isotopic data of sample f3f6 from a rod ... - nuclear … · from a rod irradiated in the swedish...
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ZC-08/001
ISOTOPIC DATA OF SAMPLE F3F6
FROM A ROD IRRADIATED
IN THE SWEDISH
BOILING WATER REACTOR
FORSMARK 3
Compilation of Data in Support of the
OECD/NEA Expert Group on
Assay Data of Spent Nuclear Fuel and the
Spent Fuel Isotopic Composition Database SFCOMPO
Hans-Urs Zwicky
September 25, 2008
Zwicky Consulting GmbH Mönthalerstr. 44 CH-5236 Remigen Switzerland Tel. +41 (0)56 284 16 94 Fax +41 (0)56 284 16 93 [email protected]
ZC-08/001
LEGAL NOTICE
Zwicky Consulting GmbH exercised its best efforts to meet the objectives sought in this assignment
and applied to the work professional personnel having the required skills, experience and competence.
It is understood that Zwicky Consulting’s liability, if any, for any damages direct or consequential
resulting therefrom, will be limited to the amount paid for this assignment.
ERROR! NO TEXT OF SPECIFIED STYLE IN DOCUMENT.
Compilation of Data in Support of the
OECD/NEA Expert Group on
Assay Data of Spent Nuclear Fuel and the
Spent Fuel Isotopic Composition Database SFCOMPO
Hans-Urs Zwicky
Error! No text of specified style in document.
ZC-08/001
II
CONTENT
Page
1. Introduction ....................................................................................................................... 1
2. Reactor and Core ............................................................................................................... 1
3. Mechanical and Nuclear Assembly and Rod Design ....................................................... 2 3.1 Assembly 14595 .................................................................................................................... 2 3.2 Surrounding Assemblies ...................................................................................................... 3 3.3 Fuel Rod in Position F6 ....................................................................................................... 4
4. Irradiation History ............................................................................................................ 5
5. Analysed Samples .............................................................................................................. 9
6. Nuclide and Burnup Analysis Performed in Harwell .................................................... 10 6.1 Experimental Deatails........................................................................................................ 10 6.2 Results ................................................................................................................................. 10
7. Nuclide and Burnup Analysis Performed in Dimitrovgrad ........................................... 12 7.1 Experimental Deatails........................................................................................................ 12 7.2 Results ................................................................................................................................. 14
8. Nuclide Analyses Performed in Studsvik ....................................................................... 15 8.1 2003 Campaign ................................................................................................................... 15
8.1.1 Dissolution ....................................................................................................................... 15 8.1.2 “Old” HPLC-ICP-MS Instrument .................................................................................... 16 8.1.3 Isotope Dilution Analysis ................................................................................................ 16
8.2 2006 Campaign ................................................................................................................... 21 8.2.1 „New“ HPLC-ICP-MS Instrument .................................................................................. 21 8.2.2 Isotope Dilution Analysis ................................................................................................ 21
9. Data Comparison and Discussion .................................................................................. 22
10. Alternative Burnup Determination ................................................................................. 27 10.1 Introductory Remarks ....................................................................................................... 27 10.2 CASMO Calculations ........................................................................................................ 27 10.3 Burnup Determination Based on 2003 Data .................................................................... 28 10.4 Method Application on Harwell, Dimitrovgrad and Studsvik 2006 Data .................... 29
11. Conclusions ..................................................................................................................... 30
12. Acknowledgements .......................................................................................................... 31
13. References ........................................................................................................................ 32
ZC-08/001
III
FIGURES
Page
Figure 1 The nuclear power plant Forsmark 3 [1] ................................................................. 1
Figure 2 Forsmark 3 core lattice dimensions ......................................................................... 2
Figure 3 SVEA-100 assembly, cross section (dimensions in mm) ........................................ 3
Figure 4 SVEA-100 assembly 14595, nuclear design ........................................................... 3
Figure 5 8x8 and SVEA-64 assembly cross sections [5] ....................................................... 4
Figure 6 Forsmark 3 reactor power during cycles 3 to 8 ....................................................... 6
Figure 7 Forsmark 3 core burnup during cycles 3 to 8 .......................................................... 6
Figure 8 Forsmark 3 primary system pressure during cycles 3 to 8 ...................................... 6
Figure 9 Position of assembly 14595 inForsmark 3 core during cycles 3 to 8 ...................... 7
Figure 10 Assembly types and exposures adjacent to assembly 14595 during cycles 3 to
8 ............................................................................................................................... 7
Figure 11 Burnup of pin 14595/F6 during cycles 3 to 8 .......................................................... 8
Figure 12 Linear heat generation rate of pin 14595/F6 during cycles 3 to 8 ........................... 8
Figure 13 Nodal void representative for analysed sample location during cycles 3 to 8 ........ 8
Figure 14 Nodal moderator temperature representative for analysed sample location
during cycles 3 to 8 ................................................................................................. 9
Figure 15 Nodal fuel temperature representative for analysed sample location during
cycles 3 to 8 ............................................................................................................. 9
Figure 16 Scheme of fuel analysis in Dimitrovgrad .............................................................. 12
Figure 17 Scheme of chemical separations (techniques by SSC RIAR) ............................... 13
Figure 18 Isotopic abundance of uranium isotopes and relative deviation from mean
value ...................................................................................................................... 25
Figure 19 Isotopic abundance of plutonium isotopes and relative deviation from mean
value ...................................................................................................................... 25
Figure 20 Isotopic abundance of neodymium isotopes and relative deviation from mean
value ...................................................................................................................... 25
Figure 21 Isotopic abundance of cerium isotopes and relative deviation from mean
value ...................................................................................................................... 26
Figure 22 Amount of nuclide nPu (weight%) relative to
238U and relative deviation from
mean value ............................................................................................................. 26
Figure 23 Amount of nuclide nNd (weight%) relative to
238U and relative deviation
from mean value .................................................................................................... 26
Figure 24 Amount of nuclide nCe (weight%) relative to
238U and relative deviation from
mean value ............................................................................................................. 26
Figure 25 Nodal power and void history based on core tracking (filled diamonds) and
used for CASMO-4 simulation (open squares) ..................................................... 28
Figure 26 Principle of burnup determination by comparing experimentally determined nNd/
238U weight ratios as well as
235U and
239Pu isotopic abundances to
corresponding CASMO data ................................................................................. 29
ZC-08/001
IV
TABLES
Page
Table 1 Characteristics of Forsmark 3 reactor ..................................................................... 2
Table 2 Characteristics of surrounding assemblies .............................................................. 4
Table 3 Zircaloy-2 cladding composition ............................................................................. 5
Table 4 UO2 fuel composition .............................................................................................. 5
Table 5 Start-up and shut-down dates as well as nominal rated power, mass power
density and initial uranium core inventory for Forsmark 3 cycles of concern ........ 5
Table 6 Isotopic composition and burnup for sample FFBU determined in Harwell ........ 11
Table 7 ASTM E244-80 calculation of 148
Nd effective fractional fission yield ................ 11
Table 8 Spike isotopes and enrichment used in analyses performed in Dimitrovgrad ...... 13
Table 9 Isotopic composition [atom%] for sample FFBU determined in Dimitrovgrad ... 14
Table 10 Element ratios for sample FFBU determined in Dimitrovgrad ............................. 15
Table 11 Fractions of fissions, effective yields and fuel burn-up for sample FFBU
determined in Dimitrovgrad .................................................................................. 15
Table 12 Abundances of spike isotope and isotope to be analysed in spike solutions ......... 17
Table 13 Errors of input data used in calculations ............................................................... 20
Table 14 Isotopic composition (atom%) of F3F6 sample, determined 2003 in Studsvik .... 20
Table 15 Amount of nuclide nX (weight%) relative to
238U, determined 2003 in
Studsvik ................................................................................................................. 21
Table 16 Isotopic composition (atom%) of F3F6 sample, determined 2006 in Studsvik .... 22
Table 17 Amount of nuclide nX (weight%) relative to
238U, determined 2006 in
Studsvik ................................................................................................................. 22
Table 18 Isotopic composition (atom%) of F3F6 sample .................................................... 24
Table 19 Amount of nuclide nX (weight%) relative to
238U in F3F6 sample ....................... 24
Table 20 Elemental ratios in F3F6 sample ........................................................................... 25
Table 21 Burnup values based on the comparison of experimentally determined
Studsvik 2003 values with values calculated by CASMO .................................... 29
Table 22 Burnup values based on the comparison of experimental values determined
by Harwell, Dimitrovgrad and Studsvik 2006 with values calculated by
CASMO ................................................................................................................. 30
ZC-08/001
1
1. INTRODUCTION
A sample from the central part of a fuel rod irradiated until June 6, 1993 in the Swedish
boiling water reactor Forsmark 3 to a burnup of about 58 MWd/kgU was dissolved in
Studsvik. Aliquots of this solution were shipped to two well-recognised independent
laboratories1 for the determination of the isotopic composition and for radiochemical burnup
analysis. In 2003, a sample adjacent to the one taken for the analyses in Harwell and
Dimitrovgrad was dissolved and analysed in Studsvik. The same solution was re-analysed
with new equipment in 2006.
