ISOTOPIC DATA OF SAMPLE F3F6 FROM A ROD IRRADIATED IN THE SWEDISH BOILING WATER REACTOR FORSMARK 3

Transcription

1 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) Fax +41 (0) hans-urs.zwicky@bluewin.ch

2 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. 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.

3 CONTENT Page 1. Introduction Reactor and Core Mechanical and Nuclear Assembly and Rod Design Assembly Surrounding Assemblies Fuel Rod in Position F Irradiation History Analysed Samples Nuclide and Burnup Analysis Performed in Harwell Experimental Deatails Results Nuclide and Burnup Analysis Performed in Dimitrovgrad Experimental Deatails Results Nuclide Analyses Performed in Studsvik Campaign Dissolution Old HPLC-ICP-MS Instrument Isotope Dilution Analysis Campaign New HPLC-ICP-MS Instrument Isotope Dilution Analysis Data Comparison and Discussion Alternative Burnup Determination Introductory Remarks CASMO Calculations Burnup Determination Based on 2003 Data Method Application on Harwell, Dimitrovgrad and Studsvik 2006 Data Conclusions Acknowledgements References II

4 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 Figure 7 Forsmark 3 core burnup during cycles 3 to Figure 8 Forsmark 3 primary system pressure during cycles 3 to Figure 9 Position of assembly inforsmark 3 core during cycles 3 to Figure 10 Assembly types and exposures adjacent to assembly during cycles 3 to Figure 11 Burnup of pin 14595/F6 during cycles 3 to Figure 12 Linear heat generation rate of pin 14595/F6 during cycles 3 to Figure 13 Nodal void representative for analysed sample location during cycles 3 to Figure 14 Nodal moderator temperature representative for analysed sample location during cycles 3 to Figure 15 Nodal fuel temperature representative for analysed sample location during cycles 3 to Figure 16 Scheme of fuel analysis in Dimitrovgrad Figure 17 Scheme of chemical separations (techniques by SSC RIAR) Figure 18 Isotopic abundance of uranium isotopes and relative deviation from mean value Figure 19 Isotopic abundance of plutonium isotopes and relative deviation from mean value Figure 20 Isotopic abundance of neodymium isotopes and relative deviation from mean value Figure 21 Isotopic abundance of cerium isotopes and relative deviation from mean value Figure 22 Amount of nuclide n Pu (weight%) relative to 238 U and relative deviation from mean value Figure 23 Amount of nuclide n Nd (weight%) relative to 238 U and relative deviation from mean value Figure 24 Amount of nuclide n Ce (weight%) relative to 238 U and relative deviation from mean value Figure 25 Nodal power and void history based on core tracking (filled diamonds) and used for CASMO-4 simulation (open squares) Figure 26 Principle of burnup determination by comparing experimentally determined n Nd/ 238 U weight ratios as well as 235 U and 239 Pu isotopic abundances to corresponding CASMO data III

5 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 UO 2 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 Table 7 ASTM E calculation of 148 Nd effective fractional fission yield Table 8 Spike isotopes and enrichment used in analyses performed in Dimitrovgrad Table 9 Isotopic composition [atom%] for sample FFBU determined in Dimitrovgrad 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 Table 12 Abundances of spike isotope and isotope to be analysed in spike solutions Table 13 Errors of input data used in calculations Table 14 Isotopic composition (atom%) of F3F6 sample, determined 2003 in Studsvik Table 15 Amount of nuclide n X (weight%) relative to 238 U, determined 2003 in Studsvik Table 16 Isotopic composition (atom%) of F3F6 sample, determined 2006 in Studsvik Table 17 Amount of nuclide n X (weight%) relative to 238 U, determined 2006 in Studsvik Table 18 Isotopic composition (atom%) of F3F6 sample Table 19 Amount of nuclide n X (weight%) relative to 238 U in F3F6 sample Table 20 Elemental ratios in F3F6 sample Table 21 Burnup values based on the comparison of experimentally determined Studsvik 2003 values with values calculated by CASMO Table 22 Burnup values based on the comparison of experimental values determined by Harwell, Dimitrovgrad and Studsvik 2006 with values calculated by CASMO IV

6 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 laboratories 1 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 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 MW th. 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 MW th. 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 1

7 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: 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 and its rod in position F6 was provided by Westinghouse Electric Sweden AB [4]. 3.1 ASSEMBLY Assembly 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 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 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. 2

8 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 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. 3

