ENVIRONMENTAL PHYSICS
CANDU RADIOTOXICITY INVENTORIES ESTIMATION –
A CALCULATED EXPERIMENT CROSS-CHECK
FOR DATA VERIFICATION AND VALIDATION
ALEXANDRU OCTAVIAN PAVELESCU, DAN GABRIEL GEPRAGA

University Politehnica of Bucharest, Bucharest, Romania, [email protected]
 ENEA FIS-MET, Bologna, Italy, [email protected]
Received October 4, 2006
This paper is related to the Clearance Potential Index, Ingestion and Inhalation
Hazard Factors of the nuclear spent fuel and radioactive wastes. This study required a
complex activity that consisted of various phases such us: the acquisition, setting up,
validation and application of procedures, codes and libraries.
The paper reflects the validation phase of this study. Its objective was to
compare the measured inventories of selected actinide and fission products
radionuclides in an element from a Pickering CANDU reactor with inventories
predicted using a recent version of the ORIGEN-ARP from SCALE 5 coupled with
the time dependent cross sections library, CANDU 28.lib, produced by the sequence
SAS2H of SCALE 4.4a.
In this way, the procedures, codes and libraries for the characterization of
radioactive material in terms of radioactive inventories, clearance, and biological
hazard factors are being qualified and validated, in support for the safety
management of the radioactive wastes.
1. INTRODUCTION
The environmental and safety assessment of the concept of geological
disposal of used CANDU UO2 fuel waste in plutonic rock involves a detailed
analysis of the pathways for radionuclides released from used fuel into ground
waters present in the disposal vault, as well as the transport of these nuclides
through the geosphere to the biosphere. These analyses provide the information
required to estimate the radiological dose consequences and risks to man and the
environment from such a disposal facility.
To perform this safety assessment it is necessary to have a detailed
knowledge of the radionuclide content in the used fuel. The radionuclide
inventories have been calculated using the radionuclide generation and depletion
code ORIGEN [1, 12] for a type of Canada Deuterium Uranium Reactor
(CANDU) [1, 4]. These calculated inventories can provide the source-term data
for other safety assessment code, such us the German GRS code, EMOS, or the
Canadian SYVAC (Systems Variability Analysis Code) [1].
Rom. Journ. Phys., Vol. 52, Nos. 1– 2 , P. 137–148, Bucharest, 2007
138
A. O. Pavelescu, D. G. Gepraga
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ORIGEN-S code was publicly released as part of the SCALE (Standardized
Computer Analyses for Licensing Evaluation) modular code system that was
developed at the Oak Ridge National Laboratory (ORNL).
Knowing the detailed radionuclide content in the used fuel, it is possible to
have a broader vision regarding the radiotoxicity of the spent fuel. In this
context, there are two major approaches:
– IAEA clearance potential index concept
– US radioactivity concentration guides (in terms of ingestion hazard
factor in m3 of water, respectively inhalation hazard factor in m 3 of air).
2. CONCEPTS DESCRIPTION
In this field it is very useful to make a comparison between the two existing
safety concepts, which are based on different approaches:
• The AIEA approach is related on Safety Guide RS-G-1.7 – 2004 [2] that uses
the concept of “clearance”. The clearance is intended to indicate which material
under regulatory control can be removed from this control. Clearance is defined
as the removal of radioactive materials or radioactive objects within authorized
practices from any further regulatory control by the regulatory body.
In summary the International Basic Safety Standards for Protection against
Ionizing Radiation and for the Safety of Radiation Sources (BSS) was
provided radiological criteria to serve as a basis for the derivation of the
clearance levels. Specific activity values in Becquerel/g or Becquerel/kg (or
clearance levels) were introduced for use in making decisions on the
exemption of bulk radioactive materials [11]. Further, the clearance potential
index is defined as ratio between specific radioactivity and clearance level, the
state in which the radioactivity material became undangerous. So, the
clearance potential index is an undimensional quantity, which, if greater than
unit, shows the measure of radiotoxicity for the radioactive material and when
is equal to unit or less means that the radioactive material became clean in the
sense that can be manipulated without restrictions.