This report compiles all isotopic data acquired so far on this particular fuel rod together with
corresponding pre-irradiation and irradiation information in support of the OECD/NEA
Expert Group on Assay Data of Spent Nuclear Fuel and the Spent Fuel Isotopic Composition
Database SFCOMPO [3].
2. REACTOR AND CORE
Figure 1 The nuclear power plant Forsmark 3 [1]
The nuclear power plant Forsmark 3 (Figure 1) operates a boiling water reactor (BWR) built
by ASEA Atom (later ABB Atom, now Westinghouse Electric Sweden). Commercial
operation started in 1985 with a thermal power of 3020 MWth. The core consists of 700
assemblies and contains 169 cruciform control rods. The plant is operated on a 12 month
cycle basis, with somewhat shifting cycle lengths and outages during the summer months.
Thermal output was increased in 1989 to 3300 MWth. Characteristic data, provided by
Vattenfall Nuclear Fuel [2], are compiled in Table 1. Figure 2 shows dimensions of the core
lattice.
1 AEA Technology, Fuel Performance Group, Harwell, United Kingdom
State Scientific Centre of Russian Federation Research Institute of Atomic Reactors (RF SSC RIAR),
Dimitrovgrad, Russia
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2
Table 1 Characteristics of Forsmark 3 reactor
Nominal thermal power: 3020 MW (Cycles 1 - 4)
3300 MW (since Cycle 5)
Nominal pressure of primary system: 70.0 bar
Maximum core flow: 13100 kg/s
Nominal coolant inlet temperature: ~278°C
Nominal coolant outlet temperature: ~286°C
Number of internal recirculation pumps: 8
Number of fuel assemblies: 700
Number of control rods: 169
(Outer channel and water gap dimensions: see Table 2)
Figure 2 Forsmark 3 core lattice dimensions
3. MECHANICAL AND NUCLEAR ASSEMBLY AND ROD DESIGN
Information on the design of assembly 14595 and its rod in position F6 was provided by
Westinghouse Electric Sweden AB [4].
3.1 ASSEMBLY 14595
Assembly 14595 is a SVEA-100 assembly with 100 fuel rods in four 5x5 sub-bundles. The
sub-bundles are free-standing in the sub-channels of a SVEA-100 fuel channel and connected
to the handle with one screw per sub-bundle. The sub-channels are separated by so called
water wings, flat internal channels, bringing non-boiling coolant even into the upper part of
the fuel assembly. Figure 3 shows a cross section and important dimensions.
Figure 4 illustrates the nuclear design of assembly 14595. Rod types 9 and 10 are spacer
capture rods. Their nuclear design is similar to rod types 17 and 16, respectively. Rod type 11
represents tie rods, with nuclear design similar to rod type 16. The enriched zone has a height
of 3450 mm. On top and at the bottom, blanket zones with natural uranium are 150 mm high,
resulting in a total active fuel height of 3750 mm.
Each sub-bundle contains six Inconel X750 spacer grids. The lower edge of the first spacer is
531.4 mm from the bottom of the fuel pellet column (533 mm from upper edge of lower tie
plate), and then the spacers follow at 568 mm intervals. The mass of a spacer is 24 g, its
dimensions 65.1 mm x 65.1 mm x 26 mm.
ZC-08/001
3
Figure 3 SVEA-100 assembly, cross section (dimensions in mm)
Figure 4 SVEA-100 assembly 14595, nuclear design
3.2 SURROUNDING ASSEMBLIES
Assembly types that had been loaded adjacent to assembly 14595 during its operation and
their characteristics are listed in Table 2. Except for the initial core 8x8 assemblies IA84 and
the demo assemblies D691, all were of the SVEA-64 type. Figure 5 shows assembly cross
sections of 8x8 and SVEA-64 geometries.
ZC-08/001
4
The Forsmark 3 lattice is basically symmetric, although Figure 2 indicates an asymmetry
(wide and narrow water gaps). This was indeed the case for the initial core assemblies,
whereas the nominal geometry of reload assemblies forms a symmetric lattice. The values
indicated in Table 2 are those used by Vattenfall Nuclear Fuel for modelling, assuming an
initial channel bow of 0.5 mm.
Table 2 Characteristics of surrounding assemblies
Assembly type IA84 E287 E388 E489 E590/E591 D691 E691
Geometry 8x8 SVEA-64 SVEA-100 SVEA-64
Rod array 8x8 4x(4x4) 4x(5x5) 4x(4x4)
Fuel UO2/Gd 235
U Enrichment [%] 2.63 3.11 3.03 3.06 3.08 3.11 3.07
# of Gd rods 3 4 5 6 7 8 7
Gd2O3 content [%] 5.5 3.15 3.15 3.5 3.5 3.95 3.15
Poisoning(a)
0.8 1 0.8 0.8 0.8 1 0.8
water rods (number) 12.25 (1x) -
Outer channel width [mm] 139 139.6 140.2
Channel wall thickness [mm] 2.5 1.1 1.4
Sub-channel size [mm] - 65.9
Rod outer diameter [mm] 12.25 (52x)
11.75 (12x) 12.25 (64x)
9.62
(100x)
12.25
(64x)
Cladding wall thickness [mm] 0.8 0.63 0.8
Rod pitch [mm] 16.3
(16.05/15.8) 15.8
12.7
(12.55) 15.8
Wide water gap [mm] 10 8.075 7.775
Narrow water gap [mm] 5.75 7.075 6.775 (a)
Poisoning: Old ASEA concept (“zebra fuel”). p = 1: all pellets in a Gd rod are Gd pellets;
p = 8.032 : every third pellet is a UO2 pellet)
Figure 5 8x8 and SVEA-64 assembly cross sections [5]
3.3 FUEL ROD IN POSITION F6
Position F6 corresponds to the corner rod towards the water cross in the sub-bundle in the
control rod assembly corner, as can be seen from Figure 4. The fuel rod contains a top plenum
with a length of (158±12.5) mm. It was filled with helium (>98%) to a pressure of
(0.4±0.05) MPa (absolute).
Cladding tube outer and inner diameters are 9.62 and 8.36 mm, respectively. Cladding
material is Zircaloy-2. Specified composition and impurities are compiled in Table 3. The
material density is 6.57 g/cm3.
The pellet diameter is 8.19 mm, which results in a diametrical gap of 0.17 mm. The pellets
have a 0.1 mm deep dish of 3 mm diameter. The fuel density is (10.47+0.15/-0.10) g/cm3.
Fuel pellet composition and impurities are listed in Table 4.
ZC-08/001
5
Table 3 Zircaloy-2 cladding composition
Main components Maximum amount of impurities
Element [wt%] Element [ppm] Element [ppm] Element [ppm]
Sn 1.20 - 1.70 Al 75 Cu 50 N 80
Fe 0.07 - 0.20 B 0.5 Hf 100 Si 200
Cr 0.05 - 0.15 Cd 0.5 H 25 Na 20
Ni 0.03 - 0.08 C 270 Mg 20 Ti 50
O 0.09 - 0.16 Cl 20 Mn 50 W 100
Zr Remainder Co 20 Mo 50 U 3.5
Fe+Cr+Ni 0.18 - 0.38
Table 4 UO2 fuel composition
Main components Maximum amount of impurities
Element [wt%] Element [ppm] Element [ppm] Element [ppm] Element [ppm]
Uranium (tot) 88.14 Ag 0.5 Co 6 Mn 10 W 50
Isotopic composition Al 50 Cr 50 Mo 100 V 1.0
Isotope [% mass] B 0.5 Cu 25 N 50 Zn 20 234
U 0.031 Bi 2.0 F 15 Ni 50 Dy 0.5 235
U 3.965 C 20 Fe 100 Pb 20 Eu 0.5 236
U 0.013 Ca 25 In 3.0 Si 100 Gd 1.0 238
U 95.991 Cd 0.5 Li 2.0 Sn 5.0 Sm 2
Cl 25 Mg 50 Ti 40 Na 70
4. IRRADIATION HISTORY
Information on the power history was provided by Vattenfall Nuclear Fuel [2]. Table 5 shows
start-up and shut-down dates as well as nominal rated power, mass power density and initial
uranium core inventory for the cycles of concern. Figure 6 depicts reactor power, Figure 7
core burnup and Figure 8 primary system pressure during the same cycles as a function of
effective full power hours.