9 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 UO 2 /Gd 235 U Enrichment [%] # of Gd rods Gd 2 O 3 content [%] Poisoning (a) water rods (number) (1x) - Outer channel width [mm] Channel wall thickness [mm] Sub-channel size [mm] Rod outer diameter [mm] (52x) (64x) (12x) (100x) (64x) Cladding wall thickness [mm] Rod pitch [mm] (16.05/15.8) (12.55) 15.8 Wide water gap [mm] Narrow water gap [mm] (a) Poisoning: Old ASEA concept ( zebra fuel ). p = 1: all pellets in a Gd rod are Gd pellets; p = : every third pellet is a UO 2 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/cm 3. 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 ( /-0.10) g/cm 3. Fuel pellet composition and impurities are listed in Table 4. 4

10 Table 3 Zircaloy-2 cladding composition Main components Maximum amount of impurities Element [wt%] Element [ppm] Element [ppm] Element [ppm] Sn Al 75 Cu 50 N 80 Fe B 0.5 Hf 100 Si 200 Cr Cd 0.5 H 25 Na 20 Ni C 270 Mg 20 Ti 50 O Cl 20 Mn 50 W 100 Zr Remainder Co 20 Mo 50 U 3.5 Fe+Cr+Ni Table 4 UO 2 fuel composition Main components Maximum amount of impurities Element [wt%] Element [ppm] Element [ppm] Element [ppm] Element [ppm] Uranium (tot) 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 U Bi 2.0 F 15 Ni 50 Dy U C 20 Fe 100 Pb 20 Eu U Ca 25 In 3.0 Si 100 Gd U Cd 0.5 Li 2.0 Sn 5.0 Sm 2 Cl 25 Mg 50 Ti 40 Na 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 is shown in Figure 9. Figure 10 contains information on assembly types and exposure in positions adjacent to assembly 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] Uranium init Core Inventory [t] 3 August 1, 1987 August 13, September 3, 1988 June 10, July 8, 1989 July 14, August 1, 1990 August 17, September 4, 1991 May 15, June 18, 1992 June 6,

11 Primary System Pressure [bar] Core Burnup [MWd/kgU] Reactor Power [% of 3020 MW] ZC-08/ Effective Full Power Hours Figure 6 Forsmark 3 reactor power during cycles 3 to Effective Full Power Hours Figure 7 Forsmark 3 core burnup during cycles 3 to Effective Full Power Hours Figure 8 Forsmark 3 primary system pressure during cycles 3 to 8 6

12 Figure 9 Position of assembly in Forsmark 3 core during cycles 3 to 8 IA Cycle 3 Cycle 4 Cycle 5 IA IA E IA IA IA E IA E E IA E Cycle 6 Cycle 7 Cycle 8 E D E E D E E E E E E (Numbers below assembly type: bundle and nodal exposure at first TIP measurement [MWd/kgU]) Figure 10 Assembly types and exposures adjacent to assembly during cycles 3 to 8 7

13 Void [%] Pin Linear Heat Generation Rate [kw/m] Pin Burnup [MWd/kgU] ZC-08/ Effective Full Power Hours Figure 11 Burnup of pin 14595/F6 during cycles 3 to Effective Full Power Hours Figure 12 Linear heat generation rate of pin 14595/F6 during cycles 3 to Effective Full Power Hours 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 8

14 Fuel Temperature [C] Moderator Temperature [C] ZC-08/ Effective Full Power Hours Figure 14 Nodal moderator temperature representative for analysed sample location during cycles 3 to Effective Full Power Hours 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 mm from the lower end plug [6]. The fuel matrix, but not alloy particles and cladding material, was dissolved in concentrated HNO 3. 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). 9

15 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 E 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 E 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]. 10

16 Table 6 Isotopic composition and burnup for sample FFBU determined in Harwell Table 7 ASTM E calculation of 148 Nd effective fractional fission yield 11

17 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 E and ASTM E 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 Cm was determined by alpha spectrometry. Figure 16 Scheme of fuel analysis in Dimitrovgrad 12

18 Figure 17 Table 8 Scheme of chemical separations (techniques by SSC RIAR) Spike isotopes and enrichment used in analyses performed in Dimitrovgrad Isotope 233 U 242 Pu 243 Am 146 Nd 140 Ce Enrichment [%] ± ± ± ± ±

19 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

20 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 CAMPAIGN Dissolution The 2 mm fuel rod slice was placed in a glass flask together with 20 ml of concentrated HNO 3 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. 15