• The US approach, described in the Code of Federal Regulations. The code
provides in Part 10, Title 20 (10CFR20 – 1991) [3] the radioactivity
concentration guides (RCG) for continuous ingestion (from water) and
inhalation (from air) in unrestricted areas, in units of curies per cubic meter
(Ci/m3). The RCG values specify the maximum permissible concentrations of
an isotope in soluble and insoluble forms, for both ingestion and inhalation,
and for occupational and unrestricted exposure. When the activity (in curies)
of a given isotope is divided by the radioactivity concentration guides for that
isotope, the result is the volume of water (or air) required to dilute that
3
CANDU radiotoxicity inventories estimation
139
quantity of the isotope to its maximum permissible concentration. The dilution
volume is a measure of the radioactive toxicity of the nuclide for cases of
direct ingestion or inhalation and is known as ingestion or inhalation hazard
factor.
3. EXPERIMENTAL DESCRIPTION AND MEASUREMENTS
(METHODS AND DATA)
The fuel chosen for this study was a Zircaloy-4 clad, 28-element bundle
(Fig. 3.1) that resided in one fuel channel from the central core region of
Pickering Nuclear Generating Station-A [5]. The relatively constant power
history provided that the modeling of the history could be accomplished with
more certainty.
Fig. 3.1 – Pickering CANDU 28 Element Fuel Bundle.
The following experimental data is taken from reference [5]:
– Gamma Scan
High resolution axial gamma scans were performed along the longitudinal
distribution of gamma emitting fission products. The revealed activity (gamma
scan) is due mostly to 134Cs and 137Cs.
140
A. O. Pavelescu, D. G. Gepraga
4
– Chemical Analysis of used fuel
Chemical analysis for actinides and fission products was performed on the
samples from the middle of the element. The fuel element was cut at the
interpellet location (determined from the gamma scan) with a tube-cutting
device, and then the processing continued with the fuel dissolution.
– Radiochemical Analysis has been performed for the following elements:
– Uranium. The uranium was determined by potentiometric titration with
standard potassium dichromate; the uranium isotopic composition was
determined by thermal ionization (TI) mass spectrometry
– Neptunium. Neptunium was determined by alpha spectrometry
– Plutonium. Isotopic analysis was performed by TI mass spectroscopy.
– Cesium, antimony, cobalt and europium. The dissolver sample was
diluted with nitric acid an analyzed for 137Cs, 134Cs, 125Sb, 60Co, 155Eu and
155Eu, by gamma spectroscopy.
– Strontium. The sample was counted by LSC (liquid scintillation counting).
All other experimental data are also found in the reference [5].
4. CALCULATION: RADIONUCLIDES INVENTORIES ESTIMATION
WITH ORIGEN
For the evaluation of spent fuel inventories of CANDU experimental
element using ORIGEN procedures we applied the following steps:
1. Fuel irradiation analysis to produce time-dependent cross sections for
CANDU 28 library with SAS2H module.
2. Burnup and decay analysis with ORIGEN: input modeling and processing of inventory results.
4.1. FUEL IRRADIATION ANALYSIS TO PRODUCE TIME-DEPENDENT CROSS
SECTIONS FOR CANDU 28 WITH SAS2H
Overview of SAS2H Modules
Five different codes plus several routines in the SCALE subroutine library
are utilized by the SAS2H control module to produce time dependent crosssections. (See the computational flow chart presented Fig. 4.1.)
The basic functions of the five functional modules, as applied by SAS2 are
described below [6]:
BONAMI applies the Bondarenko method of resonance self-shielding for
nuclides that have Bondarenko data included with their cross sections.
NITAWL-II performs the Nordheim resonance self-shielding corrections
for nuclides that have resonance parameters included with their cross sections.
5
CANDU radiotoxicity inventories estimation
141
Fig. 4.1 – Computational flow chart for the SAS2H procedure to produce the cross-sections
time-depending library [6].
XSDRNPM performs a 1-D discrete-ordinates transport calculation based
on various specified geometries requested in the data supplied by SAS2H. The
code, as applied by SAS2H, has three functions. The first of these functions, to
produce cell-weighted cross sections for fuel depletion calculations is of particular
interest for this paper.