The core position of assembly 14595 is shown in Figure 9. Figure 10 contains information on
assembly types and exposure in positions adjacent to assembly 14595 during Cycles 3 - 8.
Burnup and linear heat generation rate of pin 14595/F6 as well as nodal void, moderator
temperature and fuel temperature representative for the location of the analysed sample are
plotted in Figure 11 to Figure 15.
Table 5 Start-up and shut-down dates as well as nominal rated power, mass power
density and initial uranium core inventory for Forsmark 3 cycles of concern
Cycle Beginning
of Cycle
End of Cycle Nominal Full
Power [MW]
Mass Power
Density [W/g]
Uraniuminit Core
Inventory [t]
3 August 1, 1987 August 13, 1988 3020 24.152 125.040
4 September 3, 1988 June 10, 1989 3020 23.995 125.859
5 July 8, 1989 July 14, 1990 3300 23.805 126.862
6 August 1, 1990 August 17, 1991 3300 23.659 127.645
7 September 4, 1991 May 15, 1992 3300 23.540 128.294
8 June 18, 1992 June 6, 1993 3300 23.532 128.335
ZC-08/001
6
0
20
40
60
80
100
120
0 10000 20000 30000 40000 50000
Effective Full Power Hours
Re
ac
tor
Po
we
r [%
of
30
20
MW
]
Figure 6 Forsmark 3 reactor power during cycles 3 to 8
0
5
10
15
20
25
30
0 10000 20000 30000 40000 50000
Effective Full Power Hours
Co
re B
urn
up
[M
Wd
/kg
U]
Figure 7 Forsmark 3 core burnup during cycles 3 to 8
70.2
70.3
70.4
70.5
70.6
70.7
70.8
70.9
71
0 10000 20000 30000 40000 50000
Effective Full Power Hours
Pri
ma
ry S
ys
tem
Pre
ss
ure
[b
ar]
Figure 8 Forsmark 3 primary system pressure during cycles 3 to 8
ZC-08/001
7
Figure 9 Position of assembly 14595 in Forsmark 3 core during cycles 3 to 8
Cycle 3 Cycle 4 Cycle 5
IA84
18.56
22.23
IA84
23.50
27.00
E388
8.38
10.00
IA84
18.56
22.24
14595
0.17
0.18
IA84
18.50
21.67
IA84
26.66
30.96
14595
11.22
12.54
IA84
20.44
23.15
E388
8.79
10.61
14595
19.66
22.75
IA84
26.99
31.03
IA84
18.61
21.74
E388
0.17
0.20
E489
1.12
1.34
Cycle 6 Cycle 7 Cycle 8
E489
9.36
10.92
D691
0.38
0.44
E691
9.06
10.60
E287
17.58
19.46
14595
27.61
31.94
E590
0.17
0.19
D691
0.38
0.44
14595
37.66
43.42
E590
11.82
13.52
E591
7.66
8.70
14595
44.16
50.67
E388
32.33
36.37
E489
9.35
11.24
E590
11.87
13.53
E489
24.10
26.97
(Numbers below assembly type: bundle and nodal exposure at first TIP measurement [MWd/kgU])
Figure 10 Assembly types and exposures adjacent to assembly 14595 during cycles 3 to 8
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8
0
10
20
30
40
50
60
70
0 10000 20000 30000 40000 50000
Effective Full Power Hours
Pin
Bu
rnu
p [
MW
d/k
gU
]
Figure 11 Burnup of pin 14595/F6 during cycles 3 to 8
0
5
10
15
20
25
0 10000 20000 30000 40000 50000
Effective Full Power Hours
Pin
Lin
ea
r H
ea
t G
en
era
tio
n R
ate
[kW
/m]
Figure 12 Linear heat generation rate of pin 14595/F6 during cycles 3 to 8
0
10
20
30
40
50
60
70
80
0 10000 20000 30000 40000 50000
Effective Full Power Hours
Vo
id [
%]
Void calculated for coolant flow area,
excluding areas for internal and external bypass flow
Figure 13 Nodal void representative for analysed sample location during cycles 3 to 8
ZC-08/001
9
286
286.1
286.2
286.3
286.4
286.5
286.6
286.7
286.8
0 10000 20000 30000 40000 50000
Effective Full Power Hours
Mo
de
rato
r T
em
pe
ratu
re [
C]
Figure 14 Nodal moderator temperature representative for analysed sample location
during cycles 3 to 8
0
100
200
300
400
500
600
700
0 10000 20000 30000 40000 50000
Effective Full Power Hours
Fu
el T
em
pe
ratu
re [
C]
Figure 15 Nodal fuel temperature representative for analysed sample location during cycles
3 to 8
5. ANALYSED SAMPLES
A 10 mm long sample was cut out from fuel rod 14595/F6 at a distance of 1999 - 2009 mm
from the lower end plug [6]. The fuel matrix, but not alloy particles and cladding material,
was dissolved in concentrated HNO3. Diluted aliquots of this solution (sample designation:
FFBU) were sent to Harwell and Dimitrovgrad for radiochemical characterisation (see
Chapters 6 and 7).
The rod segment adjacent to the lower side of the dissolved sample is used as reference rod
F3F6 in gamma scans at Studsvik. A 2 mm slice was later cut off at the top of this reference
rod and dissolved (for details, see 8.1.1). Diluted aliquots of this solution were characterised
radiochemically at Studsvik in 2003 and 2006 (see Chapter 8).
ZC-08/001
10
6. NUCLIDE AND BURNUP ANALYSIS PERFORMED IN
HARWELL
The radiochemical analyses performed by AEA Technology in Harwell were described in [7].
6.1 EXPERIMENTAL DETAILS
Fuel burnup was measured by determining the 148
Nd/U ratio, using a similar method to
ASTM E321-79.
Two aliquots were taken from each sample and a known U/Pu/Nd mixed spike was added to
one. For Pu, the spike was 99.9% 242
Pu, standardised using a Pu metal alloy. For U, the spike
was 99.7% 233
U, standardised using depleted U dioxide. For Nd, the spike was 98.3% 142
Nd,
standardised using natural Nd metal. The aliquots were separated into U, Pu and Nd fractions
using ion exchange. The three elements were then analysed separately using Thermal
Ionisation Mass Spectrometry (TIMS). This allows the necessary calculation of the 148
Nd/U
ratio and the relative isotopic compositions of U, Pu and Nd. The effective fractional fission
yield of 148
Nd was calculated following ASTM E244-80.
6.2 RESULTS
The 148
Nd/U ratio and atom% burnup is presented in Table 6. Relative isotopic compositions
of U, Pu and Nd are also given. Table 7 shows values used for calculating the effective
fractional fission yield of 148
Nd. The Tables are cut out from the original report [7].
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11
Table 6 Isotopic composition and burnup for sample FFBU determined in Harwell
Table 7 ASTM E244-80 calculation of 148
Nd effective fractional fission yield
ZC-08/001
12
7. NUCLIDE AND BURNUP ANALYSIS PERFORMED IN
DIMITROVGRAD
The radiochemical analyses performed by the State Scientific Centre of Russian Federation
Research Institute of Atomic Reactors (RF SSC RIAR) in Dimitrovgrad were described in [8].
7.1 EXPERIMENTAL DETAILS
The investigations were carried out following the standards ASTM E321-90 and ASTM
E244-85 as well as techniques specifically developed at SSC RIAR. Isotopic compositions of
uranium, plutonium, americium, neodymium and cerium were determined by Thermal
Ionisation Mass Spectrometry (TIMS) after chemical separation. Data evaluation included
burnup analysis and determination of Pu/U, Am/U, Nd/U and Ce/U ratios. Figure 16
illustrates schematically the flow of fuel analyses in Dimitrovgrad. Chemical separations
applied at SSC RIAR are illustrated in Figure 17.
