A method for the identification of observable radionuclides from neutron activation analysis of spent nuclear fuel was investigated. A predictive model was formulated using ORIGEN-ARP and nuclear decay data to predict neutron activation analysis results of two spent nuclear fuel samples with variable burnup values and cooling times. Model predictions were tested by performing thermal instrumental neutron activation analysis on the spent nuclear fuel samples using both cyclic and conventional irradiation methods. Preliminary results indicate neutron activation analysis was successful in identifying several stable and long-lived radionuclides predicted via model calculations but results appear limited to sample concentration. Spent nuclear fuel samples of higher specific activity are needed to further validate model results.
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Predictive Modeling of Neutron Activation Analysis of Spent Nuclear Fuel for the Detection of Stable and Long-Lived Radionuclides
1. 1
Predictive Modeling of Neutron Activation Analysis
of Spent Nuclear Fuel for the
Detection of Stable and Long-Lived Radionuclides
R. I. Palomares, K. Dayman, S. Biegalski, S. Landsberger
Nuclear Engineering Teaching Laboratory
Nuclear and Radiation Engineering Program
The University of Texas at Austin
Research Report
September 2012
2. 2
Abstract
A method for the identification of observable radionuclides from neutron activation analysis of spent
nuclear fuel was investigated. A predictive model was formulated using ORIGEN-ARP and nuclear decay
data to predict neutron activation analysis results of two spent nuclear fuel samples with variable burnup
values and cooling times. Model predictions were tested by performing thermal instrumental neutron
activation analysis on the spent nuclear fuel samples using both cyclic and conventional irradiation
methods. Preliminary results indicate neutron activation analysis was successful in identifying several
stable and long-lived radionuclides predicted via model calculations but results appear limited to sample
concentration. Spent nuclear fuel samples of higher specific activity are needed to further validate model
results.
Introduction
Determination of radionuclide composition has proven a vital component to processes such as safety
evaluations of spent nuclear fuel systems and experimental planning [1,2]. At present, material accounting
systems of spent fuel at nuclear facilities rely heavily on the use of burn-up codes. Although continuous
development of computational methodologies for nuclear fuel analysis and revision of nuclear data have
improved accuracy of burnup calculations [1], experimental code verification remains an important step
towards the establishment of improved safety bases. Experiments of this type are often performed with
destructive analysis techniques such as mass spectrometry.
Spent fuel samples may be destructively examined by means of radiochemical analysis methods when
high accuracy measurements of isotropic compositions are desired [1]. Destructive analysis typically
involves extensive sample preparation that may involve dissolution, chemical separation, and vaporization
[1]. Although an effective method, destructive analysis is associated with long analysis times and high costs
[3]. Optimism currently lies with more conventional non-destructive analysis techniques such as passive
gamma counting.
Passive methods have proven for years to be both effective and reliable techniques for the determination
of radionuclide composition. Passive gamma-ray spectrometry systems have the benefit of being readily
available, relatively small, and providing real-time data [3]. Nonetheless, gamma-ray spectrometry relies on
the gamma emissions of the analyte thus, problems inevitably arise with the characterization of stable and
long-lived radionuclides whose very slow to non-existant decay rates cause difficulties in passive counting.
Neutron activation analysis (NAA) is an attractive technique to circumvent such complications as it
boasts non-destructive analysis independent of both sample size and activity [4]. In addition, high
3. 3
selectivity of gamma-ray spectrometry allows for the determination of various radionuclides simultaneously
with sensitivities potentially reaching parts per million (PPM) or parts per billion (PPB) [5]. Lastly, NAA is
the only nuclide analysis technique based on phonomena occuring in the atomic nucleus [4]. As such, NAA
depends on the gamma-ray signatures of the activation products rather than the unactivated analyte; this
allows for the detection and measurement of stable and long-lived nuclides whose activation products are
detectable.
The primary objective of this project was to determine if NAA provides a viable non-destructive method
for the detection of radionuclides in spent fuel. The project was structured into three parts consisting of
predictive modeling, experimental verification, and model application. We first modeled spent fuel samples
of variable burnup values and cooling times on the order of 20 - 30 years using ORIGEN-ARP to estimate
nuclide concentrations. Concentrations were used as input for activation calculations to determine the
viability of nuclide detection given assumed experimental parameters. Spent fuel parameters such as burnup
and cooling time were chosen to be representative of spent fuel samples at the University of Texas at Austin.