21 In the order of g of the original fuel solution was diluted into 100 ml of HNO 3 (7.5 M) in the hotcell. 20 ml of this solution were transferred to the laboratory. An appropriate aliquot was diluted with 100 ml HNO 3 (0.16 M) to a target uranium concentration of about 4 ppm. The uranium concentration was determined by Scintrex analysis. The Scintrex 2 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 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] 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: a spike isotope b isotope to beanalysed R R R N N N N s Sp M S a S b isotope ratio ( a / b) in sample Sp a Sp b isotope ratio in spike isotope ratio in mixture number of isotope a in sample number of isotope a in spike number of isotope b in sample number of isotope b in spike N Rs Eq. 1 N S a S b Sp a Sp b N Rsp Eq. 2 N 2 SCINTREX UA-3 Uranium Analyser, SCINTREX, Snidercroft Road, Concord Ontario Canada L4K 1B5 16

22 R M S a S b Sp a Sp b N N Eq. 3 N N By transforming Eq. 3, the following Eq. 4 can be derived N S b Sp Sp N a RM N b Eq. 4 RM Rs Sp N b can be substituted by means of Eq. 2, which leads to Eq. 5 N S b N Sp a RM 1 RSp R R M s 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 R S, the isotope ratio in the sample, is given. R Sp, the ratio in the spike is fixed as well, once the appropriate standard is chosen for a series of analyses. R M, 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 R M value: counting statistics, influencing the uncertainty of the isotopic ratio, and the factor that determines the contribution of the uncertainty in R M 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 R S values were estimated based on the result of semi-quantitative analyses and on CASMO calculations. After choosing an appropriate R M 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 [%] 233 U Pu Ce Nd Isotope to be analysed Abundance [%] 238 U Pu Ce Nd 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 17

23 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 isotopes 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. R S was determined based on the corresponding abundances; R M 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 n X in weight percent relative to 238 U 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). with a r Eq. 6 b r isotopic ratio a, b peak areas 18

24 s r 1 1 Eq. 7 r a b with s r error of r 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 10 6, 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 10 5 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, s r 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 s Sp N R S RSp sr s S R R sr a M S M Sp s S N b Eq. 8 Nb Sp N a RSp RM RM RS RM RS RSp RM RSp with s i absolute error of i For all other isotopes, Eq. 9 is applied: 2 2 s s N s b r sn N x Eq. 9 x s N b r N The relative error of the number of added spike atoms a and the relative error of R Sp Sp N a sr Sp used in the calculations are estimated as shown in Table 13. They correspond to 1σ. RSp s Sp 19

25 Table 13 Errors of input data used in calculations Parameter Relative Error (Comment) Sp N a 1% (Estimated, same value for all elements) R Sp R S,R M, r U 0.1% (Estimated) Pu 0.1% (Estimated) Ce 1% (Estimated) Nd 0.5% (Estimated) 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 n X in weight percent relative to 238 U. Table 14 Isotopic composition (atom%) of F3F6 sample, determined 2003 in Studsvik 234 U 235 U 236 U Uranium Mean Uncertainty Plutonium 238 Pu 239 Pu 240 Pu 241 Pu 242 Pu Mean Uncertainty Neodymium 142 Nd 143 Nd 144 Nd 145 Nd 146 Nd 148 Nd 150 Nd Mean Uncertainty Cerium 140 Ce 142 Ce Mean Uncertainty Date of analysis: October 8, 2003 (Pu); October 20, 2003 (Nd, Ce); November 4, 2003 (U) 238 U 20

26 Table 15 Amount of nuclide n X (weight%) relative to 238 U, determined 2003 in Studsvik 234 U 235 U 236 U Uranium 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 % 0.110% 0.294% 0.116% 0.140% 0.073% 0.036% Uncertainty % 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) CAMPAIGN 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 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 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 n X in weight percent relative to 238 U. 21

27 Table 16 Isotopic composition (atom%) of F3F6 sample, determined 2006 in Studsvik 234 U 235 U 236 U Uranium Mean Uncertainty Plutonium 238 Pu 239 Pu 240 Pu 241 Pu 242 Pu Mean Uncertainty Neodymium 142 Nd 143 Nd 144 Nd 145 Nd 146 Nd 148 Nd 150 Nd Mean Uncertainty Cerium 140 Ce 142 Ce Mean Uncertainty Date of analysis: December 14, U Table 17 Amount of nuclide n X (weight%) relative to 238 U, determined 2006 in Studsvik 234 U 235 U 236 U Uranium 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 % 0.106% 0.282% 0.116% 0.141% 0.070% 0.034% Uncertainty % 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, DATA COMPARISON AND DISCUSSION In order to compare all results on a common basis, data were decay-corrected to December 31, Isotopic compositions were re-normalised if necessary. All four sets of isotopic compositions are compiled in Table 18, all n X/ 238 U 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 140 Ce and 142 Ce abundances were simply determined by normalising the corresponding contents to 100%. 22