Spike solutions were prepared by the Scientific Production Society V.G. Khlopin “Radium
Institute” St. Petersburg. Spike isotopes and enrichment are compiled in Table 9.
244Cm was determined by alpha spectrometry.
Figure 16 Scheme of fuel analysis in Dimitrovgrad
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13
Figure 17 Scheme of chemical separations (techniques by SSC RIAR)
Table 8 Spike isotopes and enrichment used in analyses performed in Dimitrovgrad
Isotope 233
U 242
Pu 243
Am 146
Nd 140
Ce
Enrichment [%] 99.584±0.015 99.530±0.040 99.95±0.02 99.646±0.019 99.71±0.01
ZC-08/001
14
7.2 RESULTS
Table 9 shows the isotopic composition of uranium, plutonium, americium, neodymium and
cerium in sample FFBU, as it was determined by SSC RIAR. Table 10 contains the element
ratios, Table 11 the details on the burnup analysis. Burnup is not only based on 148
Nd, but on
the sum of 145
Nd and 146
Nd as well. The content of the Tables was cut out from the original
report [8].
Table 9 Isotopic composition [atom%] for sample FFBU determined in Dimitrovgrad
Date of measurements:
April 1996
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15
Table 10 Element ratios for sample FFBU determined in Dimitrovgrad
Table 11 Fractions of fissions, effective yields and fuel burn-up for sample FFBU
determined in Dimitrovgrad
Fractions of fissions [%]
Effective fractional fission yield [%]
Burnup [%FIMA]
8. NUCLIDE ANALYSES PERFORMED IN STUDSVIK
8.1 2003 CAMPAIGN
8.1.1 Dissolution
The 2 mm fuel rod slice was placed in a glass flask together with 20 ml of concentrated HNO3
and kept at 65°C for 6 h. Evaporation of liquid was avoided by means of an air-cooled reflux
cooler. Nitrogen was bubbled through the liquid in order to stir it. The fuel matrix together
with all fission products of interest went into solution. The cladding and the metallic fission
product inclusions remained undissolved.
ZC-08/001
16
In the order of 0.1-0.4 g of the original fuel solution was diluted into 100 ml of HNO3 (7.5 M)
in the hotcell. 20 ml of this solution were transferred to the laboratory. An appropriate aliquot
was diluted with 100 ml HNO3 (0.16 M) to a target uranium concentration of about 4 ppm.
The uranium concentration was determined by Scintrex analysis. The Scintrex2 UA-3 is a
uranium analyser, measuring the characteristic fluorescence of the uranyl ion in solution after
irradiation with a very short pulse of ultraviolet light from a nitrogen laser. 30 g of this mother
solution was then mixed with all necessary spike solutions.
8.1.2 “Old” HPLC-ICP-MS Instrument
A DIONEX DX300 High Performance Liquid Chromatography system with an IonPac CG10
(4 x 50 mm) guard and an IonPac CS10 (4 x 250 mm) analytical column was used for the
separations. The eluents were directly injected into a VG ELEMENTAL Plasmaquad PQ2+
Inductively Coupled Plasma Mass Spectrometer (ICP-MS), installed in a glove box. Details
can be found in [9].
8.1.3 Isotope Dilution Analysis
Basis
Isotope Dilution Analysis (IDA) is based on the addition of a known amount of an enriched
isotope (“spike”) to a sample. Isotopic ratios between the added isotope and the isotope to be
analysed are determined by mass spectrometry in the mixture of spike and sample, in the pure
sample solution, and, if not already known, in the pure spike solution. The amount of the
isotope to be determined in the sample can be calculated according to the method derived
below:
spikeinbisotopeofnumberN
sampleinbisotopeofnumberN
spikeinaisotopeofnumberN
sampleinaisotopeofnumberN
mixtureinratioisotopeR
spikeinratioisotopeR
sampleinbaratioisotopeR
analysedbetoisotopeb
isotopespikea
Sp
b
S
b
Sp
a
S
a
M
Sp
s
)/(
S
b
S
as
N
NR Eq. 1
Spb
Spa
spN
NR Eq. 2
2 SCINTREX UA-3 Uranium Analyser, SCINTREX, Snidercroft Road, Concord Ontario Canada L4K 1B5
ZC-08/001
17
Spb
Sb
Spa
Sa
MNN
NNR
Eq. 3
By transforming Eq. 3, the following Eq. 4 can be derived
sM
SpbM
SpaS
bRR
NRNN
Eq. 4
Sp
bN can be substituted by means of Eq. 2, which leads to Eq. 5
sM
Sp
M
Spa
Sb
RR
R
R
NN
1
Eq. 5
Once the amount of isotope b in the sample has been determined, all other isotopes of the
same element can easily be determined by means of the isotopic ratios measured by mass
spectrometry.
Spiking
RS, the isotope ratio in the sample, is given. RSp, the ratio in the spike is fixed as well, once the
appropriate standard is chosen for a series of analyses. RM, the isotope ratio in the mixture, on
the other hand can be influenced by the amount of spike solution that is blended with the
sample aliquot. Two aspects have to be taken into account when choosing the appropriate RM
value: counting statistics, influencing the uncertainty of the isotopic ratio, and the factor that
determines the contribution of the uncertainty in RM by error propagation to the overall error
of the analysis.
The approximate amount of the isotopes to be analysed in the sample as well as the
corresponding RS values were estimated based on the result of semi-quantitative analyses and
on CASMO calculations. After choosing an appropriate RM value, the number of spike
isotopes to be added to an aliquot of the mother solution was calculated based on Eq. 5.
Identities of spike isotopes and of isotopes to be analysed, as well as their abundance in the
corresponding spike solutions, are shown in Table 12.
Table 12 Abundances of spike isotope and isotope to be analysed in spike solutions
Spike Isotope Abundance
[%]
Isotope to be
analysed
Abundance
[%]
233U 98.043
238U 0.804
242Pu 99.903
239Pu 0.0826
140Ce 99.30
142Ce 0.70
148Nd 91.60
146Nd 2.50
IDA without Separation
Uranium isotopes were determined by IDA based on ICP-MS without separation. Aliquots of
spiked and unspiked solutions were diluted as appropriate in order to avoid too large dead
ZC-08/001
18
time corrections and were measured five times. The measurements were performed in the
peak jump mode.
HPLC-ICP-MS
Plutonium isotopes were determined by IDA based on HPLC-ICP-MS, with an elution
program separating plutonium from interfering elements, e.g. uranium and americium.
Aliquots of spiked and unspiked solutions were diluted as appropriate. Blank samples were
measured before each unspiked and spiked sample, in order to check the absence of any
memory effect.
In a separate run, the lanthanides cerium and neodymium were determined, applying the
corresponding elution method.
Data Evaluation
Count rates measured in the analysis of uranium, performed without any separation, were
dead time and blank corrected. The count rates from the unspiked and spiked samples of mass
238 were corrected for the contribution of 238
Pu, based on the count rate for mass 239 and the
ratio of 238
Pu and 239
Pu determined in the plutonium analysis. The abundance of uranium iso-
topes in the unspiked sample was determined by normalising the corresponding count rates of
five individual measurements to 100%, followed by calculating an average value for each
individual isotope. RS was determined based on the corresponding abundances; RM was
calculated directly from the corresponding count rates. The number of 238
U atoms was
calculated according to Eq. 5. For all other isotopes, the number of atoms in the sample was
calculated by means of the corresponding abundances, based on the number of atoms of the
isotope to be analysed.
HPLC-ICP-MS analyses were evaluated in the same way. Instead of count rates, peak areas
determined by a dedicated program (MassLynx) were used as input data. In the case of
HPLC-ICP-MS, only three individual measurements were performed.
The number of atoms in the sample was transformed into micrograms. Finally, the amount of
nuclide nX in weight percent relative to
238U was calculated by dividing the corresponding
amount by the amount of 238
U.
Error Estimation
The uncertainty of the number of counts in a pulse counting system like ICP-MS is given by
the square root of the number of counts, neglecting the contribution of the background signal.