BONAMI, NITAWL and XSDRNPM modules use the data described in
Table 4.1.
COUPLE updates the cross-section constants included on an ORIGEN-S
nuclear data library with data from the cell-weighted cross-section library
produced by XSDRNPM. Also, the weighting spectrum computed by
XSDRNPM is applied to update all nuclides in the ORIGEN-S library that were
not specified in the XSDRNPM analysis.
ORIGEN-S performs in this case only nuclide generation and depletion
calculations for the specified reactor fuel history. Also, the code can compute the
neutron and gamma sources generated by the fuel assembly.
The in-reactor geometry for the test element in the fuel bundle was
rigorously modeled (see Table 4.1 and 4.2) and the known power history of the
fuel was used to model the burnup (see Table 4.3)
From the above table the characteristics of these CANDU 28 libraries are:
– CANDU 28-element fuel assembly ORIGEN-S binary working library [8];
142
A. O. Pavelescu, D. G. Gepraga
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Table 4.1
Specifications for the Pickering fuel experiment. Neutron – Depletion
Analysis to produce ORIGEN libraries [7]
Fuel material
Initial uranium compositions (weight percent)
234U
235U
238U
Fuel density
Fuel temperature
Element radius
Number of fuel pins
Inner fuel ring radius (4)
Middle fuel ring radius (8)
Outer fuel ring radius (16)
Cladding material
Cladding radius
Pressure tube
Inner radius
Outer radius
Calandria tube
Inner radius
Outer radius
Coolant
Atom purity
Density
Temperature
Moderator
Atom purity
Density
Temperature
Fuel channel square pitch
Exit outer element burnup (measured)
Cooling time in days
Natural UO2
0.0054
0.7110
99.2836
10.6 g/cm3
1003 K
0.7120 cm
28
1.175 cm
2.685 cm
4.229 cm
Zircaloy-4
0.7605 cm
Zr-Nb 2.5%
5.1815 cm
5.6965 cm
Zircaloy
6.5405 cm
6.6955 cm
D2O
99.75%
0.8445 g/cm3
545 K
D2O
99.91%
1.0838 g/cm3
336 K
28.575 cm
221 MWh/kgU
1162
– Modified card-image ORIGEN-S libraries of Scale 4.4a with SAS2H procedure;
– The light element, actinide, and fission product libraries data, including
gamma and total energy, are from ENDF/B-VI;
– Include beta-delayed neutron decay data;
– Fission product yields are from ENDF/B-V libraries and use an 18-energygroup structure;
7
CANDU radiotoxicity inventories estimation
143
– Photon data are from the master photon data base, to include bremsstrahlung
from UO2 matrix;
– The burnup-dependent cross sections were generated using eight burnup
intervals that extend up to a burnup of approximately 12000 MWd/MgU.
– The cross sections for each interval are stored in positions on the ORIGEN-S
library and are accessed by referencing the position number. The burnup
values associated with each position are listed below – see Table 4.2. They are
the same for both the 37-element and 28-element libraries.
Table 4.2
The burnup values associated with each position
Library Position
Burnup [MWd/MgU]
1
2
3
4
5
6
7
8
240
720
1440
2880
4800
6720
8640
10560
4.2. BURNUP AND DECAY ANALYSIS WITH ORIGEN: INPUT MODELING
AND PROCESSING OF INVENTORY RESULTS
Modeling Conditions:
The modeling of the irradiation phase was divided into six time segments,
using an averaged power over each time segment. (Table 4.3)
Table 4.3
Specifications for the Pickering – Fuel Element Irradiation History [7]
Operating Cycle
Power [kW/kgU]
Irradiation [days]
1
2
3
4
5
6
31.4
28.9
31.1
29.5
28.7
26.7
31.33
15.38
66.67
66.67
64.55
70.78
A burnup of 221 MWh/kg U, referring to the final irradiation period, was
used in the ORIGEN-ARP calculation (see the irradiation history presented in
Table 4.3).
