
isotopic ratio as determined by mass spectroscopy; this is a relatively slow
and expensive process. It may also suffer from sampling errors because of
the financial implication of destructive testing of large samples.
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Nuclear Materials Management
The Small Sample Assay System (SSAS) was developed by Rad Tech to
provide the nuclear industry with the tool to fulfill the need described
above. The SSAS is a transportable, self-contained, compact system which
will fit easily in most existing laboratory or production spaces. It is based
on fission activation of the sample, using a relatively small "^Cf neutron
source and the detection of induced delayed gamma-ray activity. The SSAS
allows one to measure the total fissile content of relatively large numbers
of production samples, obtaining accurate results at a very reasonable cost.
2.
PRINCIPLES OF OPERATION
Up to approximately 50 samples can be loaded into a sample changer
which is used to insert them into the assay system. According to a preselected adjustable timing cycle, the loading mechanism then automatically moves
each individual sample through the assay system where three basic operations
are performed in the following order.
1.
Measurement of room temperature plus passive signal
from sample.
2.
Irradiation of sample
3.
Measurement of the signal, namely the induced fission
delayed gamma, plus the background.
The background includes any ambient background, due mostly to the source, as
well as that which may be inherent in the sample itself, such as the 185-keV
gamma activity in 235u, "400-keV groups" and other lines in 239pU) 232>jh decay
product activities, etc.
The irradiation of the sample is made in a proper neutron flux with
either soft or hard spectrum created by a relatively small 252Cf neutron
source. The irradiator and the shielding in the standard SSAS can accommodate
up to 50 yg of 252cf> xhe harder neutron spectrum is recommended for cases of
high enrichment and where substantial radial nonuniformity in the distribution
of fissile material within the sample is expected. The soft spectrum, which
delivers a much higher signal, was found entirely adequate for LWR pellets and
oxide powders. A calibration run for BWR pellets irradiated in a soft SSAS
spectrum is shown in Figure 1.
The design of the irradiator is made so that the neutron flux distribution along the sample axis will be reasonably flat. This reduces the
systematic errors due to possible nonuniformities along the axis of the
sample. Typical axial flux distributions in a standard SSAS are shown in
Figure 2. The three distributions differ mainly because of the different
materials surrounding the source (low-Z versus high-Z) and the source-tosample distance. Because one deals here with a small sample, the quantity
of interest is the width of the distribution at an axial point with an amplitude of 0.9 of the maximum (FW 0.9M). One can further flatten the distribution to accommodate longer samples. This will result in lower peak flux, and
would require a larger detector to accommodate the longer sample. The shape
of the flux distribution also indicates to the designer the requirements for
Fall 1974
93
reproducibility in the sample position in the irradiation zone. A change of
3.10"^% in the fission rate would result from a position change of 10"^ inches.
Following irradiation, the sample is moved rapidly, taking 1.5 seconds in the prototype SSAS, into a gamma-ray detector. Typically, a NaI(T£)
detector is used, but other detectors can be used if there is a need. The
delayed gamma activity from fission products is measured for the same length
of time as are the irradiation and background measurements.
The use of delayed gamma as a signature which is correlated to the
number of fissions occurring in the fissile isotopes of the sample, offers
very high sensitivities. There are several (about six) delayed gammas
(integrated over the entire time domain) per fission event. The most prominent gamma rays have a relatively long decay time; thus, their decay during
the sample travel from the irradiation location to the detector will be small.
A typical decay curve of the total delayed gamma energy deposited in the
detector versus time following an irradiation of 10 seconds, and "waiting"
time of 1.5 seconds, is shown in Figure 3. In this case, the initial decay
is about 16% per second. Thus, the basic timing and motion error should not
exceed 1 to 10 msec.
The pulse-height distribution of the delayed fission gamma rays
reveals another important and useful feature, namely, it contains a significant fraction of hard gamma components which penetrate through a LWR fuel
pellet with only a small attenuation. Thus, the gamma response is from the
bulk of the sample rather than the surface, as in the case of the passive
185 keV gamma line of 235u. A pulse-height spectrum using BWR pellets in
the 10-second cycle is shown in Figure 4. The average energy of this pulseheight distribution is 860 keV, and 45% of all pulses are above 660 keV.
Other spectra were measured in different cycle lengths, and no substantial
change in the pulse-height distribution was observed.
The overall basic physics design of the system makes efficient use
of the delayed gamma spectrum and time dependence. This yields very high
count rates, about 100,000 to 300,000 cps per gram of 235y ±n low-enriched
uranium powder, using neutron sources of modest intensities.