When applying the rules of error propagation on the simple Eq. 6 for the ratio of two isotopes
of interest, it can be demonstrated that the precision of the ratio is limited by the size of the
smaller peak (Eq. 7).
b
ar Eq. 6
with
areaspeakba
ratioisotopicr
,
ZC-08/001
19
bar
sr 11 Eq. 7
with
roferrorsr
Experience from routine analysis has shown that it is normally not possible to achieve a lower
relative standard deviation of r than about 0.1 %, even if sufficient counts are available [12].
If the number of counts in the smaller of the two peaks is significantly larger than 106, the
contribution of counting statistics is negligible. This is normally the case in HPLC-ICP-MS
analyses. In ICP-MS analyses in peak jump mode, numbers of counts may be smaller. With
105 counts in the smaller peak, the contribution of counting statistics to the relative error of r
is still below 0.5%. On the other hand, additional factors like instrument instability limit the
achievable accuracy. A possibility of assessing this scatter is calculating the relative standard
deviation of the five and three abundance values of individual isotopes, respectively, in the
unspiked samples that were determined by normalising the count rates of individual
measurements to 100%. For each isotopic ratio, sr calculated by error propagation from the
standard deviation of abundance values was compared to a value based on Eq. 7. The larger of
the two values was then used in the overall error estimation.
The equation for calculating the error of the number of atoms of the isotope to be analysed in
the sample (Eq. 8) is derived from Eq. 5 according to the general rules of error propagation.
222222
Sp
R
MSp
M
SM
R
SM
R
MSp
SpS
Spa
NSbN R
s
RR
R
RR
s
RR
s
RR
RR
N
sNs
SpSMSpa
Sb
Eq. 8
with
ioferrorabsolutesi
For all other isotopes, Eq. 9 is applied:
22
r
s
N
sNs r
sb
N
xN
sb
x Eq. 9
The relative error of the number of added spike atoms
Sp
a
N
N
s Spa and the relative error of RSp
Sp
R
R
sSp used in the calculations are estimated as shown in Table 13. They correspond to 1σ.
ZC-08/001
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Table 13 Errors of input data used in calculations
Parameter Relative Error (Comment)
Sp
aN 1% (Estimated, same value for all elements)
RSp
U 0.1% (Estimated)
Pu 0.1% (Estimated)
Ce 1% (Estimated)
Nd 0.5% (Estimated)
RS ,RM, r Determined according to the method described in the text
Results
The isotopic composition of uranium, plutonium, neodymium and cerium determined in
October 2003 by Studsvik, as it was documented in [10], is compiled in Table 14. Table 15
shows the amount of nuclide nX in weight percent relative to
238U.
Table 14 Isotopic composition (atom%) of F3F6 sample, determined 2003 in Studsvik
Uranium 234
U 235
U 236
U 238
U
Mean 0.018 0.360 0.642 98.980
Uncertainty 0.001 0.010 0.013 0.016
Plutonium 238
Pu 239
Pu 240
Pu 241
Pu 242
Pu
Mean 3.824 45.409 29.860 8.662 12.244
Uncertainty 0.112 0.454 0.286 0.287 0.136
Neodymium 142
Nd 143
Nd 144
Nd 145
Nd 146
Nd 148
Nd 150
Nd
Mean 0.851 14.400 38.200 14.919 17.929 9.255 4.446
Uncertainty 0.038 0.100 0.254 0.114 0.147 0.178 0.175
Cerium 140
Ce 142
Ce
Mean 52.496 47.504
Uncertainty 0.414 0.414
Date of analysis: October 8, 2003 (Pu); October 20, 2003 (Nd, Ce); November 4, 2003 (U)
ZC-08/001
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Table 15 Amount of nuclide nX (weight%) relative to
238U, determined 2003 in Studsvik
Uranium 234
U 235
U 236
U
Mean 0.017% 0.359% 0.643%
Uncertainty 0.001% 0.013% 0.020%
Plutonium 238
Pu 239
Pu 240
Pu 241
Pu 242
Pu
Mean 0.044% 0.524% 0.346% 0.101% 0.143%
Uncertainty 0.002% 0.017% 0.012% 0.005% 0.005%
Neodymium 142
Nd 143
Nd 144
Nd 145
Nd 146
Nd 148
Nd 150
Nd
Mean 0.0065% 0.110% 0.294% 0.116% 0.140% 0.073% 0.036%
Uncertainty 0.0003% 0.002% 0.007% 0.003% 0.003% 0.002% 0.002%
Cerium 140
Ce 142
Ce
Mean 0.254% 0.233%
Uncertainty 0.007% 0.006%
Date of analysis: October 8, 2003 (Pu); October 20, 2003 (Nd, Ce); November 4, 2003 (U)
8.2 2006 CAMPAIGN
8.2.1 „New“ HPLC-ICP-MS Instrument
In 2005, the Studsvik HPLC-ICP-MS equipment was replaced by a new instrument. A
DIONEX SP Gradient High Performance Liquid Chromatography (HPLC) system and
Autosampler Dionex AS with an IonPac CG10 (4 x 50 mm) guard and an IonPac CS10 (4 x
250 mm) analytical column is now used for the separations. Chromeleon Xpress, CHX-1
software controls the autosampler, injector and HPLC pump. The eluents are injected into a
Perkin Elmer Elan 6100 DRC II Inductively Coupled Plasma Mass Spectrometer (ICP-MS),
installed in a glove box. The ICP-MS instrument is controlled by Perkin Elmer Chromera
software. The Chromera software is also used for the collection and evaluation of the
chromatograms. Peak areas are used for the evaluation.
8.2.2 Isotope Dilution Analysis
A fresh aliquot of the same fuel solution as in 2003 was re-analysed in 2006 applying the new
equipment. Again, uranium, plutonium, cerium and neodymium nuclides were assessed.
Applied methods and data evaluation were similar to 2003, as described in 8.1.3. The isotopic
composition of uranium, plutonium, neodymium and cerium determined in December 2006
by Studsvik and documented in [11] is compiled in Table 16. Table 17 shows the amount of
nuclide nX in weight percent relative to
238U.
ZC-08/001
22
Table 16 Isotopic composition (atom%) of F3F6 sample, determined 2006 in Studsvik
Uranium 234
U 235
U 236
U 238
U
Mean 0.020 0.356 0.700 98.924
Uncertainty 0.001 0.010 0.013 0.016
Plutonium 238
Pu 239
Pu 240
Pu 241
Pu 242
Pu
Mean 4.033 45.805 30.183 7.560 12.418
Uncertainty 0.077 0.350 0.164 0.246 0.062
Neodymium 142
Nd 143
Nd 144
Nd 145
Nd 146
Nd 148
Nd 150
Nd
Mean 0.840 14.250 37.556 15.383 18.532 9.043 4.396
Uncertainty 0.009 0.142 0.099 0.165 0.058 0.050 0.066
Cerium 140
Ce 142
Ce
Mean 52.561 47.439
Uncertainty 0.132 0.132
Date of analysis: December 14, 2006
Table 17 Amount of nuclide nX (weight%) relative to
238U, determined 2006 in Studsvik
Uranium 234
U 235
U 236
U
Mean 0.020% 0.356% 0.701%
Uncertainty 0.001% 0.007% 0.013%
Plutonium 238
Pu 239
Pu 240
Pu 241
Pu 242
Pu
Mean 0.045% 0.512% 0.339% 0.085% 0.141%
Uncertainty 0.002% 0.016% 0.011% 0.004% 0.005%
Neodymium 142
Nd 143
Nd 144
Nd 145
Nd 146
Nd 148
Nd 150
Nd
Mean 0.0062% 0.106% 0.282% 0.116% 0.141% 0.070% 0.034%
Uncertainty 0.0001% 0.002% 0.004% 0.002% 0.002% 0.001% 0.001%
Cerium 140
Ce 142
Ce
Mean 0.227% 0.208%
Uncertainty 0.004% 0.003%
Date of analysis: December 14, 2006
9. DATA COMPARISON AND DISCUSSION
In order to compare all results on a common basis, data were decay-corrected to December
31, 2006. Isotopic compositions were re-normalised if necessary. All four sets of isotopic
compositions are compiled in Table 18, all nX/
238U values in Table 19. For errors, see Tables
in the corresponding Chapters. Table 20 summarises elemental ratios calculated from data in
Table 19. In particular, the following decay corrections were taken into account:
Decay of 241
Pu
Formation of 240
Pu through decay of 244
Cm, based on the 244
Cm/U ratio determined in
Dimitrovgrad
Decay of the remaining 144
Ce into 144
Nd, based on the 144
Ce content determined in
Dimitrovgrad
140Ce and
142Ce abundances were simply determined by normalising the
corresponding contents to 100%.