The model was subsequently tested by performing NAA on the spent fuel samples at the Nuclear
Engineering Teaching Laboratory (NETL) at the University of Texas at Austin.
Verification of the model would ultimately transition to model application in which the model would be
used for predicting NAA results and suggesting optimum NAA parameters for spent fuel with cooling times
on the order of three to five years.
Predictive Modeling
1. Samples
More detailed sample and preparation details are provided by Orton elsewhere [9]. The samples used in
this experiment were created from Boiling Water Reactor (BWR) spent fuel rod segments from different
nuclear reactors: ATM 105 segment A and a segment from ATM 109. Segments ATM 105 A and ATM 109
are Approved Testing Material (ATM) for the U.S. Department of Energy Office of Civilian Radioactive
Waste Management Program. Fuel dissolution and sample preparation was performed at Pacific Northwest
National Laboratory (PNNL).
ATM 105 A was created using fuel irradiated at the Cooper Nuclear Power Plant (NPP) in the Nebraska
public power district. The fuel rod was fabricated by General Electric (GE) in 1972 and was bundled for use
in a 7×7 assembly. The fuel was 2.94% enriched and irradiated for various periods between July 1974 and
May 1982 with a burnup range between 15MWd/kgU - 35MWd/kgU. The rod segment used to create
sample ATM 105 A was located approximately 52cm - 56cm from the top of the rod and had an
approximate burnup of 15.5MWd/kgU - 17.5MWd/kgU. As of May 2012, the fuel has been cooling for
approximately 30 years.
4. 4
Sample ATM 109 was created using fuel irradiated in reactor I at the Quad Cities NPP. The fuel was
fabricated by GE for use in 7×7 assemblies. The fuel was 3.0% enriched and irradiated between February
1979 and September 1987 and again between November 1989 and September 1992 with an approximate
burnup of 67MWd/kgU - 70MWd/kgU. As of July 2012, the fuel has been cooling for approximately 20
years. The total activity of each sample upon arrival at the University of Texas NETL was approximately
0.1µCi in 2010.
2. ORIGEN-ARP
ORIGEN-ARP was used to calculate nuclide-specific activity and mass compositions for the spent
fuel samples given initial conditions such as fuel type and burnup. Calculations for both ATM 105 A and
ATM 109 samples were performed using ORIGEN-ARP Express Mode using the parameters listed in
Table 1. Burnup values were taken to be the burnup range mid-values. Moderator densities were chosen
following the recommendations of Orton [6]. Model parameters specific to ATM 105 A and ATM 109 are
listed in Table 2.
Table 1: Input parameters for both sample ATM 105 A and sample ATM 109
for use with ORIGEN-ARP Express Mode.
Irradiation Cycles 3
Power History [%] 85
Average Power [MWd/MTU] 23
Fuel Type GE 7×7
Uranium Quantity [g]
1×10
6
Table 2: Unique input parameters for samples ATM 105 A and ATM 109
for use with ORIGEN-ARP Express Mode.
ATM105A ATM109
Enrichment [%] 2.94 3.00
Burnup [MWd/MTU] 16500 68500
Moderator Density [g/cc] 0.3 0.5
Activity Calculation 28 18
Cooling Time [a]
Mass Calculation 30 20
Cooling Time [a]
ORIGEN-ARP was first used to estimate the specific activity of the spent fuel samples at the time of
shipment to the NETL (2 years ago) when each sample had an activity of about 0.1µCi. The activities of one
gram of ATM 109 and ATM 105 A were calculated for cooling times of 18 and 28 years, respectively, using
the aforementioned parameters. Sample activity values from ORIGEN-ARP calculations were then scaled
to 0.1µCi to obtain scaling factors.
5. 5
ORIGEN-ARP was also used to obtain present-day mass compositions for one gram of ATM 109 and
one gram of ATM 105 A using cooling times of 20 and 30 years, respectively. Mass values were converted
to number densities and the number densities were multiplied by the scaling factors found from the activity
calculations to obtain approximations for the nuclide number densities in the samples.