144
A. O. Pavelescu, D. G. Gepraga
8
Other input modeling data:
– Initial Charge = 1 kg U (0.0054% U234 + 0.711% U235+99.2836% U238) &
134 g Oxy & 1kg Zircaloy-4
– Irradiation Power = 2.938 E–02 MW
– Burnup = 9.2150 E+00 MWd/kg U = 221.16 MWh/kg U
– Total neutron flux = 1.16 E+14 n/cm 2* sec
– The isotopic analyses were all decay corrected to 5590 days (15.3 years)
following the fuel discharge. All ORIGEN-ARP inventories were also
decayed for the equivalent period following the burn-up cycle.
5. RESULTS
The experimental and calculated results for CANDU 28 test element are
given in the following tables (Tables 5.1, Table 5.2, and Table 5.3)
Table 5.1
Comparison between calculated and measured values of major actinides concentration (expressed
in grams), after the cooling time of 5590 days
Nuclide
Experimental Data Measurements
Concentration [g/kg U]
u234
u235
u236
u238
pu238
pu239
pu240
pu241
pu242
0.0339
1.6400
0.8020
985.3000
0.0058
2.6900
1.2200
0.1340
0.0940
Uncertainty [%]
ORIGEN Calculation
Concentration [g/kg U]
(C–E)/E*100
[%]
+/– 55
+/– 2.4
+/– 3.7
+/– 0.01
+/– 5.6
+/– 2.5
+/– 3.7
+/– 9.0
+/– 6.8
4.249E–02
1.694E+00
8.230E–01
9.835E+02
5.847E–03
2.749E+00
1.282E+00
1.273E–01
9.774E–02
20.22
3.19
2.55
–0.18
0.80
2.15
4.84
–5.26
3.83
Table 5.2
Comparison between calculated and measured values of selected actinides radioactivity,
after the cooling time of 5590 days
Nuclide
np237
am241
cm244
Experimental Data Measurements
Specific Radioactivity
[Bq/kg U]
Uncertainty
[%]
9.95E+05
1.86E+10
7.12E+08
+/– 20
+/– 20
+/– 15
ORIGEN Calculation
Specific Radioactivity
(C–E)/E*100
[Bq/kg U]
[%]
9.731E+05
1.948E+10
7.951E+08
–2.20
4.71
11.68
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CANDU radiotoxicity inventories estimation
145
Table 5.3
Comparison between calculated and measured values of major fission products radioactivity, after
the cooling time of 5590 days
Nuclide
h3
Experimental Data Measurements
Specific Radioactivity
[Bq/kg U]
Uncertainty
[%]
2.07E+09
+/– 7
ORIGEN Calculation
Specific Radioactivity
(C–E)/E*100
[Bq/kg U]
[%]
2.121E+09
2.46
sr 90
4.86E+11
+/– 4
4.980E+11
2.47
tc 99
1.08E+08
+/– 10
1.510E+08
39.85
ru106
8.72E+07
+/– 5
1.446E+08
65.78
sb125
2.20E+09
+/– 18
1.842E+09
–16.28
i129
2.44E+05
+/– 0
3.553E+05
45.62
cs134
4.16E+09
+/– 7
2.837E+09
–31.80
cs137
8.05E+11
+/– 5
7.781E+11
–3.34
eu154
8.14E+09
+/– 5
8.436E+09
3.64
eu155
3.35E+09
+/– 8
2.548E+09
–23.95
The calculated clearance index [3, 9], ingestion hazard factor [10] and
inhalation hazard factor [10] of major actinides and fission products, after the
cooling time of 5590 days, are given in Table 5.4 and Table 5.5.
Table 5.4
Clearance potential index, ingestion hazard factor and inhalation hazard factor of major actinides
after the cooling time of 5590 days
Origen ARP
Specific Radioactivity
Bq/kg U
Clearance
Potential Index
Specific Rad./Clear. Lev.