ZC-08/001
23
It should be kept in mind that the collection does not consist of four completely independent
sets. The Harwell and Dimitrovgrad data are based on aliquots of the same fuel solution. The
same is true for the two sets of Studsvik data.
In Figure 18 to Figure 21, isotopic abundances are compared to each other. Every Figure
shows analysed values side by side and relative deviations from the mean value. Except for
four cases, the deviations between the four individual values (three in the case of cerium) are
small.
The 236
U value determined by Studsvik in 2006 is significantly larger than the other
three values. No obvious reason could be identified.
The 142
Nd value determined by Harwell is significantly higher than the other three
values, indicating that the sample might have been contaminated by a small amount
of natural (or spike) neodymium. If the Harwell values are corrected by subtracting
an amount of natural neodymium corresponding to the difference between the
Harwell 142
Nd value and the average of the three other ones and then normalised
again, the abundance of all other isotopes is not significantly changed, but the sum of
squares of deviations (excluding 142
Nd) between Harwell and average value of the
other three gets smaller.
The Dimitrovgrad 238
Pu value is significantly larger than the other three values.
The Harwell and Dimitrovgrad 234
U values are lower, the two Stdsvik values higher
than the mean value. The difference seems to be significant.
nX/
238U values for plutonium, neodymium and cerium are shown in Figure 22, Figure 23 and
Figure 24 together with relative deviations of individual values from the mean. When
comparing nX/
238U values with abundances, it is obvious that some systematic biases were
introduced during the analysis. A potential source impacting all nuclides of an element in the
same direction is a spiking error. Even a selective loss of material, e.g. by co-precipitation,
could be the reason for such an effect. Two cases are obvious:
Dimitrovgrad neodymium values (disregarding 142
Nd) are systematically higher than
all other data. The mean deviation from the average of the other three is more than
5%. This is also reflected in the elemental ratios (Table 20).
The Studsvik 2006 cerium values are about 10% lower than the Dimitrovgrad and the
Studsvik 2003 values.
In the case of nPu/
238U values, the Harwell and Dimitrovgrad data on one hand and the
Studsvik values on the other hand form pairs. This is also reflected in the elemental ratios
(Table 20). This picture could be explained with an erroneous plutonium spike concentration
in the Studsvik analyses. Another, speculative, explanation for such an effect could be a small
real difference of the plutonium to uranium ratio in the two sample solutions, caused by the
fact that two pellet halves had been dissolved in one case, a 2 mm slice only in the other case.
Unfortunately, the information necessary for calculating a mass balance is incomplete. The
total mass of the mother solution was not determined.
ZC-08/001
24
Table 18 Isotopic composition (atom%) of F3F6 sample
Uranium 234
U 235
U 236
U 238
U
Harwell 0.016 0.357 0.624 99.003
Dimitrovgrad 0.013 0.352 0.630 99.005
Studsvik 2003 0.018 0.360 0.642 98.980
Studsvik 2006 0.020 0.356 0.700 98.924
Plutonium(a) 238
Pu 239
Pu 240
Pu 241
Pu 242
Pu 244
Pu
Harwell 3.974 45.571 30.142 7.725 12.588 0.001
Dimitrovgrad 4.315 45.275 30.086 7.711 12.613
Studsvik 2003 3.837 45.754 30.389 7.529 12.492
Studsvik 2006 4.003 45.661 30.218 7.583 12.535
Neodymium 142
Nd 143
Nd 144
Nd 145
Nd 146
Nd 148
Nd 150
Nd
Harwell(b) 1.032 14.298 37.443 15.197 18.467 9.079 4.483
Dimitrovgrad(b) 0.876 14.297 37.493 15.177 18.399 9.159 4.599
Studsvik 2003 0.851 14.400 38.200 14.919 17.929 9.255 4.446
Studsvik 2006 0.840 14.250 37.556 15.383 18.532 9.043 4.396
Cerium 140
Ce 142
Ce
Dimitrovgrad(c) 52.860 47.140
Studsvik 2003 52.496 47.504
Studsvik 2006 52.561 47.439 (a
) Decay-corrected to December 31, 2006, renormalised (244
Cm decay into 240
Pu based on Dimitrovgrad data) (b)
Decay of remaining 144
Ce into 144
Nd taken into account (based on Dimitrovgrad analysis) and renormalized (c)
144
Ce not taken into account
Table 19 Amount of nuclide nX (weight%) relative to
238U in F3F6 sample
Uranium 234
U 235
U 236
U
Harwell 0.016% 0.356% 0.625%
Dimitrovgrad 0.013% 0.351% 0.631%
Studsvik 2003 0.017% 0.359% 0.643%
Studsvik 2006 0.020% 0.356% 0.701%
Plutonium 238
Pu 239
Pu 240
Pu(a) 241
Pu(a) 242
Pu
Harwell 0.0467% 0.535% 0.348% 0.0907% 0.148%
Dimitrovgrad 0.0504% 0.529% 0.346% 0.0901% 0.147%
Studsvik 2003 0.0439% 0.524% 0.346% 0.0862% 0.143%
Studsvik 2006 0.0449% 0.512% 0.339% 0.0851% 0.141%
Neodymium 142
Nd 143
Nd 144
Nd(b) 145
Nd 146
Nd 148
Nd 150
Nd
Harwell 0.0076% 0.107% 0.281% 0.115% 0.140% 0.0700% 0.0350%
Dimitrovgrad 0.0069% 0.113% 0.299% 0.122% 0.149% 0.0751% 0.0382%
Studsvik 2003 0.0065% 0.110% 0.294% 0.116% 0.140% 0.0732% 0.0356%
Studsvik 2006 0.0062% 0.106% 0.282% 0.116% 0.141% 0.0698% 0.0344%
Cerium 140
Ce 142
Ce
Dimitrovgrad 0.250% 0.226%
Studsvik 2003 0.254% 0.233%
Studsvik 2006 0.227% 0.208% (a
) Decay-corrected to December 31, 2006 (244
Cm decay into 240
Pu based on Dimitrovgrad data) (b)
Decay of remaining 144
Ce into 144
Nd taken into account (based on Dimitrovgrad analysis)
ZC-08/001
25
Table 20 Elemental ratios in F3F6 sample
Pu/U Nd/U Ce/U
Harwell 1.16E-02 7.48E-03
Dimitrovgrad 1.16E-02 7.95E-03 4.72E-03
Studsvik 2003 1.13E-02 7.67E-03 4.82E-03
Studsvik 2006 1.11E-02 7.48E-03 4.30E-03
0.0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
U-234 U-235 U-236
Isotope
Ab
un
da
nc
e [
%] Harwell
Dimitrovgrad
Studsvik 2003
Studsvik 2006
-25%
-20%
-15%
-10%
-5%
0%
5%
10%
15%
20%
25%
U-234 U-235 U-236
IsotopeD
evia
tio
n f
rom
mean
Harwell
Dimitrovgrad
Studsvik 2003
Studsvik 2006
Figure 18 Isotopic abundance of uranium isotopes and relative deviation from mean value
0.0
5.0
10.0
15.0
20.0
25.0
30.0
35.0
40.0
45.0
50.0
Pu-238 Pu-239 Pu-240 Pu-241 Pu-242
Isotope
Ab
un
da
nc
e [
%]
Harwell
Dimitrovgrad
Studsvik 2003
Studsvik 2006
-6%
-4%
-2%
0%
2%
4%
6%
8%
Pu-238 Pu-239 Pu-240 Pu-241 Pu-242
Isotope
Devia
tio
n f
rom
mean
Harwell
Dimitrovgrad
Studsvik 2003
Studsvik 2006
Figure 19 Isotopic abundance of plutonium isotopes and relative deviation from mean value
0.0
5.0
10.0
15.0
20.0
25.0
30.0
35.0
40.0
Nd-142 Nd-143 Nd-144 Nd-145 Nd-146 Nd-148 Nd-150
Isotope
Ab
un
da
nc
e [
%]
Harwell
Dimitrovgrad
Studsvik 2003
Studsvik 2006
-8%
-6%
-4%
-2%
0%
2%
4%
6%
8%
10%
Nd-142 Nd-143 Nd-144 Nd-145 Nd-146 Nd-148 Nd-150
Isotope
De
via
tio
n f
rom
me
an
Harwell
Dimitrovgrad
Studsvik 2003
Studsvik 2006
14.73
Figure 20 Isotopic abundance of neodymium isotopes and relative deviation from mean
value
ZC-08/001
26
44.0
45.0
46.0
47.0
48.0
49.0
50.0
51.0
52.0
53.0
54.0
Ce-140 Ce-142
Isotope
Ab
un
da
nc
e [
%] Dimitrovgrad
Studsvik 2003
Studsvik 2006
-0.6%
-0.5%
-0.4%
-0.3%
-0.2%