28 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. n X/ 238 U 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 n X/ 238 U 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 n Pu/ 238 U 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. 23

29 Table 18 Isotopic composition (atom%) of F3F6 sample 234 U 235 U 236 U Uranium Harwell Dimitrovgrad Studsvik Studsvik Plutonium (a) 238 Pu 239 Pu 240 Pu 241 Pu 242 Pu 244 Pu Harwell Dimitrovgrad Studsvik Studsvik Neodymium 142 Nd 143 Nd 144 Nd 145 Nd 146 Nd 148 Nd 150 Nd Harwell (b) Dimitrovgrad (b) Studsvik Studsvik Cerium 140 Ce 142 Ce Dimitrovgrad (c) Studsvik Studsvik (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 238 U Table 19 Amount of nuclide n X (weight%) relative to 238 U in F3F6 sample 234 U 235 U 236 U Uranium Harwell 0.016% 0.356% 0.625% Dimitrovgrad 0.013% 0.351% 0.631% Studsvik % 0.359% 0.643% Studsvik % 0.356% 0.701% Plutonium 238 Pu 239 Pu 240 Pu (a) 241 Pu (a) 242 Pu Harwell % 0.535% 0.348% % 0.148% Dimitrovgrad % 0.529% 0.346% % 0.147% Studsvik % 0.524% 0.346% % 0.143% Studsvik % 0.512% 0.339% % 0.141% Neodymium 142 Nd 143 Nd 144 Nd (b) 145 Nd 146 Nd 148 Nd 150 Nd Harwell % 0.107% 0.281% 0.115% 0.140% % % Dimitrovgrad % 0.113% 0.299% 0.122% 0.149% % % Studsvik % 0.110% 0.294% 0.116% 0.140% % % Studsvik % 0.106% 0.282% 0.116% 0.141% % % Cerium 140 Ce 142 Ce Dimitrovgrad 0.250% 0.226% Studsvik % 0.233% Studsvik % 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) 24

30 Abundance [%] Deviation from mean Abundance [%] Deviation from mean Abundance [%] Deviation from mean ZC-08/001 Table 20 Elemental ratios in F3F6 sample Pu/ U Nd/ U Ce/ U Harwell 1.16E E-03 Dimitrovgrad 1.16E E E-03 Studsvik E E E-03 Studsvik E E E Harwell Dimitrovgrad Studsvik 2003 Studsvik 2006 U-234 U-235 U % 20% 15% 10% 5% 0% -5% -10% -15% -20% -25% Harwell Dimitrovgrad Studsvik 2003 Studsvik 2006 U-234 U-235 U-236 Isotope Isotope Figure 18 Isotopic abundance of uranium isotopes and relative deviation from mean value Harwell Dimitrovgrad Studsvik 2003 Studsvik 2006 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 8% 6% 4% 2% 0% -2% -4% -6% Harwell Dimitrovgrad Studsvik 2003 Studsvik 2006 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Isotope Isotope Figure 19 Isotopic abundance of plutonium isotopes and relative deviation from mean value Harwell Dimitrovgrad Studsvik 2003 Studsvik 2006 Nd-142 Nd-143 Nd-144 Nd-145 Nd-146 Nd-148 Nd % 8% 6% 4% 2% 0% -2% -4% -6% -8% Harwell Dimitrovgrad Studsvik 2003 Studsvik 2006 Nd-142 Nd-143 Nd-144 Nd-145 Nd-146 Nd-148 Nd-150 Isotope Isotope Figure 20 Isotopic abundance of neodymium isotopes and relative deviation from mean value 25