Number densities were used to calculate activation activities at the end of irradiation using the
expression:
Aj
=Ni
φσi
(1−e−λj
tirr.) (1)
where φ is the average neutron flux during irradiation; Ni
and σi
are the number density and radiative
capture cross section of the unactivated analyte, respectively; tirr.
is the irradiation time; and λj
and Aj
are
the decay constant and activity of the activation product, respectively.
3. Nuclide Analysis
Mass composition lists of the top 200 nuclides in ATM 105 A and ATM 109 were pared to identify the
nuclides with the highest probability of detection when undergoing NAA. Composition lists were initially
reduced by eliminating gaseous nuclides and eliminating nuclides whose activation products are stable, give
off no γ-rays, or give off no γ-rays with intensities greater than 1%. Gaseous nuclides were eliminated as the
samples were not confined in sealed ampules. As such, observation of gaseous nuclides would be less likely.
Nuclides whose activation products are stable, emit no γ-radiation, or emit γ-radiation with intensity less
than 1% were eliminated as they give off little or no signal for detection with NAA. The first filtration pared
each composition list by approximately 50%.
Filtration 2 was created with the intention of identifying adequately detectable nuclides. The goal was to
incorportate all radiative capture and γ-emission characteristics to identify nuclides possessing the highest
probability of both absorbing neutrons and emitting γ-radiation at a detectable rate. A discriminating
criterion was formulated using the saturation activities (the maximum activity of the activation product) and
maximum γ-ray intensities of the activation product nuclides. Saturation activities were calculated by
allowing the irradiation time (tirr.
) in Expression 1 to approach infinity. Radiative capture cross sections and
γ-intensities were obtained from [7,8,9].
The γ-ray emission rate threshold was chosen to be 500 γs/second although the value is arbitrary and
variable for different situations. For our survey purposes, provided a detector with 1% - 3% efficiency, this
would allow for a viable detection rate of 5 - 15 γs/sec1
. The theoretical saturation activity was used as it
1
This neglects efficiency is a function of energy and optimization of signal to noise ratio
6. 6
would eliminate all nuclides that would not reach adequate detection levels (≥500γs/sec) even if irradiated
for a very long time.
The criterion used for filtration 2 is expressed as:
(Ni
σi
φ)γmax
=Asat.
γmax
≥500γs/sec (2)
where Ni
and σi
are the number density and radiative capture cross section of the parent nuclide,
respectively; φ is the average neutron flux during irradiation; Asat.
is the saturation activity; and γmax
is the
highest intensity of the activation product γ-ray(s).
Gamma intensities from both meta-stable and non-metastable activation products were used.
Meta-stable γ-ray intensities were only used when meta-stable activation products possessed half-lives in
the range of minutes to a few years. Having eliminated all but the potentially observable nuclides, model
results were dependent solely on experimental parameters such as spent fuel quantity, irradiation time, etc.
The full list of potentially detectable nuclides is given in the appendix.
4. Model Results
Lists of the most detectable nuclides in one milligram (mg) of spent fuel with characteristics similar to
that of ATM 105 A and ATM 109 were compiled for a fixed irradiation time of one hour. Irradiation time
was fixed such that results could be presented in terms of specific activities (nuclide γ-emission rates per mg
of spent fuel) for application to various mass quantites of spent fuel. Such a presentation is useful as spent
fuel mass is often limited by laboratory/facility regulations. Spent fuel characteristic of ATM 105 A and
ATM 109 would correspond to burnup values and cooling times given in Table 2.
Both thermal and epithermal neutron fluxes were considered when generating results. In the cases when
epithermal neutron fluxes were considered, radiative capture cross sections were replaced with resonance
integral values. Nuclides with the highest detection probabilities (≥500 γs/sec) for spent fuel characteristic
of ATM 105 A and ATM 109 (Burnup: 16500 MWd/MTU, 68500 MWd/MTU; Cooling: 30 yr, 20yr,
respectively) irradiated for one hour are presented in Tables 3 and 4 alongside their corresponding
theoretical γ-emission rates determined from Expression 1.
7. 7
Table 3: Theoretical γ-ray emission rates per nuclide per milligram of ATM 105 A spent fuel with burn-up of
approximately 16500 MWd/MTU and cooling time of approximately 30 years after one hour irradiation. Cells marked
with a (×) indicate predicted γ-emission rate is less than 500 γs/sec
Detectable (Unactivated) γ-Emission Rate of Activation γ-Emission Rate of Activation
Analyte Product (φ
th.