3.
DESCRIPTION OF THE DEVICE AND ITS OPERATION
A photograph of the SSAS prototype is shown in Figure 5. The SSAS
consists of a linear drive, loader, source container and irradiator, radiation detector, and control and processing electronics, with a printer as an
output.
The linear drive is loaded with the sample; it then positions the
sample inside the detector for background measurement, moves the sample to
the irradiation position, and transports the sample at the end of irradiation back to the detector. Following the end of the measurement the sample
is unloaded into an unloading tray, and a new sample is loaded onto the
linear drive. Presently, the background and signal measurements, as well as
the irradiation length, are equal, but the time length can be selected. The
motion time between detector and irradiation zone is about 1.5 seconds. Different loading and unloading mechanisms may be required in some instances to
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Nuclear Materials Management
be able to count neutrons, in addition to gamma rays. The resulting resolving
time still limited the accidental coincidence rate to 0.1% of the maximum time
coincidence rates expected for Pu loadings of approximately 20 g.
240
Since the FMD directly measures the amount of
Pu in a barrel, the
Pu isotopic ratio has to be known to determine the fissile content. If the
isotopic ratio is in doubt, an approximate indication of the fissile 239pu
content can be obtained with the Nal detector. This detector does not employ
costly and sophisticated techniques like vertical scanning, peak fitting with
spectral background subtraction, and transmission measurements. Instead, it
is meant to be used only as a diagnostic indicator.
4.
PERFORMANCE DATA
An extensive series of measurements was conducted to establish the
response characteristics of the FMD. All performance data was acquired with
spontaneously fission source material (both 239pu an(j 252cf were used) , positioned in a 55-gallon barrel specially prepared for the test program. The
barrel was filled with a cloth and paper matrix, except for three vertical
pipes, one at the center, the other two close to the periphery, into which
the source material was positioned. With this arrangement it was possible
to measure the effects on the response of the FMD due to differences in matrix
density and position of the source in the barrel.
With a 3/10 coincidence a fission efficiency of 15.6% was achieved,
at which mixed oxide fuel pellet standards containing 0.05 and 0.41 g of
240pu Were measured with a statistical accuracy of 21 and 2.5%, respectively,
in 2000-sec measurements. The fissile content of the pellets was 0.42 and
3.52 g, respectively. For a 4/10 coincidence setting the fission efficiency
dropped to 3%.
The detector response in five-minute measurements to small quantities
of 240pu £s snown in Figure 5. Even in this relatively short counting time a
small quantity, such as 0.16 g of 240pUj ^s measured with a statistical precision of 16%. The response is essentially linear up to 60 g of 240pu> with
the detection range extending down to a minimum of 0.01 g of 240pu for measurements up to one hour in duration.
The response as a function of position inside a 55-gallon drum is
shown to be uniform within 4% over most of the barrel volume (Table 1). The
maximum variation from the response is 14%, with the drum positioned on the
rotating platform. The largest change occurs at the bottom; it is caused by
attenuation effects in the relatively bulky rotating mechanism.
The fission multiplicity technique is less susceptible to attenuation effects than the method of scanning for 400 keV isotopic gamma radiation
from the fissile component (Ref. 2). The response of the latter technique
was reduced by 23% more than the response of the former when the fission
source was shielded with 0.25 inch of lead.
Fall 1974
95
enrichment during a given group of measurements. This technique is analogous
to the method which would normally be used to calibrate the machine during
typical operation, and removed any systematic errors in the preparation of
calibration standards. Thus, the successive measurements of each individual
pellet may be used to obtain the precision of the machine itself, and the
individual measurements within a set of pellets having the same nominal
enrichment can be used to determine the production-related variations.
Two equivalent approaches were taken to determine the standard
deviation, which is characteristic of the machine itself. In one case, the
six (or three) runs for a given pellet were averaged, and the standard
deviation of the remaining five was computed about this value. This was then
repeated for each of the independent combinations of pellets and the results
were averaged. It must be remembered that this latter procedure is equivalent to taking the ratio of two measurements, both of which contain similar
statistical uncertainties, and the individual measurement precision is
obtained by dividing the result by V~2. This correction has been made in
the analysis.
The composite results of the statistical analysis are presented in
Table 1 where we show the relative standard deviation (a) using both methods
described above for calculating it; these results are quite consistent. In
addition, we show a in units of absolute enrichment (AE), in which case the
result is essentially constant (a « 0.015% enrichment) for all enriched
pellets.