Ingestion
Hazard Factor
m3 H2O
Inhalation
Hazard Factor
m3 air
u234
9.779E+09
9.779E+06
8.811E+03
6.608E+10
u235
1.355E+08
1.355E+05
1.221E+02
9.158E+08
u236
1.971E+09
1.971E+05
1.775E+03
1.331E+10
u238
1.224E+10
1.224E+07
8.270E+03
1.103E+11
pu238
3.707E+12
3.707E+10
2.003E+07
1.431E+15
pu239
6.316E+12
6.316E+10
3.413E+07
2.844E+15
pu240
1.077E+13
1.077E+11
5.822E+07
4.852E+15
pu241
4.873E+14
4.873E+10
6.585E+07
4.390E+15
pu242
1.431E+10
1.431E+08
7.733E+04
6.444E+12
np237
9.731E+05
9.731E+02
8.766E+03
2.630E+11
am241
1.948E+10
1.948E+08
1.316E+08
2.632E+15
cm244
7.951E+08
7.951E+05
3.070E+06
7.163E+13
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A. O. Pavelescu, D. G. Gepraga
10
Table 5.5
Clearance potential index, ingestion hazard factor and inhalation hazard factor of fission products,
after the cooling time of 5590 days
h3
sr 90
tc 99
ru106
sb125
i129
cs134
cs137
eu154
eu155
Origen ARP
Specific Radioactivity
Bq/kg U
Clearance
Potential Index
Specific Rad./Clear. Lev.
Ingestion
Hazard Factor
m3 H2O
Inhalation
Hazard Factor
m3 air
2.121E+09
4.980E+11
1.510E+08
1.446E+08
1.842E+09
3.553E+05
2.837E+09
7.781E+11
8.436E+09
2.548E+09
2.121E+04
4.980E+08
1.510E+05
1.446E+06
1.842E+07
3.553E+04
2.837E+07
7.781E+09
8.436E+07
2.548E+06
1.911E+04
4.486E+10
2.041E+04
3.907E+05
4.978E+05
1.600E+05
8.520E+06
1.052E+09
1.140E+07
3.443E+05
2.866E+08
4.486E+14
2.041E+09
1.954E+10
5.531E+10
4.801E+08
1.917E+11
4.206E+13
2.280E+12
2.295E+10
From the above two Tables we can see that the clearance potential index is
much higher than unit for all isotopes considered and this means that the
radiotoxicity of them is still considerable. Concerning the ingestion/inhalation
hazard factors for the same isotopes we have to remark again the great values of
the dilute volumes, this being a proof of the high radiotoxicity of the mentioned
isotopes. The conclusion is that after the cooling time taken into account the
isotopes considered are still extremely radioactive, therefore cannot be
manipulated without restrictions and need to be stored.
6. CONCLUSIONS
The ORIGEN-S code has been used to calculate major radioisotope inventories in a well-characterized Pickering-A CANDU fuel element, and these inventories have been compared with inventories measured by radiochemical analysis.
The fuel element chosen for analysis was an element with an average measured
burnup of 221 MWh/kg initial U. The measured burnup agreed well with the predicted burnup of 212 MWh/kg U, based on reactor power history and residence time.
The ORIGEN-ARP from SCALE5 coupled with the time dependent cross
sections libraries, produced by the sequence SAS2H of SCALE 4.4a, for the
irradiated lattice-cell of CANDU 28 were used for inventory calculation.
All isotopes considered – major actinides and fission products – after the
cooling time of 5590 days are still extremely radioactive and have to be stored.
The discrepancies between measured and calculated (predicted) values are
the followings:
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CANDU radiotoxicity inventories estimation
147
– The measured inventories in terms of concentration (g/kg U) for the isotopes
of 235U, 236U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, agree to within 5% of the
inventories predicted by the ORIGEN-ARP code.
– The isotopes of 237Np, 241Am also agree to within 5% of the predicted values
in terms of radioactivity (Bq/kg U); the discrepancy, however, is well within
the analytical uncertainty for these isotopes.
– The isotope of 234U agrees to within 20% of the predicted values, in terms of
concentration; the discrepancy, however, is within the analytical uncertainty
for these isotopes.
– The major fission products, 137Cs and 90Sr agreed to within 3% of the
ORIGEN-S values.
– The calculated inventories for 154Eu agree within 3% of the measured values.
– The isotope of 244Cm agrees within 11% of the predicted values, also within
the analytical uncertainty.
– The l25Sb agrees within 16%; this discrepancy was within the analytical
uncertainties as well.
– The 154Eu agree within 23% of the measured values, witch is outside of the
analytical uncertainties.