-0.1%
0.0%
0.1%
0.2%
0.3%
0.4%
0.5%
Ce-140 Ce-142
Isotope
De
via
tio
n f
rom
me
an
Dimitrovgrad
Studsvik 2003
Studsvik 2006
Figure 21 Isotopic abundance of cerium isotopes and relative deviation from mean value
0.0%
0.1%
0.2%
0.3%
0.4%
0.5%
0.6%
Pu-238 Pu-239 Pu-240 Pu-241 Pu-242
Isotope
nX
/238U
[w
t%]
Harwell
Dimitrovgrad
Studsvik 2003
Studsvik 2006
-6%
-4%
-2%
0%
2%
4%
6%
8%
Pu-238 Pu-239 Pu-240 Pu-241 Pu-242
Isotope
De
via
tio
n f
rom
me
an
Harwell
Dimitrovgrad
Studsvik 2003
Studsvik 2006
Figure 22 Amount of nuclide nPu (weight%) relative to
238U and relative deviation from
mean value
0.0%
0.1%
0.1%
0.2%
0.2%
0.3%
0.3%
0.4%
Nd-142 Nd-143 Nd-144 Nd-145 Nd-146 Nd-148 Nd-150
Isotope
nX
/238U
[w
t%]
Harwell
Dimitrovgrad
Studsvik 2003
Studsvik 2006
-9%
-7%
-5%
-3%
-1%
1%
3%
5%
7%
Nd-142 Nd-143 Nd-144 Nd-145 Nd-146 Nd-148 Nd-150
Isotope
De
via
tio
n f
rom
me
an
Harwell Dimitrovgrad
Studsvik 2003 Studsvik 200612.29
Figure 23 Amount of nuclide nNd (weight%) relative to
238U and relative deviation from
mean value
0.00%
0.05%
0.10%
0.15%
0.20%
0.25%
0.30%
Ce-140 Ce-142
Isotope
nX
/238U
[w
t%]
Dimitrovgrad Studsvik 2003 Studsvik 2006
-8.0%
-6.0%
-4.0%
-2.0%
0.0%
2.0%
4.0%
6.0%
Ce-140 Ce-142
Isotope
De
via
tio
n f
rom
me
an
Dimitrovgrad Studsvik 2003 Studsvik 2006
Figure 24 Amount of nuclide nCe (weight%) relative to
238U and relative deviation from
mean value
ZC-08/001
27
10. ALTERNATIVE BURNUP DETERMINATION
10.1 INTRODUCTORY REMARKS
The most common method for determining the burnup of irradiated light water reactor (LWR)
fuel is the 148
Nd method [13] according to ASTM E-321. Probably one of the largest sources
for systematic errors in this method is the assumed fission yield, requiring knowledge of the
fraction of fissions occurring in 238
U (fast neutron fission) and 235
U, 239
Pu and 241
Pu (thermal).
Another traditional method for burnup determination is based on the uranium and plutonium
isotopic composition (ASTM E 244) [14]; however, this method is rarely used for LWR fuel
due to its rather simplified and rough assumptions regarding the neutron spectrum and fission
fractions (the standard was withdrawn in 2001). On the other hand, modern physics codes like
CASMO and HELIOS are able to calculate the amount of fission products and actinides
formed or consumed during reactor operation in a much more sophisticated way, taking
changes of irradiating conditions into account in much more detail than in the ASTM E-321
and ASTM E-244 methods. The uncertainty of these methods can therefore be eliminated to a
certain extent, if the experimentally determined amount of suitable fission products or acti-
nides is compared to the result of, for instance, CASMO calculations. In collaboration with
Vattenfall Nuclear Fuel, Studsvik has tested and implemented a corresponding alternative
burnup determination method, by comparing isotopic data from the F3F6 sample with
CASMO calculations [15].
10.2 CASMO CALCULATIONS
Assembly 14595 was never located in a control cell or on the core periphery. Axial and radial
distribution of power in the reactor core is checked at regular intervals by means of travelling
in-core probes (TIP). Power and void for the region that contained sample F3F6 were
determined for every TIP run date, based on core tracking calculations3. The lifetime
simulation was then divided into a reasonable number of periods and representative power
and void values were estimated for each period. These values served as input for a CASMO-4
infinite lattice simulation. Power and void history based on core tracking calculations are
shown in Figure 25 together with the values used in the simulation.
Number densities and relative weight percent of all uranium, plutonium and neodymium
isotopes were calculated as a function of nodal average burnup for the interval 55 –
65 MWd/kgU. The calculated values were transformed into nX/
238U weight ratios.
The following program versions were used for core-follow calculations:
CASMO-4, version 1.13.04
CORELINK, version 3.4.13
POLCA7, version 3.0.6.3
The following program version was used for the single-assembly simulation:
CASMO-4, version 2.05.06 with JEFF2.2 library
3 Sample F3F6 was located in the lowermost part of node 14 of 25 axial nodes; by mistake, CASMO calculations
were performed for node 13. As axial distributions are flat at core mid-height, this error does not significantly
impact the result.
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Figure 25 Nodal power and void history based on core tracking (filled diamonds) and used
for CASMO-4 simulation (open squares)
10.3 BURNUP DETERMINATION BASED ON 2003 DATA
Experimentally determined values for 144
Nd, 145
Nd, 146
Nd, 148
Nd and 150
Nd were compared to
the calculated nNd/
238U weight ratios, thus allowing determination of the local pellet burnup.
In addition, the local pellet burnup was determined by comparing experimentally analysed 235
U and 239
Pu isotopic abundances to abundances calculated from CASMO number densities.
Figure 26 illustrates the principle; the results are compiled in Table 21. In contrast to [15], the
plutonium abundances were corrected for 241
Pu decay and 240
Pu formation from 244
Cm decay
back to the end of irradiation, whereas 144
Nd was compared to the calculated sum of 144
Nd
and 144
Ce at the date of analysis. Moreover, the weighted average is not calculated from all
values, as the individual results based on neodymium content are not independent from each
other. Instead, a weighted average is calculated from all neodymium values first. Another
weighted average is then calculated from this neodymium value and the two values based on
the isotopic abundance of 235
U and 239
Pu, respectively.
The indicated errors were calculated according to the rules for error propagation from errors
indicated elsewhere. No error of the CASMO calculations was taken into account.
Based on ORIGEN calculations, it can be assumed that the energy released per fission in fuel
with 4% initial enrichment irradiated to 60 MWd/kgU is about 205 MeV. This corresponds to
9.63 MWd/kgU per %FIMA. Thus, the overall weighted average of (60.7±0.4) MWd/kgU
corresponds to (6.30±0.04) %FIMA, to be compared to (6.03±0.07)%FIMA determined in
Harwell and to the Dimitrovgrad values of (6.33±0.06) and (6.30±0.05) %FIMA.
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0.110%
0.113%
0.116%
0.119%
0.122%
0.125%
56 57 58 59 60 61 62
Burnup [MWd/kgU]
nN
d/2
38U
[w
t%]
Nd-143 CASMO Nd-145 CASMO Nd-145 Exp.
0.25%
0.26%
0.27%
0.28%
0.29%
0.30%
0.31%
57 58 59 60 61 62 63
Burnup [MWd/kgU]
144N
d/2
38U
[w
t%]
0.12%
0.13%
0.14%
0.15%
0.16%
0.17%
0.18%
146N
d/2
38U
[w
t%]
Nd-144 CASMO Nd-144 Exp.
Nd-146 CASMO Nd-146 Exp.
0.060%
0.064%
0.068%
0.072%
0.076%
0.080%
57 58 59 60 61 62 63
Burnup [MWd/kgU]
148N
d/2
38U
[w
t%]
0.030%
0.032%
0.034%
0.036%
0.038%
0.040%
150N
d/2
38U
[w
t%]
Nd-148 CASMO Nd-148 Exp.