To obtain an independent check of the measurement precision, an
additional comparison was made for the case of Set 1 (3.3% enrichment). These
12 pellets were individually measured to an average relative precision of 0.4%
using the TRIGA reactor and the Rad Tech delayed neutron activation analysis
(DNAA) technique. When normalized to the DNAA results as standards, the SSAS
measurements show a standard deviation (a) of 0.52%. This implies a for SSAS
to be 0.33%, if the contribution to the a from DNAA is removed. This is quite
consistent with the value of 0.30% shown in Table 1, and independently confirms the reasonableness of the analysis.
All of the results above were obtained with the SSAS operating on a
cycle of 3 x 30 seconds. That is, the background measurement, irradiation,
and counting times are each 30 seconds. To determine the performance of the
machine for faster throughput conditions, the precision was also tested for
shorter cycle times down to 3 x 2 seconds. These results are summarized in
Table 2, along with some of those from longer counting times already shown
in Table 1.
Tables 3 and 4 show as examples the net counts given by the sealer
and printer for pellet Sets 1 and 2. All count data are the output of the
dead-time-less system representing the delayed gamma dose. These numbers do
not directly yield values for the counting statistics. Typically, the total
count for the linear pulse counting system are larger by a factor of 3 to 10.
Typical relative counting statistics are between 0.1 to 0.2%.
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Nuclear Materials Management
4.2
URANIUM OXIDE SAMPLES
Twenty-five samples of uranium oxide, 21 product oxides, and four
scrap samples were available for the tests. The samples were contained in
1/2-inch polyethylene vials. Most of them filled the entire volume of the
vial; others did not, due to lower weight or higher density. Each sample
was measured at least four times to yield the machine reproducibility when
such loose-powder samples were used. The RMS relative standard error for
all the samples was found to be 0.4%.
It is anticipated that the neutron self shielding in these samples
(enrichment between approximately 2 to 4%) should be low, since they have
relatively low bulk densities (typically about 2.3 g/cm^). Therefore, a
linear least-square fit was made between the net yield and the given mass
of total fissile; this is shown in Figure 6. The line is given by the
equation
noc
U(g
o
fi
U) = 4.545 x 10
C (counts) - 3.294 x 10 .
235
The relative difference between the given mass of
U (based on a mass
spectroscopy of a small sample) and the measured one (based on the above
curve) is given in Table 5, which includes also the given mass, the measured
mass, and the relative bulk density. The RMS of the relative deviation of
the measured mass from the given 235u mass is 0.83%. This overall accuracy
for the oxide powder is very gratifying. It is clear that most of the error
is due to the availability of relatively poorly known standards. The deviations found when scrap oxide samples were used (see Table 6), was substantially higher due to large density variation and large uncertainty in the
assigned (or given) 235y value.
5.
CONCLUSIONS
The SSAS described here is a high-precision active nuclear assay
device. It is designed for rapid high-precision measurements of representative samples taken on a statistical basis from a large production. With the
basic system a good sampling of the production throughput can be achieved
with measuring time from 90 down to 10 seconds per sample. To allow sampling
of larger throughput, or to measure the entire production of pellets, SSAS
will have to be modified, using, however, the same basic unit. The enrichment in the fissile material of samples with known isotopic composition, such
as U02 fuel pellets, is determined from the accurately known total weight and
the fissile content as determined by the SSAS. Extensive performance tests
have shown that precision at 1-a level of ±0.016%, or less, in absolute enrichment in the range of 0.7 to 3.3%, can be routinely achieved. The precision at
1-a level of enrichment determination for oxide powder was found to be approximately ±0.025% absolute enrichment.
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Table 1
SSAS PERFORMANCE RESULTS
Standard Deviation
To Mean
To Individual
Absolute
% Relative
% Relative
Enrichment
Set
No.
Enrichment
(Z)
No. of
Pellets
No. of
Runs
1
3.3
12
6
0.322
0.301
0.016
2
2.78
12
6
0.350
0.337
0.014
3
0.7
10
3
0.742
0.729
0.0050
4
1.8
3
3
0.587
0.573
0.013
Table 2
SSAS MEASUREMENT PRECISION FOR A RANGE OF CYCLE TIMES
Cycle Time
(sec)
Set
No.
Enrichment
(%)
% Relative
3 x 30
1
3.0
0.301
3 x 30
2
2.78
0.337
0.016
0.014
3x5
1
3.0
0.420
0.022
2
2.78
1.3
0.054
3x2
a
Standard Deviation
A Enrichment
a
252
50-yg
Cf source; all other data obtained with 20-yg source.