31 Deviation from mean Deviation from mean Deviation from mean Abundance [%] Deviation from mean ZC-08/ Ce-140 Dimitrovgrad Studsvik 2003 Studsvik 2006 Ce % 0.4% 0.3% 0.2% 0.1% 0.0% -0.1% -0.2% -0.3% -0.4% -0.5% -0.6% Ce-140 Dimitrovgrad Studsvik 2003 Studsvik 2006 Ce-142 Isotope Isotope Figure 21 Isotopic abundance of cerium isotopes and relative deviation from mean value n X/ 238 U [wt%] 0.6% 0.5% 0.4% 0.3% 0.2% Harwell Dimitrovgrad Studsvik 2003 Studsvik % 6% 4% 2% 0% Harwell Dimitrovgrad Studsvik 2003 Studsvik % -2% 0.0% Pu-238 Pu-239 Pu-240 Pu-241 Pu-242-4% -6% Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Isotope Isotope Figure 22 Amount of nuclide n Pu (weight%) relative to 238 U and relative deviation from mean value n X/ 238 U [wt%] 0.4% 0.3% 0.3% 0.2% 0.2% 0.1% 0.1% 0.0% Harwell Dimitrovgrad Studsvik 2003 Studsvik 2006 Nd-142 Nd-143 Nd-144 Nd-145 Nd-146 Nd-148 Nd-150 7% 5% 3% 1% -1% -3% -5% -7% -9% Harwell Dimitrovgrad Studsvik 2003 Studsvik 2006 Nd-142 Nd-143 Nd-144 Nd-145 Nd-146 Nd-148 Nd-150 Isotope Isotope Figure 23 Amount of nuclide n Nd (weight%) relative to 238 U and relative deviation from mean value 0.30% Dimitrovgrad Studsvik 2003 Studsvik % Dimitrovgrad Studsvik 2003 Studsvik % 4.0% n X/ 238 U [wt%] 0.20% 0.15% 0.10% 0.05% 0.00% Ce-140 Ce % 0.0% -2.0% -4.0% -6.0% -8.0% Ce-140 Ce-142 Isotope Isotope Figure 24 Amount of nuclide n Ce (weight%) relative to 238 U and relative deviation from mean value 26

32 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 actinides 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] CASMO CALCULATIONS Assembly 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 calculations 3. 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 MWd/kgU. The calculated values were transformed into n X/ 238 U weight ratios. The following program versions were used for core-follow calculations: CASMO-4, version CORELINK, version POLCA7, version The following program version was used for the single-assembly simulation: CASMO-4, version 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. 27

33 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 n Nd/ 238 U 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. 28

34 n Nd/ 238 U [wt%] Nd-143 CASMO Nd-145 CASMO Nd-145 Exp % 0.122% 0.119% 0.116% 0.113% 0.110% Burnup [MWd/kgU] 144 Nd/ 238 U [wt%] Nd-144 CASMO Nd-144 Exp. Nd-146 CASMO Nd-146 Exp. 0.31% 0.18% 0.30% 0.17% 0.29% 0.16% 0.28% 0.15% 0.27% 0.14% 0.26% 0.13% 0.25% 0.12% Burnup [MWd/kgU] 146 Nd/ 238 U [wt%] 0.080% Nd-148 CASMO Nd-150 CASMO Nd-148 Exp. Nd-150 Exp % 0.55% U-235 CASMO U-235 Exp. Pu-239 CASMO Pu-239 Exp. 45% 148 Nd/ 238 U [wt%] 0.076% 0.072% 0.068% 0.064% 0.038% 0.036% 0.034% 0.032% 150 Nd/ 238 U [wt%] 235 U Abundance 0.50% 0.45% 0.40% 0.35% 44% 43% 42% 41% 239 Pu Abundance 0.060% 0.030% Burnup [MWd/kgU] 0.30% 40% Burnup [MWd/kgU] Figure 26 Table 21 Principle of burnup determination by comparing experimentally determined n Nd/ 238 U weight ratios as well as 235 U and 239 Pu isotopic abundances to corresponding CASMO data Burnup values based on the comparison of experimentally determined Studsvik 2003 values with values calculated by CASMO Burnup based on experimental value of 4 [MWd/kgU] 144 Nd/ 238 U: (0.294±0.007)% 61.5± Nd/ 238 U: (0.116±0.003)% 56.8± Nd/ 238 U: (0.140±0.003)% 58.6± Nd/ 238 U: (0.073±0.002)% 60.2± Nd/ 238 U: (0036±0.002)% 59.6±2.3 Weighted average of all Nd values: 59.5± U abundance: (0.360±0.010)% 61.4± Pu abundance: (42.9±0.5)% 60.6±0.9 Weighted average (Nd, 235 U and 239 Pu abund.) 60.7± 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 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 29