Nd-150 CASMO Nd-150 Exp.
0.30%
0.35%
0.40%
0.45%
0.50%
0.55%
57 58 59 60 61 62 63
Burnup [MWd/kgU]
235U
Ab
un
da
nc
e
40%
41%
42%
43%
44%
45%
239P
u A
bu
nd
an
ce
U-235 CASMO U-235 Exp.
Pu-239 CASMO Pu-239 Exp.
Figure 26 Principle of burnup determination by comparing experimentally determined nNd/
238U weight ratios as well as
235U and
239Pu isotopic abundances to
corresponding CASMO data
Table 21 Burnup values based on the comparison of experimentally determined Studsvik
2003 values with values calculated by CASMO
Burnup based on experimental value of4 [MWd/kgU]
144Nd/
238U: (0.294±0.007)% 61.5±1.0
145Nd/
238U:
(0.116±0.003)% 56.8±1.8
146Nd/
238U:
(0.140±0.003)% 58.6±0.9
148Nd/
238U: (0.073±0.002)% 60.2±1.7
150Nd/
238U: (0036±0.002)% 59.6±2.3
Weighted average of all Nd values: 59.5±0.7 235
U abundance: (0.360±0.010)% 61.4±0.4 239
Pu abundance: (42.9±0.5)% 60.6±0.9
Weighted average (Nd, 235
U and 239
Pu abund.) 60.7±0.4
10.4 METHOD APPLICATION ON HARWELL, DIMITROVGRAD AND
STUDSVIK 2006 DATA
The method described in 10.3 was applied on experimental data determined in Harwell,
Dimitrovgrad and Studsvik 2006. The results are compiled in Table 22. Keeping in mind that
the indicated errors are based on 1σ errors of the corresponding experimental data, ignoring
4 Taken from Table 14 and Table 15
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any errors of the CASMO calculations, the data give a consistent overall picture. The
maximum difference is less than the difference of burnup values determined by AEA
Technology and SCC RIAR by means of theASTM E321 148
Nd standard method.
Table 22 Burnup values based on the comparison of experimental values determined by
Harwell, Dimitrovgrad and Studsvik 2006 with values calculated by CASMO
Burnup [MWd/kgU] based on experimental value of Harwell Dimitrovgrad Studsvik 2006 144
Nd/238
U 59.1±0.4 62.3±0.3 59.6±0.7 145
Nd/238
U
56.2±0.7 61.1±0.6 57.2±1.5 146
Nd/238
U
58.8±0.5 61.6±0.3 59.0±0.7 148
Nd/238
U 57.6±0.6 61.7±0.4 57.4±0.9 150
Nd/238
U 58.7±0.5 63.2±0.4 57.7±1.1
Weighted average of all Nd values 58.3±0.2 62.0±0.2 58.4±0.4 235
U abundance 61.6±0.1 61.8±0.2 61.6±0.4 239
Pu abundance 60.1±0.1 61.6±0.1 60.8±0.7
Weighted average (Nd, 235
U and 239
Pu abund.) 60.8±0.1 61.8±0.1 60.2±0.3
%FIMA (9.63 MWd/kgU per %FIMA) 6.32±0.01 6.42±0.01 6.26±0.03
11. CONCLUSIONS
A sample from the central part of a fuel rod irradiated in the Swedish boiling water reactor
Forsmark 3 to a burnup of about 58 MWd/kgU was dissolved in Studsvik. Aliquots of this
solution were shipped to two well-recognised independent laboratories for the determination
of the isotopic composition and for radiochemical burnup analysis. Both laboratories applied
methods equivalent to the ASTM E321 standard (isotope dilution analysis, radiochemical
separations by ion exchange chromatography, thermal ionisation mass spectrometry). Later
on, a sample adjacent to the one taken for these analyses was dissolved and analysed in
Studsvik, applying isotope dilution analysis as well, but by means of HPLC-ICP-MS. Three
years later, the same solution was re-analysed in Studsvik with new equipment.
Overall, the four sets of uranium, plutonium, neodymium and cerium (three sets only) isotopic
data form a consistent package. Only in three cases, single isotopic abundance values seem to
deviate significantly from the group of the other three values. In one case (234
U abundance),
the two Studsvik values deviate significantly from the two values of the earlier analyses.
Significant systematic deviations of one set of nX/
238U values from the other ones indicate two
cases of spiking errors or selective loss of material.
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12. ACKNOWLEDGEMENTS
The exercise including isotope analyses in Harwell and Dimitrovgrad was funded by ABB
Atom AB (now Westinghouse Electric Sweden AB) and the utilities Vattenfall and OKG.
Many members of the Studsvik staff contributed to the analyses of the F3F6 samples. Special
thanks go to Ulla-Britt Eklund, involved in the inter-laboratory comparison exercise, to
Jeanett Low and Michael Granfors, who performed the isotope analyses at Studsvik, and to
Gunnar Lysell, project manager of the inter-laboratory comparison exercise, for his
contributions to the data evaluation, for all the good ideas during many discussions and for
careful review of several reports.
Many thanks go to Anders Wallander, Westinghouse Electric Sweden AB. He performed a
big effort for organising and providing all relevant fabrication and pre-irradiation data.
Active support of the present work by Vattenfall Nuclear Fuel AB, in particular by Ewa
Kurcyusz and David Schrire is much appreciated. Andreas Lidén did a great job, when
compiling all the reactor-specific data, irradiation history and modelling results. Without his
help, it would not have been possible to complete this work.
The compilation of these data and the Swedish support of the OECD/NEA Expert Group on
Assay Data of Spent Nuclear Fuel are funded by Vattenfall and SKB. Many thanks go to
Ingemar Zelbi (SKB) for his coordinating efforts.
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13. REFERENCES
[1] 2006 Erfarenheter från driften av de svenska kärnkraftwerken, ISSN-1654-0484,
Kärnkraftsäkerhet och Utbildning AB ( http://www.ksu.se/Svensk_2006.pdf )
[2] A. Lidén, Vattenfall Nuclear Fuel AB, Personal communication (e-mail “SV: Data
för SFCOMPO”), February 6, 2008
[3] http://www.nea.fr/html/science/wpncs/ADSNF/index.html
[4] A. Wallander, Westinghouse Electric Sweden AB, Personal communication (e-mail
“Re: SFCOMPO; Information on Forsmark assembly 14595 and its rod F6”), January
30, 2008
[5] F. Jatuff, Recent and future activities in PROTEUS, LWR-PROTEUS, overview and
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[6] G. Lysell, Resultat och provlägen för utbränningsprov inom projekten HCL-113, -119
och -134, Studsvik Technical Note N(H)-96/63, 1996-09-20
[7] J.A. Tibbles, P.K. Ivison, Radiochemical burnup analysis, a report for Studsvik
Nuclear AB, AEA Technology Report AEAT-0589, July 1996
[8] The determination of nuclide composition and atom percent fission in Uranium fuel,
Report of State Scientific Centre of Russian Federation Research Institute of Atomic
Reactors, undated
[9] S. Röllin et al., Determination of lanthanides in uranium materials using high
performance liquid chromatographic separation and ICP-MS detection, "Recent
Advances in Plasma Source Mass Spectrometry" (Ed. G. Holland), 28-35, 1994
[10] H.U. Zwicky, J. Low, Fuel pellet isotopic analyses of Vandellós 2 rods WMtR124
and WZR0046: Qualification of method, Studsvik Technical Note N(H)-04/002
Rev. 1, 2004-09-09
[11] H.U. Zwicky, J. Low, M. Granfors, Additional Fuel Pellet Isotopic Analyses of
Vandellós 2 Rods WZtR160 and WZR0058, Final Report, Studsvik Report N-07/140,
DRAFT
[12] K.E. Jarvis, A.L. Gray, Handbook of ICP-MS, ISBN 0-216-92912-1
[13] Standard test method for atom percent fission in uranium and plutonium fuel
(Neodymium-148 method), ASTM Standard E-321-96
[14] Standard test method for atom percent fission in uranium and plutonium fuel (mass
spectrometric method), ASTM Standard E 244-80 (1995) (Withdrawn 2001)
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irradiated fuel by means of isotopic analysis compared to CASMO calculations
2005 Water Reactor Fuel Performance Meeting, October 2-6, 2005, Kyoto, Japan