Table 3
DETAILED DATA
(30 x 30 sec runs, 20 yg 252Cf source)
Raw Counts, Background Subtracted, 6 Runs
Set 1 (3.3% enrichment)
Pellet No.
Cl
C2
C3
C4
C5
C6
1
2
3
4
5
6
7
8
9
10
11
12
74984
72945
72092
73613
71552
72527
72576
72467
73555
73412
73998
71967
74655
72857
72102
72128
71184
72898
72401
72359
73958
72862
73483
71786
74809
72866
72229
73280
71819
72047
72651
72508
73820
73350
73576
71463
74127
72608
72228
72727
71351
71869
72376
72508
73899
73659
73083
71436
73794
72167
71408
71977
70795
71292
72036
72027
72612
72717
72974
71245
74352
98
72795
71978
73233
71030
71844
72332
71977
73531
73346
72734
71312
Nuclear Materials Management
Table 4
DETAILED DATA
(30 x 30 sec runs, 20 yg 252Cf source)
Raw Counts, Background Subtracted, 6 Runs
Set 2 (2.78% enrichment)
Pellet No.
1
2
3
4
5
6
7
8
9
10
11
12
Cl
C2
C3
C4
C5
C6
64819
63275
64478
64269
65435
62589
63264
64531
64029
63498
65625
65516
64743
62516
64575
64473
65244
62497
63329
64377
64019
63266
64814
65104
64636
62893
64020
64029
65141
62578
63468
64182
64101
62988
65626
65067
64614
62296
64584
64061
65198
62629
62785
63878
64602
63239
64994
64939
63902
62117
64059
63376
64963
62108
63000
63799
63782
62918
64968
64995
64181
62384
64546
63768
65332
62682
63179
64019
64225
62989
65193
64879
Table 5
TOTAL FISSILE DETERMINATION IN URANIUM OXIDE SAMPLES
Relative Deviation
Given Mass
0.1479
0.1924
0.2084
0.2526
0.2997
0.1565
0.2119
0.2642
0.1908
0.2687
0.3243
0.1455
0.2133
0.1870
0.2539
0.2936
0.2070
0.2682
0.3243
0.1972
0.1573
Fall! 974
Relative Bulk Density
0.98
1.09
1.05
0.99
0.89
1.03
1.07
1.00
1.07
1.05
0.97
0.97
1.07
1.05
0.99
0.87
1.04
1.05
0.96
1.11
1.05
Measured Mass
0.149480
0.192554
0.207697
0.257519
0.303765
0.155647
0.211473
0.263418
0.189391
0.267308
0.322934
0.146699
0.210355
0.186369
0.254415
0.296076
0.206992
0.267549
0.320689
0.195535
0.158829
Root mean square deviation =
-1.0682
-0.08
0.34
-1.95
-1.36
0.545
0.201
0.296
0.738
0.518
0.421
-0.824
1.38
0.0337
-0 . 203
-0.843
0
0.243
1.113
0.845
-0.972
0.83
99
Table 6
TOTAL FISSILE DETERMINATION IN SCRAP URANIUM OXIDE SAMPLES
Given Mass
Relative Bulk Density
0.1274
0.1437
0.1399
0.1455
0.59
0.91
0.68
0.90
100
Measured Mass
0.1186
0.1388
0.1369
0.1380
Relative Deviation
(%)
-6.93
-3.42
-2.18
-0.052
Nuclear Materials Management
4x10"
3x10
en
CD
2x10H
CU
</i
o
o
10
RT-08283
Figure 1.
Fall 1974
ENRICHMENT (%)
Typical SSAS calibration curve for BWR fuel pellets
101
Figure 2.
Flux (or fission rate) distribution along the irradiator axis of the SSAS
F±8ure 3
' E2 ofsIS SLjT^r follo"in8 lrradlatl°n of 10 s—
Figure 4.
Delayed gamma pulse-height distribution (30 second irradiation and meaning time)
LIST OF ILLUSTRATIONS
Figure
Fall 1974
1
Typical SSAS calibration curve for BWR fuel pellets
2
Flux (or fission rate) distribution along the irradiator of
the SSAS
3
Decay of delayed gamma dose rate following irradiation of 10
seconds (with 0.5 second waiting time)
4
Delayed gamma pulse-height distribution (30 second irradiation
and meaning time)
5
SSAS prototype
105
Figure 5.
SSAS prototype