) [γ/sec/mg SNF] Product (φ
epi
) [γ/sec/mg SNF]
238
U 6.7×10
5
2.0×10
5
103
Rh 2.7×10
5
7.8×10
4
109
Ag 3.4×10
4
2.2×10
4
99
Tc 8.4×10
3
5.4×10
3
115
In 2.3×10
3
1.5×10
3
129
I 1.2×10
3 ×
139
La 9.9×10
2 ×
154
Sm 7.7×10
2 ×
152
Sm 6.3×10
2 ×
242
Pu
×
8.4×10
2
Table 4: Theoretical γ-ray emission rates per nuclide per milligram of ATM 109 spent fuel with burn-up of
approximately 68500 MWd/MTU and cooling time of approx 20 years after one hour irradiation. Cells marked
with a (×) indicate predicted γ-emission rate is less than 500 γs/sec.
Detectable (Unactivated) γ-Emission Rate of Activation γ-Emission Rate of Activation
Analyte Product (φ
th.
) [γ/sec/mg SNF] Product (φ
epi
) [γ/sec/mg SNF]
238
U 6.4×10
6
1.9×10
5
103
Rh 7.2×10
5
2.0×10
5
109
Ag 2.3×10
5
1.5×10
5
99
Tc 2.8×10
4
1.8×10
4
242
Pu 6.9×10
3
1.7×10
4
129
I 5.7×10
3 ×
154
Sm 4.9×10
3
8.4×10
2
115
In 4.5×10
3
2.9×10
3
138
Ba 9.6×10
2 ×
139
La 3.7×10
3 ×
137
Cs 2.2×10
3 ×
152
Sm 2.0×10
3
1.1×10
3
150
Nd 1.8×10
3
9.6×10
2
148
Nd 1.6×10
3 ×
108
Pd 1.5×10
3
1.8×10
3
130
Te 1.4×10
3 ×
100
Mo 1.0×10
3
8.2×10
2
127
I 7.8×10
2
7.4×10
2
8. 8
Milligram quantities of ATM 105 A and ATM 109 spent fuel correspond to activities of approximately
0.1mCi and 0.5mCi, respectively—significantly higher than the activities of the NETL samples. To more
accurately model the samples, results were generated for variable spent fuel activities as opposed to variable
spent fuel mass via manipulation of the scaling factors. Potentially detectable nuclides in samples
ATM 105 A and ATM 109 are presented as a function of sample activity for both thermal and epithermal
fluxes in Tables 5, 6, 7, and 8. Sample activity is an analog for the amount of spent fuel.
Table 5: Model prediction for detectable nuclides as a function of sample activity for
ATM 105 A irradiated in a thermal neutron flux (2.5×1012
n/cm2
/sec). Times
correspond to irradiation times.
Total Sample Activity Detectable (Unactivated) Analytes t500
Prior to NAA (µCi) (approx.)
0.1 238U 30 min
(NETL sample)
10 238U, 103Rh,109Ag 1 min
50 238U, 99Tc,103Rh109Ag, 115In 30 sec
139La, 129I 2 hrs
100 238U, 99Tc,103Rh109Ag, 115In 30 sec
139La, 129I 1 hr
152Sm, 154Sm 2 hrs
Table 6: Model prediction for detectable nuclides as a function of sample activity for
ATM 105 A irradiated in an epithermal neutron flux (1.0×10
11
n/cm
2
/sec). Times
correspond to irradiation times.
Total Sample Activity Detectable (Unactivated) Analytes t
500
Prior to NAA (µCi) (approx.)
0.1 none -
(NETL sample)
10 103
Rh,
109
Ag
30 sec
238
U
5 min
50 238
U,
99
Tc,
103
Rh
109
Ag
30 sec
100 238
U,
99
Tc,
103
Rh
109
Ag,
115
In
30 sec
242
Pu
2 hrs
9. 9
Table 7: Model prediction for detectable nuclides as a function of sample activity for
ATM 109 irradiated in a thermal neutron flux (2.5×10
12
n/cm
2
/sec). Times correspond
to irradiation times.