– The isotope of 134Cs agree to 31.9% witch is outside of the analytical
uncertainties.
– The calculated inventories for the isotopes 99Tc and 129I agree within 39% and
45% of the measured inventories respectively, outside of the analytical
uncertainties.
– The 106Ru agree within 65% of the measured values, far outside of the
analytical uncertainties.
– The calculated inventory of 3H, generated by ternary fission in the fuel, agrees
within 3% of the predicted total inventory and is within the analytical
uncertainty.
The discrepancies for the isotopes that are outside the analytical uncertainty
must be attributed to losses during chemical separation before measuring or, in
the case of Tc, to incomplete recovery due to its association with metallic
residues that could not be completely dissolved.
The large discrepancy for 106Ru can be attributed to poor counting
geometry for the solid residue, as it appears to be associated entirely with the
undissolved residue.
This benchmarking of the ORIGEN-S code represents a comparison up to
date of measured radionuclide inventories in a CANDU fuel with those predicted
by a generation/depletion code. The results have demonstrated that the
SAS2H/ORIGEN-ARP code system and nuclear data from ENDF/B-(IV to VI)
libraries provide inventory predictions that are within the range of measurement
uncertainties for the actinides and most fission products.
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A. O. Pavelescu, D. G. Gepraga
12
REFERENCES
1. A. O. Pavelescu, D. G. Cepraga, Radiotoxicity of CANDU in Terms of Clearance Potential
2.
3.
4.
5.
6.
7.
8.
9.
10.
11.
12.
Index and Biological Ingestion/Inhalation Hazard Factors, Romanian Scientist Academy
Review, No. July / 2006, Bucharest, Romania.
IAEA (International Atomic Energy Agency) Safety Standard Series, Safety Guide No. RS-G-1.7/
2004, Application of the Concepts of Exclusion, Exemption and Clearance.
United States Code of Federal Regulations, Part 10, Title 20 (10CFR20-1991).
A. A. Pasanen, Fundamentals of CANDU Nuclear Reactor Design, Atomic Energy of Canada
Limited,(AECL) , Course on Operational Physics and Power Reactors , Trieste, Italy, 1980.
J. C. Tait, I. Gauld, A. H. Kerr, Validation of the ORIGEN-S code for predicting radionuclide
inventories in used CANDU fuel, AECL Research, Whiteshell Laboratories, Pinawa, Canada,
Journal of Nuclear Materials 223 , 1995.
O. W. Hermann, C. V. Parks, SAS2H: A Coupled One-Dimensional Depletion and Shielding
Analysis Module ORNL (Oak Ridge National Laboratory) NUREG/CR-0200, SCALE 4.4a
– ORIGEN-S Manual, 1998.
I. C. Gauld, K. A. Litwin, Verification and Validation of the ORIGEN-S Code and Nuclear
Data Libraries, 1995 August, RC-1429, COG-I-95-150.
I. C. Gauld, B. D. Murphy, M. L. Williams, ORIGEN-S Data Libraries, SCALE 5 Manual,
ORNL/TM-2005/39, 2005.
D. G. Cepraga, G. Cambi, M. Frisoni, Clearance Potential of ITER Vacuum Vessel Activated
Materials, International Conference on Issues and Trends in Radioactive Waste Management, Vienna, 9–13 December 2002, IAEA-CN-90, IAEA-EC-NEA Editors, IAEA-CN-90/65,
pp. 333–339, Vienna 2002.
D. G. Cepraga, G. Cambi, M. Frisoni, G. C. Panini, ANITA-2000, NEA Data Bank Program,
NEA-1638, OECD Nuclear Energy Agency, 22 November 2000, RSICC Code Package
CCC-693, January 2002.
IAEA-TECDOC-855 (International Atomic Energy Agency), Clearance levels for
radionuclides in solid materials. Application of exemption principles.
I. C. Gauld, S. M. Bowman, J. E. Horwedel, L. C. Leal, ORIGEN-ARP: Automatic Rapid
Processing for Spent Fuel Depletion, Decay, and Source Term Analysis, Oak Ridge Nuclear
Laboratory TM-2005/39, SCALE 5 Manual, April 2005.