Total Sample Activity Detectable (Unactivated) Analytes t
500
Prior to NAA (µCi) (approx.)
0.1 238
U
30 min
(NETL sample)
10 238
U,
103
Rh,
109
Ag
30 sec
50 238
U,
99
Tc,
103
Rh
109
Ag
30 sec
100 238
U,
99
Tc,
103
Rh
109
Ag
30 sec
Table 8: Model prediction for detectable nuclides as a function of sample
activity for ATM 109 irradiated in an epithermal neutron flux (1.0×10
11
n/cm
2
/sec). Times correspond to irradiation times.
Total Sample Activity Detectable (Unactivated) Analytes t
500
Prior to NAA (µCi) (approx.)
0.1 none -
(NETL sample)
10 103
Rh,
109
Ag
30 sec
238
U
10 min
50 99
Tc,
103
Rh
109
Ag
30 sec
238
U
5 min
242
Pu
30 min
100 99
Tc,
103
Rh
109
Ag
30 sec
238
U
5 min
242
Pu
30 min
In addition to varying sample activity, irradiation time was allowed to vary in an attempt to maximize
the yield of easily detectable nuclides. Nuclides require variable amounts of irradiation time to approach
saturation activities as a result of variable decay constants (see Expression 1). Similarly, variable amounts
of irradiation time are required for different nuclides to reach an emission rate of 500γs/sec. The
approximate irradiation time required for a nuclide(s) to reach a γ-emission rate of 500γs/sec is indicated by
t500
in Tables 5, 6, 7, and 8.
To calculate t500
values, we began with Expression 2 and added an additional decay term that accounts
for the decay of the activation product during irradiation:
10. 10
Asat.
γmax
(1−e−λj
t
)=500γs/sec (3)
We then solved for the minimum irradiation time needed to obtain a γ-emission rate of 500γs/sec:
t500
=−
log(1−
500
Asat.
γmax
)
λj
(4)
It is important to note that the goal was to make the irradiation time as small as possible to prevent the
occurance of fresh fissions caused by trace amounts of fissionable material in the spent fuel samples.
Fissioning would lead to the generation of fission products which, given a long enough irradiation time, may
undergo neutron activation. Activation of fresh fission products would make determination of the original
composition more difficult as it could: (a) augment the background which would weaken detection limits,
(b) introduce interferences, and/or (c) lead to overestimation of the original composition if fission products
produce nuclides already present in the sample.
Experiment
1. Parameters
ATM 109 was irradiated during two trials in an attempt to validate model predictions. Thermal
NAA was performed using both cyclic and conventional pneumatic irradiation methods with a power level
of 950kW (whole flux of approximately 2.7×10 12
n/sec/cm 2
). Counting was performed using a
Compton-suppression HPGe detector system. In the first trial, 0.5mL of ATM 109 was irradiated using a
cyclic irradiation method. The sample was pneumatically inserted into the reactor and the sample was
continuously irradiated and counted for 3 cycles using an irradiation time of approximately 10 seconds, a
shuttle/decay time of 10 seconds, and a count time of 30 seconds. In the second trial, 0.5mL of ATM 109
was irradiated once using a pneumatic method. Trial 2 had an irradiation time of approximately 10 minutes
and a decay time of approximately 5 minutes. Eleven spectra were obtained for analysis of trial 2. Seven 10
minute counts were recorded immediately after the decay time and a one hour count was recorded thereafter.
The sample was also counted for one hour each subsequent day for three days.
2. Results
Analysis of neutron activation results is currently an ongoing process. NAA of ATM 105 A indicated a
uranium content of approximately 5µg/mL therefore, fresh fission products were expected to be present
during trial 1 and especially trial 2 when the sample was irradiated for 10 minutes. Fresh fission products
were of concern mostly for the interference they produce among spectra i.e. interference caused by fission
11. 11
product γ-lines. Analysis was therefore begun with spectra possessing the longest decay times (one hour
count times) and analysis will continue in reverse order. Such an approach is beneficial as it allows one to
deconstruct the earlier, more complex spectra (shorter decay times) via information deduced from the later,
less populated spectra (longer decay times).
As previously mentioned, spectral analysis is an ongoing process. As such, only results from select
spectra are presented. Probable nuclides from trial 2 with approximately 75 minutes of decay time are listed
in Table 9 alongside the peak energies used to characterize them.
Table 9: Detectable nuclides from sample ATM 109 after a thermal irradiation time of approximately 10 minutes, decay time of
approximately 75 minutes, and count time of one hour. Nuclides marked with (*) indicate the nuclide only has one peak with
intensity greater than 1%. A shaded cell indicates the nuclide corresponds to one of the predicted analytes from Table 4 and/or
Table A.1. Intensity values were rounded to the nearest whole integer
Observed Nuclide Peak Energies Used for Characterization [keV] (Intensity [%])
239
Np
99.525 (14), 103.374 (22), 106.123 (26), 228.183 (11), 277.599 (14)
123
Xe
28.317 (21), 28.612 (38), 148.9 (49)
137
Cs
31.817 (2), 32.194 (4), 661.657 (85)
56
Mn
846.7638 (99), 1810.726 (27), 2113.092 (14)
38
Cl
1642.43 (33), 2167.54 (44)
24
Na
1368.626 (100), 2754.007 (99)
41
Ar*
1293.64 (99)
42
K*
1524.6 (18)
128
I
442.901 (13), 526.557 (1)
82
Br
554.348 (71), 619.106 (44), 776.517 (83), 1474.88 (17)
139
Ce
33.442 (41), 165.8575 (80)
239
U
13.9 (14), 74.664 (49)
110m
Ag
657.76 (94), 884.6781 (73), 1384.2931 (25)
139
Ba
33.442 (3), 165.8575 (24)
138
Cs
462.796 (31), 1435.86 (76)
131
Te
149.716 (69), 452.323 (18)
128
I
442.901 (13), 526.557 (1)
Table 9: Detectable nuclides from sample ATM 109 after a thermal irradiation time of approximately 10
minutes, decay time of approximately 75 minutes, and count time of one hour. Nuclides marked with (*) indicate
the nuclide only has one peak with intensity greater than 1%. A shaded cell indicates the nuclide corresponds to
one of the predicted analytes from Table 4 and/or Table A.1. Intensity values were rounded to the nearest whole
integer.
Discussion
Data obtained from activity variation calculations (Tables 5, 6, 7, 8) accentuated the strong correlation
between sample activity and the quantity of easily observable nuclides. Reverting to Expressions 2 and 3, it
12. 12
became evident that the low number of detectable nuclides was the direct result of low nuclide number
densities as all other variables such as irradiation time and neutron flux were constant for all nuclides.
Number densities were estimated using the scaling factors obtained via activity calculations therefore as
sample activity was increased, number densities were also increased, and vice versa.
Model predictions for both ATM 105 A and ATM 109 were disappointing (regardless of variation of
experimental parameters) due to such low sample concentrations. The decision was thus made to irradiate
ATM 109 using a cyclic method to detect short-lived activation products and again using a longer
irradiation time to detect nuclides with slower activation rates. Results from the cyclic irradiation (trial 1)
matched predictions quite accurately as the only verifiable nuclide was 239
U (via 74.664keV(49)) which
corresponds to the analyte 238
U. Additional peaks were present throughout the spectrum but they exhibited
amplitudes too small for accurate identification of more nuclides.
Trial 2 spectra yielded more prominent peak amplitudes which permitted the identification of more
nuclides. Results from Table 9 are not yet final as further analysis (i.e. ratio analysis) is required to confirm
the presence of nuclides such as 139
Ce and 139
Ba which share identical γ-emission energies. Ratio analysis
would involve the comparison of peak area to γ-intensity for various γ-emission energies of a given nuclide.
The method relies on the assumption that the peak areas are (approximately) in the same ratio as the
γ-intensities corresponding to the peaks. Moreover, nuclides such as 41
Ar and 38
Cl were present in
relatively high quantities although it now appears that 56
Mn, 38
Cl, 24
Na, 41
Ar, and 42
K correspond to
non-spent fuel agents such as human sweat, vial plastic, and air [10] as well as impurities in anyof the
original chemicals used in the separation techniques.
Summary
Although NAA appears a viable analysis technique for the detection of stable and long lived
nuclides such as 238
U, 99
Tc, and 109
Ag in spent nuclear fuel, NAA results seem limited by sample mass
and activity. To further validate neutron activation as an effective analysis technique for spent nuclear fuel,
spent fuel samples of higher concentration are needed for experiment. Such samples would have higher
activities than the samples used to date which would also permit us to gauge the limits of NAA as a spent
fuel analysis technique as detector saturation becomes an issue following drastic increases in sample
activity. Future work might involve analysis of a broader range of spent fuel burnup and cooling time values
as well as more variation in irradiation and decay times to determine their effects on NAA results.
13. 13
Bibliography
[1] Nuclear Science Committee, WPNCS, EGADSNF, Spent Nuclear Fuel Assay Data for Isotopic
Validation
[2] R.E. Naegeli, Calculation of the Radionuclides in PWR Spent Fuel Samples for SFR Experiment
Planning
[3] I. Gauld, M. Francis, Investigation of Passive Gamma Spectroscopy to Verify Spent Nuclear Fuel
Content
[4] P. Bode, Opportunities for Innovation in Neutron Activation Analysis
[5] R.R. Greenberg, P. Bode, E.A. De Nadai Fernandes, Neutron Activation Analysis: A Primary
Method of Measurement
[6] R. Orton, The Multi-Isotope Process Monitor: Non-destructive, Near-Real-Time Nuclear
Safeguards Monitoring at a Reprocessing Facility
[7] National Nuclear Data Center, information extracted from the Chart of Nuclides database,
http://www.nndc.bnl.gov/chart/
[8] Korea Atomic Energy Research Institute, information extracted from the Table of Nuclides,
http://atom.kaeri.re.kr/ton/nuc8.html
[9] Japan Atomic Energy Agency, information extracted from the Tables of Nuclear Data,
http://wwwndc.jaea.go.jp/NuC/index.html
[10] C. Dresser, C. Henson, J. Mock, et al., Neutron Activation Analysis, A Titanium Material Study
![2
Abstract
A method for the identification of observable radionuclides from neutron activation analysis of spent
nuclear fuel was investigated. A predictive model was formulated using ORIGEN-ARP and nuclear decay
data to predict neutron activation analysis results of two spent nuclear fuel samples with variable burnup
values and cooling times. Model predictions were tested by performing thermal instrumental neutron
activation analysis on the spent nuclear fuel samples using both cyclic and conventional irradiation
methods. Preliminary results indicate neutron activation analysis was successful in identifying several
stable and long-lived radionuclides predicted via model calculations but results appear limited to sample
concentration. Spent nuclear fuel samples of higher specific activity are needed to further validate model
results.
Introduction
Determination of radionuclide composition has proven a vital component to processes such as safety
evaluations of spent nuclear fuel systems and experimental planning [1,2]. At present, material accounting
systems of spent fuel at nuclear facilities rely heavily on the use of burn-up codes. Although continuous
development of computational methodologies for nuclear fuel analysis and revision of nuclear data have
improved accuracy of burnup calculations [1], experimental code verification remains an important step
towards the establishment of improved safety bases. Experiments of this type are often performed with
destructive analysis techniques such as mass spectrometry.
Spent fuel samples may be destructively examined by means of radiochemical analysis methods when
high accuracy measurements of isotropic compositions are desired [1]. Destructive analysis typically
involves extensive sample preparation that may involve dissolution, chemical separation, and vaporization
[1]. Although an effective method, destructive analysis is associated with long analysis times and high costs
[3]. Optimism currently lies with more conventional non-destructive analysis techniques such as passive
gamma counting.
Passive methods have proven for years to be both effective and reliable techniques for the determination
of radionuclide composition. Passive gamma-ray spectrometry systems have the benefit of being readily
available, relatively small, and providing real-time data [3]. Nonetheless, gamma-ray spectrometry relies on
the gamma emissions of the analyte thus, problems inevitably arise with the characterization of stable and
long-lived radionuclides whose very slow to non-existant decay rates cause difficulties in passive counting.
Neutron activation analysis (NAA) is an attractive technique to circumvent such complications as it
boasts non-destructive analysis independent of both sample size and activity [4]. In addition, high](data:image/gif;base64,R0lGODlhAQABAIAAAAAAAP///yH5BAEAAAAALAAAAAABAAEAAAIBRAA7)